WorldWideScience

Sample records for cooled plasma facing

  1. High quality actively cooled plasma facing components for fusion

    International Nuclear Information System (INIS)

    Nygren, R.

    1993-01-01

    This paper interweaves some suggestions for developing actively-cooled PFCs (plasma facing components) for future fusion devices with supporting examples taken from the design, fabrication and operation of Tore Supra's Phase III Outboard Pump Limiter (OPL). This actively-cooled midplane limiter, designed for heat and particle removal during long pulse operation, has been operated in essentially thermally steady state conditions. From experience with testing to identify braze flaws in the OPL, recommendations are made to analyze the impact of joining flaws on thermal-hydraulic performance of PFCs and to validate a method of inspection for such flaws early in the design development. Capability for extensive in-service monitoring of future PFCs is also recommended and the extensive calorimetry and IR thermography used to confirm and update safe operating limits for power handling of the OPL are reviewed

  2. The design of actively cooled plasma-facing components

    International Nuclear Information System (INIS)

    Scheerer, M.; Smid, I.; Bolt, H.; Gervash, A.; Linke, J.

    2001-01-01

    In future fusion devices, like in the stellarator Wendelstein 7-X, the target plates of the divertor will be exposed to heat loads up to power densities of 10 MW/m 2 for 1000 s. For this purpose actively cooled target elements with an internal coolant flow return, made of 2-D CFC armor tiles brazed onto a two tube cooling structure were developed and manufactured at the Forschungszentrum Juelich. Individual bent- and coolant flow reversal elements were used to achieve a high flexibility in the shape of the target elements. A special brazing technology, using a thin layer of plasma-arc deposited titanium was used for the bonding of the cooling structure to the plasma facing armor (PFA). FEM-simulations of the thermal and mechanical behavior show that a detachment of about 25% of the bonded area between the copper tubes and the PFA can be tolerated, without exceeding the critical heat flux at 15 MW/m 2 or a surface temperature of 1400 C at 10 MW/m 2 by using twisted tape inserts with a twist ratio of 2 at a cooling water velocity of 10 m/s. Thermal cycling tests in an electron beam facility up to a power density level 10.5 MW/m 2 show a very good behavior of parts of the target elements, which confirms the performance under fusion relevant conditions. Even defected parts in the bonding interface of the target elements, known from ultrasonic inspections before, show no change in the thermal performance under cycling, which confirms also the structural integrity of partly defected regions. (orig.)

  3. High quality actively cooled plasma-facing components for fusion

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1995-01-01

    This paper interweaves some suggestions for developing actively cooled plasma-facing components (PFCs) for future fusion devices, with supporting examples taken from the design, fabrication and operation of Tore Supra's Phase III outboard pump limiter (OPL). This actively cooled midplane limiter, designed for heat and particle removal during long-pulse operation, has been operated under essentially thermally steady state conditions. Testing to identify braze flaws, analysis of the impact of joining flaws on the thermal-hydraulic performance of the OPL, and the extensive calorimetry and IR thermography used to confirm and update safe operating limits for power handling of the OPL are reviewed. This experience suggests that, for PFCs in future fusion devices, flaw-tolerant designs are possible; analyses of the impacts of flaws on performance can provide criteria for quality assurance; and validating appropriate methods of inspection for such flaws early in the design development of PFCs is prudent. The need for in-service monitoring is also discussed. (orig.)

  4. Actively cooled plasma facing components qualification, commissioning and health monitoring

    International Nuclear Information System (INIS)

    Escourbiac, F.; Durocher, A.; Grosman, A.; Courtois, X.; Farjon, J.-L.; Schlosser, J.; Merola, M.; Tivey, R.

    2006-01-01

    In modern steady state magnetic fusion devices, actively cooled plasma facing components (PFC) have to handle heat fluxes in the range of 10-20 MW/m 2 . This generates a number of engineering constraints: the armour materials must be refractory and compatible with plasma wall interaction requirements (low sputtering and/or low atomic number); the heat sink must offer high thermal conductivity, high mechanical resistance and sufficient ductility; the component cooling system -which is generally based on the circulation of pressurized water in the PFC's heat sink - must offer high thermal heat transfer efficiency. Furthermore, the assembling of the refractory armour material onto the metallic heat sink causes generic difficulties strongly depending on thermo-mechanical properties of materials and design requirements. Life time of the PFC during plasma operation are linked to their manufacturing quality, in particular they are reduced by the possible presence of flaw assembling. The fabrication of PFC in an industrial frame including their qualification and their commissioning - which consists in checking the manufacturing quality during and at the end of manufacture - is a real challenge. From experience gained at Tore Supra on carbon fibre composite flat tiles technology components, it was assessed that a set of qualifications activities must be operated during R(and)D and manufacturing phases. Dedicated Non Destructive Technique (NDT) based on advanced active infrared thermography was developed for this purpose, afterwards, correlations between NDT, high heat flux testing and thermomechanical modelling were performed to analyse damage detection and propagation, and define an acceptance criteria valuable for industrial application. Health monitoring using lock-in technique was also recently operated in-situ of the Tore Supra tokamak for detection of possible defect propagation during operations, presence of acoustic precursor for critical heat flux detection induced

  5. Advanced qualification methodology for actively cooled plasma facing components

    Science.gov (United States)

    Durocher, A.; Escourbiac, F.; Grosman, A.; Boscary, J.; Merola, M.; Cismondi, F.; Courtois, X.; Farjon, J. L.; Missirlian, M.; Schlosser, J.; Tivey, R.

    2007-12-01

    The use of high heat flux plasma facing components (PFCs) in steady state fusion devices requires high reliability. These components have to withstand heat fluxes in the range 10-20 MW m-2 involving a number of severe engineering constraints. Feedback from the experience of various industrial manufacturings showed that the bonding of the refractory armour material onto the metallic heat sink causes generic difficulties strongly depending on material qualities and specific design. As the heat exhaust capability and lifetime of PFCs during plasma operation are directly linked to the manufacturing quality, a set of qualification activities such as active infrared thermography, lock-in and acoustic measurements were performed during the component development phases following a qualification route. This paper describes the major improvements stemming from better measurement accuracy and refined data processing and analyses recent developments aimed at investigating the capability to qualify the component in situ during its lifetime.

  6. Advanced qualification methodology for actively cooled plasma facing components

    International Nuclear Information System (INIS)

    Durocher, A.; Escourbiac, F.; Grosman, A.; Boscary, J.; Merola, M.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Missirlian, M.; Schlosser, J.; Tivey, R.

    2007-01-01

    The use of high heat flux plasma facing components (PFCs) in steady state fusion devices requires high reliability. These components have to withstand heat fluxes in the range 10-20 MW m -2 involving a number of severe engineering constraints. Feedback from the experience of various industrial manufacturings showed that the bonding of the refractory armour material onto the metallic heat sink causes generic difficulties strongly depending on material qualities and specific design. As the heat exhaust capability and lifetime of PFCs during plasma operation are directly linked to the manufacturing quality, a set of qualification activities such as active infrared thermography, lock-in and acoustic measurements were performed during the component development phases following a qualification route. This paper describes the major improvements stemming from better measurement accuracy and refined data processing and analyses recent developments aimed at investigating the capability to qualify the component in situ during its lifetime

  7. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.

    2004-01-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m 2 of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m 2 TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  8. High heat flux actively cooled plasma facing components development, realization and first results in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    2004-07-01

    The development, design, manufacture and testing of actively cooled high heat flux plasma facing components (PFC) has been an essential stage towards long powerful tokamak operations for Tore-Supra, it lasted about 10 years. This paper deals with the toroidal pumped limiter (TPL) that is able to sustain up to 10 MW/m{sup 2} of nominal heat flux. This device is based on hardened copper alloy heat sink structures covered by a carbon fiber composite armour, it resulted in the manufacturing of 600 elementary components, called finger elements, to achieve the 7.6 m{sup 2} TPL. This assembly has been operating in Tore-Supra since spring 2002. Some difficulties occurred during the manufacturing phase, the valuable industrial experience is summarized in the section 2. The permanent monitoring of PFC surface temperature all along the discharge is performed by a set of 6 actively cooled infrared endoscopes. The heat flux monitoring and control issue but also the progress made in our understanding of the deuterium retention in long discharges are described in the section 3. (A.C.)

  9. CFC/Cu bond damage in actively cooled plasma facing components

    International Nuclear Information System (INIS)

    Schlosser, J; Martin, E; Henninger, C; Boscary, J; Camus, G; Escourbiac, F; Leguillon, D; Missirlian, M; Mitteau, R

    2007-01-01

    Carbon fibre composite (CFC) armours have been successfully used for actively cooled plasma facing components (PFCs) of the Tore Supra (TS) tokamak. They were also selected for the divertor of the stellarator W7-X under construction and for the vertical target of the ITER divertor. In TS and W7-X a flat tile design for heat fluxes of 10 MW m -2 has been chosen. To predict the lifetime of such PFCs, it is necessary to analyse the damage mechanisms and to model the damage propagation when the component is exposed to thermal cycling loads. Work has been performed to identify a constitutive law for the CFC and parameters to model crack propagation from the edge singularity. The aim is to predict damage rates and to propose geometric or material improvements to increase the strength and the lifetime of the interfacial bond. For ITER a tube-in-tile concept (monoblock), designed to sustain heat fluxes up to 20 MW m -2 , has been developed. The optimization of the CFC/Cu bond, proposed for flat tiles, could be adopted for the monoblock concept

  10. Damage of actively cooled plasma facing components of magnetic confinement controlled fusion machines

    Energy Technology Data Exchange (ETDEWEB)

    Chevet, G. [Association Euratom-CEA, DSM/DRFC, CEA Cadarache, Saint-Paul-Lez-Durance (France)], E-mail: gaelle.chevet@cea.fr; Schlosser, J. [Association Euratom-CEA, DSM/DRFC, CEA Cadarache, Saint-Paul-Lez-Durance (France); Martin, E.; Herb, V.; Camus, G. [Universite Bordeaux 1, UMR 5801 (CNRS-SAFRAN-CEA-UB1), Laboratoire des Composites Thermostructuraux, F-33600 Pessac (France)

    2009-03-31

    Plasma facing components (PFCs) of magnetic fusion machines have high manufactured residual stresses and have to withstand important stress ranges during operation. These actively cooled PFCs have a carbon fibre composite (CFC) armour and a copper alloy heat sink. Cracks mainly appear in the CFC near the composite/copper interface. In order to analyse damage mechanisms, it is important to well simulate the damage mechanisms both of the CFC and the CFC/Cu interface. This study focuses on the mechanical behaviour of the N11 material for which the scalar ONERA damage model was used. The damage parameters of this model were identified by similarity to a neighbour material, which was extensively analysed, according to the few characterization test results available for the N11. The finite elements calculations predict a high level of damage of the CFC at the interface zone explaining the encountered difficulties in the PFCs fabrication. These results suggest that the damage state of the CFC cells is correlated with a conductivity decrease to explain the temperature increase of the armour surface under fatigue heat load.

  11. Damage of actively cooled plasma facing components of magnetic confinement controlled fusion machines

    Science.gov (United States)

    Chevet, G.; Schlosser, J.; Martin, E.; Herb, V.; Camus, G.

    2009-03-01

    Plasma facing components (PFCs) of magnetic fusion machines have high manufactured residual stresses and have to withstand important stress ranges during operation. These actively cooled PFCs have a carbon fibre composite (CFC) armour and a copper alloy heat sink. Cracks mainly appear in the CFC near the composite/copper interface. In order to analyse damage mechanisms, it is important to well simulate the damage mechanisms both of the CFC and the CFC/Cu interface. This study focuses on the mechanical behaviour of the N11 material for which the scalar ONERA damage model was used. The damage parameters of this model were identified by similarity to a neighbour material, which was extensively analysed, according to the few characterization test results available for the N11. The finite elements calculations predict a high level of damage of the CFC at the interface zone explaining the encountered difficulties in the PFCs fabrication. These results suggest that the damage state of the CFC cells is correlated with a conductivity decrease to explain the temperature increase of the armour surface under fatigue heat load.

  12. Damage of actively cooled plasma facing components of magnetic confinement controlled fusion machines

    International Nuclear Information System (INIS)

    Chevet, G.; Schlosser, J.; Martin, E.; Herb, V.; Camus, G.

    2009-01-01

    Plasma facing components (PFCs) of magnetic fusion machines have high manufactured residual stresses and have to withstand important stress ranges during operation. These actively cooled PFCs have a carbon fibre composite (CFC) armour and a copper alloy heat sink. Cracks mainly appear in the CFC near the composite/copper interface. In order to analyse damage mechanisms, it is important to well simulate the damage mechanisms both of the CFC and the CFC/Cu interface. This study focuses on the mechanical behaviour of the N11 material for which the scalar ONERA damage model was used. The damage parameters of this model were identified by similarity to a neighbour material, which was extensively analysed, according to the few characterization test results available for the N11. The finite elements calculations predict a high level of damage of the CFC at the interface zone explaining the encountered difficulties in the PFCs fabrication. These results suggest that the damage state of the CFC cells is correlated with a conductivity decrease to explain the temperature increase of the armour surface under fatigue heat load

  13. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    Science.gov (United States)

    Missirlian, M.; Richou, M.; Riccardi, B.; Gavila, P.; Loarer, T.; Constans, S.

    2011-12-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m-2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m-2 for the CFC-armoured tiles and 15 MW m-2 for the W-armoured tiles, respectively.

  14. The heat removal capability of actively cooled plasma-facing components for the ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M; Richou, M; Loarer, T; Riccardi, B; Gavila, P; Constans, S

    2011-01-01

    Non-destructive examination followed by high-heat-flux testing was performed for different small- and medium-scale mock-ups; this included the most recent developments related to actively cooled tungsten (W) or carbon fibre composite (CFC) armoured plasma-facing components. In particular, the heat-removal capability of these mock-ups manufactured by European companies with all the main features of the ITER divertor design was investigated both after manufacturing and after thermal cycling up to 20 MW m - 2. Compliance with ITER requirements was explored in terms of bonding quality, heat flux performances and operational compatibility. The main results show an overall good heat-removal capability after the manufacturing process independent of the armour-to-heat sink bonding technology and promising behaviour with respect to thermal fatigue lifetime under heat flux up to 20 MW m - 2 for the CFC-armoured tiles and 15 MW m - 2 for the W-armoured tiles, respectively.

  15. Characterization and damaging law of CFC for high heat flux actively cooled plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Chevet, G., E-mail: gaelle.chevet@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Martin, E., E-mail: martin@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Boscary, J., E-mail: jean.boscary@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Camus, G., E-mail: camus@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Herb, V., E-mail: herb@lcts.u-bordeaux1.fr [LCTS, CNRS UMR 5801, Universite Bordeaux 1, Bordeaux (France); Schlosser, J., E-mail: jacques.schlosser@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France); Missirlian, M., E-mail: marc.missirlian@cea.fr [Association EURATOM-CEA, DSM/IRFM, CEA Cadarache, F-13108 Saint Paul lez Durance (France)

    2011-10-01

    The carbon fiber reinforced carbon composite (CFC) Sepcarb N11 has been used in the Tore Supra (TS) tokamak (Cadarache, France) as armour material for the plasma facing components. For the fabrication of the Wendelstein 7-X (W7-X) divertor (Greifswald, Germany), the NB31 material was chosen. For the fabrication of the ITER divertor, two potential CFC candidates are the NB31 and NB41 materials. In the case of Tore Supra, defects such as microcracks or debonding were found at the interface between CFC tile and copper heat sink. A mechanical characterization of the behaviour of N11 and NB31 was undertaken, allowing the identification of a damage model and finite element calculations both for flat tiles (TS and W7-X) and monoblock (ITER) armours. The mechanical responses of these CFC materials were found almost linear under on-axis tensile tests but highly nonlinear under shear tests or off-axis tensile tests. As a consequence, damage develops within the high shear-stress zones.

  16. Characterization and damaging law of CFC for high heat flux actively cooled plasma facing components

    International Nuclear Information System (INIS)

    Chevet, G.; Martin, E.; Boscary, J.; Camus, G.; Herb, V.; Schlosser, J.; Escourbiac, F.; Missirlian, M.

    2011-01-01

    The carbon fiber reinforced carbon composite (CFC) Sepcarb N11 has been used in the Tore Supra (TS) tokamak (Cadarache, France) as armour material for the plasma facing components. For the fabrication of the Wendelstein 7-X (W7-X) divertor (Greifswald, Germany), the NB31 material was chosen. For the fabrication of the ITER divertor, two potential CFC candidates are the NB31 and NB41 materials. In the case of Tore Supra, defects such as microcracks or debonding were found at the interface between CFC tile and copper heat sink. A mechanical characterization of the behaviour of N11 and NB31 was undertaken, allowing the identification of a damage model and finite element calculations both for flat tiles (TS and W7-X) and monoblock (ITER) armours. The mechanical responses of these CFC materials were found almost linear under on-axis tensile tests but highly nonlinear under shear tests or off-axis tensile tests. As a consequence, damage develops within the high shear-stress zones.

  17. Damage prediction of carbon fibre composite armoured actively cooled plasma-facing components under cycling heat loads

    International Nuclear Information System (INIS)

    Chevet, G; Schlosser, J; Courtois, X; Escourbiac, F; Missirlian, M; Herb, V; Martin, E; Camus, G; Braccini, M

    2009-01-01

    In order to predict the lifetime of carbon fibre composite (CFC) armoured plasma-facing components in magnetic fusion devices, it is necessary to analyse the damage mechanisms and to model the damage propagation under cycling heat loads. At Tore Supra studies have been launched to better understand the damage process of the armoured flat tile elements of the actively cooled toroidal pump limiter, leading to the characterization of the damageable mechanical behaviour of the used N11 CFC material and of the CFC/Cu bond. Up until now the calculations have shown damage developing in the CFC (within the zone submitted to high shear stress) and in the bond (from the free edge of the CFC/Cu interface). Damage is due to manufacturing shear stresses and does not evolve under heat due to stress relaxation. For the ITER divertor, NB31 material has been characterized and the characterization of NB41 is in progress. Finite element calculations show again the development of CFC damage in the high shear stress zones after manufacturing. Stresses also decrease under heat flux so the damage does not evolve. The characterization of the CFC/Cu bond is more complex due to the monoblock geometry, which leads to more scattered stresses. These calculations allow the fabrication difficulties to be better understood and will help to analyse future high heat flux tests on various mock-ups.

  18. Improvement of non destructive infrared test bed SATIR for examination of actively cooled tungsten armour Plasma Facing Components

    Energy Technology Data Exchange (ETDEWEB)

    Vignal, N., E-mail: nicolas.vignal@cea.fr; Desgranges, C.; Cantone, V.; Richou, M.; Courtois, X.; Missirlian, M.; Magaud, Ph.

    2013-10-15

    Highlights: • Non destructive infrared techniques for control ITER like PFCs. • Reflective surface such as W induce a measurement temperature error. • Numerical data processing by evaluation of the local emissivity. • SATIR test bed can control metallic surface with low and variable emissivity. -- Abstract: For steady state (magnetic) thermonuclear fusion devices which need large power exhaust capability and have to withstand heat fluxes in the range 10–20 MW m{sup −2}, advanced Plasma Facing Components (PFCs) have been developed. The importance of PFCs for operating tokamaks requests to verify their manufacturing quality before mounting. SATIR is an IR test bed validated and recognized as a reliable and suitable tool to detect cooling defaults on PFCs with CFC armour material. Current tokamak developments implement metallic armour materials for first wall and divertor; their low emissivity causes several difficulties for infrared thermography control. We present SATIR infrared thermography test bed improvements for W monoblocks components without defect and with calibrated defects. These results are compared to ultrasonic inspection. This study demonstrates that SATIR method is fully usable for PFCs with low emissivity armour material.

  19. Improvement of non destructive infrared test bed SATIR for examination of actively cooled tungsten armour Plasma Facing Components

    International Nuclear Information System (INIS)

    Vignal, N.; Desgranges, C.; Cantone, V.; Richou, M.; Courtois, X.; Missirlian, M.; Magaud, Ph.

    2013-01-01

    Highlights: • Non destructive infrared techniques for control ITER like PFCs. • Reflective surface such as W induce a measurement temperature error. • Numerical data processing by evaluation of the local emissivity. • SATIR test bed can control metallic surface with low and variable emissivity. -- Abstract: For steady state (magnetic) thermonuclear fusion devices which need large power exhaust capability and have to withstand heat fluxes in the range 10–20 MW m −2 , advanced Plasma Facing Components (PFCs) have been developed. The importance of PFCs for operating tokamaks requests to verify their manufacturing quality before mounting. SATIR is an IR test bed validated and recognized as a reliable and suitable tool to detect cooling defaults on PFCs with CFC armour material. Current tokamak developments implement metallic armour materials for first wall and divertor; their low emissivity causes several difficulties for infrared thermography control. We present SATIR infrared thermography test bed improvements for W monoblocks components without defect and with calibrated defects. These results are compared to ultrasonic inspection. This study demonstrates that SATIR method is fully usable for PFCs with low emissivity armour material

  20. Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

    International Nuclear Information System (INIS)

    Ritz, G; Pintsuk, G; Linke, J; Hirai, T; Norajitra, P; Reiser, J; Giniyatulin, R; Makhankov, A; Mazul, I

    2009-01-01

    Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ∼14 MW m -2 , the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

  1. Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

    Science.gov (United States)

    Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

    2009-12-01

    Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

  2. ITER plasma facing components

    International Nuclear Information System (INIS)

    Kuroda, T.; Vieider, G.; Akiba, M.

    1991-01-01

    This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

  3. Leak tightness tests on actively cooled plasma facing components: Lessons learned from Tore Supra experience and perspectives for the new fusion machines

    Energy Technology Data Exchange (ETDEWEB)

    Chantant, M., E-mail: michel.chantant@cea.fr; Lambert, R.; Gargiulo, L.; Hatchressian, J.-C.; Guilhem, D.; Samaille, F.; Soler, B.

    2015-10-15

    Highlights: • Test procedures for the qualification of the tightness of actively cooled plasma facing components were defined. • The test is performed after the component manufacturing and before its set-up in the vacuum vessel. • It allows improving the fusion machine availability. • The lessons of tests over 20 years at Tore Supra are presented. - Abstract: The fusion machines under development or construction (ITER, W7X) use several hundreds of actively cooled plasma facing components (ACPFC). They are submitted to leak tightness requirements in order to get an appropriate vacuum level in the vessel to create the plasma. During the ACPFC manufacturing and before their installation in the machine, their leak tightness performance must be measured to check that they fulfill the vacuum requirements. A relevant procedure is needed which allows to segregate potential defects. It must also be optimized in terms of test duration and costs. Tore Supra, as an actively cooled Tokamak, experienced several leaks on ACPFCs during the commissioning and during the operation of the machine. A test procedure was then defined and several test facilities were set-up. Since 1990 the tightness of all the new ACPFCs is systematically tested before their installation in Tore Supra. During the qualification test, the component is set up in a vacuum test tank, and its cooling circuits are pressurized with helium. It is submitted to 3 temperature cycles from room temperature up to the baking temperature level in Tore Supra (200 °C) and two pressurization tests are performed (6 MPa at room temperature and 4 MPa at 200 °C) at each stage. At the end of the last cycle when the ACPFC is at room temperature and pressurized with helium at 6 MPa, the measured leak rate must be lower than 5 × 10{sup −11} Pa m{sup 3} s{sup −1}, the pressure in the test tank being <5 × 10{sup −5} Pa. A large experience has been gained on ACPFCs with carbon parts on stainless steel and Cu

  4. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F.; Missirlian, M.; Schlosser, J. [Association EURATOM-CEA Cadarache, Departement de Recherches sur la Fusion Controlee, 13 - Saint Paul lez Durance (France); Bobin-Vastra, I. [AREVA Centre Technique de Framatome, 71 - Le Creusot (France); Kuznetsov, V. [Efremov Institute, Doroga na Metallostroy, St. Petersburg (Russian Federation); Schedler, B. [Plansee AG, Reutte (Austria)

    2004-07-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m{sup 2} with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m{sup 2}. These results highlight the high potential of this technology for ITER divertor application.

  5. A mature industrial solution for ITER divertor plasma facing components: hypervapotron cooling concept adapted to Tore Supra flat tile technology

    International Nuclear Information System (INIS)

    Escourbiac, F.; Missirlian, M.; Schlosser, J.; Bobin-Vastra, I.; Kuznetsov, V.; Schedler, B.

    2004-01-01

    The use of flat tile technology to handle heat fluxes in the range of 20 MW/m 2 with components relevant for fusion experiment applications is technically possible with the hypervapotron cooling concept. This paper deals with recent high heat flux performances operated with success on 2 identical mock-ups, based on this concept, that were tested in 2 different electron gun facilities. Each mock-up consisted of a CuCrZr heat sink armored with 25 flat tiles of the 3D carbon fibre composite material SEPcarb NS31 assembled with pure copper by active metal casting (AMC). The AMC tiles were electron beam welded on the CuCrZr bar, fins and slots on the neutral beam JET design were machined into the bar, then the bar was closed with a thick CuCrZr rear plug including hydraulic connections then the bar was electron beam welded onto the sidewalls. The testing results show that full ITER design specifications were achieved with margins, the critical heat flux limit was even higher than 30 MW/m 2 . These results highlight the high potential of this technology for ITER divertor application

  6. Tungsten covered graphite and copper elements and ITER-like actively cooled tungsten divertor plasma facing units for the WEST project

    International Nuclear Information System (INIS)

    Guilhem, D; Bucalossi, J; Burles, S; Corre, Y; Ferlay, F; Firdaouss, M; Languille, P; Lipa, M; Martinez, A; Missirlian, M; Proust, M; Richou, M; Samaille, F; Tsitrone, E

    2016-01-01

    After a brief introduction giving some insight of the WEST project, we present the three types of plasma facing units (PFUs) developed for the WEST project taking into account the envisaged main scenarios: (1) high power short pulse scenario (a few seconds) where the objective is to maximize the power handling of the PFUs, up to 20 MW m −2 , (2) high fluence scenario (a few 100 s) on actively cooled ITER-like tungsten (W) PFUs, up to 10 MW m −2 during 1000 s. For the graphite PFUs, the high heat flux tests have been done at GLADIS (ion beam test facility), and for the CuCrZr PFUs on the JUDITH (electron beam test facility). The tests were successful, as no damage occurred for the different load cases. This confirms that the modelling done during the design phase is appropriate to describe these PFUs. Series productions are expected to be achieved by the end of 2015 for the graphite and CuCrZr PFUs, and few ITER-like W PFUs are expected at the beginning of 2016. The lower divertor will be complemented with ITER-like W PFUs as soon as available from our partners so that different fabrication procedures could be evaluated in a real industrial process and a real tokamak environment. (paper)

  7. A new vision of plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Nygren, Richard E., E-mail: renygre@sandia.gov [Sandia National Laboratories, Albuquerque, NM (United States); Youchison, Dennis L. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Wirth, Brian D. [University of Tennessee, Knoxville, TN (United States); Snead, Lance L.

    2016-11-01

    Highlights: • New approach recommended to develop refractory fusion plasma facing components. • Need to develop engineered materials architecture with nano-features. • Need to develop PFCs with gas jet cooling with very fine scale for jet arrays. • Emphasis on role of additive manufacturing as needed method for fabrication. - Abstract: This paper advances a vision for plasma facing components (PFCs) that includes the following points. The solution for plasma facing materials likely consists of engineered structures in which the layer of plasma facing material (PFM) is integrated with an engineered structure that cools the PFM and may also transition with graded composition. The key to achieving this PFC architecture will likely lie in advanced manufacturing methods, e.g., additive manufacturing, that can produce layers with controlled porosity and features such as micro-fibers and/or nano-particles that can collect He and transmutation products, limit tritium retention, and do all this in a way that maintains adequate robustness for a satisfactory lifetime. This vision has significant implications for how we structure a development program.

  8. Plasma facing device of thermonuclear device

    International Nuclear Information System (INIS)

    Sumita, Hideo; Ioki, Kimihiro.

    1993-01-01

    The present invention improves integrity of thermal structures of a plasma facing device. That is, in the plasma facing device, an armour block portion from a metal cooling pipe to a carbon material comprises a mixed material of the metal as the constituent material of the cooling pipe and ceramics. Then, the mixing ratio of the composition is changed continuously or stepwise to suppress peakings of remaining stresses upon production and thermal stresses upon exertion of thermal loads. Accordingly, thermal integrity of the structural materials can further be improved. In this case, a satisfactory characteristic can be obtained also by using ceramics instead of carbon for the mixed material, and the characteristic such as heat expansion coefficient is similar to that of the armour tile. (I.S.)

  9. Plasma facing parts and repairing method

    International Nuclear Information System (INIS)

    Fuse, Toshiaki; Tachikawa, Nobuo.

    1994-01-01

    Plasma facing parts of the present invention are constituted by joining an armour comprising a material having a high melting point and a cooling member comprising copper or the like. A metal member having good solderability with the cooling member is disposed on the joined surface of the armor member. In addition, the joined surface of the cooling member is provided with a barrier layer for preventing invasion of a solder. A solder having a low melting point is interposed between the armour and the cooling member. If they are heated entirely, the solder having low melting point is melted, so that the metal member having good solderability disposed on the armor member is soldered with the barrier layer for the cooling member. Upon exchange of the armour, the joint is heated again. Then, the solder having a low melting point is melted and the armour member and the cooling member are separated. If a solder is put on the cooling member and a new armour is placed and then heated, repairing is completed. (I.S.)

  10. Plasma-wall interaction and plasma facing materials

    International Nuclear Information System (INIS)

    Tanabe, Tetsuo; Miyahara, Akira.

    1990-01-01

    The recognition that plasma-wall interaction plays the essential role from both standpoints of energy balance and particle balance for realizing nuclear fusion reactors has become to prevail. However, on how each elementary process acts and what competitive effect the synthetic action brings about, the stage of doing the qualitative discussion has just come, and the quantitative investigation is the problem for the future. In this paper, the plasma-wall interaction as seen from the research field of plasma-facing materials is discussed centering around graphite materials which have been mostly used at present, and the present status of the research and development on the problems of impurities, hydrogen recycling and heat resistance and radiation resistance is mentioned. Moreover, the problems are pointed out, and the course for the future is looked for. The recent experiment with large tokamaks adopted graphite or carbon as the plasma-facing materials, and the reduction of metallic impurities in plasma showed the clear improvement of plasma confinement characteristics. However, for the next device which requires forced cooling, the usability of graphite is doubtful. (K.I.) 51 refs

  11. Tungsten fibre-reinforced composites for advanced plasma facing components

    OpenAIRE

    Neu, R.; Riesch, J.; Müller, A.v.; Balden, M.; Coenen, J.W.; Gietl, H.; Höschen, T.; Li, M.; Wurster, S.; You, J.-H.

    2016-01-01

    The European Fusion Roadmap foresees water cooled plasma facing components in a first DEMO design in order to provide enough margin for the cooling capacity and to only moderately extrapolate the technology which was developed and tested for ITER. In order to make best use of the water cooling concept copper (Cu) and copper-chromium-zirconium alloy (CuCrZr) are envisaged as heat sink whereas as armour tungsten (W) based materials will be used. Combining both materials in a high heat flux comp...

  12. Plasma facing components design of KT-2 tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

  13. Integrated modelling of the edge plasma and plasma facing components

    International Nuclear Information System (INIS)

    Coster, D.P.; Bonnin, X.; Mutzke, A.; Schneider, R.; Warrier, M.

    2007-01-01

    Modelling of the interaction between the edge plasma and plasma facing components (PFCs) has tended to place more emphasis on either the plasma or the PFCs. Either the PFCs do not change with time and the plasma evolution is studied, or the plasma is assumed to remain static and the detailed interaction of the plasma and the PFCs are examined, with no back-reaction on the plasma taken into consideration. Recent changes to the edge simulation code, SOLPS, now allow for changes in both the plasma and the PFCs to be considered. This has been done by augmenting the code to track the time-development of the properties of plasma facing components (PFCs). Results of standard mixed-materials scenarios (base and redeposited C; Be) are presented

  14. Design of plasma facing components for the SST-1 tokamak

    International Nuclear Information System (INIS)

    Jacob, S.; Chenna Reddy, D.; Choudhury, P.; Khirwadkar, S.; Pragash, R.; Santra, P.; Saxena, Y.C.; Sinha, P.

    2000-01-01

    Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m 2 . (author)

  15. Manufacturing technology development for vacuum vessel and plasma facing components

    International Nuclear Information System (INIS)

    Laitinen, Arttu; Liimatainen, Jari; Hallila, Pentti

    2005-01-01

    Vacuum vessel and plasma facing components of the ITER construction including shield modules and primary first wall panels have great impact on the production costs and reliability of the installation. From the manufacturing technology point of view, accuracy of shape, properties of the various austenitic stainless steel/austenitic stainless steel interfaces or CuCrZr/austenitic stainless steel interfaces as well as those of the base materials are crucial for technical reliability of the construction. The current approach in plasma facing components has been utilisation of solid-HIP technology and solid-powder-HIP technology. Due to the large size of especially shield modules shape, control of the internal cavities and cooling channels is extremely demanding. This requires strict control of the raw materials and manufacturing parameters

  16. Net erosion measurements on plasma facing components of Tore Supra

    International Nuclear Information System (INIS)

    Tsitrone, E.; Chappuis, P.; Corre, Y.; Gauthier, E.; Grosman, A.; Pascal, J.Y.

    2001-01-01

    Erosion of the plasma facing components is a crucial point of investigation in long pulse operation of future fusion devices. Therefore erosion measurements have been undertaken in the Tore Supra tokamak. After each experimental campaign, different plasma facing components have been monitored in situ by non-destructive means, in order to evaluate their net erosion following a long plasma exposure. This paper presents the results obtained over three experimental campaigns on the Tore Supra ergodic divertor B 4 C-coated neutralisers and CFC Langmuir probes. The erosion on the Langmuir probes after one year of plasma exposure can reach 100 μm, leading to an effective erosion coefficient of around 5x10 -3 to 10 -2 , in reasonable agreement with values found on other tokamaks. The erosion of the ergodic divertor neutraliser plates is lower (10 μm). This is coherent with the attenuated particle flux due to a lower incidence angle, and might also be due to some surface temperature effect, since the neutralisers are actively cooled while the Langmuir probes are not. Moreover, the profile along the neutraliser shows net erosion in zones wetted by the plasma and net redeposition in shadowed zones

  17. Selection of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.; Barabash, V.; Chiocchio, S.

    1996-01-01

    ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA

  18. Neutron irradiation effects on plasma facing materials

    Science.gov (United States)

    Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

    2000-12-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

  19. ITER plasma facing materials. Some critical considerations

    International Nuclear Information System (INIS)

    Barabash, V.; Dietz, K.J.; Federici, G.; Janeschitz, G.; Matera, R.; Tanaka, S.

    1995-01-01

    The description of current status with the choice of materials for ITER plasma facing components is presented. The main problem with lifetime of divertor elements is the particle and energy-induced erosion of armour materials. A solution for the first operation phase consists in using Be as an armour for the first wall and the divertor, however other possible materials (e.g. W) could be considered. (orig.)

  20. Neutron irradiation effects on plasma facing materials

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

    2000-01-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

  1. Design of the ITER Plasma-Facing Components

    Energy Technology Data Exchange (ETDEWEB)

    Merola, M.

    2009-07-01

    The ITER plasma-facing components cover an area of about 850 m{sup 2} and consist of the Divertor, the Blanket and the Test Blanket Modules (TBMs) with their corresponding frames. The Divertor is located at the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimizing the helium and impurity content in the plasma. It consists of 54 cassette assemblies. Each assembly has 3 plasma-facing components (PFCs), namely the inner and outer target and the dome, which are mounted onto a steel support structure, the cassette body. The targets directly intercept the magnetic field lines and are designed to withstand heat fluxes as high as 20 MW/m{sup 2}. CFC is the reference design solution for the armour of the lower part of the targets. However, the resultant high erosion rate could potentially limit machine operation in the DT phase (due to co-deposition with T). Therefore, prior to the DT phase, the divertor PFCs will be replaced with a new set entirely covered with W armour. The Divertor is a RH Class 1 component, which is planned to be replaced 3 times during the 20 years of the ITER operation. The construction phase of the ITER Divertor is being launched. The Blanket covers the largest fraction of the plasma-facing surface. Each of the 440 Blanket modules consists of a first wall (FW) panel, which is mechanically attached onto a Shield Module (SM). The design heat flux is set up to 1 or 5 MW/m{sup 2}. The FW panels are covered by Be tiles, which are joined onto a copper alloy (CuCrZr) heat sink, which is in turn intimately joined onto a 316L(N) stainless steel part. The SM is a block of 316L(N)-IG steel, where an array of cooling channels are obtained by machining and welding. The TBMs are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to TBM testing, each of them allocating two TBMs, inserted in a thick steel frame. The frame is a water-cooled 316L

  2. Heat transfer for plasma facing components

    International Nuclear Information System (INIS)

    Boyd, R.D.; Meng, X.; Maughan, H.

    1995-01-01

    Although the high heat flux requirements for plasma-facing components have been reduced drastically from 40.0 MW/m 2 to near 10.0 MW/m 2 , there are still some refinements needed. This paper highlights: (1) recent accomplishments and pinpoints new thermal solutions and problem areas of immediate concern to the development of plasma-facing components, and (2) next generation thermal hydraulic problems which must be addressed to insure safety and reliability in component operation. More specifically the near-term thermal hydraulic problems entail: (1) generating an appropriate data base to insure the development of single-side heat flux correlations; and (2) adapting the existing vast uniform heat flux literature to the case of non-uniform heat flux distributions found in plasma facing components in fusion reactors. Results are presented for the latter task which includes: (a) an accurate subcooled flow boiling curve correlation for the partial nucleate boiling regime which can be adapted using previously proposed correlations relating single-side boundary heat flux to heat transfer, in uniformly heated channels, (b) the evaluation of the possibility of using the existing literature directly with redefined parameters, and (c) an estimation of circumferential variations in the heat transfer coefficient

  3. ITER plasma facing components, design and development

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Akiba, M.; Matera, R.; Watson, R.

    1991-01-01

    The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) The definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R and D work giving already first results, and the definition of the required further R and D program to support the contemplated ITER Engineering Design Activity (EDA). From the ITER CDA effort on PFC it is mainly concluded that: (a) The expected PFC operating conditions lead to design solutions at the limit of present technology in particular for the divertor, which may constrain the overall machine performance, (b) the development of convincing PFC designs requires an intensified R and D effort both on PFC technology and plasma physics. (orig.)

  4. Material Challenges For Plasma Facing Components in Future Fusion Reactors

    International Nuclear Information System (INIS)

    Linke, J; Pintsuk, G.; Rödig, M.

    2013-01-01

    Increasing attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO2-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm-2; the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm-2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm-2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions

  5. Plasma edge physics in an actively cooled tokamak

    International Nuclear Information System (INIS)

    Gunn, J.P.; Adamek, A.; Boucher, C.

    2005-01-01

    Tore Supra is a large tokamak with a plasma of circular cross section (major radius 2.4 m and minor radius 0.72 m) lying on a toroidal limiter. Tore Supra's main mission is the development of technology to inject up to 25 MW of microwave heating power and extract it continuously for up to 1000 s in steady state without uncontrolled overheating of, or outgassing from, plasma-facing components. The entire first wall of the tokamak is actively cooled by a high pressure water loop and special carbon fiber composite materials have been designed to handle power fluxes up to 10 MW/m 2 . The edge plasma on open magnetic flux surfaces that intersect solid objects plays an important role in the overall behaviour of the plasma. The transport of sputtered impurity ions and the fueling of the core plasma are largely governed by edge plasma density, temperature, and flow profiles. Measurements of these quantities are becoming more reliable and frequent in many tokamaks, and it has become clear that we do not understand them very well. Classical two-dimensional fluid modelling fails to reproduce many aspects of the experimental observations such as the significant thickness of the edge plasma, and the near-sonic flows that occur where none should be expected. It is suspected that plasma turbulence is responsible for these anomalies. In the Tore Supra tokamak, various kinds of Langmuir probes are used to characterize the edge plasma. We will present original measurements that demonstrate the universality of many phenomena that have been observed in X-point divertor tokamaks, especially concerning the ion flows. As in the JET tokamak, surprisingly large values of parallel Mach number are measured midway between the two strike zones, where one would expect to find nearly stagnant plasma if the particle source were poloidally uniform. We will present results of a novel experiment that provides evidence for a poloidally localized particle and energy source on the outboard midplane of

  6. Charge exchange cooling in the tandem mirror plasma confinement apparatus

    Science.gov (United States)

    Logan, B. Grant

    1978-01-01

    Method and apparatus for cooling a plasma of warm charged species confined in the center mirror cell of the tandem mirror apparatus by injecting cold neutral species of the plasma into at least one mirroring region of the center mirror cell, the cooling due to the loss of warm charged species through charge exchange with the cold neutral species with resulting diffusion of the warm neutral species out of the plasma.

  7. Charge exchange cooling in the tandem mirror plasma confinement apparatus

    International Nuclear Information System (INIS)

    Logan, B.G.

    1978-01-01

    A method and apparatus are described for cooling a plasma confined in the center mirror cell of the tandem mirror apparatus by injecting cold neutral species of the plasma into at least one mirroring region of the center mirror cell. The cooling is due to the loss of warm charged species through charge exchange with the cold neutral species with resulting diffusion of the warm neutral species out of the plasma

  8. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

    Skinner, C.H.; Coad, J.P.; Federici, G.

    2003-01-01

    Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating

  9. Tungsten thick coatings for plasma facing components

    International Nuclear Information System (INIS)

    Riccardi, B.; Pizzuto, A.; Orsini, A.; Libera, S.; Visca, E.; Bertamini, L.; Casadei, F.; Severini, E.; Montanari, R.; Litunovsky, N.

    1998-01-01

    The aim of the R and D activity was to realize thick W coatings on CuCrZr hollow bars and to test the mock ups with respect to thermal fatigue. Eight mock ups provided of 4 mm thick W coating were finally manufactured. The bonding integrity between coating and substrate was checked by means of an Ultrasonic apparatus. Characterisation of coatings was performed in order to assess microstructure, impurity content, density, tensile strength, adhesion strength, thermal conductivity and thermal expansion coefficient. Macroscopic residual strain measurements were performed by means of 'hole drilling' technique. The activities performed demonstrated the feasibility of thick Tungsten coatings on geometries with more complex residual strain distribution. These coatings are reliable armour of medium heat flux plasma facing component. (author)

  10. Time-dependent Cooling in Photoionized Plasma

    Energy Technology Data Exchange (ETDEWEB)

    Gnat, Orly, E-mail: orlyg@phys.huji.ac.il [Racah Institute of Physics, The Hebrew University, Jerusalem 91904 (Israel)

    2017-02-01

    I explore the thermal evolution and ionization states in gas cooling from an initially hot state in the presence of external photoionizing radiation. I compute the equilibrium and nonequilibrium cooling efficiencies, heating rates, and ion fractions for low-density gas cooling while exposed to the ionizing metagalactic background radiation at various redshifts ( z = 0 − 3), for a range of temperatures (10{sup 8}–10{sup 4} K), densities (10{sup −7}–10{sup 3} cm{sup −3}), and metallicities (10{sup −3}–2 times solar). The results indicate the existence of a threshold ionization parameter, above which the cooling efficiencies are very close to those in photoionization equilibrium (so that departures from equilibrium may be neglected), and below which the cooling efficiencies resemble those in collisional time-dependent gas cooling with no external radiation (and are thus independent of density).

  11. Fatigue life of the plasma-facing components in PULSAR

    International Nuclear Information System (INIS)

    Crowell, J.A.; Blanchard, J.P.

    1994-01-01

    The PULSAR project is a multi-institutional effort to determine the advantages that can be gained by building a tokamak without current drive. This machine would reduce the capital and operating costs of the machine by avoiding the need for complex current drive hardware but it must compensate for this with an energy storage scheme and with increased structural requirements due to cyclic fatigue. This paper presents the results of the fatigue analysis for the plasma-facing components of PULSAR. The structural analysis is carried out using two-dimensional finite element models and a variety of boundary conditions to account for the third dimension. In some cases the temperature distribution is modified to simulate behaviors which cannot normally be modeled with two-dimensional finite element models. PULSAR features two major engineering designs: a liquid metal-cooled vanadium design and a helium-cooled SiC/SiC design. Results are given for each. It is shown that the superior thermal and strength properties of the vanadium alloy simplify the component design process significantly. The SiC composite properties cause significantly more difficulty for the designer and, in particular, no credible design is found for a divertor fabricated solely from the SiC composite. This conclusion is based on current data for the thermophysical properties and fatigue strength of SiC fiber composites, so developments in these areas could allow the fabrication of a SiC/SiC divertor for a pulsed tokamak

  12. Beryllium application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Barabash, V.; Cardella, A.; Jakeman, R.; Ioki, K.; Janeschitz, G.; Parker, R.; Tivey, R.; Pacher, H.D.; Wu, C.H.; Bartels, H.W.

    1997-01-01

    Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R and D program are highlighted. (orig.)

  13. Qualification, commissioning and in situ monitoring of high heat flux plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)], E-mail: frederic.escourbiac@cea.fr; Durocher, A.; Grosman, A.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France); Merola, M.; Tivey, R. [ITER Team, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)

    2007-10-15

    Up-to-date development of actively cooled high heat flux (HHF) plasma facing components (PFC) prototypes only allows reduced margins with regards to the ITER thermal requirements. Additionally, perfect quality cannot be ensured along series manufacturing: the presence of flaws which impair the heat transfer capability of the component, in particular at the interface armour/heat sink appears to be statistically unavoidable. In order to ensure a successful series production, a qualification methodology of actively cooled high heat flux plasma facing components is proposed. Secondly, advanced non-destructive techniques developed for HHF PFC commissioning are detailed with definition of acceptance criteria. Finally, innovative diagnostics for in situ monitoring during plasma operations or tokamak shutdowns are investigated in order to prevent immediate damage (safety monitoring); or evaluate component degradation (health monitoring). This work takes into account the relevance to Tore Supra, and is applied to W7X and ITER Divertor HHF PFC.

  14. Qualification, commissioning and in situ monitoring of high heat flux plasma facing components

    International Nuclear Information System (INIS)

    Escourbiac, F.; Durocher, A.; Grosman, A.; Cismondi, F.; Courtois, X.; Farjon, J.L.; Schlosser, J.; Merola, M.; Tivey, R.

    2007-01-01

    Up-to-date development of actively cooled high heat flux (HHF) plasma facing components (PFC) prototypes only allows reduced margins with regards to the ITER thermal requirements. Additionally, perfect quality cannot be ensured along series manufacturing: the presence of flaws which impair the heat transfer capability of the component, in particular at the interface armour/heat sink appears to be statistically unavoidable. In order to ensure a successful series production, a qualification methodology of actively cooled high heat flux plasma facing components is proposed. Secondly, advanced non-destructive techniques developed for HHF PFC commissioning are detailed with definition of acceptance criteria. Finally, innovative diagnostics for in situ monitoring during plasma operations or tokamak shutdowns are investigated in order to prevent immediate damage (safety monitoring); or evaluate component degradation (health monitoring). This work takes into account the relevance to Tore Supra, and is applied to W7X and ITER Divertor HHF PFC

  15. Tungsten fibre-reinforced composites for advanced plasma facing components

    Directory of Open Access Journals (Sweden)

    R. Neu

    2017-08-01

    Full Text Available The European Fusion Roadmap foresees water cooled plasma facing components in a first DEMO design in order to provide enough margin for the cooling capacity and to only moderately extrapolate the technology which was developed and tested for ITER. In order to make best use of the water cooling concept copper (Cu and copper-chromium-zirconium alloy (CuCrZr are envisaged as heat sink whereas as armour tungsten (W based materials will be used. Combining both materials in a high heat flux component asks for an increase of their operational range towards higher temperature in case of Cu/CuCrZr and lower temperatures for W. A remedy for both issues- brittleness of W and degrading strength of CuCrZr- could be the use of W fibres (Wf in W and Cu based composites. Fibre preforms could be manufactured with industrially viable textile techniques. Flat textiles with a combination of 150/70 µm W wires have been chosen for layered deposition of tungsten-fibre reinforced tungsten (Wf/W samples and tubular multi-layered braidings with W wire thickness of 50 µm were produced as a preform for tungsten-fibre reinforced copper (Wf /Cu tubes. Cu melt infiltration was performed together with an industrial partner resulting in sample tubes without any blowholes. Property estimation by mean field homogenisation predicts strongly enhanced strength of the Wf/CuCrZr composite compared to its pure CuCrZr counterpart. Wf /W composites show very high toughness and damage tolerance even at room temperature. Cyclic load tests reveal that the extrinsic toughening mechanisms counteracting the crack growth are active and stable. FEM simulations of the Wf/W composite suggest that the influence of fibre debonding, which is an integral part of the toughening mechanisms, and reduced thermal conductivity of the fibre due to the necessary interlayers do not strongly influence the thermal properties of future components.

  16. Plasma facing components integration studies for the WEST divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ferlay, Fabien, E-mail: fabien.ferlay@cea.fr; Missirlian, Marc; Guilhem, Dominique; Firdaouss, Mehdi; Richou, Marianne; Doceul, Louis; Faisse, Frédéric; Languille, Pascal; Larroque, Sébastien; Martinez, André; Proust, Maxime; Louison, Céphise; Jeanne, Florian; Saille, Alain; Samaille, Frank; Verger, Jean-Marc; Bucalossi, Jérôme

    2015-10-15

    Highlights: • The divertor PFU integration has been studied regarding existing environment. • Magnetic, electric, thermal, hydraulic, mechanical loads and assembly are considered. - Abstract: In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m{sup 2} heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

  17. Plasma spraying process of disperse carbides for spraying and facing

    International Nuclear Information System (INIS)

    Blinkov, I.V.; Vishnevetskaya, I.A.; Kostyukovich, T.G.; Ostapovich, A.O.

    1989-01-01

    A possibility to metallize carbides in plasma of impulsing capacitor discharge is considered. Powders granulation occurs during plasma spraying process, ceramic core being completely capped. X-ray phase and chemical analyses of coatings did not show considerable changes of carbon content in carbides before and after plasma processing. This distinguishes the process of carbides metallization in impulsing plasma from the similar processing in arc and high-frequency plasma generator. Use of powder composites produced in the impulsing capacitor discharge, for plasma spraying and laser facing permits 2-3 times increasing wear resistance of the surface layer as against the coatings produced from mechanical powders mixtures

  18. Plasma Wind Tunnel Testing of Electron Transpiration Cooling Concept

    Science.gov (United States)

    2017-02-28

    Colorado State University ETC Electron Transpiration Cooling LHTS Local Heat Transfer Simulation LTE Local Thermodynamic Equilibrium RCC Reinforced...ceramic electric material testing in plasma environment (not performed), 4. measurements and analysis of the Electron Transpiration Cooling (Sec. 4.2). 2...VKI 1D boundary layer code for computation of enthalpy and boundary layer parameters: a) iterate on ’virtually measured ’ heat flux, b) once enthalpy

  19. Silicon Carbide as a tritium permeation barrier in tungsten plasma-facing components

    Science.gov (United States)

    Wright, G. M.; Durrett, M. G.; Hoover, K. W.; Kesler, L. A.; Whyte, D. G.

    2015-03-01

    The control of tritium inventory is of great importance in future fusion reactors, not only from a safety standpoint but also to maximize a reactor's efficiency. Due to the high mobility of hydrogenic species in tungsten (W) one concern is the loss of tritium from the system via permeation through the tungsten plasma-facing components (PFC). This can lead to loss of tritium through the cooling channels of the wall thereby mandating tritium monitoring and recovery methods for the cooling system of the first wall. The permeated tritium is then out of the fuel cycle and cannot contribute to energy production until it is recovered and recycled into the system.

  20. Plasma edge cooling during RF heating

    International Nuclear Information System (INIS)

    Suckewer, S.; Hawryluk, R.J.

    1978-01-01

    A new approach to prevent the influx of high-Z impurities into the core of a tokamak discharge by using RF power to modify the edge plasma temperature profile is presented. This concept is based on spectroscopic measurements on PLT during ohmic heating and ATC during RF heating. A one dimensional impurity transport model is used to interpret the ATC results

  1. Materials for the plasma-facing components of steady state stellarators

    International Nuclear Information System (INIS)

    Bolt, H.; Boscary, J.; Greuner, H.; Grigull, P.; Maier, H.; Streibl, B.

    2005-01-01

    The specific advantage of current-free stellarators is their inherent capability for full steady-state operation. This will lead to long discharges and the corresponding stationary plasma exposure of the plasma-facing materials. Further to this, the absence of disruptions relaxes the requirements to the plasma-facing materials in terms of thermal shock stability, although ELM activity occurs also in stellarators and leads to fast transient surface loads on the ms-time scale. Another aspect regarding the plasma-material interactions in stellarators is the sensitivity to impurity accumulation in the core plasma. Thus, it is preferred to apply low-Z materials until operation scenarios are established which do not lead to this accumulation process. In the case of high-Z materials impurity accumulation will lead to a radiative plasma collapse. For the stellarator W7-X low-Z plasma-facing materials have been selected to protect the divertor and the wall surfaces. Due to the stationary operation, the plasma-facing materials have to be bonded or clamped to actively water-cooled substrates to remove the incident heat fluxes. The following materials have been selected to fulfil the operational requirements: 1. A three directionally carbon fibre reinforced carbon composite (CFC) with very high thermal conductivity bonded to a water cooled CuCrZr heat sink for the divertor which will be exposed to heat fluxes up to 10MW/m 2 . 2. Isotropic fine grain graphite tiles mechanically clamped to a CuCrZr heat sink which is brazed to a stainless steel cooling tube for the areas of moderate heat fluxes up to 0.5 MW/m 2 (baffles, inner wall). 3. Thick boron carbide coating on water cooled steel panels for the outer wall surfaces with low heat fluxes up to 0.2 MW/m 2 . This coating would be applied on most surfaces only after the initial operation. In the presentation the properties of these materials will be discussed with a view to the plasma-wall interaction in W7-X. In fusion reactors

  2. Thermographic analysis of plasma facing components covered by carbon surface layer in tokamaks

    International Nuclear Information System (INIS)

    Gardarein, Jean-Laurent

    2007-01-01

    Tokamaks are reactors based on the thermonuclear fusion energy with magnetic confinement of the plasma. In theses machines, several MW are coupled to the plasma for about 10 s. A large part of this power is directed towards plasma facing components (PFC). For better understanding and control the heat flux transfer from the plasma to the surrounding wall, it is very important to measure the surface temperature of the PFC and to estimate the imposed heat flux. In most of tokamaks using carbon PFC, the eroded carbon is circulating in the plasma and redeposited elsewhere. During the plasma operations, this leads at some locations to the formation of thin or thick carbon layers usually poorly attached to the PFC. These surface layers with unknown thermal properties complicate the calculation of the heat flux from IR surface temperature measurements. To solve this problem, we develop first, inverse method to estimate the heat flux using thermocouple (not sensitive to the carbon surface layers) temperature measurements. Then, we propose a front face pulsed photothermal method allowing an estimation of layers thermal diffusivity, conductivity, effusivity and the thermal contact resistance between the layer and the tile. The principle is to study with an infrared sensor, the cooling of the layer surface after heating by a short laser pulse, this cooling depending on the thermal properties of the successive layers. (author) [fr

  3. Analytical method for thermal stress analysis of plasma facing materials

    Science.gov (United States)

    You, J. H.; Bolt, H.

    2001-10-01

    The thermo-mechanical response of plasma facing materials (PFMs) to heat loads from the fusion plasma is one of the crucial issues in fusion technology. In this work, a fully analytical description of the thermal stress distribution in armour tiles of plasma facing components is presented which is expected to occur under typical high heat flux (HHF) loads. The method of stress superposition is applied considering the temperature gradient and thermal expansion mismatch. Several combinations of PFMs and heat sink metals are analysed and compared. In the framework of the present theoretical model, plastic flow and the effect of residual stress can be quantitatively assessed. Possible failure features are discussed.

  4. Analytical method for thermal stress analysis of plasma facing materials

    International Nuclear Information System (INIS)

    You, J.H.; Bolt, H.

    2001-01-01

    The thermo-mechanical response of plasma facing materials (PFMs) to heat loads from the fusion plasma is one of the crucial issues in fusion technology. In this work, a fully analytical description of the thermal stress distribution in armour tiles of plasma facing components is presented which is expected to occur under typical high heat flux (HHF) loads. The method of stress superposition is applied considering the temperature gradient and thermal expansion mismatch. Several combinations of PFMs and heat sink metals are analysed and compared. In the framework of the present theoretical model, plastic flow and the effect of residual stress can be quantitatively assessed. Possible failure features are discussed

  5. FEM-DBEM approach to analyse crack scenarios in a baffle cooling pipe undergoing heat flux from the plasma

    Directory of Open Access Journals (Sweden)

    R. Citarella

    2017-02-01

    Full Text Available Wendelstein 7-X is the world’s largest nuclear fusion experiment of stellarator type, in which a hydrogen plasma is confined by a magnet field generated with external superconducting coils, allowing the plasma to be heated up to the fusion temperature. The water-cooled Plasma Facing Components (PFC protect the Plasma Vessel (PV against radiative and convective heat from the plasma. After the assembly process of heat shields and baffles, several cracks were found in the braze and cooling pipes. Due to heat load cycles occurring during each Operational Phase (OP, thermal stresses are generated in the heat sinks, braze root and cooling pipes, capable to drive fatigue crack-growth and, possibly, a water leak through the pipe thickness. The aim of this study is to assess the most dangerous initial crack configurations in one of the most critical baffles by using numerical models based on a FEM-DBEM approach.

  6. RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Pt. II. Analysis of ITER plasma facing components

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.

    1997-01-01

    For pt.I see ibid., p.85-100, 1997. The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the various ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness. (orig.)

  7. RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part II: Analysis of ITER plasma facing components

    Science.gov (United States)

    Federici, Gianfranco; Raffray, A. René

    1997-04-01

    The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the variuos ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness.

  8. Self Passivating W-based Alloys as Plasma Facing Material

    International Nuclear Information System (INIS)

    Koch, F.; Koeppl, S.; Bolt, H.

    2007-01-01

    Full text of publication follows: Tungsten (W) is presently the main candidate material for the plasma-facing protection of future fusion power reactors due to the low sputter erosion under bombardment by energetic D, T and He ions. Thus a W-based protection material may provide a wall erosion lifetime of the order of five years which is a pre-requisite for economic fusion reactor operation. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO 3 compounds and their potential release under accidental conditions. A loss-of-coolant event in a He-cooled reactor would lead to a temperature rise to 1100 deg. C after approx. 10 to 30 days due to the nuclear decay heat of the in-vessel components. In such a situation additional accidental intense air ingress into the reactor vessel would lead to the formation of WO 3 and subsequent evaporation of radioactive (WO 3 ) x -clusters. The use of self passivating W alloys either as bulk material or as thick coating on the steel wall may be a passively safe alternative for the plasma-facing protection. The use of this material would eliminate the above mentioned concern related to pure W. To enable the formation of a protective film in oxidizing atmosphere which seals the tungsten surface from further oxidation, different elements have been investigated as corrosion protection additives. Therefore binary and ternary tungsten alloys were synthesised using magnetron sputtering. The oxidation behaviour of films deposited on inert substrates was measured with a thermo-balance set up under synthetic air at temperatures up to 1000 deg. C. Binary alloys of W-Si showed good self passivation properties by forming a SiO 2 film at the surface. The oxidation rate of a compound containing 11 wt.% Si was reduced by a factor of 10 2 compared to pure tungsten between 800 deg. C and 1000 deg. C. Using ternary alloys the oxidation behaviour could be further improved. A compound of W

  9. Tritium saturation in plasma-facing materials surfaces

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.; Causey, R.A.; Federici, G.; Haasz, A.A.

    1998-01-01

    Plasma-facing components in the international thermonuclear experimental reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10 20 -10 23 particles/m 2 s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments. (orig.)

  10. Investigation of Plasma Facing Components in Plasma Focus Operation

    Science.gov (United States)

    Roshan, M. V.; Babazadeh, A. R.; Kiai, S. M. Sadat; Habibi, H.; Mamarzadeh, M.

    2007-09-01

    Both aspects of the plasma-wall interactions, counter effect of plasma and materials, have been considered in our experiments. The AEOI plasma focus, Dena, has Filippov-type electrodes. The experimental results verify that neutron production increases using tungsten as an anode insert material, compared to the copper one. The experiments show decrement of the hardness of Aluminum targets outward the sides, from 135 to 78 in Vickers scale. The sputtering yield is about 0.0065 for deuteron energy of 50 keV.

  11. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  12. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

  13. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  14. Numerical simulation of runaway electron effect on Plasma Facing Components

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Suzuki, Satoshi; Akiba, Masato; Kunugi, Tomoaki

    1998-07-01

    The runaway electron effects on Plasma Facing Components (PFCs) are studied by the numerical analyses. The present study is the first investigation of time-dependent thermal response of PFCs caused by runaway electron impact. For this purpose, we developed a new integrated numerical code, which consists of the Monte Carlo code for the coupled electrons and photons transport analysis and the finite element code for the thermo-mechanical analysis. In this code, we apply the practical incident parameters and distribution of runaway electrons recently proposed by S. Putvinski, which can express the time-dependent behavior of runaway electrons impact. The incident parameters of electrons in this study are the energy density ranging from 10 to 75 MJ/m 2 , the average electrons' energy of 12.5 MeV, the incident angle of 0.01deg and the characteristic time constant for decay of runaway electrons event of 0.15sec. The numerical results showed that the divertor with CFC (Carbon-Fiber-Composite) armor did not suffer serious damage. On the other hand, maximum temperatures at the surface of the divertor with tungsten armor and the first wall with beryllium armor exceed the melting point in case of the incident energy density of 20 and 50 MJ/m 2 . Within the range of the incident condition of runaway electrons, the cooling pipe of each PFCs can be prevented from the melting or burn-out caused by runaway electrons impact, which is one of the possible consequences of runaway electrons event so far. (author)

  15. Platelet-cooled plasma arc torch. Final report

    International Nuclear Information System (INIS)

    1995-10-01

    In this 12-month program sponsored by the DOE Morgantown Energy Technology Center, Aerojet designed, fabricated, and tested six platelet cooled electrodes for a Retech 75T (90 MW) plasma arc torch capable of processing mixed radioactive waste. Two of the electrodes with gas injection through the electrode wall demonstrated between eight and forty times the life of conventional water cooled electrodes. If a similar life increase can be produced in a 1 Mw size electrode, then electrodes possessing thousands, rather than hundreds, of hours of life will be available to DOE for potential application to mixed radioactive waste processing

  16. Counter-facing plasma guns for efficient extreme ultra-violet plasma light source

    Science.gov (United States)

    Kuroda, Yusuke; Yamamoto, Akiko; Kuwabara, Hajime; Nakajima, Mitsuo; Kawamura, Tohru; Horioka, Kazuhiko

    2013-11-01

    A plasma focus system composed of a pair of counter-facing coaxial guns was proposed as a long-pulse and/or repetitive high energy density plasma source. We applied Li as the source of plasma for improvement of the conversion efficiency, the spectral purity, and the repetition capability. For operation of the system with ideal counter-facing plasma focus mode, we changed the system from simple coaxial geometry to a multi-channel configuration. We applied a laser trigger to make synchronous multi-channel discharges with low jitter. The results indicated that the configuration is promising to make a high energy density plasma with high spectral efficiency.

  17. Effect of disruptions on plasma-facing components

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Bourham, M.A.; Tucker, E.C.

    1995-01-01

    Erosion of plasma-facing components during disruptions is a limiting factor in the design of large tokamaks like ITER. During a disruption, much of the stored thermal energy of the plasma will be dumped onto divertor plates, resulting in local heat fluxes, which may exceed 100 GW/m 2 over a period of about 0.1--1.0 msec. Melted and/or vaporized material is produced which is redistributed in the divertor region. Simulation of disruption damage is summarized from code results and from experimental exposure of materials to high heat-flux plasmas in plasma guns. In the US several codes have been used to predict both melt/vaporization and heat transfer on surfaces as well as energy and momentum transport in the vapor/plasma shield produced at the surface

  18. Flaw detection device for plasma facing wall in thermonuclear device

    International Nuclear Information System (INIS)

    Doi, Akira.

    1996-01-01

    The present invention concerns plasma facing walls of a thermonuclear device and provides a device for detecting a thickness of amour tiles accurately and efficiently with no manual operation. Namely, the position of the plasma facing surface of the amour tile is measured using a structure to which the amour tiles are to be disposed as a reference. Also in a case of disposing new armor tiles, the position of the plasma facing surface of the armor tiles is measured to thereby measure the wearing amount of the amour tiles based on the difference between the reference and the measured value. If a measuring means capable of measuring a plurality of amour tiles at once is used efficiency of the measurement and the detection can be enhanced. Several ten thousands of amour tiles are disposed to the plasma facing wall in a large scaled thermonuclear device, and a plenty of time was required for the detection. However, the present invention can improve the accuracy for the measurement and detection and provide time and labors-saving. (I.S.)

  19. European development of carbon armoured plasma facing components for ITER

    International Nuclear Information System (INIS)

    Merola, M.; Vieider, G.; Wu, C.; Schedler, B.; Chappuis, P.; Escourbiac, F.; Schlosser, J.; Duwe, R.; Roedig, M.; Febvre, M.; Grattarola, M.; Tahtinen, S.; Vesprini, R.

    2001-01-01

    After a brief description of the rationale of the material and geometry selection for each carbon armoured plasma facing components, this paper describes the European development of the two basic geometries, namely the monoblock and the flat tile. An overview of the non-destructive inspection techniques specifically developed for these components is also presented. (orig.)

  20. Brazing and machining of carbon based materials for plasma facing components

    International Nuclear Information System (INIS)

    Brossa, M.; Guerreschi, U.; Rossi, M.

    1994-01-01

    Carbon based materials in the recent years have often been considered and used as armour material in plasma facing components for several fusion devices, because of their low Z and good high temperature characteristics that are compatible with the operation of nuclear reactors. These materials are often connected (mechanically or by brazing) to metals, that allow the support and the cooling functions (heat sink materials). In the following the experience of Ansaldo Ricerche about the study and the manufacturing of plasma facing components and mockups is described with reference to the influence of the carbon materials in performing brazing junction with metals. It is interesting to observe how the different characteristics of the carbon materials influence the brazing process. ((orig.))

  1. Tungsten-microdiamond composites for plasma facing components

    International Nuclear Information System (INIS)

    Livramento, V.; Nunes, D.; Correia, J.B.; Carvalho, P.A.; Mardolcar, U.; Mateus, R.; Hanada, K.; Shohoji, N.; Fernandes, H.; Silva, C.; Alves, E.

    2011-01-01

    Tungsten is considered as one of promising candidate materials for plasma facing component in nuclear fusion reactors due to its resistance to sputtering and high melting point. High thermal conductivity is also a prerequisite for plasma facing components under the unique service environment of fusion reactor characterised by the massive heat load, especially in the divertor area. The feasibility of mechanical alloying of nanodiamond and tungsten, and the consolidation of the composite powders with Spark Plasma Sintering (SPS) was previously demonstrated. In the present research we report on the use of microdiamond instead of nanodiamond in such composites. Microdiamond is more favourable than nanodiamond in view of phonon transport performance leading to better thermal conductivity. However, there is a trade off between densification and thermal conductivity as the SPS temperature increases tungsten carbide formation from microdiamond is accelerated inevitably while the consolidation density would rise.

  2. Analysis of singular interface stresses in dissimilar material joints for plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    You, J.H. E-mail: jeong-ha.you@ipp.mpg.de; Bolt, H

    2001-10-01

    Duplex joint structures are typical material combinations for the actively cooled plasma facing components of fusion devices. The structural integrity under the incident heat loads from the plasma is one of the most crucial issues in the technology of these components. The most critical domain in a duplex joint component is the free surface edge of the bond interface between heterogeneous materials. This is due to the fact that the thermal stress usually shows a singular intensification in this region. If the plasma facing armour tile consists of a brittle material, the existence of the stress singularity can be a direct cause of failure. The present work introduces a comprehensive analytical tool to estimate the impact of the stress singularity for duplex PFC design and quantifies the relative stress intensification in various materials joints by use of a model formulated by Munz and Yang. Several candidate material combinations of plasma facing armour and metallic heat sink are analysed and the results are compared with each other.

  3. Analysis of singular interface stresses in dissimilar material joints for plasma facing components

    International Nuclear Information System (INIS)

    You, J.H.; Bolt, H.

    2001-01-01

    Duplex joint structures are typical material combinations for the actively cooled plasma facing components of fusion devices. The structural integrity under the incident heat loads from the plasma is one of the most crucial issues in the technology of these components. The most critical domain in a duplex joint component is the free surface edge of the bond interface between heterogeneous materials. This is due to the fact that the thermal stress usually shows a singular intensification in this region. If the plasma facing armour tile consists of a brittle material, the existence of the stress singularity can be a direct cause of failure. The present work introduces a comprehensive analytical tool to estimate the impact of the stress singularity for duplex PFC design and quantifies the relative stress intensification in various materials joints by use of a model formulated by Munz and Yang. Several candidate material combinations of plasma facing armour and metallic heat sink are analysed and the results are compared with each other

  4. Investigation of plasma facing components in JT-60U operation

    International Nuclear Information System (INIS)

    Masaki, K.; Ando, T.; Kodama, K.; Arai, T.; Neyatani, Y.; Yoshino, R.; Tsuji, S.; Yagyu, J.; Kaminaga, A.; Sasajima, T.; Ouchi, Y.; Koike, T.; Shimizu, M.

    1995-01-01

    The mechanical fracture of three carbon fiber composite (CFC) first wall tiles was observed. This damage was probably caused by the electromagnetic force due to halo current during disruption. The required current to break the CFC tile is estimated to be 25 kA. The broken tile was rotated poloidally around the plasma with a speed of about 10 m/s during the following discharge. A possible driving force of this rotation might be the electromagnetic force due to the scrape-off layer (SOL) current. The required current to rotate the piece of the broken tile is 1 kA. These results indicate that electromagnetic interaction between SOL plasma and the plasma facing components is important in the research on the plasma wall interactions in fusion devices. ((orig.))

  5. Tritium saturation in plasma-facing materials surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J. [Idaho Nat. Eng. and Environ. Lab., Idaho Falls, ID (United States); Causey, R.A. [Sandia National Labs., Livermore, CA (United States); Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Haasz, A.A. [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

    1998-10-01

    Plasma-facing components in the international thermonuclear experimental reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10{sup 20}-10{sup 23} particles/m{sup 2}s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments. (orig.) 39 refs.

  6. Interaction of stochastic boundary layer with plasma facing components

    International Nuclear Information System (INIS)

    Nguyen, F.; Ghendrih, P.; Grosman, A.

    1997-01-01

    To alleviate the plasma-wall interaction problems in magnetic confinement devices, a stochastic layer is used at the edge of the Tore Supra tokamak (ergodic divertor). A very important point is to determine the power deposition on the plasma facing components. Two different kinds of transport can be identified in such a configuration: Stochastic transport surrounding the confined plasma, with a random walk process, and scrape-off layer (SOL) like transport, a laminar transport, near the plasma facing components. The laminar regime is investigated in terms of a simple criterion, namely that the power deposition is proportional to the radial penetration of the laminar zone flux tubes over a finite parallel length. The magnetic connection properties of the first wall components are then determined. The connection lengths are quantified with two characteristic scales. The larger corresponds to one poloidal turn and appears to be the characteristic parallel length for laminar transport. A field line tracing code MASTOC (magnetic stochastic configuration) is used to computer the complex topology and the statistics of the connection in the real tokamak geometry. The numerical simulations are then compared with the experimental heat deposition on the modules and neutralizer plates of the Tore Supra ergodic divertor. Good agreement is found. Further evidence of laminar transport is also provided by the tangential view of such structures revealed from H α structures in detached plasma experiments. (author). 27 refs, 14 figs

  7. Dynamic behavior of plasma-facing materials during plasma instabilities in tokamak reactors

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1997-01-01

    Damage to plasma-facing and nearby components due to plasma instabilities remains a major obstacle to a successful tokamak concept. The high energy deposited on facing materials during plasma instabilities can cause severe erosion, plasma contamination, and structural failure of these components. Erosion damage can take various forms such as surface vaporization, spallation, and liquid ejection of metallic materials. Comprehensive thermodynamic and radiation hydrodynamic codes have been developed, integrated, and used to evaluate the extent of various damage to plasma-facing and nearby components. The eroded and splashed materials will be transported and then redeposited elsewhere on other plasma-facing components. Detailed physics of plasma/solid-liquid/vapor interaction in a strong magnetic field have been developed, optimized, and implemented in a self-consistent model. The plasma energy deposited in the evolving divertor debris is quickly and intensely reradiated, which may cause severe erosion and melting of other nearby components. Factors that influence and reduce vapor-shielding efficiency such as vapor diffusion and turbulence are also discussed and evaluated

  8. Development of plasma facing components with functionally gradient layers

    Energy Technology Data Exchange (ETDEWEB)

    Morimoto, M.; Kudough, F. [Mitsubishi Atomic Power Industries, Inc., Yokohama (Japan); Onozuka, M.; Tsunoda, H.; Toyoda, M. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan)

    1994-11-01

    The use of functionally graded layers (FGLs) for plasma facing components (PFCs), owing to moderate or piecewise transition in material properties from low-Z surface materials to metal substrates, can provide reduction in thermal stresses, and also provide high thermal load resistance to PFCs. This article deals with the comparison of high heat flux testing and thermal stress analysis results on PFCs. Thermal stress analyses confirmed the thermal loading test results.

  9. ALPS - advanced limiter-divertor plasma-facing systems

    International Nuclear Information System (INIS)

    Allain, J. P.; Bastasz, R.; Brooks, J. N.; Evans, T.; Hassanein, A.; Luckhardt, S.; Maingi, R.; Mattas, R. F.; McCarthy, K.; Mioduszewski, P.; Mogahed, E.; Moir, R.; Molokov, S.; Morely, N.; Nygren, R.; Reed, C.; Rognlien, T.; Ruzic, D.; Sviatoslavsky, I.; Sze, D.; Tillack, M.; Ulrickson, M.; Wade, P. M.; Wong, C.; Wooley, R.

    1999-01-01

    The Advanced Limiter-divertor Plasma-facing Systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and diverters are a peak heat flux of >50 MW/m 2 ,elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (approximately40%). The evaluation of various options is being conducted through a combination of laboratory experiments, modeling of key processes, and conceptual design studies. The current emphasis for the work is on the effects of free surface liquids on plasma edge performance

  10. Heat loads on plasma facing components during disruptions on JET

    International Nuclear Information System (INIS)

    Arnoux, G.; Riccardo, V.; Fundamenski, W.; Loarte, A.; Huber, A.

    2009-01-01

    For the first time, fast measurements of heat loads on the main chamber plasma facing components (about 1 ms time resolution) during disruptions are taken on JET. The timescale of energy deposition during the thermal quench is estimated and compared with the timescale of the core plasma collapse measured with soft x-ray diagnostic. The energy deposition time is 3-8 times longer than the plasma energy collapse during density limit disruptions or radiative limit disruptions. This factor is rather in the range 1.5-4 for vertical displacement events. The heat load profiles measured during the thermal quench show substantial broadening of the power footprint on the upper dump plate. The scrape-off layer power width is increased by a factor of 3 for the density limit disruptions. The far scrape-off layer is characterized by a steeper gradient which could be explained by shadowing of the dump plate by other main chamber plasma facing components such as the outer limiter.

  11. Towards intelligent video understanding applied to plasma facing component monitoring

    International Nuclear Information System (INIS)

    Martin, V.; Travere, J.M.; Moncada, V.; Bremond, F.

    2011-01-01

    In this paper, we promote intelligent plasma facing component video monitoring for both real-time purposes (machine protection issues) and post event analysis purposes (plasma-wall interaction understanding). We propose a vision-based system able to automatically detect and classify into different pre-defined categories thermal phenomena such as localized hot spots or transient thermal events (e.g. electrical arcing) from infrared imaging data of PFCs. This original computer vision system is made intelligent by endowing it with high level reasoning (i.e. integration of a priori knowledge of thermal event spatio-temporal properties to guide the recognition), self-adaptability to varying conditions (e.g. different thermal scenes and plasma scenarios), and learning capabilities (e.g. statistical modelling of event behaviour based on training samples). (authors)

  12. Microscopic Motion of Liquid Metal Plasma Facing Components In A Diverted Plasma

    International Nuclear Information System (INIS)

    Jaworski, M.A.; Gerhardt, S.P.; Morley, N.B.; Abrams, T.; Kaita, R.; Kallman, J.; Kugel, H.; Majeski, R.; Ruzic, D.N.

    2010-01-01

    Liquid metal plasma facing components (PFCs) have been identified as an alternative material for fusion plasma experiments. The use of a liquid conductor where significant magnetic fields are present is considered risky, with the possibility of macroscopic fluid motion and possible ejection into the plasma core. Analysis is carried out on thermoelectric magnetohydrodynamic (TEMHD) forces caused by temperature gradients in the liquid-container system itself in addition to scrape-off-layer currents interacting with the PFC from a diverted plasma. Capillary effects at the liquid-container interface will be examined which govern droplet ejection criteria. Stability of the interface is determined using linear stability methods. In addition to application to liquidmetal PFCs, thin film liquidmetal effects have application to current and future devices where off-normal events may liquefy portions of the first wall and other plasma facing components.

  13. Counter-facing plasma guns for efficient extreme ultra-violet plasma light source

    Directory of Open Access Journals (Sweden)

    Kuroda Yusuke

    2013-11-01

    Full Text Available A plasma focus system composed of a pair of counter-facing coaxial guns was proposed as a long-pulse and/or repetitive high energy density plasma source. We applied Li as the source of plasma for improvement of the conversion efficiency, the spectral purity, and the repetition capability. For operation of the system with ideal counter-facing plasma focus mode, we changed the system from simple coaxial geometry to a multi-channel configuration. We applied a laser trigger to make synchronous multi-channel discharges with low jitter. The results indicated that the configuration is promising to make a high energy density plasma with high spectral efficiency.

  14. Development of advanced high heat flux and plasma-facing materials

    Science.gov (United States)

    Linsmeier, Ch.; Rieth, M.; Aktaa, J.; Chikada, T.; Hoffmann, A.; Hoffmann, J.; Houben, A.; Kurishita, H.; Jin, X.; Li, M.; Litnovsky, A.; Matsuo, S.; von Müller, A.; Nikolic, V.; Palacios, T.; Pippan, R.; Qu, D.; Reiser, J.; Riesch, J.; Shikama, T.; Stieglitz, R.; Weber, T.; Wurster, S.; You, J.-H.; Zhou, Z.

    2017-09-01

    Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling

  15. Low-pressure water-cooled inductively coupled plasma torch

    Science.gov (United States)

    Seliskar, Carl J.; Warner, David K.

    1988-12-27

    An inductively coupled plasma torch is provided which comprises an inner tube, including a sample injection port to which the sample to be tested is supplied and comprising an enlarged central portion in which the plasma flame is confined; an outer tube surrounding the inner tube and containing water therein for cooling the inner tube, the outer tube including a water inlet port to which water is supplied and a water outlet port spaced from the water inlet port and from which water is removed after flowing through the outer tube; and an r.f. induction coil for inducing the plasma in the gas passing into the tube through the sample injection port. The sample injection port comprises a capillary tube including a reduced diameter orifice, projecting into the lower end of the inner tube. The water inlet is located at the lower end of the outer tube and the r.f. heating coil is disposed around the outer tube above and adjacent to the water inlet.

  16. Influence of electron evaporative cooling on ultracold plasma expansion

    International Nuclear Information System (INIS)

    Wilson, Truman; Chen, Wei-Ting; Roberts, Jacob

    2013-01-01

    The expansion of ultracold neutral plasmas (UCP) is driven primarily by the thermal pressure of the electron component and is therefore sensitive to the electron temperature. For typical UCP spatial extents, evaporative cooling has a significant influence on the UCP expansion rate at lower densities (less than 10 8 /cm 3 ). We studied the effect of electron evaporation in this density range. Owing to the low density, the effects of three-body recombination were negligible. We modeled the expansion by taking into account the change in electron temperature owing to evaporation as well as adiabatic expansion and found good agreement with our data. We also developed a simple model for initial evaporation over a range of ultracold plasma densities, sizes, and electron temperatures to determine over what parameter range electron evaporation is expected to have a significant effect. We also report on a signal calibration technique, which relates the signal at our detector to the total number of ions and electrons in the ultracold plasma

  17. Plasma facing surface composition during NSTX Li experiments

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.H., E-mail: cskinner@pppl.gov [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States); Sullenberger, R. [Department of Mechanical and Aerospace Engineering, Princeton University, NJ 08540 (United States); Koel, B.E. [Department of Chemical and Biological Engineering, Princeton University, NJ 08540 (United States); Jaworski, M.A.; Kugel, H.W. [Princeton Plasma Physics Laboratory, POB 451, Princeton, NJ 08543 (United States)

    2013-07-15

    Lithium conditioned plasma facing surfaces have lowered recycling and enhanced plasma performance on many fusion devices. However, the nature of the plasma–lithium surface interaction has been obscured by the difficulty of in-tokamak surface analysis. We report laboratory studies of the chemical composition of lithium surfaces exposed to typical residual gases found in tokamaks. Solid lithium and a molybdenum alloy (TZM) coated with lithium have been examined using X-ray photoelectron spectroscopy, temperature programmed desorption, and Auger electron spectroscopy both in ultrahigh vacuum conditions and after exposure to trace gases. Lithium surfaces near room temperature were oxidized after exposure to 1–2 Langmuirs of oxygen or water vapor. The oxidation rate by carbon monoxide was four times less. Lithiated PFC surfaces in tokamaks will be oxidized in about 100 s depending on the tokamak vacuum conditions.

  18. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  19. Carbon fiber composites application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Akiba, M.; Nakamura, K.; Bonal, J.P.; Pacher, H.D.; Roedig, M.; Vieider, G.; Wu, C.H.

    1998-01-01

    Carbon fiber composites (CFCs) are one of the candidate armour materials for the plasma facing components of the international thermonuclear experimental reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R and D needs are critically discussed. (orig.)

  20. Heat Loads On Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Chantant, M.; Beaumont, B.; Ekedahl, A.; Goniche, M.; Moreau, P.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency launchers plasma facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. Lessons are drawned both with regards to Tore Supra possible operational limits and to ITER ICRF launcher design

  1. Armour Materials for the ITER Plasma Facing Components

    Science.gov (United States)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

    The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

  2. Armour materials for the ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A.R.

    1999-01-01

    The selection of the armour materials for the plasma facing components (PFCs) of the international thermonuclear experimental reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-a-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R and D. (orig.)

  3. Plasma facing components: a conceptual design strategy for the first wall in FAST tokamak

    Science.gov (United States)

    Labate, C.; Di Gironimo, G.; Renno, F.

    2015-09-01

    Satellite tokamaks are conceived with the main purpose of developing new or alternative ITER- and DEMO-relevant technologies, able to contribute in resolving the pending issues about plasma operation. In particular, a high criticality needs to be associated to the design of plasma facing components, i.e. first wall (FW) and divertor, due to physical, topological and thermo-structural reasons. In such a context, the design of the FW in FAST fusion plant, whose operational range is close to ITER’s one, takes place. According to the mission of experimental satellites, the FW design strategy, which is presented in this paper relies on a series of innovative design choices and proposals with a particular attention to the typical key points of plasma facing components design. Such an approach, taking into account a series of involved physical constraints and functional requirements to be fulfilled, marks a clear borderline with the FW solution adopted in ITER, in terms of basic ideas, manufacturing aspects, remote maintenance procedure, manifolds management, cooling cycle and support system configuration.

  4. Heat loads on Tore Supra ICRF Launchers Plasma Facing Components

    International Nuclear Information System (INIS)

    Bremond, S.; Colas, L.; Beaumont, B.; Chantant, M.; Goniche, M.; Mitteau, R.

    2005-01-01

    Understanding the heat loads on Ion Cyclotron Range of Frequency (ICRF) launchers plasma-facing components is a crucial task both for operating present tokamaks and for designing ITER ICRF launchers as these loads may limit the RF power coupling capability. Tore Supra facility is particularly well suited to take this issue. Parametric studies have been performed which enables to get an overall detailed picture of the different heat loads on several areas, pointing to different mechanisms at the origin of the heat power fluxes. It is found that the most critical items for Tore-Supra operation are localized heat loads on the Faraday screen top left corner and vertical edges. Warming up close to maximum temperature limit originally set for protection of the plasma-facing components is found of high power pulses, but no erosion was observed after detailed inspection of the launcher in Tore-Supra vessel. Yet, the associated heat loads could be limiting for Tore-Supra operation in the future, and some dedicated work is under progress to improve the understanding of these power fluxes, pointing out the importance of getting a better knowledge of particle flows in the scrape of layer

  5. Interaction of plasma-facing materials with air and steam

    International Nuclear Information System (INIS)

    Druyts, F.; Fays, J.; Wu, C.H.

    2002-01-01

    In the design of ITER-FEAT, several candidate materials are foreseen for plasma-facing components of the divertor (tungsten, carbon fibre-reinforced composites (CFC), molybdenum) and the first wall (beryllium). In the view of accidental scenarios such as a loss of coolant accident or a loss of vacuum accident the reaction between these materials and steam or air remains a safety concern. To provide kinetic data, describing the chemical reactivity of plasma-facing materials in air and steam, we used coupled thermogravimetry/quadrupole mass spectrometry. In this paper we present the results of a screening investigation that compares the oxidation rates of tungsten, molybdenum, CFC and beryllium in the temperature range 300-700 deg. C. From the thermogravimetry and mass spectrometry results we obtained the reaction rates as a function of temperature. For the metals tungsten, molybdenum and beryllium, a transition is observed between protective oxidation at lower temperatures and non-protective oxidation at higher temperatures. This transition temperature lies in the range 500-550 deg. C for tungsten and molybdenum, which is lower than for beryllium. At above temperatures 550 deg. C, the oxides formed on molybdenum and tungsten volatilise. This increases the oxidation rate dramatically and can lead to mobilisation of activation products in a fusion reactor. We also performed experiments on both undoped CFC and CFC doped with 8-10% silicon. The influence of silicon doping on the chemical reactivity of CFC's in air is discussed

  6. Hydrogen transport behavior of metal coatings for plasma facing components

    International Nuclear Information System (INIS)

    Anderl, R.A.; Holland, D.F.; Longhurst, G.R.

    1990-01-01

    Plasma-facing components for experimental and commercial fusion reactor studies may include cladding or coatings of refractory metals like tungsten on metallic structural substrates such as copper, vanadium alloys and austenitic stainless steel. Issues of safety and fuel economy include the potential for inventory buildup and permeation of tritium implanted into the plasma-facing surface. This paper reports on laboratory-scale studies with 3-keV D 3 + ion beams to investigate the hydrogen transport behavior in tungsten coatings on substrates of copper. These experiments entailed measurements of the deuterium re-emission and permeation rates for tungsten, copper, and tungsten-coated copper specimens at temperatures ranging from 638 K to 825 K and implanting particle fluxes of approximately 5 x 10 19 D/m 2 s. Diffusion constants and surface recombination coefficients with enhancement factors due to sputtering were obtained from these measurements. These data may be used in calculations to estimate permeation rates and inventory buildups for proposed diverter designs. 18 refs., 3 figs., 3 tabs

  7. Engineering design and thermal hydraulics of plasma facing components of SST-1

    International Nuclear Information System (INIS)

    Pragash, N. Ravi; Chaudhuri, P.; Santra, P.; Chenna Reddy, D.; Khirwadkar, S.; Saxena, Y.C.

    2001-01-01

    SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (∼1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m 2 . In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 degree sign C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a FORTRAN code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ANSYS software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper

  8. Evaluation of surface fractal dimension of carbon for plasma-facing material damaged by hydrogen plasma

    International Nuclear Information System (INIS)

    Nishino, Nobuhiro

    1997-01-01

    The surface structure of the plasma facing materials (PFM) changes due to plasma-surface interaction in a nuclear fusion reactor. Usually B 4 C coated graphite block are used as PFM. In this report, the surface fractal was applied to study the surface structure of plasma-damaged PFM carbon. A convenient flow-type adsorption apparatus was developed to evaluate the surface fractal dimension of materials. Four branched alkanol molecules with different apparent areas were used as the probe adsorbates. The samples used here were B 4 C coated isotopic graphite which were subjected to hydrogen plasma for various periods of exposure. The monolayer capacities of these samples for alkanols were determined by applying BET theory. The surface fractal dimension was calculated using the monolayer capacities and molecular areas for probe molecules and was found to increase from 2 to 3 with the plasma exposure time. (author)

  9. Overview of decade-long development of plasma-facing components at ASIPP

    Science.gov (United States)

    Luo, G.-N.; Liu, G. H.; Li, Q.; Qin, S. G.; Wang, W. J.; Shi, Y. L.; Xie, C. Y.; Chen, Z. M.; Missirlian, M.; Guilhem, D.; Richou, M.; Hirai, T.; Escourbiac, F.; Yao, D. M.; Chen, J. L.; Wang, T. J.; Bucalossi, J.; Merola, M.; Li, J. G.; EAST Team

    2017-06-01

    The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m-2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m-2 and 1000 cycles at 20 MW m-2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m-2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.

  10. HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

    International Nuclear Information System (INIS)

    Visca, Eliseo; Roccella, S.; Candura, D.; Palermo, M.; Rossi, P.; Pizzuto, A.; Sanguinetti, G.P.; Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G.

    2015-01-01

    Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m 2 but the capability to remove up to 20 MW/m 2 during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

  11. HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: eliseo.visca@enea.it [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Roccella, S. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Candura, D.; Palermo, M. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Rossi, P.; Pizzuto, A. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Sanguinetti, G.P. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy)

    2015-10-15

    Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m{sup 2} but the capability to remove up to 20 MW/m{sup 2} during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

  12. Conditionings for boron-carbon plasma facing wall

    International Nuclear Information System (INIS)

    Hino, Tomoaki; Yamauchi, Yuji; Yamashina, Toshiro

    1994-01-01

    For plasma facing material with components of boron and carbon, the method of conditionings due to He discharge cleaning and baking is considered. The conditioning time required to suppress the hydrogen recycling is discussed. It is shown that the hydrogen trapped by the boron can be relatively easily removed only by the baking at 300degC or only by He discharge cleaning with current density of 0.1 mA/cm 2 . It is not easy to remove the hydrogen trapped by the carbon by the baking since the temperature required becomes 500degC. The current density required also becomes high, 1 mA/cm 2 , for the reduction of the hydrogen trapped by the carbon. (author)

  13. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    International Nuclear Information System (INIS)

    Bisio, M.; Branca, V.; Marco, M. Di; Federici, A.; Grattarola, M.; Gualco, G.; Guarnone, P.; Luconi, U.; Merola, M.; Ozzano, C.; Pasquale, G.; Poggi, P.; Rizzo, S.; Varone, F.

    2005-01-01

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions

  14. Safety characteristics of options for plasma-facing components for ITER and beyond

    International Nuclear Information System (INIS)

    Piet, S.J.; McCarthy, K.A.; Holland, D.F.; Longhurst, G.R.; Merrill, B.J.

    1991-01-01

    Plasma-facing components (PFC) likely dominate the safety hazards of the International Thermonuclear Experimental Reactor (ITER) and post-ITER machines. To gain regulatory approval and for fusion energy to fulfill its ultimate attractive safety and environmental potential, safety must be considered when selecting among PFC options. This paper summarizes current PFC safety information. PFC safety issues fall into seven areas: disruption tolerance, disruption severity, tritium inventory and permeation, accidental energy release, activation/toxin hazards, cooling disturbances, and system issues. RFC options include current ITER mainline options (Be or W coating, C tiles), variants on current ITER options, and liquid metal (LM) divertors. No PFC option that we have examined is free of critical safety concerns. There are also innovative ideas that may improve any PFC's performance -- super-permeable vacuum ducts, helium self-pumping, and gaseous divertors. We conclude with recommendations and a future strategy. 17 refs., 1 fig., 3 tabs

  15. Towards intelligent video understanding applied to plasma facing component monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Martin, V.; Bremond, F. [INRIA, Pulsa team-project, Sophia Antipolis (France); Travere, J.M. [CEA IRFM, Saint Paul-lez-Durance (France); Moncada, V.; Dunand, G. [Sophia Conseil Company, Sophia Antipolis (France)

    2011-07-01

    Infrared thermography has become a routine diagnostic in many magnetic fusion devices to monitor the heat loads on the plasma facing components (PFCs) for both physics studies and machine protection. The good results of the developed systems obtained so far motivate the use of imaging diagnostics for control, especially during long pulse tokamak operation (e.g. lasting several minutes). In this paper, we promote intelligent monitoring for both real-time purposes (machine protection issues) and post event analysis purposes (PWI understanding). We propose a vision-based system able to automatically detect and classify into different pre-defined categories phenomena as localized hot spots, transient thermal events (e.g. electrical arcing), and unidentified flying objects (UFOs) as dusts from infrared imaging data of PFCs. This original vision system is made intelligent by endowing it with high-level reasoning (i.e. integration of a priori knowledge of thermal event spatial and temporal properties to guide the recognition), self-adaptability to varying conditions (e.g. different plasma scenarios), and learning capabilities (e.g. statistical modelling of thermal event behaviour based on training samples). This approach has been already successfully applied to the recognition of one critical thermal event at Tore Supra. We present here latest results of its extension for the recognition of others thermal events (e.g., B{sub 4}C flakes, impact of fast particles, UFOs) and show how extracted information can be used during plasma operation at Tore Supra to improve the real time control system, and for further analysis of PFC aging. This document is composed of an abstract followed by the slides of the presentation. (authors)

  16. Liquid-metal plasma-facing component research on the National Spherical Torus Experiment

    International Nuclear Information System (INIS)

    Jaworski, M A; Khodak, A; Kaita, R

    2013-01-01

    Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. Liquid metals face critical issues in three key areas: free-surface stability, material migration and demonstration of integrated scenarios. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The liquid lithium divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m −2 , no macroscopic ejection events were observed. The stability can be understood from a Rayleigh–Taylor instability analysis. Capillary restraint and thermal-hydraulic considerations lead to a proposed liquid-metal PFCs scheme of actively-supplied, capillary-restrained systems. Even with state-of-the-art cooling techniques, design studies indicate that the surface temperature with divertor-relevant heat fluxes will still reach temperatures above 700 °C. At this point, one would expect significant vapor production from a liquid leading to a continuously vapor-shielded regime. Such high-temperature liquid lithium PFCs may be possible on the basis of momentum-balance arguments. (paper)

  17. Liquid-metal plasma-facing component research on the National Spherical Torus Experiment

    Science.gov (United States)

    Jaworski, M. A.; Khodak, A.; Kaita, R.

    2013-12-01

    Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation of economically viable fusion power reactors. Liquid metals face critical issues in three key areas: free-surface stability, material migration and demonstration of integrated scenarios. To date, few demonstrations exist of this approach in a diverted tokamak and we here provide an overview of such work on the National Spherical Torus Experiment (NSTX). The liquid lithium divertor (LLD) was installed and operated for the 2010 run campaign using evaporated coatings as the filling method. Despite a nominal liquid level exceeding the capillary structure and peak current densities into the PFCs exceeding 100 kA m-2, no macroscopic ejection events were observed. The stability can be understood from a Rayleigh-Taylor instability analysis. Capillary restraint and thermal-hydraulic considerations lead to a proposed liquid-metal PFCs scheme of actively-supplied, capillary-restrained systems. Even with state-of-the-art cooling techniques, design studies indicate that the surface temperature with divertor-relevant heat fluxes will still reach temperatures above 700 °C. At this point, one would expect significant vapor production from a liquid leading to a continuously vapor-shielded regime. Such high-temperature liquid lithium PFCs may be possible on the basis of momentum-balance arguments.

  18. Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C.

    2010-01-01

    sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. To test and to demonstrate the acceptability of plasma facing materials and components special high heat flux test facilities based on intense ion or electron beams are being used routinely to demonstrate the heat removal efficiency and the lifetime under fusion specific loading conditions. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm -2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions. Here a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone; the time scale of these events is typically in the order of 1 ms. As a consequence, thermal shock induced crack formation, vaporization, surface melting and droplet ejection as well as particle emission induced by brittle destruction processes will limit the lifetime of the components. This is also valid for instabilities in the plasma positioning (vertical displacement events) which cause irreversible damage to plasma facing components, particularly to the metallic wall armour. Moreover, dust particles (neutron activated or toxic metals or tritium enriched carbon) are a serious concern from a safety point of view. In order to investigate the thermally induced plasma wall interaction under fusion specific thermal loads, high heat flux simulation tests are performed in electron or ion beam test facilities as well as in quasi stationary plasma devices. These experiments cover thermal fatigue loads and/or thermal shock tests with relevant operational loading conditions. Furthermore, the wall bombardment

  19. Development and evaluation of plasma facing materials for future thermonuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Pintsuk, G.; Roedig, M.; Schmidt, A.; Thomser, C. [Forschungszentrum Juelich GmbH, EURATOM Association, Juelich (Germany)

    2010-07-01

    and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. To test and to demonstrate the acceptability of plasma facing materials and components special high heat flux test facilities based on intense ion or electron beams are being used routinely to demonstrate the heat removal efficiency and the lifetime under fusion specific loading conditions. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm{sup -2} are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions. Here a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone; the time scale of these events is typically in the order of 1 ms. As a consequence, thermal shock induced crack formation, vaporization, surface melting and droplet ejection as well as particle emission induced by brittle destruction processes will limit the lifetime of the components. This is also valid for instabilities in the plasma positioning (vertical displacement events) which cause irreversible damage to plasma facing components, particularly to the metallic wall armour. Moreover, dust particles (neutron activated or toxic metals or tritium enriched carbon) are a serious concern from a safety point of view. In order to investigate the thermally induced plasma wall interaction under fusion specific thermal loads, high heat flux simulation tests are performed in electron or ion beam test facilities as well as in quasi stationary plasma devices. These experiments cover thermal fatigue loads and/or thermal shock tests with relevant operational loading conditions. Furthermore, the

  20. ITER vacuum vessel, in vessel components and plasma facing materials

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Enoeda, M.; Federici, G.

    2007-01-01

    Design of the NB ports including duct liners under heat loads of the neutral beams has been developed. Design of the in-wall shielding has been developed in more details considering the supporting structure and the assembly method. The ferromagnetic inserts have previously not been installed in the outboard midplane region due to irregularity caused by the tangential ports for NB injection. Due to this configuration, the maximum ripple is relatively large (∝1 %) in a limited region of the plasma and the toroidal field flux lines fluctuate ∝10 mm in the FW region. To avoid these problems, additional ferromagnetic inserts are to be installed in the equatorial port region. Detailed studies were carried out on the ITER vacuum vessel to define appropriate codes and standards in the context of the ITER licensing in France. A set of draft documents regarding the ITER vacuum vessel structural code were prepared including an RCC-MR Addendum for the ITER VV with justified exceptions or modifications. The main deviation from the base Code is the extensive use of UT in lieu of radiography for the volumetric examination of all one-side access welds of the outer shell and field joint. The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. Development of new beryllium materials is progressing in China and Russia. The ITER limiters will be installed in equatorial ports at two toroidal locations. The limiter plasma-facing surface protrudes ∝8 cm from the FW during the start-up and shutdown phase. In the new limiter concept, the limiters are retracted by ∝8 cm during the plasma flat top phase. This concept gives important advantages; (i) mitigation of the particle and heat loads due to disruptions, ELMs and blobs, (ii) improvement of the power coupling with the ICRH antenna

  1. Hydrogen in tungsten as plasma-facing material

    Science.gov (United States)

    Roth, Joachim; Schmid, Klaus

    2011-12-01

    Materials facing plasmas in fusion experiments and future reactors are loaded with high fluxes (1020-1024 m-2 s-1) of H, D and T fuel particles at energies ranging from a few eV to keV. In this respect, the evolution of the radioactive T inventory in the first wall, the permeation of T through the armour into the coolant and the thermo-mechanical stability after long-term exposure are key parameters determining the applicability of a first wall material. Tungsten exhibits fast hydrogen diffusion, but an extremely low solubility limit. Due to the fast diffusion of hydrogen and the short ion range, most of the incident ions will quickly reach the surface and recycle into the plasma chamber. For steady-state operation the solute hydrogen for the typical fusion reactor geometry and wall conditions can reach an inventory of about 1 kg. However, in short-pulse operation typical of ITER, solute hydrogen will diffuse out after each pulse and the remaining inventory will consist of hydrogen trapped in lattice defects, such as dislocations, grain boundaries and irradiation-induced traps. In high-flux areas the hydrogen energies are too low to create displacement damage. However, under these conditions the solubility limit will be exceeded within the ion range and the formation of gas bubbles and stress-induced damage occurs. In addition, simultaneous neutron fluxes from the nuclear fusion reaction D(T,n)α will lead to damage in the materials and produce trapping sites for diffusing hydrogen atoms throughout the bulk. The formation and diffusive filling of these different traps will determine the evolution of the retained T inventory. This paper will concentrate on experimental evidence for the influence different trapping sites have on the hydrogen inventory in W as studied in ion beam experiments and low-temperature plasmas. Based on the extensive experimental data, models are validated and applied to estimate the contribution of different traps to the tritium inventory in

  2. Improved CuCrZr/316L transition for plasma facing components

    International Nuclear Information System (INIS)

    Tabernig, Bernhard; Rainer, Florian; Scheiber, Karl-Heinz; Schedler, Bertram

    2007-01-01

    Different welding strategies were investigated to improve the tubular transition of CuCrZr to 316L in cooling pipes for actively cooled plasma facing components. Electron beam welding experiments have been carried out on tubular samples using different filler and adapter materials. After non-destructive testing by dye penetrant and He-leak tight testing samples were tensile tested at RT and 400 deg. C to down-select promising candidates. Furthermore samples were taken for a metallographic examination in order to determine the integrity of the welds, the depth of penetration and the hardness profile across the weld. In the scanning electron microscope the weld microstructure and the formation of phases were studied. Good results were obtained by the use of a Ni-filler, an Inconel and explosive welded adapter. The tested samples of these variations fulfilled the strength requirements according to the ITER specification and showed an improved transition compared with the current solution of a pure Ni-adapter. The final down-selection will be based on the results of fatigue and torsion testing

  3. Thermal-hydraulic and thermo-mechanical design of plasma facing components for SST-1 tokamak

    International Nuclear Information System (INIS)

    Chaudhuri, Paritosh; Santra, P.; Chenna Reddy, D.; Parashar, S.K.S.

    2014-01-01

    The Plasma Facing Components (PFCs) are one of the major sub-systems of ssT-1 tokamak. PFC of ssT-1 consisting of divertors, passive stabilizers, baffles and limiters are designed to be compatible for steady state operation. The main consideration in the design of the PFC cooling is the steady state heat removal of up to 1 MW/m 2 . The PFC has been designed to withstand the peak heat fluxes and also without significant erosion such that frequent replacement of the armor is not necessary. Design considerations included 2-D steady state and transient tile temperature distribution and resulting thermal loads in PFC during baking, and cooling, coolant parameters necessary to maintain optimum thermal-hydraulic design, and tile fitting mechanism. Finite Element (FE) models using ANSYS have been developed to carry out the heat transfer and stress analyses of the PFC to understand its thermal and mechanical behaviors. The results of the calculation led to a good understanding of the coolant flow behavior and the temperature distribution in the tube wall and the different parts of the PFC. Thermal analysis of the PFC is carried out with the purpose of evaluating the thermal mechanical behavior of PFCs. The detailed thermal-hydraulic and thermo-mechanical designs of PFCs of ssT-1 are discussed in this paper. (authors)

  4. Comprehensive simulation of vertical plasma instability events and their serious damage to ITER plasma facing components

    International Nuclear Information System (INIS)

    Hassanein, A.; Sizyuk, T.

    2008-01-01

    Safe and reliable operation is still one of the major challenges in the development of the new generation of ITER-like fusion reactors. The deposited plasma energy during major disruptions, edge-localized modes (ELMs) and vertical displacement events (VDEs) causes significant surface erosion, possible structural failure and frequent plasma contamination. While plasma disruptions and ELM will have no significant thermal effects on the structural materials or coolant channels because of their short deposition time, VDEs having longer-duration time could have a destructive impact on these components. Therefore, modelling the response of structural materials to VDE has to integrate detailed energy deposition processes, surface vaporization, phase change and melting, heat conduction to coolant channels and critical heat flux criteria at the coolant channels. The HEIGHTS 3D upgraded computer package considers all the above processes to specifically study VDE in detail. Results of benchmarking with several known laboratory experiments prove the validity of HEIGHTS implemented models. Beryllium and tungsten are both considered surface coating materials along with copper structure and coolant channels using both smooth tubes with swirl tape insert. The design requirements and implications of plasma facing components are discussed along with recommendations to mitigate and reduce the effects of plasma instabilities on reactor components.

  5. Development of Si–W transient tolerant plasma facing material

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C., E-mail: wongc@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Chen, B. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Hollmann, E.M.; Rudakov, D.L. [University of California, San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Wall, D.; Tao, R. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Wright, M. [Ultramet, 12173 Montague Street, Pacoima, CA 91331 (United States)

    2013-07-15

    Solid W is projected as the preferred plasma facing material. Unfortunately, W surfaces could suffer radiation damage under DT operation and will melt under Type-I edge localized modes and disruption events. A possible approach is the use of a low-Z sacrificial material, like Si deposited on the W-surface to withstand a few type-I ELMs and/or disruptions via the vapor shielding effect. Accordingly, sets of Si–W test buttons were fabricated and exposed in the DIII-D lower divertor. We found that when the Si–W buttons were exposed to a few DIII-D vertical displacement event disruptions, tungsten–silicide was formed which melts at 1414 °C. This clearly indicates that the Si–W combination cannot be used as a transient tolerance surface material, since the W surface can be damaged. Even when Si is used as a wall conditioning material the Si–W surface temperature should be operated at much lower than 1400 °C.

  6. Technological challenges at ITER plasma facing components production in Russia

    Energy Technology Data Exchange (ETDEWEB)

    Mazul, I.V., E-mail: mazuliv@niiefa.spb.su [Efremov Institute, 196641 St. Petersburg (Russian Federation); Belyakov, V.A.; Gervash, A.A.; Giniyatulin, R.N.; Guryeva, T.M.; Kuznetsov, V.E.; Makhankov, A.N.; Okunev, A.A. [Efremov Institute, 196641 St. Petersburg (Russian Federation); Sevryukov, O.N. [MEPhI, 115409 Moscow (Russian Federation)

    2016-11-01

    Highlights: • Technological aspects of ITER PFC manufacturing in Russia are presented. • Range of technologies to be used during manufacturing of ITER PFC at Efremov Institute has been, in general, defined and their complexity, originality and difficulty are described. • Some features and challenges of welding, brazing and various tests are discussed. - Abstract: Major part of ITER plasma facing components will be manufactured in the Russian Federation (RF). Operational conditions and other requirements to these components, as well as the scale of production, are quite unique. These unique features and related technological solutions found in the frame of the project are discussed. Procedure breakdown and results of qualification for the proposed technologies and potential producers are presented, based on mockups production and testing. Design of qualification mockups and prototypes, testing programs and results are described. Basic quantitative and qualitative parameters of manufactured components and methods of quality control are presented. Critical manufacturing issues and prospects for unique production for future fusion needs are discussed.

  7. Surface temperature measurement of plasma facing components in tokamaks

    International Nuclear Information System (INIS)

    Amiel, Stephane

    2014-01-01

    During this PhD, the challenges on the non-intrusive surface temperature measurements of metallic plasma facing components in tokamaks are reported. Indeed, a precise material emissivity value is needed for classical infrared methods and the environment contribution has to be known particularly for low emissivities materials. Although methods have been developed to overcome these issues, they have been implemented solely for dedicated experiments. In any case, none of these methods are suitable for surface temperature measurement in tokamaks.The active pyrometry introduced in this study allows surface temperature measurements independently of reflected flux and emissivities using pulsed and modulated photothermal effect. This method has been validated in laboratory on metallic materials with reflected fluxes for pulsed and modulated modes. This experimental validation is coupled with a surface temperature variation induced by photothermal effect and temporal signal evolvement modelling in order to optimize both the heating source characteristics and the data acquisition and treatment. The experimental results have been used to determine the application range in temperature and detection wavelengths. In this context, the design of an active pyrometry system on tokamak has been completed, based on a bicolor camera for a thermography application in metallic (or low emissivity) environment.The active pyrometry method introduced in this study is a complementary technique of classical infrared methods used for thermography in tokamak environment which allows performing local and 2D surface temperature measurements independently of reflected fluxes and emissivities. (author) [fr

  8. Technologies for ITER divertor vertical target plasma facing components

    International Nuclear Information System (INIS)

    Schlosser, J.; Escourbiac, F.; Merola, M.; Fouquet, S.; Bayetti, P.; Cordier, J.J.; Grosman, A.; Missirlian, M.; Tivey, R.; Roedig, M.

    2005-01-01

    The ITER divertor vertical target has to sustain heat fluxes up to 20 MW m -2 . The concept developed for this plasma facing component working at steady state is based on carbon fibre composite armour for the lower straight part and tungsten for the curved upper part. The main challenges involved in the use of such components include the removal of the high heat fluxes deposited and mechanically and thermally joining the armour to the metallic heat sink, despite the mismatch in the thermal expansions. Two solutions based on the use of a CuCrZr hardened copper alloy and an active metal casting (AMC (registered) ) process were investigated during the ITER EDA phase: the first one called 'flat tile geometry' was mainly developed for the Tore Supra pumped limiter, the second one called 'monoblock geometry' was developed by the EU Participating Team for the ITER project. This paper presents a review of these two solutions and analyses their assets and drawbacks: pressure drop, critical heat flux, surface temperature and expected behaviour during operation, risks during the manufacture, control of the armour defects during the manufacture and at the reception, and the possibility of repairing defective tiles

  9. Laser cooled ion beams and strongly coupled plasmas for precision experiments

    International Nuclear Information System (INIS)

    Bussmann, Michael

    2008-01-01

    This cumulative thesis summarizes experimental and theoretical results on cooling of ion beams using single-frequency, single-mode tabletop laser systems. It consists of two parts. One deals with experiments on laser-cooling of ion beams at relativistic energies, the other with simulations of stopping and sympathetic cooling of ions for precision in-trap experiments. In the first part, experimental results are presented on laser-cooling of relativistic C 3+ ion beams at a beam energy of 122 MeV/u, performed at the Experimental Storage Ring (ESR) at GSI. The main results presented in this thesis include the first attainment of longitudinally space-charge dominated relativistic ion beams using pure laser-cooling. The second part lists theoretical results on stopping and sympathetic cooling of ions in a laser-cooled one-component plasma of singly charged 24 Mg ions, which are confined in a three-dimensional harmonic trap potential. (orig.)

  10. Interaction of candidate plasma facing materials with tokamak plasma in COMPASS

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Macková, Anna; Malinský, Petr; Havránek, Vladimír; Naydenkova, Diana; Klevarová, Veronika; Petersson, P.; Gasior, P.; Hakola, A.; Rubel, M.; Fortuna, E.; Kolehmainen, J.; Tervakangas, S.

    2017-01-01

    Roč. 493, September (2017), s. 102-119 ISSN 0022-3115. [International Conference on Plasma-Facing Materials and Components for Fusion Applications/15./. Aix-en-Provence, 18.05.2015-22.05.2015] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045; GA MŠk LM2015056 Institutional support: RVO:61389021 ; RVO:61389005 Keywords : erosion * COMPASS tokamak * plasma-material interaction * ion beam analysis Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (UJF-V) OBOR OECD: Nuclear related engineering ; Nuclear related engineering (UJF-V) Impact factor: 2.048, year: 2016 http://www.sciencedirect.com/science/ article /pii/S0022311517301708

  11. Advanced solutions for beryllium and tungsten plasma-facing components

    International Nuclear Information System (INIS)

    Ibbott, C.; Jakeman, R.; Ando, T.; Chiocchio, S.; Federici, G.; Heidl, H.; Tivey, R.; Falter, H.; Ciric, D.; Merola, M.; Vieider, G.; Ploechl, L.; Roedig, M.

    1998-01-01

    Beryllium and tungsten are candidate plasma-facing armour materials for the International Thermonuclear Experimental Reactor (ITER). These armours are proposed for areas with low heat flux (≤5 MW m -2 ); however, in the divertor, surface melting during abnormal events may occur. This paper reports the progress made in developing novel approaches to solving the difficulties posed in designing with these armours. A Be monoblock brazed to an OFHC 10 mm ID Cu tube using InCuSil 'ABA' braze alloy has survived 130 cycles of 10-11 MW m -2 for 6 s, with surface temperatures of 1250 C. No visible surface cracking occurred. The same monoblock was then exposed to several cycles of 20-22 MW m -2 for 8 s, creating a 2 mm deep molten layer. High cycle fatigue was then performed. The test results are detailed in this paper. Comparison between experimental and theoretical results are made. W and Cu have a large mismatch in their thermal expansion coefficients and two designs are proposed that minimise the interface stresses. These are: a 'brush'-like structure with rectangular fibres set in a Cu substrate using the 'active metal casting' (AMC) technique; and thin monoblocks (or lamellae) brazed or active metal cast onto a Cu tube. Analyses of the lamellae concept for steady-state heat loads of 5 MW m -2 are presented. Fatigue analyses show that both solutions are theoretically viable (∝10 4 cycles). A 'brush' mock-up has been manufactured and progress on its testing is reported. Results of all tests and their relevance to the ITER design are discussed. (orig.)

  12. Selection of the optimal hard facing (HF technology of damaged forging dies based on cooling time t8/5

    Directory of Open Access Journals (Sweden)

    D. Arsić

    2016-01-01

    Full Text Available In exploitation, the forging dies are exposed to heating up to very high temperatures, variable loads: compressive, impact and shear. In this paper, the reparatory hard facing of the damaged forging dies is considered. The objective was to establish the optimal reparatory technology based on cooling time t8/5 . The verification of the adopted technology was done by investigation of the hard faced layers microstructure and measurements of hardness within the welded layers’ characteristic zones. Cooling time was determined theoretically, numerically and experimentally.

  13. The quantity of algae colonizing the inside face of cooling towers and the consequences for wear of the shell

    International Nuclear Information System (INIS)

    Aprosi, G.; Chauvel, D.

    1990-01-01

    These studies are part of the interdirectorate working group's mandate relating to lifespan project on cooling towers. Involving the collaboration of several divisions of Electricity de France: the Construction Division (SEPTEN). The Generation and Transmission Division (SPT) and the Research and Development Division (EAA). Among the biological colonies which proliferate in the cooling circuits of power stations, algae are broadly represented in the form of wall coatings which cover the inside face of cooling towers: shell algae. They can also grow at other points in the cooling circuit; in the cold water basin, in the fill, and, in some cooling towers, in the hot water basin. These plant organisms hamper the operation of power stations by clogging the grids located in the pipe from the cold water basin to the condenser. In addition, when algae come free of the shell, they remove micro-fragments of the concrete, which could accelerate wear. This paper presents the findings of studies conducted by the Aquatic and Atmospheric Department on the infestation of cooling towers by algae. In particular, the results of studies to evaluate the quantity of algae on the inside face of the shell of cooling towers. Many scenarios will be proposed, linked to the operation of the plant and to the local meteorological conditions

  14. Soft X-ray spectroscopy of high-Z ions in a cool dense plasma

    International Nuclear Information System (INIS)

    Presnyakov, L.P.; Shevelko, A.P.; Uskov, D.B.

    1991-01-01

    Spectra of multiply-charged ions with one and two electrons are investigated when a laser-produced plasma interacts with a solid obstacle. The level population densities of the Mg ions, recombining in the region of dense cool plasma near the obstacle, are studied both experimentally and theoretically. Values of the gain coefficient are calculated for the case of carbon ions. (orig.)

  15. Spark plasma sintering of pure and doped tungsten as plasma facing material

    Science.gov (United States)

    Autissier, E.; Richou, M.; Minier, L.; Naimi, F.; Pintsuk, G.; Bernard, F.

    2014-04-01

    In the current water cooled divertor concept, tungsten is an armour material and CuCrZr is a structural material. In this work, a fabrication route via a powder metallurgy process such as spark plasma sintering is proposed to fully control the microstructure of W and W composites. The effect of chemical composition (additives) and the powder grain size was investigated. To reduce the sintering temperature, W powders doped with a nano-oxide dispersion of Y2O3 are used. Consequently, the sintering temperature for W-oxide dispersed strengthened (1800 °C) is lower than for pure W powder. Edge localized mode tests were performed on pure W and compared to other preparation techniques and showed promising results.

  16. Recombining processes in a cooling plasma by mixing of initially heated gas

    International Nuclear Information System (INIS)

    Furukane, Utaro; Sato, Kuninori; Takiyama, Ken; Oda, Toshiatsu.

    1992-03-01

    A numerical investigation of recombining process in a high temperature plasma in a quasi-steady state is made in a gas contact cooling, in which the initial temperature effect of contact gas heated up by the hot plasma is considered as well as the gas cooling due to the surrounding neutral particles freely coming into the plasma. The calculation has shown that the electron temperature relaxes in accord with experimental results and that the occurrence of recombining region and the inverted populations almost agree with the experimental ones. (author)

  17. The tritium confinement and surface chemistry of plasma facing materials in controlled D-T fusion devices

    International Nuclear Information System (INIS)

    Wu, C.H.

    1987-01-01

    Tritium permeation through first walls, limiters or divertors subjected to energetic tritium charge exchange neutral bombardment is a potentially serious problem area for advanced D-T reactors operating at elevated temperatures. High concentrations of tritium in the near surface region can be reached by implantation of the charge neutral flux combined with a relatively slow recombination of these atoms into molecules at the plasma/ first wall interface. A concentration gradient is established, causing tritium to diffuse into the bulk and essentially to the outer wall surface where it can enter the first wall coolant. Since tritium separation from cooling water is very costly, release of even a small fraction of tritium to the environment could pose undesirable safety problems. Therefore, it is necessary to reduce the tritium permeation. An analysis of the way of inhibition has been made. The tritium interacts with the solid surface of the plasma facing components, resulting in trapping and material erosion, and posing problems with respect to plasma density control. The erosion of the plasma facing component materials is mainly caused by physical and chemical erosion. A detailed analysis of chemical erosion by tritium has been performed and the results are described. (author)

  18. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    International Nuclear Information System (INIS)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won; Cho, Seungyon

    2014-01-01

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity

  19. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

  20. Plasma Facing Components Generic Facilities Review Panel (PFC-GFRP): Final report

    International Nuclear Information System (INIS)

    McGrath, R.; Allen, S.; Hill, D.; Brooks, J.; Mattas, R.; Davis, J.; Lipschultz, B.; Ulrickson, M.

    1993-10-01

    The Plasma Facing Components (PFC) Facilities Review Panel was chartered by the US Department of Energy, Office of Fusion Energy, ITER (International Thermonuclear Experimental Reactor) and Technology Division, to outline the program plan and identify the supporting test facilities that lead to reliable, long-lived plasma facing components for ITER. This report summarizes the panel's findings and identifies the necessary and sufficient set of test facilities required for ITER PFC development

  1. Data merging of infrared and ultrasonic images for plasma facing components inspection

    Energy Technology Data Exchange (ETDEWEB)

    Richou, M. [CEA, IRFM, F-13108 Saint Paul-lez-Durance (France)], E-mail: marianne.richou@cea.fr; Durocher, A. [CEA, IRFM, F-13108 Saint Paul-lez-Durance (France); Medrano, M. [Association EURATOM - CIEMAT, Avda. Complutense 22, 28040 Madrid (Spain); Martinez-Ona, R. [Tecnatom, 28703 S. Sebastian de los Reyes, Madrid (Spain); Moysan, J. [LCND, Universite de la Mediterranee, F-13625 Aix-en-Provence (France); Riccardi, B. [Fusion For Energy, 08019 Barcelona (Spain)

    2009-06-15

    For steady-state magnetic thermonuclear fusion devices which need large power exhaust capability, actively cooled plasma facing components have been developed. In order to guarantee the integrity of these components during the required lifetime, their thermal and mechanical behaviour must be assessed. Before the procurement of the ITER Divertor, the examination of the heat sink to armour joints with non-destructive techniques is an essential topic to be addressed. Defects may be localised at different bonding interfaces. In order to improve the defect detection capability of the SATIR technique, the possibility of merging the infrared thermography test data coming from SATIR results with the ultrasonic test data has been identified. The data merging of SATIR and ultrasonic results has been performed on Carbon Fiber Composite (CFC) monoblocks with calibrated defects, identified by their position and extension. These calibrated defects were realised with machining, with 'stop-off' or by a lack of CFC activation techniques, these last two representing more accurately a real defect. A batch of 56 samples was produced to simulate each possibility of combination with regards to interface location, position and extension and way of realising the defect. The use of a data merging method based on Dempster-Shafer theory improves significantly the detection sensibility and reliability of defect location and size.

  2. Data merging of infrared and ultrasonic images for plasma facing components inspection

    International Nuclear Information System (INIS)

    Richou, M.; Durocher, A.; Medrano, M.; Martinez-Ona, R.; Moysan, J.; Riccardi, B.

    2009-01-01

    For steady-state magnetic thermonuclear fusion devices which need large power exhaust capability, actively cooled plasma facing components have been developed. In order to guarantee the integrity of these components during the required lifetime, their thermal and mechanical behaviour must be assessed. Before the procurement of the ITER Divertor, the examination of the heat sink to armour joints with non-destructive techniques is an essential topic to be addressed. Defects may be localised at different bonding interfaces. In order to improve the defect detection capability of the SATIR technique, the possibility of merging the infrared thermography test data coming from SATIR results with the ultrasonic test data has been identified. The data merging of SATIR and ultrasonic results has been performed on Carbon Fiber Composite (CFC) monoblocks with calibrated defects, identified by their position and extension. These calibrated defects were realised with machining, with 'stop-off' or by a lack of CFC activation techniques, these last two representing more accurately a real defect. A batch of 56 samples was produced to simulate each possibility of combination with regards to interface location, position and extension and way of realising the defect. The use of a data merging method based on Dempster-Shafer theory improves significantly the detection sensibility and reliability of defect location and size.

  3. Manufacturing study of Be, W and CFC bonded structures for plasma-facing components

    International Nuclear Information System (INIS)

    Onozuka, M.; Hirai, S.; Kikuchi, K.; Oda, Y.; Shimizu, K.

    2004-01-01

    A manufacturing study has been conducted for Be, W, and CFC bonded structures employed in plasma-facing components for the ITER. For Be tiles bonded to the Cu-Cr-Zr alloy heat sink with stainless-steel cooling pipes, a one-axis hot press with two heating processes has been used to bond the three materials. An Al-Si base interlayer has been used to bond Be to the Cu-alloy. The heating processes have been selected to match the required heat treatment conditions for the Cu-alloy. Because of the limited heat processes using a conventional hot press, the manufacturing cost can be minimized. For both the W and CFC tiles, the materials have been brazed at the same time to the Cu-alloy. Ni-Cu-Mn and Cu-Ti brazing materials have been used for the W and CFC tiles, respectively. Using the above bonding techniques, partial mockups of a blanket first-wall panel and divertor target have been successfully manufactured

  4. Manufacturing study of Be, W and CFC bonded structures for plasma-facing components

    Science.gov (United States)

    Onozuka, M.; Hirai, S.; Kikuchi, K.; Oda, Y.; Shimizu, K.

    2004-08-01

    A manufacturing study has been conducted for Be, W, and CFC bonded structures employed in plasma-facing components for the ITER. For Be tiles bonded to the Cu-Cr-Zr alloy heat sink with stainless-steel cooling pipes, a one-axis hot press with two heating processes has been used to bond the three materials. An Al-Si base interlayer has been used to bond Be to the Cu-alloy. The heating processes have been selected to match the required heat treatment conditions for the Cu-alloy. Because of the limited heat processes using a conventional hot press, the manufacturing cost can be minimized. For both the W and CFC tiles, the materials have been brazed at the same time to the Cu-alloy. Ni-Cu-Mn and Cu-Ti brazing materials have been used for the W and CFC tiles, respectively. Using the above bonding techniques, partial mockups of a blanket first-wall panel and divertor target have been successfully manufactured.

  5. Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

    International Nuclear Information System (INIS)

    Maddaluno, G.; Maruccia, G.; Merola, M.; Rollet, S.

    2003-01-01

    The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime

  6. Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

    Science.gov (United States)

    Maddaluno, G.; Maruccia, G.; Merola, M.; Rollet, S.

    2003-03-01

    The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime.

  7. Effects of face/head and whole body cooling during passive heat stress on human somatosensory processing.

    Science.gov (United States)

    Nakata, Hiroki; Namba, Mari; Kakigi, Ryusuke; Shibasaki, Manabu

    2017-06-01

    We herein investigated the effects of face/head and whole body cooling during passive heat stress on human somatosensory processing recorded by somatosensory-evoked potentials (SEPs) at C4' and Fz electrodes. Fourteen healthy subjects received a median nerve stimulation at the left wrist. SEPs were recorded at normothermic baseline (Rest), when esophageal temperature had increased by ~1.2°C (heat stress: HS) during passive heating, face/head cooling during passive heating (face/head cooling: FHC), and after HS (whole body cooling: WBC). The latencies and amplitudes of P14, N20, P25, N35, P45, and N60 at C4' and P14, N18, P22, and N30 at Fz were evaluated. Latency indicated speed of the subcortical and cortical somatosensory processing, while amplitude reflected the strength of neural activity. Blood flow in the internal and common carotid arteries (ICA and CCA, respectively) and psychological comfort were recorded in each session. Increases in esophageal temperature due to HS significantly decreased the amplitude of N60, psychological comfort, and ICA blood flow in the HS session, and also shortened the latencies of SEPs (all, P body temperature. Copyright © 2017 the American Physiological Society.

  8. A saw-tooth plasma actuator for film cooling efficiency enhancement of a shaped hole

    Science.gov (United States)

    Li, Guozhan; Yu, Jianyang; Liu, Huaping; Chen, Fu; Song, Yanping

    2017-08-01

    This paper reports the large eddy simulations of the effects of a saw-tooth plasma actuator and the laidback fan-shaped hole on the film cooling flow characteristics, and the numerical results are compared with a corresponding standard configuration (cylindrical hole without the saw-tooth plasma actuator). For this numerical research, the saw-tooth plasma actuator is installed just downstream of the cooling hole and a phenomenological plasma model is employed to provide the 3D plasma force vectors. The results show that thanks to the downward force and the momentum injection effect of the saw-tooth plasma actuator, the cold jet comes closer to the wall surface and extends further downstream. The saw-tooth plasma actuator also induces a new pair of vortex which weakens the strength of the counter-rotating vortex pair (CRVP) and entrains the coolant towards the wall, and thus the diffusion of the cold jet in the crossflow is suppressed. Furthermore, the laidback fan-shaped hole reduces the vertical jet velocity causing the disappearance of downstream spiral separation node vortices, this compensates for the deficiency of the saw-tooth plasma actuator. Both effects of the laidback fan-shaped hole and the saw-tooth plasma actuator effectively control the development of the CRVP whose size and strength are smaller than those of the anti-counter rotating vortex pair in the far field, thus the centerline and the spanwise-averaged film cooling efficiency are enhanced. The average film cooling efficiency is the biggest in the Fan-Dc = 1 case, which is 80% bigger than that in the Fan-Dc = 0 case and 288% bigger than that in the Cyl-Dc = 0 case.

  9. Estimation of Nuclei Cooling Time by Electrons in Superdense Nonequilibrium Plasma

    CERN Document Server

    Kostenko, B F

    2004-01-01

    Estimations of nuclei cooling time by electrons in superdense nonequilibrium plasma formed at cavitation bubble collapse in deuterated acetone have been carried out. The necessity of these computations was stipulated by using in the latest theoretical calculations of nuclear reaction rate in these processes one poorly grounded assumption that electron temperatures remain essentially lower than nuclei ones during thermonuclear synthesis time t_s. The estimations have shown that the initial electron temperatures at the moment of superdense plasma formation with \\rho =100 g/cm^3 turn out to be appreciably lower than the nuclear temperatures, while the nuclei cooling time is of the same order as t_s.

  10. Development and application of W/Cu flat-type plasma facing components at ASIPP

    International Nuclear Information System (INIS)

    Li, Q; Sun, Z X; Xu, Y; Li, B; Wei, R; Wang, W J; Xie, C Y; Wang, J C; Wang, X L; Yang, Z S; Luo, G-N; Zhao, S X; Qin, S G; Shi, Y L; Liu, G H; Missirlian, M; Guilhem, D

    2017-01-01

    W/Cu flat-type plasma facing components (PFCs) were widely used in divertor of fusion device because of its advantages, such as low cost, light in weight and good machinability. However, it is very difficult to manufacture them due to the large mismatch between the thermo-mechanical properties of W and Cu. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) has successfully developed W/Cu flat-type PFCs for EAST W/Cu divertor project by hot isostatic pressing (HIP) technology. This paper presents the development and application of W/Cu flat-type PFCs at ASIPP. The optimized manufacturing process is to cast pure copper onto the rear side of W tiles at temperature of 1200 °C firstly, and then to HIP the W/Cu tiles onto CuCrZr heat sink at temperature of 600 °C, pressure of 150 MPa and duration of 3 h. W/Cu flat-type testing mock-up for EAST survived 1000 cycles at heat load of 5 MW m −2 in high heat flux tests. And then ASIPP prepared two mock-ups for CEA’s tungsten environment in steady-state tokamak (WEST) project. One mock-up withstood successfully 302 cycles of 20 MW m −2 , which are far beyond the design requirement. Since 2014, W/Cu flat-type PFCs were wildly used in EAST upper divertor as baffle and dome components which showed excellent performance in 2015 and 2016 campaigns. Given the success in EAST upper divertor, W/Cu flat-type concept is as well applied in the design of actively cooled Langmuir probes which will be mounted onto EAST divertor targets soon. (paper)

  11. Development and application of W/Cu flat-type plasma facing components at ASIPP

    Science.gov (United States)

    Li, Q.; Zhao, S. X.; Sun, Z. X.; Xu, Y.; Li, B.; Wei, R.; Wang, W. J.; Qin, S. G.; Shi, Y. L.; Xie, C. Y.; Wang, J. C.; Wang, X. L.; Missirlian, M.; Guilhem, D.; Liu, G. H.; Yang, Z. S.; Luo, G.-N.

    2017-12-01

    W/Cu flat-type plasma facing components (PFCs) were widely used in divertor of fusion device because of its advantages, such as low cost, light in weight and good machinability. However, it is very difficult to manufacture them due to the large mismatch between the thermo-mechanical properties of W and Cu. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) has successfully developed W/Cu flat-type PFCs for EAST W/Cu divertor project by hot isostatic pressing (HIP) technology. This paper presents the development and application of W/Cu flat-type PFCs at ASIPP. The optimized manufacturing process is to cast pure copper onto the rear side of W tiles at temperature of 1200 °C firstly, and then to HIP the W/Cu tiles onto CuCrZr heat sink at temperature of 600 °C, pressure of 150 MPa and duration of 3 h. W/Cu flat-type testing mock-up for EAST survived 1000 cycles at heat load of 5 MW m-2 in high heat flux tests. And then ASIPP prepared two mock-ups for CEA’s tungsten environment in steady-state tokamak (WEST) project. One mock-up withstood successfully 302 cycles of 20 MW m-2, which are far beyond the design requirement. Since 2014, W/Cu flat-type PFCs were wildly used in EAST upper divertor as baffle and dome components which showed excellent performance in 2015 and 2016 campaigns. Given the success in EAST upper divertor, W/Cu flat-type concept is as well applied in the design of actively cooled Langmuir probes which will be mounted onto EAST divertor targets soon.

  12. An operational non destructive examination for ITER divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Durocher, A.; Escourbiac, F.; Farjon, J.L.; Vignal, N.; Cismondi, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [ITER International Team, Cadarache, 13 - St Paul Lez Durance (France); Riccardi, B. [CEFDA CSU-Garching, Garching bei Munchen (Germany)

    2007-07-01

    Full text of publication follows: To meet the power exhaust - heat flux of 20 MW/m{sup 2} - requirements of Plasma Facing Components (PFCs) during plasma operation requires control of their thermal and mechanical integrity. As heat exhaust capability and lifetime of PFCs during in-situ operation are linked to the manufacturing quality, it is an absolute requirement to develop reliable nondestructive examination methods, in particular of the CFC-CuCrZr joint, throughout the manufacturing process. Within the framework of Tokamak Tore Supra upgrade, a pioneering activity has been developed to evaluate the capability of the PFC to be efficiently cooled. In 1998 a test bed - so called SATIR - based on the heat transient method was developed by the CEA and is used today as an inspection tool in order to guarantee the PFCs performances. The technical procurement plan of ITER Divertor targets stated that all Cu cast layers on CFC armour should be subjected to 100% thermographic examination. Each ITER Party should demonstrate its technical capability to carry out the PFC with the required cooling efficiently. The ITER Divertor PFCs pose new challenges especially for the mono-block CFC thickness, and the number of full scale units to be tested which is higher than on any existing or under construction fusion machine. The SATIR method as functional inspection has been identified as the basis test to decide upon the final acceptance of the Divertor PFCs. In order to increase the detection sensitivity of SATIR test bed, several possibilities have been assessed i) the increase of the convective heat transfer coefficient, which improved in a significant way the sensitivity of SATIR diagnostic on ITER components. ii) the installation of a digital infrared camera and the improvement of the thermal signal processing, has led to a considerable increase of performances iii) an innovative process based on spatial image autocorrelation will allow to localize the interlayer defect

  13. Progress on Beam-Plasma Effect Simulations in Muon Ionization Cooling Lattices

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, James [IIT, Chicago; Snopok, Pavel [Fermilab

    2017-05-01

    New computational tools are essential for accurate modeling and simulation of the next generation of muon-based accelerators. One of the crucial physics processes specific to muon accelerators that has not yet been simulated in detail is beam-induced plasma effect in liquid, solid, and gaseous absorbers. We report here on the progress of developing the required simulation tools and applying them to study the properties of plasma and its effects on the beam in muon ionization cooling channels.

  14. Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.

    1996-01-01

    Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig

  15. Face-to-face interaction of multisolitons in spin-1/2 quantum plasma

    Indian Academy of Sciences (India)

    2016-12-13

    Dec 13, 2016 ... tems [14]. When de Broglie wavelength of charge car- riers becomes comparable to the system scales (such as interparticle distances), the quantum effects should be taken into account. In quantum plasma, Fermi–. Dirac distribution is used to describe the system rather than Maxwell–Boltzmann distribution.

  16. Radiative losses and electron cooling rates for carbon and oxygen plasma impurities

    International Nuclear Information System (INIS)

    Marchand, R.; Bonnin, X.

    1992-01-01

    Radiative losses and electron cooling rates are calculated for carbon and oxygen ions under conditions relevant to fusion plasmas. Both rates are calculated with the most recent recommended atomic data. A modified coronal model which includes the effects of metastable states is described and used to calculate the rates. Comparisons with other approaches are also discussed. (author). 36 ref, figs

  17. The plasma facing components of the Tore Supra ICRF antenna

    International Nuclear Information System (INIS)

    Beaumont, B.; Agarici, G.; Gauthier, E.; Kuus, H.; Schlosser, J.

    1994-01-01

    Two generations of Faraday shields for the Tore Supra ICRH antennas interacting with the edge plasma are presented. The last one, using a film of boron carbide as protective material performs well, proving the relevance of this technique for in vessel equipment submitted to low power fluxes. The different lateral protections used on Tore Supra are submitted to high power fluxes. Finite element calculations allow to assess their performances. One type, using Boron Carbide, can be used to measure the local heat flux. The estimation of this flux confirm the specificity of the edge/RF interaction, which is more than one order of magnitude above the exponential decay observed in ohmic plasmas. (author) 11 refs.; 1 fig

  18. Novel Approach to Plasma Facing Materials in Nuclear Fusion Reactors

    International Nuclear Information System (INIS)

    Livramento, V.; Correia, J. B.; Shohoji, N.; Osawa, E.; Nunes, D.; Carvalho, P. A.; Fernandes, H.; Silva, C.; Hanada, K.

    2008-01-01

    A novel material design in nuclear fusion reactors is proposed based on W-nDiamond nanostructured composites. Generally, a microstructure refined to the nanometer scale improves the mechanical strength due to modification of plasticity mechanisms. Moreover, highly specific grain-boundary area raises the number of sites for annihilation of radiation induced defects. However, the low thermal stability of fine-grained and nanostructured materials demands the presence of particles at the grain boundaries that can delay coarsening by a pinning effect. As a result, the concept of a composite is promising in the field of nanostructured materials. The hardness of diamond renders nanodiamond dispersions excellent reinforcing and stabilization candidates and, in addition, diamond has extremely high thermal conductivity. Consequently, W-nDiamond nanocomposites are promising candidates for thermally stable first-wall materials. The proposed design involves the production of W/W-nDiamond/W-Cu/Cu layered castellations. The W, W-nDiamond and W-Cu layers are produced by mechanical alloying followed by a consolidation route that combines hot rolling with spark plasma sintering (SPS). Layer welding is achieved by spark plasma sintering. The present work describes the mechanical alloying processsing and consolidation route used to produce W-nDiamond composites, as well as microstructural features and mechanical properties of the material produced Long term plasma exposure experiments are planned at ISTTOK and at FTU (Frascati)

  19. Thermal shock problems of bonded structure for plasma facing components

    International Nuclear Information System (INIS)

    Shibui, M.; Kuroda, T.; Kubota, Y.

    1991-01-01

    Thermal shock tests have been performed on W(Re)/Cu and Mo/Cu duplex structures with a particular emphasis on two failure modes: failure on the heated surface and failure near the bonding interface. The results indicate that failure of the duplex structure largely depends on the constraint of thermal strain on the heated surface and on the ductility changes of armour materials. Rapid debonding of the bonding interface may be attributed to the yielding of armour materials. This leads to a residual bending deformation when the armour cools down. Arguments are also presented in this paper on two parameter characterization of the failure of armour materials and on stress distribution near the free edge of the bonding interface. (orig.)

  20. An assessment of the tritium inventory in, permeation through and releases from the NET plasma facing materials

    International Nuclear Information System (INIS)

    Wu, C.H.

    1986-01-01

    The tritium retention, permeation and release characteristics of D-T tokamaks are extremely important from both an environmental and a plasma physics point of view. Tokamak measurements have demonstrated that release of retained hydrogen isotopes by plasma-wall interactions play a dominant role in fuel recycling during a discharge. In addition, retained tritium in the plasma facing materials may contribute substantially to the on-site tritium inventory of D-T devices. Austenitic and martensitic steels are being considered as first wall materials. Tungsten and molybdenum will be possibly used as divertor armour materials for NET. By using a computer code, the tritium inventory in, permeation through and release from these materials have been calculated as functions of material thickness, temperature and impinging fluxes. It is shown that the tritium inventory in the first wall will be strongly affected by the temperature gradient in the materials. It is evident, that the tritium permeation as well as the tritium inventory can be reduced appropriately by controlling the temperatures at the plasma and cooling sides of the first wall. The results are discussed and the possible consequences are analysed. (author)

  1. Empirical evaluation of the radiative cooling coefficient for krypton gas in the FTU plasma

    International Nuclear Information System (INIS)

    Fournier, K.B.; Pacella, D.; Mazzitelli, G.; Stutman, D.; Soukanovskii, V.; Goldstein, W.H.

    1997-01-01

    For future fusion reactors, a careful balance must be achieved between the cooling of the outer plasma via impurity radiation and the deleterious effects of inevitable core penetration by impurity ions. We have injected krypton gas into the Frascati Tokamak Upgrade (FTU) plasma. The measured visible bremsstrahlung and bolometric signals from krypton have been inverted and the resulting radial impurity density profile and power loss profile for krypton gas are extracted. Using the measured electron density and temperature profiles, the radiative cooling coefficient for krypton is derived. The level of intrinsic impurities (Mo, Cr, Mn and Fe) in the plasma during the krypton puffing is monitored with a VUV SPRED spectrometer. Models for krypton emissivity from the literature are compared to our measured results. 7 figs

  2. Radiation cooling and gain calculation for C VI 182 A line in C/Se plasma

    International Nuclear Information System (INIS)

    Nam, C.H.; Valeo, E.; Suckewer, S.; Feldman, U.

    1986-04-01

    A model is developed which is capable of describing the evolution of gain resulting from both rapid radiative and expansion cooling of a recombining, freely expanding plasma. It is demonstrated for the particular case of a carbon/selenium plasma that the cooling rate which leads to optimal gain can be achieved by adjusting the admixture of an efficiently radiating material (selenium) in the gain medium (carbon). Comparison is made to a recent observation of gain in a recent NRL/Rochester experiment with carbon/selenium plasma for the n = 3 → 2 transition in C VI occurring at 182 A. The predicted maximum gain is approx.10 cm -1 , as compared to observation of 2 to 3 cm -1

  3. Laser cooled ion beams and strongly coupled plasmas for precision experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bussmann, Michael

    2008-03-17

    This cumulative thesis summarizes experimental and theoretical results on cooling of ion beams using single-frequency, single-mode tabletop laser systems. It consists of two parts. One deals with experiments on laser-cooling of ion beams at relativistic energies, the other with simulations of stopping and sympathetic cooling of ions for precision in-trap experiments. In the first part, experimental results are presented on laser-cooling of relativistic C{sup 3+} ion beams at a beam energy of 122 MeV/u, performed at the Experimental Storage Ring (ESR) at GSI. The main results presented in this thesis include the first attainment of longitudinally space-charge dominated relativistic ion beams using pure laser-cooling. The second part lists theoretical results on stopping and sympathetic cooling of ions in a laser-cooled one-component plasma of singly charged {sup 24}Mg ions, which are confined in a three-dimensional harmonic trap potential. (orig.)

  4. Mixed plasma-facing materials research at INEEL

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.

    2001-01-01

    Mixed-materials research at the Idaho National Engineering and Environmental Laboratory (INEEL) has focused on Be-C and W-C systems. The purpose of this work was to investigate hydrogen isotope retention in these systems. Plasma-mixed material layers using carbon coated Be and W specimens that were heat-treated and tungsten carbide specimens prepared by chemical vapor deposition (CVD) were simulated. Hydrogen isotope retention was investigated by means of thermal desorption spectroscopy (TDS) measurements on deuterium implanted samples

  5. Modeling tritium processes in plasma-facing beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Dolan, T.J.; Mulock, M.J.

    1995-01-01

    In this paper we present techniques and recommended parameters for modeling tritium implantation, trapping and release, and permeation, in beryllium-clad structures adjacent to the plasma. Among the features that should be considered are the effects of surface films, the mobility of beryllium through those films, damage caused by ion implantation, especially in regions where pitting may be expected, and bubble formation. Tritium transport parameters recommended are based on fits with experimental data and available theory. Estimates of inventories in ITER using these parameters are also given. 31 refs., 2 figs., 1 tab

  6. Behavior of plasma facing surface in the large helical device

    International Nuclear Information System (INIS)

    Hino, T.; Nobuta, Y.; Sagara, A.

    2002-01-01

    Material probes have been installed at the inner walls along poloidal direction in LHD from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was considerably cleaned by helium glow discharge conditionings. For the 3rd and 4th campaigns, graphite tiles were installed at entire divertor strike region, and then the wall condition significantly changed compared to the case of stainless steel wall. The erosion of graphite took place during the main discharges and the eroded carbon deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, amount of retained discharge gas such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristics of the LHD wall is a large retention of helium gas since the wall temperature is limited below 368 K. In order to reduce the recycling of discharge gas, the wall heating before the experimental campaign and the surface heating between the main discharge shots are planned. (author)

  7. Behavior of plasma facing surface in the large helical device

    Energy Technology Data Exchange (ETDEWEB)

    Hino, T.; Nobuta, Y. [Hokkaido Univ., Dept. of Nuclear Engineering, Sapporo, Hokkaido (Japan); Sagara, A. [National Inst. for Fusion Science, Toki, Gifu (Japan)] [and others

    2002-11-01

    Material probes have been installed at the inner walls along poloidal direction in LHD from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was considerably cleaned by helium glow discharge conditionings. For the 3rd and 4th campaigns, graphite tiles were installed at entire divertor strike region, and then the wall condition significantly changed compared to the case of stainless steel wall. The erosion of graphite took place during the main discharges and the eroded carbon deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, amount of retained discharge gas such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristics of the LHD wall is a large retention of helium gas since the wall temperature is limited below 368 K. In order to reduce the recycling of discharge gas, the wall heating before the experimental campaign and the surface heating between the main discharge shots are planned. (author)

  8. Behavior of plasma facing surfaces in the large helical device

    International Nuclear Information System (INIS)

    Hino, T.; Nobuta, Y.; Sagara, A.

    2003-01-01

    Material probes have been installed at the inner walls along the poloidal direction in LHD from the first experimental campaign. After each campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was thoroughly cleaned by helium glow discharge conditioning. For the 3rd and 4th campaigns, graphite tiles were installed over the entire divertor strike region, and then the wall condition was significantly changed compared to the case of a stainless steel wall. Graphite erosion took place during the main discharges and the eroded carbon was deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, the amount of retained discharge gases such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristic of the LHD wall is a large retention of helium gas since the wall temperature is limited to below 368 K. In order to reduce the recycling of discharge gas, wall heating before the experimental campaign and surface heating between the main discharge shots are planned. (author)

  9. Behavior of plasma facing surface in the large helical device

    International Nuclear Information System (INIS)

    Hino, T.; Nobuta, Y.; Sagara, A.

    2002-10-01

    Material probes have been installed at the inner walls along poloidal direction in LHD from the first experimental campaign. After each the campaign, the impurity deposition and the gas retention have been examined to clarify the plasma surface interaction and the degree of wall cleaning. In the 2nd campaign, the entire wall was considerably cleaned by helium glow discharge conditionings. For the 3rd and 4th campaigns, graphite tiles were installed at entire divertor strike region, and then the wall condition significantly changed compared to the case of stainless steel wall. The erosion of graphite took place during the main discharges and the eroded carbon deposited on the entire wall. In particular, the deposition thickness was large at the wall far from the plasma. Since the entire wall was well carbonized, amount of retained discharge gas such as H and He became large. In particular, the helium retention was large at the position close to the anodes used for helium glow discharge cleanings. One characteristics of the LHD wall is a large retention of helium gas since the wall temperature is limited below 368 K. In order to reduce the recycling of discharge gas, the wall heating before the experimental campaign and the surface heating between the main discharge shots are planned. (author)

  10. Vacuum Plasma Spraying W-coated Reduced Activation Structural Steels for Fusion Plasma Facing Components

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Sanghoon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Tungsten (W) and its alloys are considered as candidate materials for plasma facing materials of the first wall and diverter components in fusion reactor systems because of high sputtering resistance and low tritium retention in a fusion environment. Therefore, it is considered that the joining between W and reduced activation structural steels, and its evaluation, are critical issues for the development of fusion reactors. However, the joining between these materials is a very challenging process because of significant differences in their physical properties, particularly the mismatch of coefficients of thermal expansion (CTE). For instance, the CTE of pure W is known to be about 4.3Χ10{sup -6}K{sup -1}; however, that of martensitic steels reaches over three times, about 12-14Χ10{sup -6}K{sup -1} at room temperature even up to 373K. Nevertheless, several joining techniques have been developed for joining between W and structural steels, such as a vapor deposition method, brazing and diffusion bonding. Meanwhile, vacuum plasma spraying (VPS) is supposed to be one of the prospective methods to fabricate a sufficient W layer on the steel substrates because of the coating of a large area with a relatively high fabricating rate. In this study, the VPS method of W powders on reduced activation steels was employed, and its microstructure and hardness distribution were investigated. ODS ferritic steels and F82H steel were coated by VPS-W, and the microstructure and hardness distribution were investigated. A microstructure analysis revealed that pure W was successfully coated on steel substrates by the VPS process without an intermediate layer, in spite of a mismatch of the CTE between dissimilar materials. After neutron irradiation, irradiation hardening significantly occurred in the VPSW. However, the hardening of VPS-W was lesser than that of bulk W irradiated HFIR at 773K. Substrate materials, ODS ferritic steels, and F82H steel, did not show irradiation hardening

  11. Hot ion plasma production in HIP-1 using water-cooled hollow cathodes

    Science.gov (United States)

    Reinmann, J. J.; Lauver, M. R.; Patch, R. W.; Layman, R. W.; Snyder, A.

    1975-01-01

    A steady-state ExB plasma was formed by applying a strong radially inward dc electric field near the mirror throats. Most of the results were for hydrogen, but deuterium and helium plasmas were also studied. Three water-cooled hollow cathodes were operated in the hot-ion plasma mode with the following results: (1) thermally emitting cathodes were not required to achieve the hot-ion mode; (2) steady-state operation (several minutes) was attained; (3) input powers greater than 40 kW were achieved; (4) cathode outside diameters were increased from 1.2 cm (uncooled) to 4.4 cm (water-cooled); (5) steady-state hydrogen plasma with ion temperatures from 185 to 770 eV and electron temperatures from 5 to 21 eV were produced. Scaling relations were empirically obtained for discharge current, ion temperature, electron temperature, and relative ion density as a function of hydrogen gas feed rate, magnetic field, and cathode voltage. Neutrons were produced from deuterium plasma, but it was not established whether thay came from the plasma volume or from the electrode surfaces.

  12. Concept of a staged FEL enabled by fast synchrotron radiation cooling of laser-plasma accelerated beam by solenoidal magnetic fields in plasma bubble

    Science.gov (United States)

    Seryi, Andrei; Lesz, Zsolt; Andreev, Alexander; Konoplev, Ivan

    2017-03-01

    A novel method for generating GigaGauss solenoidal fields in a laser-plasma bubble, using screw-shaped laser pulses, has been recently presented. Such magnetic fields enable fast synchrotron radiation cooling of the beam emittance of laser-plasma accelerated leptons. This recent finding opens a novel approach for design of laser-plasma FELs or colliders, where the acceleration stages are interleaved with laser-plasma emittance cooling stages. In this concept paper, we present an outline of what a staged plasma-acceleration FEL could look like, and discuss further studies needed to investigate the feasibility of the concept in detail.

  13. IAEA consultants' meeting on thermal response of plasma facing materials and components

    International Nuclear Information System (INIS)

    Janev, R.K.

    1990-07-01

    The present Summary Report contains brief proceedings and the main conclusions and recommendations of the IAEA Consultants' Meeting on ''Thermal Response of Plasma Facing Materials and Components'', which was organized by the IAEA Atomic and Molecular Data Unit and held on June 11-13, 1990, in Vienna, Austria. The Report also includes a categorization and assessment of currently studied plasma facing materials, a classification scheme of material properties data, required in fusion reactor design, and a survey of the urgently needed material properties data. (author)

  14. A fatigue lifetime assessment of WEST ITER Like Plasma Facing Unit

    International Nuclear Information System (INIS)

    Languille, P.; Missirlian, M.; Guilhem, D.; Ferlay, F.; Batal, T.; Bucalossi, J.; Firdaouss, M.; Larroque, S.; Martinez, A.; Richou, M.

    2016-01-01

    Highlights: • ITER plasma facing component divertor technology is integrated in WEST. • ITER Like attachments in WEST has been optimised. • The ITER Like PFU is compatible with a wide range of plasma scenarios. - Abstract: Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m 2 . The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

  15. New electron beam facility for irradiated plasma facing materials testing in hot cell

    International Nuclear Information System (INIS)

    Sakamoto, N.; Kawamura, H.; Akiba, M.

    1995-01-01

    Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop of plasma facing components which can resist these. Then, we have established electron beam heat facility (open-quotes OHBISclose quotes, Oarai Hot-cell electron Beam Irradiating System) at a hot cell in JMTR (Japan Materials Testing Reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30kV (constant) and 1.7A, respectively. The loading time of electron beam is more than 0.1ms. The shape of vacuum vessel is cylindrical, and the mainly dimensions are 500mm in inner diameter, 1000mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for thermal shock test has been established in a hot cell. And performance estimation on the electron beam is being conducted. Presently, the devices for heat loading tests under steady state will be added to this facility

  16. A fatigue lifetime assessment of WEST ITER Like Plasma Facing Unit

    Energy Technology Data Exchange (ETDEWEB)

    Languille, P., E-mail: pascal.languille@gmail.com; Missirlian, M.; Guilhem, D.; Ferlay, F.; Batal, T.; Bucalossi, J.; Firdaouss, M.; Larroque, S.; Martinez, A.; Richou, M.

    2016-11-01

    Highlights: • ITER plasma facing component divertor technology is integrated in WEST. • ITER Like attachments in WEST has been optimised. • The ITER Like PFU is compatible with a wide range of plasma scenarios. - Abstract: Based on a monoblock concept (e.g. a tube-in-tile concept), each elementary tungsten plasma facing component (called Plasma-Facing Unit PFU) of the WEST lower divertor follows as closely as possible the same monoblock geometry, materials and bonding technology that is envisaged for ITER. A fatigue simulation of W PFU was used to validate its specific integration into WEST. The complex design, the material heterogeneities and the usage outside operational load design envelope are all possible causes of fatigue failure. This paper shows how the ITER like monoblocks and its U-shaped attachments technology are integrated into the WEST divertor by performing finite element analysis. The WEST lower divertor is designed to withstand 15 MW steady-state of injected power, with peaked heat fluxes up to 20 MW/m{sup 2}. The integration and the design choices of a WEST ITER Like Plasma Facing Unit inside the WEST vacuum chamber is valid for an “expected life time” of repeated inter ELMs thermal steady state (>10 s) cycles and for 300 off-normal vertical displacement events.

  17. New electron beam facility for irradiated plasma facing materials testing in hot cell

    International Nuclear Information System (INIS)

    Shimakawa, S.; Akiba, M.; Kawamura, H.

    1996-01-01

    Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop plasma facing components which can resist these. We have established electron beam heat facility ('OHBIS', Oarai hot-cell electron beam irradiating system) at a hot cell in JMTR (Japan materials testing reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50 kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30 kV (constant) and 1.7 A, respectively. The loading time of the electron beam is more than 0.1 ms. The shape of vacuum vessel is cylindrical, and the main dimensions are 500 mm in inside diameter, 1000 mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for the thermal shock test has been established in a hot cell. The performance of the electron beam is being evaluated at this time. In the future, the equipment for conducting static heat loadings will be incorporated into the facility. (orig.)

  18. Self-limitation of impurity production by radiation cooling at the edge of a fusion plasma

    International Nuclear Information System (INIS)

    Neuhauser, J.; Lackner, K.; Wunderlich, R.

    1982-04-01

    The influence of radiation cooling at the edge of a fusion plasma on the plasma-wall interaction is numerically studied for parameters typical of the ZEPHYR ignition experiment. Various transport and impurity influx models and different external heating methods are studied using the 1D tokamak transport code BALDUR developed at Princeton. The results demonstrate the self-consistent formation of a radiating boundary layer (photosphere) for a wide range of parameters, limiting the impurity concentration in the plasma to a tolerable value. While the plasma behaviour is rather insensitive to model assumptions, the sputtering rate and the corresponding wall erosion depend on various parameters. Methods for external control of the photosphere and - more important - of the wall erosion are also discussed. (orig.)

  19. On thermionic emission from plasma-facing components in tokamak-relevant conditions.

    Czech Academy of Sciences Publication Activity Database

    Komm, Michael; Ratynskaia, S.; Tolias, P.; Cavalier, Jordan; Dejarnac, Renaud; Gunn, J. P.; Podolník, Aleš

    2017-01-01

    Roč. 59, č. 9 (2017), č. článku 094002. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : thermionic * PIC * tungsten * tokamak * thermionic emission * plasma facing components * particle-in-cell Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/1361-6587/aa78c4/pdf

  20. Engineering solutions for components facing the plasma in experimental power reactors

    International Nuclear Information System (INIS)

    Casini, G.; Farfaletti-Casali, F.

    1985-01-01

    A review of the engineering problems related to the structures in front of the plasma of experimental Tokamak-type reactors is made. Attention is focused on the so-named ''first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system, in particular for the case of the single-null poloidal divertor. Even if the uncertainties related to the plasma-wall interaction are stil relevant, some engineering solutions which look manageable are identified and described. (orig.)

  1. An EDDY/particle-in-cell simulation of erosion of plasma facing walls bombarded by a collisional plasma

    International Nuclear Information System (INIS)

    Inai, Kensuke; Ohya, Kaoru

    2011-01-01

    To investigate the erosion of a plasma-facing wall intersecting an oblique magnetic field, we performed a kinetic particle-in-cell (PIC) simulation of magnetized plasma, in which collision processes between charged and neutral particles were taken into account. Sheath formation and local physical quantities, such as the incident angle and energy distributions of plasma ions at the wall, were examined at a plasma density of 10 18 m -3 , a temperature of 10 eV, and a magnetic field strength of 5 T. The erosion rate of a carbon wall was calculated using the ion-solid interaction code EDDY. At a high neutral density (>10 20 m -3 ), the impact energy of the ions dropped below the threshold for physical sputtering, so that the sputtering yield was drastically decreased and wall erosion was strongly suppressed. Sputter erosion was also suppressed when the angle of the magnetic field with respect to the surface normal was sufficiently large. (author)

  2. Artificial cooling due to quiet injection in bounded plasma particle simulations

    International Nuclear Information System (INIS)

    Lawson, W.S.

    1988-01-01

    An explanation is proposed for an artificial cooling effect seen in electrostatic particle-in-cell plasma simulations. The effect hinges on heat transport from the trapped electrons to fluctuations of the electric field, which are kept at a sub-thermal level through the continuous ''quiet'' injection of passing electrons. Further simulations are done which test and support the explanation. copyright 1988 Academic Press, Inc

  3. Short-Lived Electronically-Excited Diatomic Molecules Cooled via Supersonic Expansion from a Plasma Microjet

    Science.gov (United States)

    Houlahan, Thomas J., Jr.; Su, Rui; Eden, Gary

    2014-06-01

    Using a pulsed plasma microjet to generate short-lived, electronically-excited diatomic molecules, and subsequently ejecting them into vacuum to cool via supersonic expansion, we are able to monitor the cooling of molecules having radiative lifetimes as low as 16 ns. Specifically, we report on the rotational cooling of He_2 molecules in the d^3Σ_u^+, e^3Π_g, and f^3Σ_u^+ states, which have lifetimes of 25 ns, 67 ns, and 16 ns, respectively. The plasma microjet is driven with a 2.6 kV, 140 ns high-voltage pulse (risetime of 20 ns) which, when combined with a high-speed optical imaging system, allows the nonequilibrium rotational distribution for these molecular states to be monitored as they cool from 1200 K to below 250 K with spatial and temporal resolutions of below 10 μm and 10 ns, respectively. The spatial and temporal resolution afforded by this system also allows the observation of excitation transfer between the f^3Σ_u^+ state and the lower lying d^3Σ_u^+ and e^3Π_g states. The extension of this method to other electronically excited diatomics with excitation energies >5 eV will also be discussed.

  4. Faces

    DEFF Research Database (Denmark)

    Mortensen, Kristine Køhler; Brotherton, Chloe

    2018-01-01

    for the face the be put into action. Based on an ethnographic study of Danish teenagers’ use of SnapChat we demonstrate how the face is used as a central medium for interaction with peers. Through the analysis of visual SnapChat messages we investigate how SnapChat requires the sender to put an ‘ugly’ face...... already secured their popular status on the heterosexual marketplace in the broad context of the school. Thus SnapChat functions both as a challenge to beauty norms of ‘flawless faces’ and as a reinscription of these same norms by further manifesting the exclusive status of the popular girl...

  5. TOWARD TUNGSTEN PLASMA-FACING COMPONENTS IN KSTAR: RESEARCH ON PLASMA-METAL WALL INTERACTION

    Czech Academy of Sciences Publication Activity Database

    Hong, S.-H.; Kim, K.M.; Song, J.-H.; Bang, E.-N.; Kim, H.-T.; Lee, K.-S.; Litnovsky, A.; Hellwig, M.; Seo, D.C.; Lee, H.H.; Kang, C.S.; Lee, H.-Y.; Hong, J.-H.; Bak, J.-G.; Kim, H.-S.; Juhn, J.-W.; Son, S.-H.; Kim, H.-K.; Douai, D.; Grisolia, C.; Wu, J.; Luo, G.-N.; Choe, W.-H.; Komm, Michael; van den Berg, M.; De Temmerman, G.; Pitts, R.

    2015-01-01

    Roč. 68, č. 1 (2015), s. 36-43 ISSN 1536-1055. [International Conference on Open Magnetic Systems for Plasma Confinement (OS 2014)/10./. Daejeon, 26.08.2014-29.08.2014] Institutional support: RVO:61389021 Keywords : Plasma-metal wall interaction * Tungsten technology Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.799, year: 2015 http://dx.doi.org/10.13182/FST14-897

  6. Beryllium assessment and recommendation for application in ITER plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V.; Tanaka, S.; Matera, R. [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    The design status of the ITER Plasma Facing Components (PFC) is presented. The operational conditions of the armour material for the different components are summarized. Beryllium is the reference armour material for the Primary Wall, Baffle and Limiter and the back-up material for the Divertor Dome. The activities on the selection of the Be grades and the joining technologies are reviewed. (author)

  7. The feasibility of beryllium as structural material for the ITER plasma-facing components (PFC)

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Gorenflo, H.

    1993-01-01

    Be as plasma-facing armour has attractive features including excellent plasma compatibility, no T-retention via co-deposition and the potential for in-situ repair via plasma spraying. In order to avoid the bonding of the Be-armour to a heatsink structure in e.g., Cu-alloys, the ITER Joint Central Team (JCT) proposed for the divertor tubular elements with monolithic Be, both as plasma-facing and structural material. The analysis of these Be-tubes with 5 mm wall thickness at a heat load of 5 MW/m 2 showed that even for the most favourable assumptions thermal stresses exceed by far the allowed values according to design codes. Damage by neutrons and disruptions would worsen further the case for Be as monolithic plasma-facing and structural material. For PFC at heat flux significantly above 1 MW/m 2 it appears evident that Be should be used merely as armour bonded to a suitable structural material as heatsink. (orig.)

  8. Data combination of infrared thermography images and lock-in thermography images for NDE of plasma facing components

    International Nuclear Information System (INIS)

    Moysan, J.; Gueudre, C.; Corneloup, G.; Durocher, A.

    2006-01-01

    A pioneering activity has been developed by CEA and the European industry in the field of actively cooled high heat flux plasma facing components (PFC) from the very beginning of Tore Supra project. These components have been developed in order to enable a large power exhaust capability. The goal of this study is to improve the Non Destructive Evaluation (NDE) of these components. The difficulty encountered is the evaluation of the junction between a carbon and a metallic substrate. This was even more difficult when complex designs have to be implemented. A first NDE solution was based on the so called SATIR test. The method is based on infrared measurements of tile surface temperatures during a thermal transient produced by hot/cold water flowing in the heat sink cooling channel. In order to improve the definition of acceptance rules for the PFCs, a second NDE method based on Lock-in Thermography is developed. In this work we present how we can combine the two resulting images in order to accept or to reject a component. This prospective study allows improving the experimental setup and the definition of acceptance criteria. The experimental study was conducted on trial components for the Wendelstein 7X stellarator. The conclusions will also influence future non destructive projects dedicated to the ITER project. (orig.)

  9. Observation of plasma-facing-wall via high dynamic range imaging

    International Nuclear Information System (INIS)

    Villamayor, Michelle Marie S.; Rosario, Leo Mendel D.; Viloan, Rommel Paulo B.

    2013-01-01

    Pictures of plasmas and deposits in a discharge chamber taken by varying shutter speeds have been integrated into high dynamic range (HDR) images. The HDR images of a graphite target surface of a compact planar magnetron (CPM) discharge device have clearly indicated the erosion pattern of the target, which are correlated to the light intensity distribution of plasma during operation. Based upon the HDR image technique coupled to colorimetry, a formation history of dust-like deposits inside of the CPM chamber has been recorded. The obtained HDR images have shown how the patterns of deposits changed in accordance with discharge duration. Results show that deposition takes place near the evacuation ports during the early stage of the plasma discharge. Discoloration of the plasma-facing-walls indicating erosion and redeposition eventually spreads at the periphery after several hours of operation. (author)

  10. Plasma-wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

    Science.gov (United States)

    Brezinsek, S.; Coenen, J. W.; Schwarz-Selinger, T.; Schmid, K.; Kirschner, A.; Hakola, A.; Tabares, F. L.; van der Meiden, H. J.; Mayoral, M.-L.; Reinhart, M.; Tsitrone, E.; Ahlgren, T.; Aints, M.; Airila, M.; Almaviva, S.; Alves, E.; Angot, T.; Anita, V.; Arredondo Parra, R.; Aumayr, F.; Balden, M.; Bauer, J.; Ben Yaala, M.; Berger, B. M.; Bisson, R.; Björkas, C.; Bogdanovic Radovic, I.; Borodin, D.; Bucalossi, J.; Butikova, J.; Butoi, B.; Čadež, I.; Caniello, R.; Caneve, L.; Cartry, G.; Catarino, N.; Čekada, M.; Ciraolo, G.; Ciupinski, L.; Colao, F.; Corre, Y.; Costin, C.; Craciunescu, T.; Cremona, A.; De Angeli, M.; de Castro, A.; Dejarnac, R.; Dellasega, D.; Dinca, P.; Dittmar, T.; Dobrea, C.; Hansen, P.; Drenik, A.; Eich, T.; Elgeti, S.; Falie, D.; Fedorczak, N.; Ferro, Y.; Fornal, T.; Fortuna-Zalesna, E.; Gao, L.; Gasior, P.; Gherendi, M.; Ghezzi, F.; Gosar, Ž.; Greuner, H.; Grigore, E.; Grisolia, C.; Groth, M.; Gruca, M.; Grzonka, J.; Gunn, J. P.; Hassouni, K.; Heinola, K.; Höschen, T.; Huber, S.; Jacob, W.; Jepu, I.; Jiang, X.; Jogi, I.; Kaiser, A.; Karhunen, J.; Kelemen, M.; Köppen, M.; Koslowski, H. R.; Kreter, A.; Kubkowska, M.; Laan, M.; Laguardia, L.; Lahtinen, A.; Lasa, A.; Lazic, V.; Lemahieu, N.; Likonen, J.; Linke, J.; Litnovsky, A.; Linsmeier, Ch.; Loewenhoff, T.; Lungu, C.; Lungu, M.; Maddaluno, G.; Maier, H.; Makkonen, T.; Manhard, A.; Marandet, Y.; Markelj, S.; Marot, L.; Martin, C.; Martin-Rojo, A. B.; Martynova, Y.; Mateus, R.; Matveev, D.; Mayer, M.; Meisl, G.; Mellet, N.; Michau, A.; Miettunen, J.; Möller, S.; Morgan, T. W.; Mougenot, J.; Mozetič, M.; Nemanič, V.; Neu, R.; Nordlund, K.; Oberkofler, M.; Oyarzabal, E.; Panjan, M.; Pardanaud, C.; Paris, P.; Passoni, M.; Pegourie, B.; Pelicon, P.; Petersson, P.; Piip, K.; Pintsuk, G.; Pompilian, G. O.; Popa, G.; Porosnicu, C.; Primc, G.; Probst, M.; Räisänen, J.; Rasinski, M.; Ratynskaia, S.; Reiser, D.; Ricci, D.; Richou, M.; Riesch, J.; Riva, G.; Rosinski, M.; Roubin, P.; Rubel, M.; Ruset, C.; Safi, E.; Sergienko, G.; Siketic, Z.; Sima, A.; Spilker, B.; Stadlmayr, R.; Steudel, I.; Ström, P.; Tadic, T.; Tafalla, D.; Tale, I.; Terentyev, D.; Terra, A.; Tiron, V.; Tiseanu, I.; Tolias, P.; Tskhakaya, D.; Uccello, A.; Unterberg, B.; Uytdenhoven, I.; Vassallo, E.; Vavpetič, P.; Veis, P.; Velicu, I. L.; Vernimmen, J. W. M.; Voitkans, A.; von Toussaint, U.; Weckmann, A.; Wirtz, M.; Založnik, A.; Zaplotnik, R.; PFC contributors, WP

    2017-11-01

    The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W

  11. Fabrication of gas turbine water-cooled composite nozzle and bucket hardware employing plasma spray process

    Science.gov (United States)

    Schilke, Peter W.; Muth, Myron C.; Schilling, William F.; Rairden, III, John R.

    1983-01-01

    In the method for fabrication of water-cooled composite nozzle and bucket hardware for high temperature gas turbines, a high thermal conductivity copper alloy is applied, employing a high velocity/low pressure (HV/LP) plasma arc spraying process, to an assembly comprising a structural framework of copper alloy or a nickel-based super alloy, or combination of the two, and overlying cooling tubes. The copper alloy is plamsa sprayed to a coating thickness sufficient to completely cover the cooling tubes, and to allow for machining back of the copper alloy to create a smooth surface having a thickness of from 0.010 inch (0.254 mm) to 0.150 inch (3.18 mm) or more. The layer of copper applied by the plasma spraying has no continuous porosity, and advantageously may readily be employed to sustain a pressure differential during hot isostatic pressing (HIP) bonding of the overall structure to enhance bonding by solid state diffusion between the component parts of the structure.

  12. Performance of plasma facing materials under intense thermal loads in tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Hirai, T.; Roedig, M.; Singheiser, L. [Forschungszentrum Juelich GmbH, EURATOM Association, Juelich (Germany)

    2003-07-01

    Beside quasi-stationary plasma operation, short transient thermal pulses with deposited energy densities in the order of several ten MJm{sup -2} are a serious concern for next step devices, in particular for tokamak devices such as ITER. The most serious of these transient events are plasma disruptions. Here a considerable fraction of the plasma energy is deposited on a localized surface area in the divertor strike zone region; the time scale of these events is typically in the order of 1 ms. In spite of the fact that a dense cloud of ablation vapour will form above the strike zone, only partial shielding of the divertor armour from incident plasma particles will occur. As a consequence, thermal shock induced crack formation, vaporization, surface melting, melt layer ejection, and particle emission induced by brittle destruction processes will limit the lifetime of the components. In addition, dust particles (neutron activated metals or tritium enriched carbon) are a serious concern form a safety point of view. Other transient heat loads which occasionally occur in magnetic confinement experiments such as instabilities in the plasma positioning (vertical displacement events) also may cause irreversible damage to plasma facing components (PFC), particularly to metals such as beryllium and tungsten. Another serious damage to PFCs is due to intense fluxes of 14 MeV neutrons in D-T-burning plasma devices. Integrated neutron fluence of several ten dpa in future thermonuclear fusion reactors will degrade essential physical properties of the components (e.g. thermal conductivity); another serious concern is the embrittlement of the heat sink and the plasma facing materials (PFM). (orig.)

  13. Thermoelectric-Driven Liquid-Metal Plasma-Facing Structures (TELS) Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Ruzic, David [Univ. of Illinois, Urbana-Champaign, IL (United States)

    2016-12-17

    The Thermoelectric-Driven Liquid-Metal Plasma-Facing Structures (TELS) project was able to establish the experimental conditions necessary for flowing liquid metal surfaces in order to be utilized as surfaces facing fusion relevant energetic plasma flux. The work has also addressed additional developments along with progressing along the timeline detailed in the proposal. A no-cost extension was requested to conduct other relevant experiment- specifically regarding the characterization droplet ejection during energetic plasma flux impact. A specially designed trench module, which could accommodate trenches with different aspect ratios was fabricated and installed in the TELS setup and plasma gun experiments were performed. Droplet ejection was characterized using high speed image acquisition and also surface mounted probes were used to characterize the plasma. The Gantt chart below had been provided with the original proposal, indicating the tasks to be performed in the third year of funding. These tasks are listed above in the progress report outline, and their progress status is detailed below.

  14. Behavior of liquid Li-Sn alloy as plasma facing material on ISTTOK

    Energy Technology Data Exchange (ETDEWEB)

    Loureiro, J.P.S., E-mail: jpsloureiro@ipfn.tecnico.ulisboa.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Tabarés, F.L. [Laboratorio Nacional de Fusion, Ciemat, Avenida Complutense 22, E-28040 Madrid (Spain); Fernandes, H.; Silva, C.; Gomes, R.; Alves, E.; Mateus, R.; Pereira, T.; Alves, H.; Figueiredo, H. [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal)

    2017-04-15

    The high power loads impinging on the first wall and particularly the divertor of fusion reactors is a decisive factor to the success of nuclear fusion. An alternative to solid plasma facing components is the use of liquid metals such as lithium or tin due to the regenerative properties of the liquid surface. Another suitable candidate is the eutectic lithium tin alloy (30 at.% Li) which is suggested to display beneficial properties of both its constituent elements. The application of these materials as liquid metal plasma facing components depends on several factors such as their affinity to retain hydrogenic isotopes and the discharge performance degradation induced by the enhanced impurity contamination, among others. An experimental setup has been developed to produce and expose samples to ISTTOK plasmas on both liquid and solid states. Samples of Li-Sn alloy were exposed at ISTTOK to deuterium plasmas. Post-mortem analysis of the samples was performed by means of ion beam diagnostics. To quantify the fuel retention on the samples the nuclear reaction analysis (NRA) technique was applied. Complementary, Rutherford backscattering spectrometry (RBS) was used for determination material composition, particularly of impurities, on the samples. Regardless of the high sensitivity of these techniques no deuterium was detected in the samples. Emission of the Li-I 670.7 nm line indicates that there was interaction of the plasma with the samples. Alternative reasons for the low retention of this material are discussed. Lithium segregation to the surface of the sample was observed.

  15. PFMC-16. 16th international conference on plasma-facing materials and components for fusion applications. Abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2017-07-01

    The performances of fusion devices and of future fusion power plants strongly depend on the plasma-facing materials and components. Resistance to heat and particle loads, compatibility in plasma operations, thermo-mechanical properties, as well as the response to neutron irradiation are critical parameters which need to be understood and tailored from atomistic to component levels. The 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues.

  16. Progress of research on plasma facing materials in University of Science and Technology Beijing

    International Nuclear Information System (INIS)

    Ge, Chang-Chun; Zhou, Zhang-Jian; Song, Shu-Xiang; Du, Juan; Zhong, Zhi-Hong

    2007-01-01

    In this paper, we report some new progress on plasma facing materials in University of Science and Technology Beijing (USTB), China. They include fabrication of tungsten coating with ultra-fine grain size by atmosphere plasma spraying; fabrication of tungsten with ultra-fine grain size by a newly developed method named as resistance sintering under ultra-high pressure; using the concept of functionally graded materials to join tungsten to copper based heat sink; joining silicon doped carbon to copper by brazing using a Ti based amorphous filler and direct casting

  17. The temporal behaviour of MHD waves in a partially ionized prominence-like plasma: Effect of heating and cooling

    Science.gov (United States)

    Ballester, J. L.; Carbonell, M.; Soler, R.; Terradas, J.

    2018-01-01

    Context. During heating or cooling processes in prominences, the plasma microscopic parameters are modified due to the change of temperature and ionization degree. Furthermore, if waves are excited on this non-stationary plasma, the changing physical conditions of the plasma also affect wave dynamics. Aims: Our aim is to study how temporal variation of temperature and microscopic plasma parameters modify the behaviour of magnetohydrodynamic (MHD) waves excited in a prominence-like hydrogen plasma. Methods: Assuming optically thin radiation, a constant external heating, the full expression of specific internal energy, and a suitable energy equation, we have derived the profiles for the temporal variation of the background temperature. We have computed the variation of the ionization degree using a Saha equation, and have linearized the single-fluid MHD equations to study the temporal behaviour of MHD waves. Results: For all the MHD waves considered, the period and damping time become time dependent. In the case of Alfvén waves, the cut-off wavenumbers also become time dependent and the attenuation rate is completely different in a cooling or heating process. In the case of slow waves, while it is difficult to distinguish the slow wave properties in a cooling partially ionized plasma from those in an almost fully ionized plasma, the period and damping time of these waves in both plasmas are completely different when the plasma is heated. The temporal behaviour of the Alfvén and fast wave is very similar in the cooling case, but in the heating case, an important difference appears that is related with the time damping. Conclusions: Our results point out important differences in the behaviour of MHD waves when the plasma is heated or cooled, and show that a correct interpretation of the observed prominence oscillations is very important in order to put accurate constraints on the physical situation of the prominence plasma under study, that is, to perform prominence

  18. Evaluation of runaway-electron effects on plasma-facing components for NET

    Science.gov (United States)

    Bolt, H.; Calén, H.

    1991-03-01

    Runaway electrons which are generated during disruptions can cause serious damage to plasma facing components in a next generation device like NET. A study was performed to quantify the response of NET plasma facing components to runaway-electron impact. For the determination of the energy deposition in the component materials Monte Carlo computations were performed. Since the subsurface metal structures can be strongly heated under runaway-electron impact from the computed results damage threshold values for the thermal excursions were derived. These damage thresholds are strongly dependent on the materials selection and the component design. For a carbonmolybdenum divertor with 10 and 20 mm carbon armour thickness and 1 degree electron incidence the damage thresholds are 100 MJ/m 2 and 220 MJ/m 2. The thresholds for a carbon-copper divertor under the same conditions are about 50% lower. On the first wall damage is anticipated for energy depositions above 180 MJ/m 2.

  19. Recent progress in R and D on tungsten alloys for divertor structural and plasma facing materials

    Energy Technology Data Exchange (ETDEWEB)

    Wurster, S., E-mail: stefan.wurster@oeaw.ac.at [Erich Schmid Institute of Materials Science, Austria and Association EURATOM-ÖAW, Jahnstrasse 12, A-8700 Leoben (Austria); Baluc, N.; Battabyal, M. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Villigen PSI (Switzerland); Crosby, T. [University of California, Mechanical and Aerospace Engineering Department, Los Angeles, CA (United States); Du, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); García-Rosales, C. [Centro de Estudios e Investigaciones Técnicas de Gipuzkoa (CEIT), San Sebastián (Spain); Hasegawa, A. [Department of Quantum Science and Energy Engineering, Faculty of Engineering, Tohoku University (Japan); Hoffmann, A. [Plansee Metall GmbH, Reutte (Austria); Kimura, A. [Institute of Advanced Energy, Kyoto University (Japan); Kurishita, H. [International Research Center for Nuclear Material Science, Institute for Materials Research, Tohoku University (Japan); Kurtz, R.J. [Pacific Northwest National Laboratory, Richland, WA (United States); Li, H. [Erich Schmid Institute of Materials Science, Austria and Association EURATOM-ÖAW, Jahnstrasse 12, A-8700 Leoben (Austria); Chair of Atomistic Modelling and Design of Materials, University of Leoben, Leoben (Austria); Noh, S.; Reiser, J. [Karlsruhe Institute of Technology, Karlsruhe (Germany); Riesch, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe (Germany); Setyawan, W. [Pacific Northwest National Laboratory, Richland, WA (United States); Walter, M. [Karlsruhe Institute of Technology, Karlsruhe (Germany); You, J.-H. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); and others

    2013-11-15

    Tungsten materials are candidates for plasma-facing components for the International Thermonuclear Experimental Reactor and the DEMOnstration power plant because of their superior thermophysical properties. Because these materials are not common structural materials like steels, knowledge and strategies to improve the properties are still under development. These strategies discussed here, include new alloying approaches and microstructural stabilization by oxide dispersion strengthened as well as TiC stabilized tungsten based materials. The fracture behavior is improved by using tungsten laminated and tungsten wire reinforced materials. Material development is accompanied by neutron irradiation campaigns. Self-passivation, which is essential in case of loss-of-coolant accidents for plasma facing materials, can be achieved by certain amounts of chromium and titanium. Furthermore, modeling and computer simulation on the influence of alloying elements and heat loading and helium bombardment will be presented.

  20. Thermal loads on tokamak plasma-facing components during normal operation and disruptions

    International Nuclear Information System (INIS)

    McGrath, R.T.

    1990-01-01

    Power loadings experienced by tokamak plasma-facing components during normal operation and during off-normal events are discussed. A model for power and particle flow in the tokamak boundary layer is presented and model predictions are compared to infrared measurements of component heating. The inclusion of the full three-dimensional geometry of the components and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armour tile surfaces by only a millimeter can lead to significant localized heating. An application to the design of plasma-facing components for future machines is presented. Finally, thermal loads expected during tokamak disruptions are discussed. The primary problems are surface melting and vaporization due to localized intense heating during the disruption thermal quench and volumetric heating of the component armour and structure due to localised impact of runaway electrons. (author)

  1. Application of lock-in thermography non destructive technique to CFC armoured plasma facing components

    International Nuclear Information System (INIS)

    Escourbiac, F.; Constans, S.; Courtois, X.; Durocher, A.

    2007-01-01

    A non destructive testing technique - so called modulated photothermal thermography or lock-in thermography - has been set-up for plasma facing components examination. Reliable measurements of phase contrast were obtained on 8 mm carbon fiber composite (CFC) armoured W7-X divertor component with calibrated flaws. A 3D finite element analysis allowed the correlation of the measured phase contrast and showed that a 4 mm strip flaw can be detected at the CFC/copper interface

  2. Magnetic field effects on runaway electron energy deposition in plasma facing materials and components

    International Nuclear Information System (INIS)

    Niemer, K.A.; Gilligan, J.G.

    1992-01-01

    This paper reports magnetic field effects on runaway electron energy deposition in plasma facing materials and components is investigated using the Integrated TIGER Series. The Integrated TIGER Series is a set of time-independent coupled electron/photon Monte Carlo transport codes which perform photon and electron transport, with or without macroscopic electric and magnetic fields. A three-dimensional computational model of 100 MeV electrons incident on a graphite block was used to simulate runawayelectrons striking a plasma facing component at the edge of a tokamak. Results show that more energy from runaway electrons will be deposited in a material that is in the presence of a magnetic field than in a material that is in the presence of no field. For low angle incident runaway electrons in a strong magnetic field, the majority of the increased energy deposition is near the material surface with a higher energy density. Electrons which would have been reflected with no field, orbit the magnetic field lines and are redeposited in the material surface, resulting in a substantial increase in surface energy deposition. Based on previous studies, the higher energy deposition and energy density will result in higher temperatures which are expected to cause more damage to a plasma facing component

  3. Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W

    International Nuclear Information System (INIS)

    Skinner, C.H.; Haasz, A.A.; Alimov, V.Kh.; Bekris, N.; Causey, R.A.; Clark, R.E.H.; Coad, J.P.; Davis, J.W.; Doerner, R.P.; Mayer, M.; Pisarev, A.; Roth, J.; Tanabe, T.

    2008-01-01

    Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described

  4. Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W.

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C H; Alimov, Kh; Bekris, N; Causey, R A; Clark, R.E.H.; Coad, J P; Davis, J W; Doerner, R P; Mayer, M; Pisarev, A; Roth, J

    2008-03-29

    Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.

  5. Molecular dynamics simulations of interactions between energetic dust and plasma-facing materials

    International Nuclear Information System (INIS)

    Niu, Guo-jian; Li, Xiao-chun; Xu, Qian; Yang, Zhong-shi; Luo, Guang-nan

    2015-01-01

    The interactions between dust and plasma-facing material (PFM) relate to the lifetime of PFM and impurity production. Series results have been obtained theoretically and experimentally but more detailed studies are needed. In present research, we investigate the evolution of kinetic, potential and total energy of plasma-facing material (PFM) in order to understand the dust/PFM interaction process. Three typical impacting energy are selected, i.e., 1, 10 and 100 keV/dust for low-, high- and hyper-energy impacting cases. For low impacting energy, dust particles stick on PFM surface without damaging it. Two typical time points exist and the temperature of PFM grows all the time but PFM structure experience a modifying process. Under high energy case, three typical points appear. The temperature curve fluctuates in the whole interaction process which indicates there are dust/PFM and kinetic/potential energy exchanges. In the hyper-energy case in present simulation, the violence dust/PFM interactions cause sputtering and crater investigating on energy evolution curves. We further propose the statistics of energy distribution. Results show that about half of impacting energy consumes on heating plasma-facing material meanwhile the other half on PFM structure deformation. Only a small proportion becomes kinetic energy of interstitial or sputtering atoms.

  6. Molecular dynamics simulations of interactions between energetic dust and plasma-facing materials

    Energy Technology Data Exchange (ETDEWEB)

    Niu, Guo-jian, E-mail: niugj@ipp.ac.cn [Institute of Plasma Physics Chinese Academy of Sciences, Hefei (China); Li, Xiao-chun; Xu, Qian; Yang, Zhong-shi [Hefei Center Physical Science and Technology, Hefei (China); Luo, Guang-nan [Institute of Plasma Physics Chinese Academy of Sciences, Hefei (China); Hefei Center Physical Science and Technology, Hefei (China); Hefei Science Center of CAS, Hefei (China)

    2015-11-15

    The interactions between dust and plasma-facing material (PFM) relate to the lifetime of PFM and impurity production. Series results have been obtained theoretically and experimentally but more detailed studies are needed. In present research, we investigate the evolution of kinetic, potential and total energy of plasma-facing material (PFM) in order to understand the dust/PFM interaction process. Three typical impacting energy are selected, i.e., 1, 10 and 100 keV/dust for low-, high- and hyper-energy impacting cases. For low impacting energy, dust particles stick on PFM surface without damaging it. Two typical time points exist and the temperature of PFM grows all the time but PFM structure experience a modifying process. Under high energy case, three typical points appear. The temperature curve fluctuates in the whole interaction process which indicates there are dust/PFM and kinetic/potential energy exchanges. In the hyper-energy case in present simulation, the violence dust/PFM interactions cause sputtering and crater investigating on energy evolution curves. We further propose the statistics of energy distribution. Results show that about half of impacting energy consumes on heating plasma-facing material meanwhile the other half on PFM structure deformation. Only a small proportion becomes kinetic energy of interstitial or sputtering atoms.

  7. Remote-LIBS characterization of ITER-like plasma facing materials

    International Nuclear Information System (INIS)

    Almaviva, S.; Caneve, L.; Colao, F.; Fantoni, R.; Maddaluno, G.

    2012-01-01

    Graphical abstract: Display Omitted Highlights: ► Description of a LIBS set-up as remote diagnostics in new generation fusion machines. ► Identification of the atomic composition of samples simulating plasma facing components. ► Submicrometric resolution in depth profiling the elemental composition of the samples. ► Identification of elements present in traces or as impurities on the sample surface. ► Discussion on the applicability of the Calibration Free method for quantitative analysis. - Abstract: The occurrence of several plasma-wall interaction processes, eventually affecting the overall system performances, is expected in a working fusion device chamber. Monitoring the changes in the composition of the plasma facing component (PFC) surface layer, as a result of erosion and redeposition mechanisms, can provide useful information on the possible plasma pollution and fuel retention. To this aim, suitable diagnostic techniques able to perform depth profiling analysis of the superficial layers on the PFCs have been developed. Due to the constraints commonly found in fusion devices, the measuring apparatus must be non invasive, remote and sensitive to light elements. These requirements make LIBS (Laser Induced Breakdown Spectroscopy) an ideal candidate for on-line monitoring the walls of current and of next generation (as ITER) fusion devices. LIBS is a well established tool for qualitative, semi-quantitative and quantitative analysis of surfaces, with micro-destructive characteristics and some capabilities for stratigraphy. In this work, LIBS depth profiling capability has been verified for the determination of the composition of multilayer structures simulating plasma facing components covered with deposited impurity layers. A new experimental setup has been designed and realized in order to optimize the characteristics of a LIBS system working in vacuum conditions and remotely, two noticeable properties for an ITER-relevant diagnostics. A quantitative

  8. Parametrization of the average ionization and radiative cooling rates of carbon plasmas in a wide range of density and temperature

    International Nuclear Information System (INIS)

    Gil, J.M.; Rodriguez, R.; Florido, R.; Rubiano, J.G.; Mendoza, M.A.; Nuez, A. de la; Espinosa, G.; Martel, P.; Minguez, E.

    2013-01-01

    In this work we present an analysis of the influence of the thermodynamic regime on the monochromatic emissivity, the radiative power loss and the radiative cooling rate for optically thin carbon plasmas over a wide range of electron temperature and density assuming steady state situations. Furthermore, we propose analytical expressions depending on the electron density and temperature for the average ionization and cooling rate based on polynomial fittings which are valid for the whole range of plasma conditions considered in this work. -- Highlights: ► We compute the average ionization, cooling rates and emissivities of carbon plasmas. ► We compare LTE and NLTE calculations of these magnitudes. ► We perform a parametrization of these magnitudes in a wide range of plasma conditions. ► We provide information about where LTE regime assumption is accurate

  9. Direct measurements of particle flux along gap sides in castellated plasma facing component in COMPASS

    International Nuclear Information System (INIS)

    Dejarnac, Renaud; Dimitrova, Miglena; Komm, Michael; Schweer, Bernd; Terra, Alexis; Martin, Aurelien; Boizante, Gontran; Gunn, James P.; Panek, Radomir

    2014-01-01

    Highlights: •We designed a probe to measure plasma deposition into gaps during tokamak discharges. •Isat profiles are measured on both side of the gap for different gap orientations. •Ion current is measured at the bottom of the gap in the toroidal orientation. •Kinetic simulations reproduce well experimental profiles qualitatively. -- Abstract: In this paper, we report results of a dedicated experiment that gives the plasma penetration profiles inside a gap of a tokamak castellated plasma-facing component. A specially designed probe that recreates a gap between two tiles has been built for the purpose of this study. It allows to measure ion saturation profiles along the 2 sides and at the bottom of the gap for both poloidal and toroidal orientations. The novelty of such experiment is the real time measurement of the plasma flux inside the gap during a tokamak D-shaped discharge compared to previous experimental studies which were mainly post-mortem. This experiment was performed in the COMPASS tokamak and results are compared with particle-in-cell simulations. The plasma deposition is found to be asymmetric in both orientations with a stronger effect in poloidal gaps. The Larmor radius of the incoming ions plays a role in the plasma penetration only in poloidal gaps but seems to have little impact in toroidal gaps. Profiles are qualitatively well reproduced by simulations. Ion current is recorded at the bottom of a toroidal gap under certain conditions

  10. Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Longhurst, G.R. [Lockheed Martin Idaho Technol. Co., Idaho Falls, ID (United States). Idaho Nat. Eng. and Environ. Lab.; Causey, R.A.; Wampler, W.R.; Wilson, K.L. [Sandia National Laboratories, Livermore, CA (United States)]|[Sandia National Labs., Albuquerque, NM (United States); Davis, J.W.; Haasz, A.A. [Institute for Aerospace Studies, University of Toronto, Toronto (Canada); Doerner, R.P. [California Univ., San Diego, La Jolla, CA (United States). Center for Magnetic Recording Research; Federici, G. [ITER JWS Garching Co-center, Garching (Germany)

    1999-06-01

    Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions, show a difference of several orders of magnitude in both inventory and permeation rate to the coolant. (orig.) 78 refs.

  11. Tritium recycling and inventory in eroded debris of plasma-facing materials

    International Nuclear Information System (INIS)

    Hassanein, A.

    1999-01-01

    Damage to plasma-facing components (PFCs) and structural materials due to loss of plasma confinement in magnetic fusion reactors remains one of the most serious concerns for safe, successful, and reliable tokamak operation. High erosion losses due to surface vaporization, spallation, and melt-layer splashing are expected during such an event. The eroded debris and dust of the PFCs, including trapped tritium, will be contained on the walls or within the reactor chamber therefore, they can significantly influence plasma behavior and tritium inventory during subsequent operations. Tritium containment and behavior in PFCS and in the dust and debris is an important factor in evaluating and choosing the ideal plasma-facing materials (PFMs). Tritium buildup and release in the debris of candidate materials is influenced by the effect of material porosity on diffusion and retention processes. These processes have strong nonlinear behavior due to temperature, volubility, and existing trap sites. A realistic model must therefore account for the nonlinear and multidimensional effects of tritium diffusion in the porous-redeposited and neutron-irradiated materials. A tritium-transport computer model, TRAPS (Tritium Accumulation in Porous Structure), was developed and used to evaluate and predict the kinetics of tritium transport in porous media. This model is coupled with the TRICS (Tritium In Compound Systems) code that was developed to study the effect of surface erosion during normal and abnormal operations on tritium behavior in PFCS

  12. Hydrodynamic effects of eroded materials on response of plasma-facing component during a tokamak disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1999-01-01

    Loss of plasma confinement causes surface and structural damage to plasma-facing materials (PFMs) and remains a major obstacle for tokamak reactors. The deposited plasma energy results in surface erosion and structural failure. The surface erosion consists of vaporization, spallation, and liquid splatter of metallic materials, while the structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. Comprehensive models (contained in the HEIGHTS computer simulation package) are being used self-consistently to evaluate material damage. Splashing mechanisms occur as a result of volume bubble boiling and liquid hydrodynamic instabilities and brittle destruction mechanisms of nonmelting materials. The effect of macroscopic erosion on total mass losses and lifetime is evaluated. The macroscopic erosion products may further protect PFMs from severe erosion (via the droplet-shielding effect) in a manner similar to that of the vapor shielding concept

  13. Critical plasma-wall interaction issues for plasma-facing materials and components in near-term fusion devices

    International Nuclear Information System (INIS)

    Federici, G.; Coad, J.P.; Haasz, A.A.; Janeschitz, G.; Noda, N.; Philipps, V.; Roth, J.; Skinner, C.H.; Tivey, R.; Wu, C.H.

    2000-01-01

    The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma-material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R and D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R and D work is urgently needed

  14. Production of radiatively cooled hypersonic plasma jets and links to astrophysical jets

    International Nuclear Information System (INIS)

    Lebedev, S V; Ciardi, A; Ampleford, D J; Bland, S N; Bott, S C; Chittenden, J P; Hall, G N; Rapley, J; Jennings, C; Sherlock, M; Frank, A; Blackman, E G

    2005-01-01

    We present results of high energy density laboratory experiments on the production of supersonic radiatively cooled plasma jets with dimensionless parameters (Mach number ∼30, cooling parameter ∼1 and density contrast ρ j /ρ a ∼ 10) similar to those in young stellar objects jets. The jets are produced using two modifications of wire array Z-pinch driven by 1 MA, 250 ns current pulse of MAGPIE facility at Imperial College, London. In the first set of experiments the produced jets are purely hydrodynamic and are used to study deflection of the jets by the plasma cross-wind, including the structure of internal oblique shocks in the jets. In the second configuration the jets are driven by the pressure of the toroidal magnetic field and this configuration is relevant to the astrophysical models of jet launching mechanisms. Modifications of the experimental configuration allowing the addition of the poloidal magnetic field and angular momentum to the jets are also discussed. We also present three-dimensional resistive magneto-hydrodynamic simulations of the experiments and discuss the scaling of the experiments to the astrophysical systems

  15. Development of ZL400 Mine Cooling Unit Using Semi-Hermetic Screw Compressor and Its Application on Local Air Conditioning in Underground Long-Wall Face

    Science.gov (United States)

    Chu, Zhaoxiang; Ji, Jianhu; Zhang, Xijun; Yan, Hongyuan; Dong, Haomin; Liu, Junjie

    2016-12-01

    Aiming at heat injuries occurring in the process of deep coal mining in China, a ZL400 mine-cooling unit employing semi-hermetic screw compressor with a cooling capacity of 400 kW is developed. This paper introduced its operating principle, structural characteristics and technical indexes. By using the self-built testing platform, some parameters for indication of its operation conditions were tested on the ground. The results show that the aforementioned cooling unit is stable in operation: cooling capacity of the unit was 420 kW underground-test conditions, while its COP (coefficient of performance) reached 3.4. To address the issue of heat injuries existing in No. 16305 U-shaped long-wall ventilation face of Jining No. 3 coal mine, a local air conditioning system was developed with ZL400 cooling unit as the system's core. The paper presented an analysis of characteristics of the air current flowing in the air-mixing and cooling mode of ZL400 cooling unit used in air intake way. Through i-d patterns we described the process of the airflow treatment, such as cooling, mixing and heating, etc. The cooling system decreased dry bulb temperature on working face by 3°C on average and 3.8°C at most, while lowered the web bulb temperature by 3.6°C on average and 4.8°C at most. At the same time, it reduced relative humidity by 5% on average and 8.6% at most. The field application of the ZL400 cooling unit had gain certain effects in air conditioning and provided support for the solution of mine heat injuries in China in terms of technology and equipment.

  16. Material testing facilities and programs for plasma-facing component testing

    Science.gov (United States)

    Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

    2017-09-01

    Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

  17. Development of actively cooled windows for plasma observation during quasi-continuous operation of the W7-X stellarator

    International Nuclear Information System (INIS)

    Konig, R.; Grosser, K.; Hildebrandt, D.; Pasch, E.; Werner, T.; Klinger, T.; Ogorodnikova, O.

    2005-01-01

    With the stellarator W7-X a step to quasi-continuous plasma operation will be made. The cooling system of the machine is designed such that two 30 min discharges can be run per day. Right from the start of operation 10 MW of ECRH heating power will be available for quasi-continuous operation. A working group 'Plasma Facing Optical Components' has been formed which presently concentrates on the development of water cooled windows for UV/VR/IR periscopes which can withstand the expected maximum heat loads of up to 50 kW/m 2 which due to the predominantly short wavelength nature of the radiation emitted by the plasma will be absorbed within the first millimeter of any window. We will report on the detailed Finite Element (ANSYS R ) calculations of the heat and stress distribution across the windows. Calculations have been undertaken for a large number of different window materials which are required for the various spectral regions covered by the miscellaneous diagnostics, so that the most suitable material for each application can easily be identified. Also the dependence of the cooling rate on the window diameter and thickness has been studied. The calculations show that at a power load of 50 kW/m 2 cooled sapphire windows can be used for window sizes up to ∼200 mm diameter but that for many of the other materials like ZnSe, ZnS, CaF 2 , MgF 2 and quartz window sizes need to be limited to considerably smaller sizes. Detailed simulations of the local radiation power load distribution demonstrate that by careful design the load on individual optical components can be considerably reduced. A vacuum test chamber, equipped with a vacuum compatible IR heater has been build. In this chamber a low cost, easily exchangeable window design using Helicoflex gaskets on either side of a 60 mm exposed diameter quartz window have been successfully tested over 70 heat cycles up to a maximum temperature of 450 o C at power loads of 15 kW/m 2 . The design proved to be water and

  18. High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

    Science.gov (United States)

    Neu, R.

    2006-04-01

    In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures.

  19. High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

    International Nuclear Information System (INIS)

    Neu, R.

    2006-01-01

    In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials (PFMs). However, in a future fusion reactor, for economic reasons, a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which have to be fulfilled simultaneously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFMs but also point to operational restrictions, when using them as PFCs. These are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor-relevant scenarios which incorporate the experiences and measures

  20. High Heat Flux Interactions and Tritium Removal from Plasma Facing Components by a Scanning Laser

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Hassanein, A.

    2002-01-01

    A new technique for studying high heat flux interactions with plasma facing components is presented. The beam from a continuous wave 300 W neodymium laser was focused to 80 W/mm2 and scanned at high speed over the surface of carbon tiles. These tiles were previously used in the TFTR [Tokamak Fusion Test Reactor] inner limiter and have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations. Laser scanning released up to 84% of the codeposited tritium. The temperature rise of the codeposit on the tiles was significantly higher than that of the manufactured material. In one experiment, the codeposit surface temperature rose to 1,770 C while for the same conditions, the manufactured surface increased to only 1,080 C. The peak temperature did not follow the usual square-root dependence on heat pulse duration. Durations of order 100 ms resulted in brittle destruction and material loss from the surface, while a duration of approximately 10 ms showed minimal change. A digital microscope imaged the codeposit before, during, and after the interaction with the laser and revealed hot spots on a 100-micron scale. These results will be compared to analytic modeling and are relevant to the response of plasma facing components to disruptions and vertical displacement events (VDEs) in next-step magnetic fusion devices

  1. Multi parametric sensitivity study applied to temperature measurement of metallic plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Aumeunier, M-H.; Corre, Y.; Firdaouss, M.; Gauthier, E.; Loarer, T.; Travere, J-M.; Gardarein, J-L.; EFDA JET Contributor

    2013-06-01

    In nuclear fusion experiments, the protection system of the Plasma Facing Components (PFCs) is commonly ensured by infrared (IR) thermography. Nevertheless, the surface monitoring of new metallic plasma facing component, as in JET and ITER is being challenging. Indeed, the analysis of infrared signals is made more complicated in such a metallic environment since the signals will be perturbed by the reflected photons coming from high temperature regions. To address and anticipate this new measurement environment, predictive photonic models, based on Monte-Carlo ray tracing (SPEOS R CAA V5 Based), have been performed to assess the contribution of the reflective part in the total flux collected by the camera and the resulting temperature error. This paper deals with the effects of metals features, as the emissivity and reflectivity models, on the accuracy of the surface temperature estimation. The reliability of the features models is discussed by comparing the simulation with experimental data obtained with the wide angle IR thermography system of JET ITER like wall. The impact of the temperature distribution is studied by considering two different typical plasma scenarios, in limiter (ITER start-up scenario) and in X-point configurations (standard divertor scenario). The achievable measurement performances of IR system and risks analysis on its functionalities are discussed. (authors)

  2. Parametrization of the average ionization and radiative cooling rates of carbon plasmas in a wide range of density and temperature

    OpenAIRE

    Gil de la Fe, Juan Miguel; Rodriguez Perez, Rafael; Florido, Ricardo; Garcia Rubiano, Jesus; Mendoza, M.A.; Nuez, A. de la; Espinosa, G.; Martel Escobar, Carlos; Mínguez Torres, Emilio

    2013-01-01

    In this work we present an analysis of the influence of the thermodynamic regime on the monochromatic emissivity, the radiative power loss and the radiative cooling rate for optically thin carbon plasmas over a wide range of electron temperature and density assuming steady state situations. Furthermore, we propose analytical expressions depending on the electron density and temperature for the average ionization and cooling rate based on polynomial fittings which are valid for the whole range...

  3. Manufacturing and High Heat Flux Testing of Brazed Flat-Type W/CuCrZr Plasma Facing Components

    Science.gov (United States)

    Lian, Youyun; Liu, Xiang; Feng, Fan; Chen, Lei; Cheng, Zhengkui; Wang, Jin; Chen, Jiming

    2016-02-01

    Water-cooled flat-type W/CuCrZr plasma facing components with an interlayer of oxygen-free copper (OFC) have been developed by using vacuum brazing route. The OFC layer for the accommodation of thermal stresses was cast onto the surface of W at a temperature range of 1150 °C-1200 °C in a vacuum furnace. The W/OFC cast tiles were vacuum brazed to a CuCrZr heat sink at 940 °C using the silver-free filler material CuMnSiCr. The microstructure, bonding strength, and high heat flux properties of the brazed W/CuCrZr joint samples were investigated. The W/Cu joint exhibits an average tensile strength of 134 MPa, which is about the same strength as pure annealed copper. High heat flux tests were performed in the electron beam facility EMS-60. Experimental results indicated that the brazed W/CuCrZr mock-up experienced screening tests of up to 15 MW/m2 and cyclic tests of 9 MW/m2 for 1000 cycles without visible damage. supported by National Natural Science Foundation of China (No. 11205049) and the National Magnetic Confinement Fusion Science Program of China (No. 2011GB110004)

  4. Electron cooling and finite potential drop in a magnetized plasma expansion

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Sanchez, M. [Department of Aeronautics and Astronautics, Massachusetts Institute of Technology, 77 Massachusetts Avenue, Cambridge, Massachusetts 02139 (United States); Navarro-Cavallé, J. [Escuela Técnica Superior de Ingeniería Aeronáutica y del Espacio, Universidad Politécnica de Madrid, Plaza Cardenal Cisneros 3, Madrid 28040 (Spain); Ahedo, E. [Departamento de Bioingeniería e Ingeniería Aeroespacial, Universidad Carlos III de Madrid, Avda. Universidad 30, Leganés 28911, Madrid (Spain)

    2015-05-15

    The steady, collisionless, slender flow of a magnetized plasma into a surrounding vacuum is considered. The ion component is modeled as mono-energetic, while electrons are assumed Maxwellian upstream. The magnetic field has a convergent-divergent geometry, and attention is restricted to its paraxial region, so that 2D and drift effects are ignored. By using the conservation of energy and magnetic moment of particles and the quasi-neutrality condition, the ambipolar electric field and the distribution functions of both species are calculated self-consistently, paying attention to the existence of effective potential barriers associated to magnetic mirroring. The solution is used to find the total potential drop for a set of upstream conditions, plus the axial evolution of various moments of interest (density, temperatures, and heat fluxes). The results illuminate the behavior of magnetic nozzles, plasma jets, and other configurations of interest, showing, in particular, in the divergent plasma the collisionless cooling of electrons, and the generation of collisionless electron heat fluxes.

  5. Preliminary assessment of the tritium inventory and permeation in the plasma facing components of ITER

    International Nuclear Information System (INIS)

    Federici, G.; Holland, D.; Brooks, J.; Causey, R.; Dolan, T.J.; Longhurst, G.

    1995-01-01

    This paper discusses preliminary quantitative predictions for the tritium inventory in- and permeation through the first-wall and divertor PFC's of ITER. The primary plasma facing material under consideration is beryllium, with possible use of tungsten or carbon fiber composites (CFC's) on high-heat-flux surfaces. They use state-of-the-art tritium transport models, in conjunction with design parameters, and loading conditions anticipated for the first-wall, baffle, limiter and divertor. The analysis includes the synergistic effects of erosion on tritium implantation and trapping, which are expected to play a key role, particularly in the divertor regions where the interaction of the plasma with the surfaces will be most severe. The influence of several key parameters that strongly affect tritium build-up and release is assessed. Finally, they discuss the uncertainties in materials properties under ITER operating conditions and the R and D needed to resolve these uncertainties

  6. Remote Metrology, Mapping, and Motion Sensing of Plasma Facing Components Using FM Coherent Laser Radar

    International Nuclear Information System (INIS)

    Menon, M.M.; Barry, R.E.; Slotwinsky, A.; Kugel, H.W.; Skinner, C.H.

    2000-01-01

    Metrology inside a D/T burning fusion reactor must necessarily be conducted remotely since the in-vessel environment would be highly radioactive due to neutron activation of the torus walls. A technique based on frequency modulated coherent laser radar (FM CLR) for such remote metrology is described. Since the FM CLR relies on frequency shift to measure distances, the results are largely insensitive to surface reflectance characteristics. Results of measurements in TFTR and NSTX fusion devices using a prototype FM CLR unit, capable of remotely measuring distances (range) up to 22 m with better than 0.1-mm precision, are provided. These results illustrate that the FM CLR can be used for precision remote metrology as well as viewing. It is also shown that by conducting Doppler corrected range measurements using the CLR, the motion of objects can be tracked. Thus, the FM CLR has the potential to remotely measure the motion of plasma facing components (PFCs) during plasma disruptions

  7. A possible method of carbon deposit mapping on plasma facing components using infrared thermography

    International Nuclear Information System (INIS)

    Mitteau, R.; Spruytte, J.; Vallet, S.; Travere, J.M.; Guilhem, D.; Brosset, C.

    2007-01-01

    The material eroded from the surface of plasma facing components is redeposited partly close to high heat flux areas. At these locations, the deposit is heated by the plasma and the deposition pattern evolves depending on the operation parameters. The mapping of the deposit is still a matter of intense scientific activity, especially during the course of experimental campaigns. A method based on the comparison of surface temperature maps, obtained in situ by infrared cameras and by theoretical modelling is proposed. The difference between the two is attributed to the thermal resistance added by deposited material, and expressed as a deposit thickness. The method benefits of elaborated imaging techniques such as possibility theory and fuzzy logics. The results are consistent with deposit maps obtained by visual inspection during shutdowns

  8. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report

  9. Proceedings of the 4th International Workshop on Tritium Effects in Plasma Facing Components

    International Nuclear Information System (INIS)

    Causey, R. A.

    1999-01-01

    The 4th International Workshop on Tritium Effects in Plasma Facing Components was held in Santa Fe, New Mexico on May 14-15, 1998. This workshop occurs every two years, and has previously been held in Livermore/California, Nagoya/Japan, and the JRC-Ispra Site in Italy. The purpose of the workshop is to gather researchers involved in the topic of tritium migration, retention, and recycling in materials used to line magnetic fusion reactor walls and provide a forum for presentation and discussions in this area. This document provides an overall summary of the workshop, the workshop agenda, a summary of the presentations, and a list of attendees

  10. Status of R and D of the plasma facing components for the ITER divertor

    International Nuclear Information System (INIS)

    Mazul, I.V.; Akiba, M.; Arkhipov, I.

    2001-01-01

    The paper reports the progress made by the ITER Home Teams in the development of robust carbon and tungsten armoured plasma facing components for the ITER divertor. The activities on the development and study of armour materials, joining technologies, non-destructive evaluation techniques, high heat flux testing of manufactured components and neutron irradiation resistance studies are presented. The results of these activities confirm the feasibility of the main divertor components. Examples of the fruitful collaboration between Parties and future R and D needs are also described. (author)

  11. Simulated plasma facing component measurements for an in situ surface diagnostic on Alcator C-Moda)

    Science.gov (United States)

    Hartwig, Z. S.; Whyte, D. G.

    2010-10-01

    The ideal in situ plasma facing component (PFC) diagnostic for magnetic fusion devices would perform surface element and isotope composition measurements on a shot-to-shot (˜10 min) time scale with ˜1 μm depth and ˜1 cm spatial resolution over large areas of PFCs. To this end, the experimental adaptation of the customary laboratory surface diagnostic—nuclear scattering of MeV ions—to the Alcator C-Mod tokamak is being guided by ACRONYM, a Geant4 synthetic diagnostic. The diagnostic technique and ACRONYM are described, and synthetic measurements of film thickness for boron-coated PFCs are presented.

  12. Design of Plasma Facing Components for Superconducting Modification of JT-60

    International Nuclear Information System (INIS)

    Shinji Sakurai; Kei Masaki; Yusuke-Kudo Shibama; Hiroshi Tamai; Makoto Matsukawa; Cordier, J.J.

    2006-01-01

    JT-60 is planning to modify the machine as a fully superconducting coil tokamak (JT-60 Super Advanced, the former JT-60SC and NCT) to establish scientific and technological bases for an economically and environmentally attractive DEMO reactor. It will be also a satellite tokamak in a part of broader approach for ITER. It is designed for high beta (betaN = 3.5-5.5) and steady-state research in a break-even class DD plasma for 100 s or longer. Nominal plasma parameters are I p =5.5 MA, B t =2.7 T, R=3.01 m, a=1.14 m with double-null configuration. An ITER-like single-null configuration with I p =3.5 MA, B t =2.6 T can be also operated. In order to study the ITER-relevant high confinement plasma with high density, designed plasma heating power was enhanced from 25 MW to 41 MW for 100 s through the design review with EU and Japan. The heat flux onto outer divertor target exceeds 10 MW/m 2 with moderate radiative fraction of 50-60% in single-null configuration. Therefore, the ITER-like mono-block CFC target will be adopted to aim at power handling of 15 MW/m 2 . A cooling water system should be reinforced 3 times from original design for double null divertor with high coolant flow velocity of ∼10 m/s. The peak heat flux onto the neutral beam armor for perpendicular injected positive NB is evaluated to be 2 MW/m 2 , which needs to be actively water-cooled. A bolt-fixed CFC tile was tested at the heat flux of 1-3 MW/m 2 and will be applied to the NB armor. In order to improve plasma beta value by enhancing wall stabilization effect, passive-stabilizing plates, which are electrically and mechanically connected in poloidal and toroidal direction, will be installed near the plasma surface (r wall /a=1.1-1.3) at the outboard side. Stabilizing plate has double-wall ribbed structure and can be operated at 573 K with heating nitrogen gas instead of cooling water between double walls. It has crank-type support legs to allow thermal expansion at high temperature operation. The

  13. Hydrogen transport behavior of metal coatings for plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Holland, D.F.; Longhurst, G.R. (Idaho National Engineering Lab., Idaho Falls (USA))

    1990-12-01

    Plasma-facing components for experimental and commercial fusion reactor studies may include cladding or coatings of refractory metals like tungsten on metallic structural substrates such as copper, vanadium alloys and austenitic stainless steel. Issues of safety and fuel economy include the potential for inventory buildup and permeation of tritium implanted into the plasma-facing surface. This paper reports on laboratory-scale studies with 3 keV D{sub 3}{sup +} ion beams to investigate the hydrogen transport behavior in tungsten coatings on substrates of copper. These experiments entailed measurements of the deuterium re-emission and permeation rates of tungsten, copper, and tungsten-coated copper specimens at temperatures ranging from 638 to 825 K and implanting particle fluxes of approximately 5x10{sup 19} D/m{sup 2} s. Diffusion constants and surface recombination coefficients with enhancement factors due to sputtering were obtained from these measurements. These data may be used in calculations to estimate permeation rates and inventory buildups for proposed diverter designs. (orig.).

  14. Hydrogen transport behavior of metal coatings for plasma-facing components

    Science.gov (United States)

    Anderl, R. A.; Holland, D. F.; Longhurst, G. R.

    1990-12-01

    Plasma-facing components for experimental and commercial fusion reactor studies may include cladding or coatings of refractory metals like tungsten on metallic structural substrates such as copper, vanadium alloys and austenitic stainless steel. Issues of safety and fuel economy include the potential for inventory buildup and permeation of tritium implanted into the plasma-facing surface. This paper reports on laboratory-scale studies with 3 keV D +3 ion beams to investigate the hydrogen transport behavior in tungsten coatings on substrates of copper. These experiments entailed measurements of the deuterium re-emission and permeation rates for tungsten, copper, and tungsten-coated copper specimens at temperatures ranging from 638 to 825 K and implanting particle fluxes of approximately 5 × 10 19 D/m 2 s. Diffusion constants and surface recombination coefficients with enhancement factors due to sputtering were obtained from these measurements. These data may be used in calculations to estimate permeation rates and inventory buildups for proposed diverter designs.

  15. Synergistic effects of surface erosion on tritium inventory and permeation in metallic plasma facing armours

    Science.gov (United States)

    Federici, G.; Holland, D. F.; Matera, R.

    1996-10-01

    In the next generation of DT fuelled tokamaks, i.e., the International Thermonuclear Experimental Reactor (ITER) implantation of energetic DT particles on some portions of the plasma facing components (PFCs) will take place along with significant erosion of the armour surfaces. As a result of the simultaneous removal of material from the front surface, the build-up of tritium inventory and the start of permeation originating in the presence of large densities of neutron-induced traps is expected to be influenced considerably and special provisions could be required to minimise the consequences on the design. This paper reports on the results of a tritium transport modelling study based on a new model which describes the migration of implanted tritium across the bulk of metallic plasma facing materials containing neutron-induced traps which can capture it and includes the synergistic effects of surface erosion. The physical basis of the model is summarised, but emphasis is on the discussion of the results of a comparative study performed for beryllium and tungsten armours for ranges of design and operation conditions similar to those anticipated in the divertor of ITER.

  16. Synergistic effects of surface erosion on tritium inventory and permeation in metallic plasma facing armours

    International Nuclear Information System (INIS)

    Federici, G.; Holland, D.F.; Matera, R.

    1996-01-01

    In the next generation of DT fuelled tokamaks, i.e., the international thermonuclear experimental reactor (ITER) implantation of energetic DT particles on some portions of the plasma facing components (PFCs) will take place along with significant erosion of the armour surfaces. As a result of the simultaneous removal of material from the front surface, the build-up of tritium inventory and the start of permeation originating in the presence of large densities of neutron-induced traps is expected to be influenced considerably and special provisions could be required to minimise the consequences on the design. This paper reports on the results of a tritium transport modelling study based on a new model which describes the migration of implanted tritium across the bulk of metallic plasma facing materials containing neutron-induced traps which can capture it and includes the synergistic effects of surface erosion. The physical basis of the model is summarised, but emphasis is on the discussion of the results of a comparative study performed for beryllium and tungsten armours for ranges of design and operation conditions similar to those anticipated in the divertor of ITER. (orig.)

  17. Boron carbide-coated carbon material, manufacturing method therefor and plasma facing material

    International Nuclear Information System (INIS)

    Suzuki, Takayuki; Kikuchi, Yoshihiro; Hyakki, Yasuo.

    1997-01-01

    The present invention concerns a plasma facing material suitable to a thermonuclear device. The material comprises a carbon material formed by converting the surface of a carbon fiber-reinforced carbon material comprising a carbon matrix and carbon fibers to a boron carbide, the material has a surface comprising vertically or substantially vertically oriented carbon fibers, and the thickness of the surface converted to boron carbide is reduced in the carbon fiber portion than in the carbon matrix portion. Alternatively, a carbon fiber-reinforced carbon material containing carbon fibers having a higher graphitizing degree than the carbon matrix is converted to boron carbide on the surface where the carbon fibers are oriented vertically or substantially vertically. The carbon fiber-reinforced material is used as a base material, and a resin material impregnated into a shaped carbon fiber product is carbonized or thermally decomposed carbon is filled as a matrix. The material of the present invention has high heat conduction and excellent in heat resistance thereby being suitable to a plasma facing material for a thermonuclear device. Electric specific resistivity of the entire coating layer can be lowered, occurrence of arc discharge is prevented and melting can be prevented. (N.H.)

  18. Simulation of damage to tokamaks plasma facing components during intense abnormal power deposition

    International Nuclear Information System (INIS)

    Genco, F.; Hassanein, A.

    2014-01-01

    Highlights: • HEIGHTS-PIC a new technique based on particle in cell method to study disruptions events, ELMS and VDE is benchmarked in this paper with the use of the MK-200 experiments. • Disruptions simulations results for erosion and erosion rate are proposed showing good agreement with published experimental available data for such conditions. • Results are also compared with other published results produced by FOREV1/FOREV2 computer package and the original HEIGHTS computer package. • Accuracy of the simulations results is proposed with specific aim to address the use of number of super particles adopted versus computational time. - Abstract: Intense power deposition on plasma facing components (PFC) is expected in tokamaks during loss of confinement events such as disruptions, vertical displacement events (VDE), runaway electrons (RE), or during normal operating conditions such as edge-localized modes (ELM). These highly energetic events are damaging enough to hinder long term operation and may not be easily mitigated without loss of structural or functional performance of the PFC. Surface erosion, melted/ablated-vaporized material splashing, and material transport into the bulk plasma are reliability-threatening for the machine and system performance. A novel particle-in-cell (PIC) technique has been developed and integrated into the existing HEIGHTS package in order to obtain a global view of the plasma evolution upon energy impingement. This newly developed PIC technique is benchmarked against plasma gun experimental data, the original HEIGHTS computer package, and laser experiments. Benchmarking results are shown in this paper for various relevant reactor and experimental devices. The evolution of the plasma vapor cloud is followed temporally and results are explained and commented as a function of the computational time needed and the accuracy of the calculation

  19. Non-uniform Erosion and Surface Evolution of Plasma-Facing Materials for Electric Propulsion

    Science.gov (United States)

    Matthes, Christopher Stanley Rutter

    A study regarding the surface evolution of plasma-facing materials is presented. Experimental efforts were performed in the UCLA Pi Facility, designed to explore the physics of plasma-surface interactions. The influence of micro-architectured surfaces on the effects of plasma sputtering is compared with the response of planar samples. Ballistic deposition of sputtered atoms as a result of geometric re-trapping is observed. This provides a self-healing mechanism of micro-architectured surfaces during plasma exposure. This result is quantified using a QCM to demonstrate the evolution of surface features and the corresponding influence on the instantaneous sputtering yield. The sputtering yield of textured molybdenum samples exposed to 300 eV Ar plasma is found to be roughly 1 of the 2 corresponding value of flat samples, and increases with ion fluence. Mo samples exhibited a sputtering yield initially as low as 0.22+/-8%, converging to 0.4+/-8% at high fluence. Although the yield is dependent on the initial surface structure, it is shown to be transient, reaching a steady-state value that is independent of initial surface conditions. A continuum model of surface evolution resulting from sputtering, deposition and surface diffusion is also derived to resemble the damped Kuramoto-Sivashinsky (KS) equation of non-linear dynamics. Linear stability analysis of the evolution equation provides an estimate of the selected wavelength, and its dependence on the ion energy and angle of incidence. The analytical results are confirmed by numerical simulations of the equation with a Fast Fourier Transform method. It is shown that for an initially flat surface, small perturbations lead to the evolution of a selected surface pattern that has nano- scale wavelength. When the surface is initially patterned by other means, the final resulting pattern is a competition between the "templated" pattern and the "self-organized" structure. Potential future routes of research are also

  20. Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Sugandhi, Ritesh, E-mail: ritesh@ipr.res.in; Swamy, Rajamannar, E-mail: rajamannar@ipr.res.in; Khirwadkar, Samir, E-mail: sameer@ipr.res.in

    2016-11-15

    Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

  1. Use of EPICS and Python technology for the development of a computational toolkit for high heat flux testing of plasma facing components

    International Nuclear Information System (INIS)

    Sugandhi, Ritesh; Swamy, Rajamannar; Khirwadkar, Samir

    2016-01-01

    Highlights: • An integrated approach to software development for computational processing and experimental control. • Use of open source, cross platform, robust and advanced tools for computational code development. • Prediction of optimized process parameters for critical heat flux model. • Virtual experimentation for high heat flux testing of plasma facing components. - Abstract: The high heat flux testing and characterization of the divertor and first wall components are a challenging engineering problem of a tokamak. These components are subject to steady state and transient heat load of high magnitude. Therefore, the accurate prediction and control of the cooling parameters is crucial to prevent burnout. The prediction of the cooling parameters is based on the numerical solution of the critical heat flux (CHF) model. In a test facility for high heat flux testing of plasma facing components (PFC), the integration of computations and experimental control is an essential requirement. Experimental physics and industrial control system (EPICS) provides powerful tools for steering controls, data simulation, hardware interfacing and wider usability. Python provides an open source alternative for numerical computations and scripting. We have integrated these two open source technologies to develop a graphical software for a typical high heat flux experiment. The implementation uses EPICS based tools namely IOC (I/O controller) server, control system studio (CSS) and Python based tools namely Numpy, Scipy, Matplotlib and NOSE. EPICS and Python are integrated using PyEpics library. This toolkit is currently under operation at high heat flux test facility at Institute for Plasma Research (IPR) and is also useful for the experimental labs working in the similar research areas. The paper reports the software architectural design, implementation tools and rationale for their selection, test and validation.

  2. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    Energy Technology Data Exchange (ETDEWEB)

    H.W.Kugel, M.G.Bell, H.Schneider, J.P.Allain, R.E.Bell, R Kaita, J.Kallman, S. Kaye, B.P. LeBlanc, D. Mansfield, R.E. Nygen, R. Maingi, J. Menard, D. Mueller, M. Ono, S. Paul, S.Gerhardt, R.Raman, S.Sabbagh, C.H.Skinner, V.Soukhanovskii, J.Timberlake, L.E.Zakharov, and the NSTX Research Team

    2010-01-25

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  3. Lithium coatings on NSTX plasma facing components and its effects on boundary control, core plasma performance, and operation

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Bell, M.G.; Schneider, H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, School of Nuclear Engineering, West Lafayette, IN 47907 (United States); Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Raman, R. [University of Washington, Seattle, WA 98195 (United States); Sabbagh, S. [Columbia University, New York, NY 10027 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

    2010-11-15

    NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.

  4. Disruption simulation experiments in a pulsed plasma accelerator - energy absorption and damage evolution on plasma facing materials

    International Nuclear Information System (INIS)

    Bolt, H.; Barabash, V.; Gervash, A.; Linke, J.; Lu, L.P.; Ovchinnikov, I.; Roedig, M.

    1995-01-01

    Plasma accelerators are used as test beds for disruption simulation experiments on plasma facing materials, because the incident energy fluxes and the discharge duration are of similar order as those expected during disruptions in ITER. The VIKA facility was used for the testing of materials under incident energies up to 5 kJ/cm 2 . Different carbon materials, SiC, stainless steel, TZM and tungsten have been tested. From the experimental results a scaling of the ablation with incident energy density was derived. The resulting ablation depth on carbon materials is roughly 2 μm per kJcm -2 of incident energy density. For metals this ablation is much higher due to the partial loss of the melt layer from splashing. For stainless steel an ablation depth of 9.5 μm per kJcm -2 was determined. The result of a linear scaling of the ablation depth with incident energy density is consistent with a previous calorimetric study. (orig.)

  5. Lithium Coatings on NSTX Plasma Facing Components and Its Effects On Boundary Control, Core Plasma Performance, and Operation

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Schneider, H.; Allain, J.P.; Bell, R.E.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P.; Mansfield, D.; Nygen, R.E.; Maingi, R.; Menard, J.; Mueller, D.; Ono, M.; Paul, S.; Gerhardt, S.; Raman, R.; Sabbagh, S.; Skinner, C.H.; Soukhanovskii, V.; Timberlake, J.; Zakharov, L.E.; NSTX Research Team

    2010-01-01

    NSTX high-power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a Liquid Lithium Divertor surface on the outer part of the lower divertor.

  6. Numerical studies of fast ion slowing down rates in cool magnetized plasma using LSP

    Science.gov (United States)

    Evans, Eugene S.; Kolmes, Elijah; Cohen, Samuel A.; Rognlien, Tom; Cohen, Bruce; Meier, Eric; Welch, Dale R.

    2016-10-01

    In MFE devices, rapid transport of fusion products from the core into the scrape-off layer (SOL) could perform the dual roles of energy and ash removal. The first-orbit trajectories of most fusion products from small field-reversed configuration (FRC) devices will traverse the SOL, allowing those particles to deposit their energy in the SOL and be exhausted along the open field lines. Thus, the fast ion slowing-down time should affect the energy balance of an FRC reactor and its neutron emissions. However, the dynamics of fast ion energy loss processes under the conditions expected in the FRC SOL (with ρe code, to examine the effects of SOL density and background B-field on the slowing-down time of fast ions in a cool plasma. As we use explicit algorithms, these simulations must spatially resolve both ρe and λDe, as well as temporally resolve both Ωe and ωpe, increasing computation time. Scaling studies of the fast ion charge (Z) and background plasma density are in good agreement with unmagnetized slowing down theory. Notably, Z-scaling represents a viable way to dramatically reduce the required CPU time for each simulation. This work was supported, in part, by DOE Contract Number DE-AC02-09CH11466.

  7. The Plasma Parameters and Geometry of Cool and Warm Active Region Loops

    Energy Technology Data Exchange (ETDEWEB)

    Xie, Haixia; Li, Bo; Huang, Zhenghua; Xia, Lidong; Fu, Hui; Mou, Chaozhou [Shandong Provincial Key Laboratory of Optical Astronomy and Solar-Terrestrial Environment, Institute of Space Sciences, Shandong University, Weihai, 264209 Shandong (China); Madjarska, Maria S.; Wiegelmann, Thomas [Max Planck Institute for Solar System Research, Justus-von-Liebig-Weg 3, D-37077, Göttingen (Germany)

    2017-06-10

    How the solar corona is heated to high temperatures remains an unsolved mystery in solar physics. In the present study we analyze observations of 50 whole active region loops taken with the Extreme-ultraviolet Imaging Spectrometer on board the Hinode satellite. Eleven loops were classified as cool loops (<1 MK) and 39 as warm loops (1–2 MK). We study their plasma parameters, such as densities, temperatures, filling factors, nonthermal velocities, and Doppler velocities. We combine spectroscopic analysis with linear force-free magnetic field extrapolation to derive the 3D structure and positioning of the loops, their lengths and heights, and the magnetic field strength along the loops. We use density-sensitive line pairs from Fe xii, Fe xiii, Si x, and Mg vii ions to obtain electron densities by taking special care of intensity background subtraction. The emission measure loci method is used to obtain the loop temperatures. We find that the loops are nearly isothermal along the line of sight. Their filling factors are between 8% and 89%. We also compare the observed parameters with the theoretical Rosner–Tucker–Vaiana (RTV) scaling law. We find that most of the loops are in an overpressure state relative to the RTV predictions. In a follow-up study, we will report a heating model of a parallel-cascade-based mechanism and will compare the model parameters with the loop plasma and structural parameters derived here.

  8. Baking and helium glow discharge cleaning of SST-1 tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, Pratibha; Khan, Ziauddin; Raval, Dilip

    2015-01-01

    Graphite plasma facing components (PFCs) were installed inside SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 X 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium (He) glow discharge cleaning (GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nanometers from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.48. In this paper, the results of effect of baking and He-GDC experiments of SST-1 will be presented in detail. (author)

  9. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    International Nuclear Information System (INIS)

    Semwal, P; Khan, Z; Raval, D C; Dhanani, K R; George, S; Paravastu, Y; Prakash, A; Thankey, P; Ramesh, G; Khan, M S; Saikia, P; Pradhan, S

    2017-01-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10 -5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail. (paper)

  10. Baking and helium glow discharge cleaning of SST-1 Tokamak with graphite plasma facing components

    Science.gov (United States)

    Semwal, P.; Khan, Z.; Raval, D. C.; Dhanani, K. R.; George, S.; Paravastu, Y.; Prakash, A.; Thankey, P.; Ramesh, G.; Khan, M. S.; Saikia, P.; Pradhan, S.

    2017-04-01

    Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 °C in a vacuum furnace operated below 1.0 × 10-5 mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at ≥ 250 °C. In SST-1 the PFCs were baked at 250 °C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail.

  11. Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

    International Nuclear Information System (INIS)

    Laan, J.G. van der; Akiba, M.; Seki, M.; Hassanein, A.; Tanchuk, V.

    1991-01-01

    An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating paramters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in the thermal quench phase. (orig.)

  12. Numerical simulation of strong evaporation and condensation for plasma-facing materials

    International Nuclear Information System (INIS)

    Kunugi, T.; Yasuda, H.

    1994-01-01

    The thermal response of the divertor plate to the hard plasma disruptions had been analyzed numerically by the two dimensional transient heat transfer code. There are several studies of the vapor shielding effects on the thermal response to the plasma disruption. However, it was pointed out some discrepancies among the numerical results calculated by U.S., EC and Japan for the same disruption conditions by van der Laan. One of the authors studied the sensitivity of some parameters (i.e., the temperature dependency of the thermal properties, an evaporation coefficient and a saturated condensation ratio) of disruption erosion analysis. Though the authors expected that the variations in evaporation models lead to the large variety of the erosion, they gave no significant effects on the surface temperature, the evaporation and melt-layer thickness. In this paper, the authors will describe the development of the numerical simulation codes for the strong evaporation and condensation from the plasma facing materials (PFMs) such as carbon, tungsten and beryllium

  13. Radiation damage and redeposited-layer formation on plasma facing materials in the TRIAM-1M

    International Nuclear Information System (INIS)

    Hirai, Takeshi; Tokunaga, Kazutoshi; Fujiwara, Tadashi; Yoshida, Naoaki; Itoh, Satoshi

    1997-01-01

    As an aim to obtain some informations of material damage at long time discharge and redeposited-layer formed by scrape off layer (SOL), two collector probe experiments were conducted by using Tokamak of Research Institute for Applied Mechanics (TRIAM-IM). As a result, radiation damage due to charge exchange neutral particles of more than 2 MeV high energy component flying from plasma was observed. And in either experiment, redeposited-layer formation due to deposite of impurity atoms in the plasma could be observed. In the first experiment, a redeposited-layer with fine crystalline particles was observed, which was formed to contain multi-component system of Fe, Cr and Ni and light elements O and C. And, in the second experiment, a redeposited-layer grain-grown in which main component was Mo was observed. Surface modification of plasma facing material such as above-mentioned damage induction, redeposited-layer formation, and so on, was thought to much affect deterioration of materials and recycling of hydrogen. (G.K.)

  14. 2D surface temperature measurement of plasma facing components with modulated active pyrometry

    International Nuclear Information System (INIS)

    Amiel, S.; Loarer, T.; Pocheau, C.; Roche, H.; Gauthier, E.; Aumeunier, M.-H.; Courtois, X.; Jouve, M.; Balorin, C.; Moncada, V.; Le Niliot, C.; Rigollet, F.

    2014-01-01

    In nuclear fusion devices, such as Tore Supra, the plasma facing components (PFC) are in carbon. Such components are exposed to very high heat flux and the surface temperature measurement is mandatory for the safety of the device and also for efficient plasma scenario development. Besides this measurement is essential to evaluate these heat fluxes for a better knowledge of the physics of plasma-wall interaction, it is also required to monitor the fatigue of PFCs. Infrared system (IR) is used to manage to measure surface temperature in real time. For carbon PFCs, the emissivity is high and known (ε ∼ 0.8), therefore the contribution of the reflected flux from environment and collected by the IR cameras can be neglected. However, the future tokamaks such as WEST and ITER will be equipped with PFCs in metal (W and Be/W, respectively) with low and variable emissivities (ε ∼ 0.1–0.4). Consequently, the reflected flux will contribute significantly in the collected flux by IR camera. The modulated active pyrometry, using a bicolor camera, proposed in this paper allows a 2D surface temperature measurement independently of the reflected fluxes and the emissivity. Experimental results with Tungsten sample are reported and compared with simultaneous measurement performed with classical pyrometry (monochromatic and bichromatic) with and without reflective flux demonstrating the efficiency of this method for surface temperature measurement independently of the reflected flux and the emissivity

  15. Tungsten: An option for divertor and main chamber plasma facing components in future fusion devices

    International Nuclear Information System (INIS)

    Neu, R.; Dux, R.; Kallenbach, A.; Maggi, C.F.; Puetterich, T.; Balden, M.; Eich, T.; Fuchs, J.C.; Gruber, O.; Herrmann, A.; Maier, H.; Mueller, H.W.; Pugno, R.; Radivojevic, I.; Rohde, V.; Sips, A.C.C.; Suttrop, W.; Ye, M.Y.; O'Mullane, M.; Whiteford, A.

    2005-01-01

    The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic and the cooling factor of W have been extended and refined. The W-coated surfaces represent now a fraction of 65% (24.8 m2). The only two major components which are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. A very successful remedy is the addition of central RF heating at the 20-30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10 -3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper, W coated divertor do not show higher W concentrations than comparable discharges in the lower C-based divertor. (author)

  16. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J. [Forschungszentrum Juelich (Germany). Inst. fuer Plasmaphysik

    2006-04-15

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

  17. Plasma facing materials and components for future fusion devices - development, characterization and performance under fusion specific loading conditions

    International Nuclear Information System (INIS)

    Linke, J.

    2006-01-01

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive RandD. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation

  18. Plasma facing materials and components for future fusion devices—development, characterization and performance under fusion specific loading conditions

    Science.gov (United States)

    Linke, J.

    2006-04-01

    The plasma exposed components in existing and future fusion devices are strongly affected by the plasma material interaction processes. These mechanisms have a strong influence on the plasma performance; in addition they have major impact on the lifetime of the plasma facing armour and the joining interface between the plasma facing material (PFM) and the heat sink. Besides physical and chemical sputtering processes, high heat quasi-stationary fluxes during normal and intense thermal transients are of serious concern for the engineers who develop reliable wall components. In addition, the material and component degradation due to intense fluxes of energetic neutrons is another critical issue in D-T-burning fusion devices which requires extensive R&D. This paper presents an overview on the materials development and joining, the testing of PFMs and components, and the analysis of the neutron irradiation induced degradation.

  19. Comparison of two cooling protocols for llama semen: with and without collagenase and seminal plasma in the medium.

    Science.gov (United States)

    Carretero, M I; Giuliano, S M; Arraztoa, C C; Santa Cruz, R C; Fumuso, F G; Neild, D M

    2017-08-01

    Seminal plasma (SP) of South American Camelids could interfere with the interaction of spermatozoa with the extenders; therefore it becomes necessary to improve semen management using enzymatic treatment. Our objective was to compare two cooling protocols for llama semen. Twelve ejaculates were incubated in 0.1% collagenase and then were divided into two aliquots. One was extended in lactose and egg yolk (LEY) (Protocol A: collagenase and SP present). The other aliquot was centrifuged, and the pellet was resuspended in LEY (Protocol B: collagenase and SP absent). Both samples were maintained at 5°C during 24 hr. Routine and DNA evaluations were carried out in raw and cooled semen. Both cooling protocols maintained sperm viability, membrane function and DNA fragmentation, with Protocol A showing a significantly lowered total and progressive motility (p semen samples prior to either cooling or freeze-thawing. © 2016 Blackwell Verlag GmbH.

  20. Evaluation of energy and particle impact on the plasma facing components in DEMO

    International Nuclear Information System (INIS)

    Igitkhanov, Yuri; Bazylev, Boris

    2012-01-01

    removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.

  1. Evaluation of energy and particle impact on the plasma facing components in DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Igitkhanov, Yuri, E-mail: juri.gitkhanov@ihm.fzk.de [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany); Bazylev, Boris [Karlsruhe Institute of Technology, IHM, Karlsruhe (Germany)

    2012-08-15

    -state operation heat transfer into the coolant must remain below the critical heat flux (CHF) to avoid the possible severe degradation of the coolant heat removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.

  2. Liquid Metals as Plasma-facing Materials for Fusion Energy Systems: From Atoms to Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stone, Howard A. [Princeton Univ., NJ (United States); Koel, Bruce E. [Princeton Univ., NJ (United States); Bernasek, Steven L. [Princeton Univ., NJ (United States); Carter, Emily A. [Princeton Univ., NJ (United States); Debenedetti, Pablo G. [Princeton Univ., NJ (United States); Panagiotopoulos, Athanassios Z. [Princeton Univ., NJ (United States)

    2017-06-23

    The objective of our studies was to advance our fundamental understanding of liquid metals as plasma-facing materials for fusion energy systems, with a broad scope: from atoms to tokamaks. The flow of liquid metals offers solutions to significant problems of the plasma-facing materials for fusion energy systems. Candidate metals include lithium, tin, gallium, and their eutectic combinations. However, such liquid metal solutions can only be designed efficiently if a range of scientific and engineering issues are resolved that require advances in fundamental fluid dynamics, materials science and surface science. In our research we investigated a range of significant and timely problems relevant to current and proposed engineering designs for fusion reactors, including high-heat flux configurations that are being considered by leading fusion energy groups world-wide. Using experimental and theoretical tools spanning atomistic to continuum descriptions of liquid metals, and bridging surface chemistry, wetting/dewetting and flow, our research has advanced the science and engineering of fusion energy materials and systems. Specifically, we developed a combined experimental and theoretical program to investigate flows of liquid metals in fusion-relevant geometries, including equilibrium and stability of thin-film flows, e.g. wetting and dewetting, effects of electromagnetic and thermocapillary fields on liquid metal thin-film flows, and how chemical interactions and the properties of the surface are influenced by impurities and in turn affect the surface wetting characteristics, the surface tension, and its gradients. Because high-heat flux configurations produce evaporation and sputtering, which forces rearrangement of the liquid, and any dewetting exposes the substrate to damage from the plasma, our studies addressed such evaporatively driven liquid flows and measured and simulated properties of the different bulk phases and material interfaces. The range of our studies

  3. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    International Nuclear Information System (INIS)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A; Constans, S; Merola, M; Riccardi, B

    2009-01-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  4. Repair of manufacturing defects in the armor of plasma facing units of the ITER Divertor Dome

    International Nuclear Information System (INIS)

    Litunovsky, Nikolay; Alekseenko, Evgeny; Kuznetsov, Vladimir; Lyanzberg, Dmitriy; Makhankov, Aleksey; Rulev, Roman

    2013-01-01

    Highlights: • Sporadic manufacturing defects in ITER Divertor Dome PFUs may be repaired. • We have developed a repair technique for ITER Divertor Dome PFUs. • Armor repair technique for ITER Divertor Dome PFUs is successfully tested. -- Abstract: The paper describes the repair procedure developed for removal of manufacturing defects occurring sporadically during armoring of plasma facing units (PFUs) of the ITER Divertor Dome. Availability of armor repair technique is prescribed by the procurement arrangement for the ITER Divertor Dome concluded in 2009 between the ITER Organization and the ITER Domestic Agency of Russia. The paper presents the detailed description of the procedure, data on its effect on the joints of the rest part of the armor and on the grain structure of the PFU heat sink. The results of thermocycling of large-scale Dome PFU mock-ups manufactured with demonstration of armor repair are also given

  5. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Science.gov (United States)

    Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

    2009-12-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  6. Development of novel tungsten processing technologies for electro-chemical machining (ECM) of plasma facing components

    International Nuclear Information System (INIS)

    Holstein, Nils; Krauss, Wolfgang; Konys, Juergen

    2011-01-01

    Plasma facing components for fusion applications must exhibit long-term stability under extreme conditions, and therefore material imperfections cannot be tolerated due to a high risk of technical failures. To prevent or abolish defects in refractory metals components during the manufacturing process, some methods of electro-chemical machining as S-ECM and C-ECM were developed, enabling both the processing of smooth plain defect-free surfaces of different geometry and the removal of bulk material for the shaping of three-dimensional structures, also without cracks. It is discussed, that tungsten ablation with accurate electro-chemical molding is very sensitive to the kind of electric current, and therefore current investigations focused also on the effects of frequency profiles on the sharpness of edge rounding.

  7. Repair of manufacturing defects in the armor of plasma facing units of the ITER Divertor Dome

    Energy Technology Data Exchange (ETDEWEB)

    Litunovsky, Nikolay, E-mail: nlitunovsky@sintez.niiefa.spb.su; Alekseenko, Evgeny; Kuznetsov, Vladimir; Lyanzberg, Dmitriy; Makhankov, Aleksey; Rulev, Roman

    2013-10-15

    Highlights: • Sporadic manufacturing defects in ITER Divertor Dome PFUs may be repaired. • We have developed a repair technique for ITER Divertor Dome PFUs. • Armor repair technique for ITER Divertor Dome PFUs is successfully tested. -- Abstract: The paper describes the repair procedure developed for removal of manufacturing defects occurring sporadically during armoring of plasma facing units (PFUs) of the ITER Divertor Dome. Availability of armor repair technique is prescribed by the procurement arrangement for the ITER Divertor Dome concluded in 2009 between the ITER Organization and the ITER Domestic Agency of Russia. The paper presents the detailed description of the procedure, data on its effect on the joints of the rest part of the armor and on the grain structure of the PFU heat sink. The results of thermocycling of large-scale Dome PFU mock-ups manufactured with demonstration of armor repair are also given.

  8. Thermal shock tests to qualify different tungsten grades as plasma facing material

    Science.gov (United States)

    Wirtz, M.; Linke, J.; Loewenhoff, Th; Pintsuk, G.; Uytdenhouwen, I.

    2016-02-01

    The electron beam device JUDITH 1 was used to establish a testing procedure for the qualification of tungsten as plasma facing material. Absorbed power densities of 0.19 and 0.38 GW m-2 for an edge localized mode-like pulse duration of 1 ms were chosen. Furthermore, base temperatures of room temperature, 400 °C and 1000 °C allow investigating the thermal shock performance in the brittle, ductile and high temperature regime. Finally, applying 100 pulses under all mentioned conditions helps qualifying the general damage behaviour while with 1000 pulses for the higher power density the influence of thermal fatigue is addressed. The investigated reference material is a tungsten product produced according to the ITER material specifications. The obtained results provide a general overview of the damage behaviour with quantified damage characteristics and thresholds. In particular, it is shown that the damage strongly depends on the microstructure and related thermo-mechanical properties.

  9. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A [CEA/IRFM, F-13108, Saint-Paul-lez-Durance (France); Constans, S [AREVA-NP, Le Creusot (France); Merola, M [ITER Organization, Cadarache (France); Riccardi, B [Fusion For Energy, Barcelona (Spain)], E-mail: frederic.escourbiac@cea.fr

    2009-12-15

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  10. Mechanical characterization of W-armoured plasma-facing components after thermal fatigue

    International Nuclear Information System (INIS)

    Serret, D; Richou, M; Missirlian, M; Loarer, T

    2011-01-01

    The future fusion device ITER is aimed at demonstrating the scientific and technical feasibility of fusion power. Tens of thousands of W-armoured plasma-facing components (PFCs) will be installed in the vertical targets of the ITER divertor and subjected to a high heat flux. The purpose of this paper is to present the results of mechanical and microstructural characterization of tungsten PFCs after thermal fatigue tests. On each component, Vickers hardness measurements are made. In parallel, the mean grain diameter in the corresponding zone of tungsten material is determined. The empirical Hall-Petch relation was adapted to experimental data. However, due to the plateau effect on recrystallization hardness, this relation does not seem to be relevant once recrystallization is complete: a new approach is proposed for predicting the margin to the tungsten melting onset.

  11. Near-surface thermal characterization of plasma facing components using the 3-omega method

    International Nuclear Information System (INIS)

    Dechaumphai, Edward; Barton, Joseph L.; Tesmer, Joseph R.; Moon, Jaeyun; Wang, Yongqiang; Tynan, George R.; Doerner, Russell P.; Chen, Renkun

    2014-01-01

    Near-surface regime plays an important role in thermal management of plasma facing components in fusion reactors. Here, we applied a technique referred to as the ‘3ω’ method to measure the thermal conductivity of near-surface regimes damaged by ion irradiation. By modulating the frequency of the heating current in a micro-fabricated heater strip, the technique enables the probing of near-surface thermal properties. The technique was applied to measure the thermal conductivity of a thin ion-irradiated layer on a tungsten substrate, which was found to decrease by nearly 60% relative to pristine tungsten for a Cu ion dosage of 0.2 dpa

  12. The manufacture of carbon armoured plasma-facing components for fusion devices

    International Nuclear Information System (INIS)

    Schedler, B.; Huber, T.; Zabernig, A.; Rainer, F.; Scheiber, K.H.; Schedle, D.

    2001-01-01

    Within the last decade Plansee has been active in the development and manufacture of different plasma-facing-components for nuclear fusion experiments consisting in a tungsten or CFC-armor joined onto metallic substrates like TZM, stainless steel or copper-alloys. The manufacture of these components requires unique joining technologies in order to obtain reliable thermo mechanical stable joints able to withstand highest heat fluxes without any deterioration of the joint. In an overview the different techniques will be presented by some examples of components already manufactured and successfully tested under high heat flux conditions. Furthermore an overview will be given on the manufacture of different high heat flux components for TORE SUPRA, Wendelstein 7-X and ITER. (author)

  13. The impact of transient thermal loads on beryllium as plasma facing material

    Energy Technology Data Exchange (ETDEWEB)

    Spilker, Benjamin Christof

    2017-01-24

    The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO{sub 2} free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m{sup -2} range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby

  14. The impact of transient thermal loads on beryllium as plasma facing material

    International Nuclear Information System (INIS)

    Spilker, Benjamin Christof

    2017-01-01

    The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO_2 free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m"-"2 range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby, the

  15. Evaporation of Droplets in Plasma Spray-Physical Vapor Deposition Based on Energy Compensation Between Self-Cooling and Plasma Heat Transfer

    Science.gov (United States)

    Liu, Mei-Jun; Zhang, Meng; Zhang, Qiang; Yang, Guan-Jun; Li, Cheng-Xin; Li, Chang-Jiu

    2017-10-01

    In the plasma spray-physical vapor deposition process (PS-PVD), there is no obvious heating to the feedstock powders due to the free molecular flow condition of the open plasma jet. However, this is in contrast to recent experiments in which the molten droplets are transformed into vapor atoms in the open plasma jet. In this work, to better understand the heating process of feedstock powders in the open plasma jet of PS-PVD, an evaporation model of molten ZrO2 is established by examining the heat and mass transfer process of molten ZrO2. The results reveal that the heat flux in PS-PVD open plasma jet (about 106 W/m2) is smaller than that in the plasma torch nozzle (about 108 W/m2). However, the flying distance of molten ZrO2 in the open plasma jet is much longer than that in the plasma torch nozzle, so the heating in the open plasma jet cannot be ignored. The results of the evaporation model show that the molten ZrO2 can be partly evaporated by self-cooling, whereas the molten ZrO2 with a diameter <0.28 μm and an initial temperature of 3247 K can be completely evaporated within the axial distance of 450 mm by heat transfer.

  16. Developments toward the use of tungsten as armour material in plasma facing components promoted by Euratom-CEA Association

    International Nuclear Information System (INIS)

    Mitteau, R.; Missiaen, J.M.; Brustolin, P.

    2006-01-01

    Tungsten is increasingly considered as a prime candidate armour material facing the plasma in fusion experiments (ASDEX, JET, ITER). This material is, however, a challenge for the engineers due to its brittleness at room temperature. Its bonding to structural or cooled substrates is a critical issue. The Euratom-CEA Association promotes the development of evolutionary techniques aiming to produce high performance assemblies between tungsten and various substrates. These are 1) functionally graded tungsten to copper, 2) direct electron beam welding of tungsten to Mo-alloy TZM and 3) the characterisation of tungsten coatings deposited on carbon fibre composite by high energy deposition processes. 1) A functionally graded material eliminates the singular point which weakens the heterogeneous assembly, reducing the stresses and allowing a better behaviour. The sintering of submicronic W-Cu powders is investigated. The green shape is processed from W-CuO powder, which is reduced by a hydrogen flow. The compaction and sintering of layers of various compositions (10 to 30 % Cu) produces an assembly (density of ∼ 94%) with a good cohesion. However, the gradient is not effectively controlled, because of the migration of melt copper during the sintering. Future work aims to improve the process by using spark or microwave assisted sintering. 2) Electron beam welding of Mo-alloy TZM is investigated, to produce high temperature components required by radiation cooled PFCs. They require only mechanical properties and no vacuum sealing. The driving line is to use simple tungsten shapes to reduce the milling cost. In spite of low weldable properties of the refractory alloys, a good bonding up to a depth of 5 mm is obtained. Hardness measurements show that the melt area and the heat affected zone are harder than TZM, the weakest materials at 230 Hv. Quench tests in water from up to 2000 o C are done without apparent crack formation. 3) Finally, characterisation techniques are

  17. Design, fabrication and testing of an improved high heat flux element, experience feedback on steady state plasma facing components in Tore Supra

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Chatelier, M.; Durocher, A.; Guilheim, D.; Lipa, M.; Mitteau, R.; Tonon, G.; Tsitrone, E.

    1998-01-01

    Actively cooled plasma facing components (PFC) have been developed and used in Tore Supra since 1985. One of the main technological problem is due to the expansion mismatch between graphite armour and metallic heat sink material. A first technology used graphite tiles with or without a reinforcement and a compliant layer, brazed with titanium copper-silver (TiCuAg) alloy. The next technology used carbon fiber material (CFC) tiles with a 2 mm pure copper compliant layer, since the good mechanical strength of the CFC allowed the reinforcement layer to be suppressed. No destructive inspection during the manufacturing procedure was found to be essential to insure a good reliability of the elements. (orig.)

  18. Design and development of a LIBS system on linear plasma device PSI-2 for in situ real-time diagnostics of plasma-facing materials

    Directory of Open Access Journals (Sweden)

    X. Jiang

    2017-08-01

    Full Text Available Laser induced breakdown spectroscopy (LIBS is a strong candidate for detecting and monitoring the H/D/T content on the surface of plasma facing components (PFCs due to its capability of fast direct in situ measurement in extreme environment (e.g., vacuum, magnetic field, long distance, complex geometry. To study the feasibilities and encounter the challenges of LIBS on plasma devices, a LIBS system has been set up on the linear plasma device PSI-2. A number of key parameters including laser energy, the influence of magnetic field and the persistence of laser induced plasma are studied. Real-time measurements of deuterium outgassing on tungsten samples exposed to deuterium plasma of 1025 D/m2 are performed in the first 40–130 min after plasma exposure. The experimental results are compared to the calculations in the literature.

  19. Proceedings of 2nd Internaitonal workshop on tritium effects in plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Morita, Kenji [Nagoya Univ. (Japan). School of Engineering; Noda, Nobuaki [eds.

    1994-08-01

    This workshop was held at Nagoya University on May 19 and 20, 1994. Approximately 1/3 of the lectures discussed the migration and retention of tritium in graphite and the other forms of carbon. As to this topic, most of the different aspects of the tritium reactions with carbon were generally agreed on. At the temperature lower than 800 K, tritium plasma interacts with graphite by forming a saturated layer on the surface, by forming a codeposited layer of sputtered carbon and tritium, and by allowing tritium diffusion through Pores. At the temperature higher than 800 K, the principal reaction of tritium with carbon is intergranular diffusion with high energy trapping. Because beryllium is the reference plasma-facing material for the ITER, several presentations on the reactions of tritium with beryllium were made. Also the tritium permeation through other metals was the topics. The results of TFTR D-T experiment were reported in the first talk. In this book, the gists of these lectures are collected. (K.I.).

  20. Irradiation-induced structure and property changes in tokamak plasma-facing, carbon-carbon composites

    International Nuclear Information System (INIS)

    Burchell, T.D.

    1994-01-01

    Carbon-carbon composites are an attractive choice for fusion reactor plasma-facing components because of their low atomic number, superior thermal shock resistance, and low neutron activation. Next generation plasma fusion reactors, such as the International Thermonuclear Experimental Reactor (ITER), will require advanced carbon-carbon composite materials possessing high thermal conductivity to manage the anticipated severe heat loads. Moreover, ignition machines such as ITER will produce large neutron fluxes. Consequently, the influence of neutron damage on the structure and properties of carbon-carbon composite materials must be evaluated. Data from two irradiation experiments are reported and discussed here. Carbon-carbon composite materials were irradiated in target capsules in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). A peak damage dose of 4.7 displacements per atom (dpa) at 600 degree C was attained. The carbon materials irradiated included uni-directional, two-directional, and three-directional carbon-carbon composites. Dimensional changes are reported for the composite materials and are related to single crystal dimensional changes through fiber and composite structural models. Moreover, the irradiation-induced dimensional changes are reported and discussed in terms of their architecture, fiber type, and graphitization temperature. The effect of neutron irradiation on thermal conductivity of two three-directional, carbon-carbon composites is reported and the recovery of thermal conductivity due to thermal annealing is discussed

  1. Electromagnetic and structural analyses of the vacuum vessel and plasma facing components for EAST

    International Nuclear Information System (INIS)

    Xu, Weiwei; Liu, Xufeng; Song, Yuntao; Li, Jun; Lu, Mingxuan

    2013-01-01

    Highlights: • The electromagnetic and structural responses of VV and PFCs for EAST are analyzed. • A detailed finite element model of the VV including PFCs is established. • The two most dangerous scenarios, major disruptions and downward VDEs are considered. • The distribution patterns of eddy currents, EMFs and torques on PFCs are analyzed. -- Abstract: During plasma disruptions, time-varying eddy currents are induced in the vacuum vessel (VV) and Plasma Facing Components (PFCs) of EAST. Additionally, halo currents flow partly through these structures during the vertical displacement events (VDEs). Under the high magnetic field circumstances, the resulting electromagnetic forces (EMFs) and torques are large. In this paper, eddy currents and EMFs on EAST VV, PFCs and their supports are calculated by analytical and numerical methods. ANSYS software is employed to evaluate eddy currents on VV, PFCs and their structural responses. To learn the electromagnetic and structural response of the whole structure more accurately, a detailed finite element model is established. The two most dangerous scenarios, major disruptions and downward VDEs, are examined. It is found that distribution patterns of eddy currents for various PFCs differ greatly, therefore resulting in different EMFs and torques. It can be seen that for certain PFCs the transient reaction force are severe. Results obtained here may set up a preliminary foundation for the future dynamic response research of EAST VV and PFCs which will provide a theoretical basis for the future engineering design of tokamak devices

  2. On the generation of runaway electrons and their impact to plasma facing components

    International Nuclear Information System (INIS)

    Kawamura, Takaichi; Obayashi, Haruo; Miyahara, Akira.

    1988-06-01

    Runaway electrons accompanied by inductive or non-inductive plasma currents in a tokamak have severe interactions with plasma facing materials of a first wall, and influence the first wall structure due to activation and damage. In this paper, modelling of runaway electron generation near the wall in a tokamak is carried out. This includes the evaluation of acceleration along magnetic surfaces for relativistic electrons with energies larger than the runaway threshold. Penetration of runaway electrons of energy ranges from a few MeV to several ten MeV leads to gamma ray photon production by bremsstrahlung. One of the specific features of the impact on the first wall technology is that they give rise to activation due to giant resonance of the (γ,n) nuclear reaction and, as a consequence, cause a requirement of remote maintenance. The other is that they bring energy deposition at brazing areas between low Z material and metal, or at a metal itself, and they result in melting, cracking and grain growth. The methods to estimate these effects using nuclear data and material data on the basis of runaway flux modelling are introduced and examples of estimation are given. (author)

  3. Lifetime evaluation of plasma-facing materials during a tokamak disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1995-09-01

    Erosion losses of plasma-facing materials in a tokamak reactor during major disruptions, giant ELMS, and large power excursions are serious concerns that influence component survivability and overall lifetime. Two different mechanisms lead to material erosion during these events: surface vaporization and loss of the melt layer. Hydrodynamics and radiation transport in the rapidly developed vapor-cloud region above the exposed area are found to control and determine the net erosion thickness from surface vaporization. A comprehensive self-consistent kinetic model has been developed in which the time-dependent optical properties and the radiation field of the vapor cloud are calculated in order to correctly estimate the radiation flux at the divertor surface. The developed melt layer of metallic divertor materials will, however, be free to move and can be eroded away due to various forces. , Physical mechanisms that affect surface vaporization and cause melt layer erosion are integrated in a comprehensive model. It is found that for metallic components such as beryllium and tungsten, lifetime due to these abnormal events will be controlled and dominated by the evolution and hydrodynamics of the melt layer during the disruption. The dependence of divertor plate lifetime on various aspects of plasma/material interaction physics is discussed

  4. Proceedings of 2nd International workshop on tritium effects in plasma facing components

    International Nuclear Information System (INIS)

    Morita, Kenji; Noda, Nobuaki

    1994-08-01

    This workshop was held at Nagoya University on May 19 and 20, 1994. Approximately 1/3 of the lectures discussed the migration and retention of tritium in graphite and the other forms of carbon. As to this topic, most of the different aspects of the tritium reactions with carbon were generally agreed on. At the temperature lower than 800 K, tritium plasma interacts with graphite by forming a saturated layer on the surface, by forming a codeposited layer of sputtered carbon and tritium, and by allowing tritium diffusion through Pores. At the temperature higher than 800 K, the principal reaction of tritium with carbon is intergranular diffusion with high energy trapping. Because beryllium is the reference plasma-facing material for the ITER, several presentations on the reactions of tritium with beryllium were made. Also the tritium permeation through other metals was the topics. The results of TFTR D-T experiment were reported in the first talk. In this book, the gists of these lectures are collected. (K.I.)

  5. The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Garrison, L. M., E-mail: garrisonlm@ornl.gov; Egle, B. J. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, Tennessee 37831 (United States); Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States); Zenobia, S. J.; Kulcinski, G. L.; Santarius, J. F. [Fusion Technology Institute, University of Wisconsin-Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States)

    2016-08-15

    The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10{sup 14} ions/(cm{sup 2} s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

  6. Experimental studies of lithium-based surface chemistry for fusion plasma-facing materials applications

    International Nuclear Information System (INIS)

    Allain, J.P.; Rokusek, D.L.; Harilal, S.S.; Nieto-Perez, M.; Skinner, C.H.; Kugel, H.W.; Heim, B.; Kaita, R.; Majeski, R.

    2009-01-01

    Lithium has enhanced the operational performance of fusion devices such as: TFTR, CDX-U, FTU, T-11 M, and NSTX. Lithium in the solid and liquid state has been studied extensively in laboratory experiments including its erosion and hydrogen-retaining properties. Reductions in physical sputtering up to 40-60% have been measured for deuterated solid and liquid lithium surfaces. Computational modeling indicates that up to a 1:1 deuterium volumetric retention in lithium is possible. This paper presents the results of systematic in situ laboratory experimental studies on the surface chemistry evolution of ATJ graphite under lithium deposition. Results are compared to post-mortem analysis of similar lithium surface coatings on graphite exposed to deuterium discharge plasmas in NSTX. Lithium coatings on plasma-facing components in NSTX have shown substantial reduction of hydrogenic recycling. Questions remain on the role lithium surface chemistry on a graphite substrate has on particle sputtering (physical and chemical) as well as hydrogen isotope recycling. This is particularly due to the lack of in situ measurements of plasma-surface interactions in tokamaks such as NSTX. Results suggest that the lithium bonding state on ATJ graphite is lithium peroxide and with sufficient exposure to ambient air conditions, lithium carbonate is generated. Correlation between both results is used to assess the role of lithium chemistry on the state of lithium bonding and implications on hydrogen pumping and lithium sputtering. In addition, reduction of factors between 10 and 30 reduction in physical sputtering from lithiated graphite compared to pure lithium or carbon is also measured.

  7. Elaboration of functionally graded materials for plasma facing components of the thermonuclear machines

    International Nuclear Information System (INIS)

    Autissier, Emmanuel

    2014-01-01

    The objective of this study was to develop a Functionally Graded Material (FGM) W/Cu to replace the compliance layer (Cu-OFHC) in the plasma facing components of thermonuclear fusion reactor like ITER. The peculiarity of this work is to elaborate these materials without exceeding the melting temperature of copper in order to control its microstructure. The co-sintering is the most attractive solution to achieve this goal. The first phase of this study has been to decrease the sintering temperature of the tungsten to achieve this co-sintering. The elaboration of a Functionally Graded Materials being delicate, thermomechanical calculations were performed in order to determine the number and chemical composition in order to increase the lifespan of Plasma Facing Components. Spark Plasma Sintering conditions were optimized in order to achieve maximum density of W x Cu 1-x composites. The effect of copper content and density of the W x Cu 1-x composites on thermal and mechanical properties was investigated. The SPS conditions were applied for W/CuCrZr assemblies with a compliance layer composed of several interlayers. The importance of time for the integrity of assemblies thereof has been highlighted. The study of the dwell time during W/CuCrZr assembly leads to identify a parameter to characterize the integrity of the interface regardless of the composition and the nature of the layer of compliance. Moreover, the phenomena associated with the formation of the interface assembly have been identified. The interface W/W x Cu 1-x is formed by the extrusion of the copper layer of the W x Cu 1-x inside the tungsten porosities. The W y Cu 1-y /CuCrZr interface is formed by copper migration of CuCrZr layer inside the W y Cu 1-y layer. Finally optimization assembly conditions showed that the mechanical stresses due to the densification of the Functionally Graded Materials can be limited by sintering the FGM before the assembly. (author)

  8. Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

    International Nuclear Information System (INIS)

    Lee, Youngmin; Ku, Duck Young; Ahn, Mu-Young; Cho, Seungyon; Park, Yi-Hyun; Lee, Dong Won

    2015-01-01

    Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

  9. Electromagnetic analysis on Korean Helium Cooled Ceramic Reflector (HCCR) TBM during plasma major disruption

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youngmin, E-mail: ymlee@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young; Ahn, Mu-Young; Cho, Seungyon; Park, Yi-Hyun [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) will be installed at the #18 equatorial port of the Vaccum Vessel in order to test the feasibility of the breeding blanket performance for forthcoming fusion power plant in the ITER TBM Program. Since ITER tokamak contains Vaccum Vessel and set of electromagnetic coils, the TBM as well as other components is greatly influenced by magnetic field generated by these coils. By the electromagnetic (EM) fast transient events such as major disruption (MD), vertical displacement event (VDE) or magnet fast discharge (MFD) occurred in tokamak system, the eddy current can be induced eventually in the conducting components. As a result, the magnetic field and induced eddy current produce extremely huge EM load (force and moment) on the TBM. Therefore, EM load calculation is one of the most important analyses for optimized design of TBM. In this study, a 20-degree sector model for tokamak system including central solenoid (CS) coil, poloidal field (PF) coil, toroidal field (TF) coil, vaccum vessel, shield blankets and TBM set (TBM, TBM key, TBM shield, TBM frame) is prepared for analysis by ANSYS-EMAG tool. Concerning the installation location of the TBM, a major disruption scenario is particularly applied for fast transient analysis. The final goal of this study is to evaluate the EM load on HCCR TBM during plasma major disruption.

  10. Particle exhaust of helium plasmas with actively cooled outboard pump limiter on Tore Supra

    International Nuclear Information System (INIS)

    Uckan, T.; Mioduszewski, P.K.; Loarer, T.; Chatelier, M.; Guilhem, D.; Lutz, T.; Nygren, R.E.; Mahdavi, M.A.

    1995-08-01

    The superconducting tokamak Tore Supra was designed for long-pulse (30-s) high input power operation. Here observations on the particle-handling characteristics of the actively cooled modular outboard pump limiter (OPL) are presented for helium discharges. The important experimental result was that a modest pumping speed (1 m 3 /s) of the OPL turbomolecular pump (TMP) provided background helium exhaust. This result came about due to a well-conditioned vessel wall with helium discharges that caused no wall outgasing. The particle accountability in these helium discharges was excellent, and the well-conditioned wall did not play a significant role in the particle balance. The helium density control, 25% density drop with OPL exhaust efficiency of ∼1%, was possible with TMP although this may not be the case with reactive gases such as deuterium. The observed quadratic increase of the OPL neutral pressure with helium density was consistent with an improvement of the particle control with increasing plasma density

  11. Principle of a 'retro reflecting' or 'backfire' LH antenna. A robust efficiently cooled launching structure for LH waves in reactor grade plasmas

    International Nuclear Information System (INIS)

    Bibet, Ph.; Litaudon, X.; Moreau, D.

    1994-01-01

    Up to now, lower hybrid current drive launchers have been made of waveguide arrays. To extrapolate present techniques despite the thermal power load and the mechanical stress, it is necessary to develop an advanced launcher concept which is simpler and more robust. To simplify the antennas between RF vacuum windows and plasma, studies have been carried, for example, the hyperguide for JET, the toroidal oversized waveguide for JT-60U and the poloidal mode converter for Tore Supra. For plasma-facing components, it has been proposed to use diffraction through rod arrays although this suffers from bad coupling per pass. Here a new concept is given which seems to be robust, and should leave enough space for efficient water cooling. The front part near plasma can be easily replaced by remote handling. Its principle is explained, and the description is made. The coupling properties of the proposed structure have been studied by using the SWAN code, and good agreement between the SWAN calculation and the coupling measurement performed at Tore Supra has been reported previously. (K.I.)

  12. Liquid metals as a divertor plasma-facing material explored using the Pilot-PSI and Magnum-PSI linear devices

    Science.gov (United States)

    Morgan, T. W.; Rindt, P.; van Eden, G. G.; Kvon, V.; Jaworksi, M. A.; Lopes Cardozo, N. J.

    2018-01-01

    For DEMO and beyond, liquid metal plasma-facing components are considered due to their resilience to erosion through flowed replacement, potential for cooling beyond conduction and inherent immunity to many of the issues of neutron loading compared to solid materials. The development curve of liquid metals is behind that of e.g. tungsten however, and tokamak-based research is currently somewhat limited in scope. Therefore, investigation into linear plasma devices can provide faster progress under controlled and well-diagnosed conditions in assessing many of the issues surrounding the use of liquid metals. The linear plasma devices Magnum-PSI and Pilot-PSI are capable of producing DEMO-relevant plasma fluxes, which well replicate expected divertor conditions, and the exploration of physics issues for tin (Sn) and lithium (Li) such as vapour shielding, erosion under high particle flux loading and overall power handling are reviewed here. A deeper understanding of erosion and deposition through this work indicates that stannane formation may play an important role in enhancing Sn erosion, while on the other hand the strong hydrogen isotope affinity reduces the evaporation rate and sputtering yields for Li. In combination with the strong redeposition rates, which have been observed under this type of high-density plasma, this implies that an increase in the operational temperature range, implying a power handling range of 20-25 MW m-2 for Sn and up to 12.5 MW m-2 for Li could be achieved. Vapour shielding may be expected to act as a self-protection mechanism in reducing the heat load to the substrate for off-normal events in the case of Sn, but may potentially be a continual mode of operation for Li.

  13. Characterization of a segmented plasma torch assisted High Heat Flux (HHF) system for performance evaluation of plasma facing components in fusion devices

    International Nuclear Information System (INIS)

    Ngangom, Aomoa; Sarmah, Trinayan; Sah, Puspa; Kakati, Mayur; Ghosh, Joydeep

    2015-01-01

    A wide variety of high heat and particle flux test facilities are being used by the fusion community to evaluate the thermal performance of plasma facing materials/components, which includes electron beam, ion beam, neutral beam and thermal plasma assisted sources. In addition to simulate heat loads, plasma sources have the additional advantage of reproducing exact fusion plasma like conditions, in terms of plasma density, temperature and particle flux. At CPP-IPR, Assam, we have developed a high heat and particle flux facility using a DC, non-transferred, segmented thermal plasma torch system, which can produce a constricted, stabilized plasma jet with high ion density. In this system, the plasma torch exhausts into a low pressure chamber containing the materials to be irradiated, which produces an expanded plasma jet with more uniform profiles, compared to plasma torches operated at atmospheric pressure. The heat flux of the plasma beam was studied by using circular calorimeters of different diameters (2 and 3 cm) for different input power (5-55 kW). The effect of the change in gas (argon) flow rate and mixing of gases (argon + hydrogen) was also studied. The heat profile of the plasma beam was also studied by using a pipe calorimeter. From this, the radial heat flux was calculated by using Abel inversion. It is seen that the required heat flux of 10 MW/m 2 is achievable in our system for pure argon plasma as well as for plasma with gas mixtures. The plasma parameters like the temperature, density and the beam velocity were studied by using optical emission spectroscopy. For this, a McPherson made 1.33 meter focal length spectrometer; model number 209, was used. A plane grating with 1800 g/mm was used which gave a spectral resolution of 0.007 nm. A detailed characterization with respect to these plasma parameters for different gas (argon) flow rate and mixing of gases (argon+hydrogen) for different input power will be presented in this paper. The plasma

  14. PFMC14. 14th international conference on plasma-facing materials and components for fusion applications. Book of abstracts

    International Nuclear Information System (INIS)

    2013-01-01

    The performance of fusion devices and of a future fusion power plant critically depends on the plasma facing materials and components. Resistance to local heat and particle loads, thermo-mechanical properties, as well as the response to neutron damage of the selected materials are critical parameters which need to be understood and tailored from atomistic to component levels. The 14th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues. Among the topics of the joint conference recent developments and research results in the following fields are addressed: - Tungsten and tungsten alloys - Low-Z materials - Mixed materials - Erosion, redeposition and fuel retention - Materials under extreme thermal loads - Technology and testing of plasma-facing components - Neutron effects in plasma-facing materials - Advanced characterization of materials and components. Selected international speakers present overview lectures and treat detailed aspects of the given topics. Contributed papers to the subjects of the meeting are solicited for oral and poster presentations.

  15. RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part I: Theory and description of model capabilities

    Science.gov (United States)

    Raffray, A. René; Federici, Gianfranco

    1997-04-01

    RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case.

  16. RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Pt. I. Theory and description of model capabilities

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.

    1997-01-01

    For pt.II see ibid., p.101-30, 1997. RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case. (orig.)

  17. Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

    Science.gov (United States)

    Klimov, N. S.; Putrik, A. B.; Linke, J.; Pitts, R. A.; Zhitlukhin, A. M.; Kuprianov, I. B.; Spitsyn, A. V.; Ogorodnikova, O. V.; Podkovyrov, V. L.; Muzichenko, A. D.; Ivanov, B. V.; Sergeecheva, Ya. V.; Lesina, I. G.; Kovalenko, D. V.; Barsuk, V. A.; Danilina, N. A.; Bazylev, B. N.; Giniyatulin, R. N.

    2015-08-01

    PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic-martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, "corrugated" surface, with hills and valleys spaced by 0.2-2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

  18. Ti-doped isotropic graphite: A promising armour material for plasma-facing components

    Science.gov (United States)

    García-Rosales, C.; López-Galilea, I.; Ordás, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

    2009-04-01

    Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of ˜200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

  19. Ti-doped isotropic graphite: A promising armour material for plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Rosales, C. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain)], E-mail: cgrosales@ceit.es; Lopez-Galilea, I.; Ordas, N. [CEIT and Tecnun (University of Navarra), Paseo de Manuel Lardizabal, 15, E-20018 San Sebastian (Spain); Adelhelm, C.; Balden, M. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Pintsuk, G. [Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany); Grattarola, M.; Gualco, C. [Ansaldo Ricerche S.p.A., I-16152 Genoa (Italy)

    2009-04-30

    Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of {approx}200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

  20. Ti-doped isotropic graphite: A promising armour material for plasma-facing components

    International Nuclear Information System (INIS)

    Garcia-Rosales, C.; Lopez-Galilea, I.; Ordas, N.; Adelhelm, C.; Balden, M.; Pintsuk, G.; Grattarola, M.; Gualco, C.

    2009-01-01

    Finely dispersed Ti-doped isotropic graphites with 4 at.% Ti have been manufactured using synthetic mesophase pitch 'AR' as raw material. These new materials show a thermal conductivity at room temperature of ∼200 W/mK and flexural strength close to 100 MPa. Measurement of the total erosion yield by deuterium bombardment at ion energies and sample temperatures for which pure carbon shows maximum values, resulted in a reduction of at least a factor of 4, mainly due to dopant enrichment at the surface caused by preferential erosion of carbon. In addition, ITER relevant thermal shock loads were applied with an energetic electron beam at the JUDITH facility. The results demonstrated a significantly improved performance of Ti-doped graphite compared to pure graphite. Finally, Ti-doped graphite was successfully brazed to a CuCrZr block using a Mo interlayer. These results let assume that Ti-doped graphite can be a promising armour material for divertor plasma-facing components.

  1. Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component

    International Nuclear Information System (INIS)

    Martinez-Ona, Rafael; Garcia, Monica; Medrano, Mercedes

    2009-01-01

    The divertor is one of the most challenging components of ITER machine. Its plasma facing components contain thousands of joints that should be assessed to demonstrate their integrity during the required lifetime. Ultrasonic (US) techniques have been developed to study the capability of defect detection and to control the quality and degradation of these interfaces after the manufacturing process. Three types of joints made of carbon fibre composite to copper alloy, tungsten to copper alloy, and copper-to-copper alloy with two types of configurations have been studied. More than 100 samples representing these configurations and containing implanted flaws of different sizes have been examined. US techniques developed are detailed and results of validation samples examination before and after high heat flux (HHF) tests are presented. The results show that for W monoblocks the US technique is able to detect, locate and size the degradations in the two sample joints; for CFC monoblocks, the US technique is also able to detect, locate and size the calibrated defects in the two joints before the HHF, however after the HHF test the technique is not able to reliably detect defects in the CFC/Cu joint; finally, for the W flat tiles the US technique is able to detect, locate and size the calibrated defects in the two joints before HHF test, nevertheless defect location and sizing are more difficult after the HHF test.

  2. Non-destructive testing of bonded structures for plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan)]. E-mail: masanori_onozuka@mhi.co.jp; Kikuchi, K. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Kirihigashi, A. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Oda, Y. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Shimizu, K. [Mitsubishi Heavy Industries Ltd., Nuclear Systems Engineering Department, Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan)

    2005-11-15

    A preliminary investigation has been conducted to examine the applicability of the ultrasonic testing (UT) inspection technique for bonded structures in plasma facing components. In this study, existing UT probes have been used. Three test samples to simulate the blanket first-wall panel were fabricated. Artificial defects were applied along the diffusively bonded interfaces of the samples. Three types of UT probes have been tested. A vertical UT probe with 10 MHz, and a phased-array UT probe with 5 MHz, were used to detect defects between the Cu-alloy plates, and between the Cu-alloy plate and the stainless-steel (SS) block. The test results show that defects as small as 2 mm in size could be detected at a signal versus noise (S/N) ratio of more than 2. To detect defects along the SS pipes, a beam-focused-type UT probe with 20 MHz, has been applied. It was found that defects as small as 1 mm were identified at an S/N ratio of more than 2. While the results of the tested techniques were good, optimization of the probe systems is required before it can be concluded that such methods are most applicable for use on the bonded structures.

  3. Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Klimov, N.S., E-mail: klimov@triniti.ru [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Putrik, A.B. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Linke, J. [Forschungszentrum Jülich GmbH, EURATOM Association, Jülich D-52425 (Germany); Pitts, R.A. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Zhitlukhin, A.M. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kuprianov, I.B. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Spitsyn, A.V. [NRC «Kurchatov Institute», Akademika Kurchatova pl., 1, Moscow 123182 (Russian Federation); Ogorodnikova, O.V. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Ivanov, B.V.; Sergeecheva, Ya.V.; Lesina, I.G. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Kovalenko, D.V.; Barsuk, V.A.; Danilina, N.A. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Bazylev, B.N. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Giniyatulin, R.N. [Efremov Institute, Doroga na Metallostroy, 3 bld., Metallostroy, Saint-Petersburg 196641 (Russian Federation)

    2015-08-15

    PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic–martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, “corrugated” surface, with hills and valleys spaced by 0.2–2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

  4. Study on Energetic Ions Behavior in Plasma Facing Materials at Lower Temperature

    International Nuclear Information System (INIS)

    Morimoto, Y.; Sugiyama, T.; Akahori, S.; Kodama, H.; Tega, E.; Sasaki, M.; Oyaidu, M.; Kimura, H.; Okuno, K.

    2003-01-01

    An apparatus equipped with X-ray Photoelectron Spectroscopy (XPS) and Thermal Desorption Spectroscopy (TDS) was constructed to study interactions of energetic hydrogen isotopes with plasma facing materials. It is a remarkable feature of the apparatus that energetic ion implantation is carried out at around 150K to study reactions of energetic ions with matrix by suppressing the reactions of thermalized ions. Using this apparatus, TDS experiments for pyrolytic graphite implanted with energetic D 2 ions at 173 and 373K were carried out. The experimental results suggest that the deuterium implanted was released through a four-step release processes, involving three D 2 and one CD x (x = 2, 3 and 4) desorption processes. Two deuterium and CD x desorption processes were observed in the temperature range from 700 to 1200 K. In addition, a new deuterium desorption process was observed for the deuterium-implanted sample at 173 K. This has never been observed for deuterium-implanted graphite implanted at temperatures higher than room temperature

  5. Evaluation of thermo-mechanical properties data of carbon-based plasma facing materials

    International Nuclear Information System (INIS)

    Ulrickson, M.; Barabash, V.R.; Matera, R.; Roedig, M.; Smith, J.J.; Janev, R.K.

    1991-03-01

    This Report contains the proceedings, results and conclusions of the work done and the analysis performed during the IAEA Consultants' Meeting on ''Evaluation of thermo-mechanical properties data of carbon-based plasma facing materials'', convened on December 17-21, 1990, at the IAEA Headquarters in Vienna. Although the prime objective of the meeting was to critically assess the available thermo-mechanical properties data for certain types of carbon-based fusion relevant materials, the work of the meeting went well beyond this task. The meeting participants discussed in depth the scope and structure of the IAEA material properties database, the format of data presentation, the most appropriate computerized system for data storage, retrieval, exchange and management. The existing IAEA ALADDIN system was adopted as a convenient tool for this purpose and specific ALADDIN labelling schemes and dictionaries were established for the material properties data. An ALADDIN formatted test-file for the thermo-physical and thermo-mechanical properties of pyrolytic graphite is appended to this Report for illustrative purposes. (author)

  6. Optimization of tungsten-steel joints for plasma facing components in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Heuer, Simon; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, Juelich (Germany); Weber, Thomas; Linke, Jochen [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Werkstoffstruktur und -eigenschaften, Juelich (Germany); Matejicek, Jiri [Institute of Plasma Physics, Academy of Sciences of the Czech Republic, Prague (Czech Republic)

    2015-07-01

    Tungsten, joint to a martensitic-ferritic EUROFER97 structure, is a promising plasma facing material composite for fusion reactors. Due to the effect of mismatch in thermo-mechanical properties direct bonding is not feasible. Current research is therefore ongoing on interlayer systems. While the adhesion was already improved by the utilization of a discrete Cu, Ti or V interlayer, that is able to relax stresses by plastic deformation, joints still do not resist the expected load cycles in a fusion reactor. Therefore, alternatives for the interface are needed. This contribution presents research on functionally graded materials (FGM). The particular microstructure of a graded interlayer allows re-distributing macro stresses from a discrete interface to a greater volume while avoiding in particular Cu which tends to swell under neutron irradiation. A parameter study on the basis of finite element analysis will be presented as well as first results of several processing routes for FGM that shall be evaluated and benchmarked by mechanical as well as thermal testing.

  7. First wall and divertor plate disposed facing to plasma of thermonuclear device

    International Nuclear Information System (INIS)

    Araki, Masanori; Suzuki, Satoshi; Akiba, Masato; Hayata, Yoshiho; Inoue, Taiji; Hayashi, Yukihiro; Kude, Yukinori

    1998-01-01

    In order to make the most of characteristics of each ingredient of carbon fiber-reinforced composite materials, carbon fiber unidirectionally reinforced materials and a carbon fiber three-directionally reinforced material are laminated in the direction of the thickness to form a carbon fiber-reinforced carbon composite material. In this case, the carbon fibers are continuously oriented in the direction of the thickness to constitute the carbon fiber reinforced carbon composite materials integrally. In addition, a carbon fiber-reinforced carbon composite material prepared by bonding a metal on one surface in adjacent with the unidirectional carbon fiber reinforced portion and substantially in perpendicular to the direction of the thickness of the unidirectional carbon fiber reinforced portion is used as a main constitutional material. Further, a metal tube is buried in the carbon fiber three-directionally reinforced carbon composite material. Then, a first wall and a divertor plate excellent in thermal impact resistance to be disposed facing to plasmas of a thermonuclear device can be provided. (N.H.)

  8. Toward sub-Kelvin resistive cooling and non destructive detection of trapped non-neutral electron plasma

    Science.gov (United States)

    Di Domizio, S.; Krasnický, D.; Lagomarsino, V.; Testera, G.; Vaccarone, R.; Zavatarelli, S.

    2015-01-01

    A resonant circuit tuned to a particular frequency of the motion of charged particles stored in a Penning trap and connected to a low noise amplifier allows, at the same time, cooling and non destructive detection of the particles. Its use is widely diffused when single or few particles are stored near the centre of a hyperbolic Penning trap. We present a consistent model that predicts the shape of the induced signal when the tuned circuit is used to detect and cool the axial motion of a cold non neutral plasma stored in an open-ended cylindrical Penning trap. The model correctly accounts for the not negligible axial plasma size. We show that the power spectrum of the signal measured across the tuned circuit provides information about the particle number and insights about the plasma temperature. We report on the design of a HEMT-based cryogenic amplifier working at 14.4 MHz and 4.2 K and the results of the noise measurements. We have measured a drain current noise in the range from 6 to 17 pA/√Hz, which corresponds to an increase of the tuned circuit equivalent temperature of at maximum 0.35 K. The cryogenic amplifier has a very low power consumption from few tens to few hundreds of μW corresponding to a drain current in the range 100-800 μ A. An additional contribution due to the gate noise has been identified when the drain current is below 300 μA above that value an upper limit of the increase of the equivalent tuned circuit temperature due to this contribution of 0.02 K has been obtained. These features make the tuned circuit connected to this amplifier a promising device for detecting and cooling the axial motion of an electron plasma when the Penning trap is mounted inside a dilution refrigerator.

  9. Exploring liquid metal plasma facing component (PFC) concepts-Liquid metal film flow behavior under fusion relevant magnetic fields

    International Nuclear Information System (INIS)

    Narula, M.; Abdou, M.A.; Ying, A.; Morley, N.B.; Ni, M.; Miraghaie, R.; Burris, J.

    2006-01-01

    The use of fast moving liquid metal streams or 'liquid walls' as a plasma contact surface is a very attractive option and has been looked upon with considerable interest over the past several years, both by the plasma physics and fusion engineering programs. Flowing liquid walls provide an ever replenishing contact surface to the plasma, leading to very effective particle pumping and surface heat flux removal. A key feasibility issue for flowing liquid metal plasma facing component (PFC) systems, pertains to their magnetohydrodynamic (MHD) behavior under the spatially varying magnetic field environment, typical of a fusion device. MHD forces hinder the development of a smooth and controllable liquid metal flow needed for PFC applications. The present study builds up on the ongoing research effort at UCLA, directed towards providing qualitative and quantitative data on liquid metal free surface flow behavior under fusion relevant magnetic fields

  10. Feasibility of arc-discharge and plasma-sputtering methods in cleaning plasma-facing and diagnostics components of fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hakola, Antti, E-mail: antti.hakola@vtt.fi [VTT Technical Research Centre of Finland, VTT (Finland); Likonen, Jari [VTT Technical Research Centre of Finland, VTT (Finland); Karhunen, Juuso; Korhonen, Juuso T. [Department of Applied Physics, Aalto University (Finland); Aints, Märt; Laan, Matti; Paris, Peeter [Department of Physics, University of Tartu (Estonia); Kolehmainen, Jukka; Koskinen, Mika; Tervakangas, Sanna [DIARC-Technology Oy, Espoo (Finland)

    2015-10-15

    Highlights: • Feasibility of the arc-discharge and plasma-sputtering techniques in removing deposited layers from ITER-relevant samples demonstrated. • Samples with the size of an A4 paper can be cleaned from 1-μm thick deposited layers in 10–20 minutes by the arc-discharge method. • The plasma-sputtering method is 5–10 times slower but the resulting surfaces are very smooth. • Arc-discharge method could be used for rapid cleaning of plasma-facing components during maintenance shutdowns of ITER, plasma sputtering is preferred for diagnostics mirrors. - Abstract: We have studied the feasibility of arc-discharge and plasma-sputtering methods in removing deposited layers from ITER-relevant test samples. Prototype devices have been designed and constructed for the experiments and the cleaning process is monitored by a spectral detection system. The present version of the arc-discharge device is capable of removing 1-μm thick layers from 350-mm{sup 2} areas in 4–8 s, but due to the increased roughness of the cleaned surfaces and signs of local melting, mirror-like surfaces cannot be treated by this technique. The plasma-sputtering approach, for its part, is some 5–10 times slower in removing the deposited layers but no changes in surface roughness or morphology of the samples could be observed after the cleaning phase. The arc-discharge technique could therefore be used for rapid cleaning of plasma-facing components during maintenance shutdowns of ITER while in the case of diagnostics mirrors plasma sputtering is preferred.

  11. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    International Nuclear Information System (INIS)

    Greuner, Henri; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-01-01

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m"2 heat load for e.g. DEMO. • HHF tests up to 20 MW/m"2 were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m"2 and screening tests up to 20 MW/m"2 were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  12. Results of high heat flux testing of W/CuCrZr multilayer composites with percolating microstructure for plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Greuner, Henri, E-mail: henri.greuner@ipp.mpg.de; Zivelonghi, Alessandro; Böswirth, Bernd; You, Jeong-Ha

    2015-10-15

    Highlights: • Improvement of the performance of plasma-facing components made of W and CuCrZr. • Functionally graded composite at the interface of W and CuCrZr to mitigate the CTE. • A three-layer composite system (W volume fraction: 70/50/30%) was developed. • Design of water-cooled divertor components up to 20 MW/m{sup 2} heat load for e.g. DEMO. • HHF tests up to 20 MW/m{sup 2} were successfully performed. - Abstract: Reliable joining of tungsten to copper is a major issue in the design of water-cooled divertor components for future fusion reactors. One of the suggested advanced engineering solutions is to use functionally graded composite interlayers. Recently, the authors have developed a novel processing route for fabricating multi-layer graded W/CuCrZr composites. Previous characterization confirmed that the composite materials possess enhanced strength compared to the matrix alloy and shows reasonable ductility up to 300 °C indicating large potential to extend the operation temperature limit. Furthermore, a three-layer composite system (W volume fraction: 70/50/30%) was developed as a graded interlayer between the W armour and CuCrZr heat sink. In this study, we investigated the structural performance of the graded joint. Three water-cooled mock-ups of a flat tile type component were fabricated using electron beam welding and thermally loaded at the hydrogen neutral beam test facility GLADIS. Cycling tests at 10 MW/m{sup 2} and screening tests up to 20 MW/m{sup 2} were successfully performed and confirmed the expected thermal performance of the compound. The measured temperature values were in good agreement with the prediction of finite element analysis. Microscopic investigation confirmed the structural integrity of the newly developed functionally graded composite after these tests.

  13. 1st IAEA research co-ordination meeting on 'plasma-material interaction data for mixed plasma facing materials in fusion reactors'. Summary report

    International Nuclear Information System (INIS)

    Janev, R.K.; Longhurst, G.

    1998-12-01

    The proceedings and conclusions of the 1st IAEA Research Co-ordination Meeting on 'Plasma-Material Interaction Data for Mixed Plasma Facing Materials in Fusion Reactors', held on December 19 and 20, 1998 at the IAEA Headquarters in Vienna, are briefly described. This report includes a summary of the presentations made by meeting participants, a review of the data availability and data needs in the areas from the scope of the Co-ordinated Research Project (CRP) on the subject of the meeting, and recommendations regarding the future work within this CRP. (author)

  14. 2nd (final) IAEA research co-ordination meeting on 'plasma-material interaction data for mixed plasma facing materials in fusion reactors'. Summary report

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2001-11-01

    The proceedings and conclusions of the 2nd Research Co-ordination Meeting on 'Plasma-Material Interaction Data for Mixed Plasma Facing Materials in Fusion Reactors', held on October 16 and 17, 2000 at the IAEA Headquarters in Vienna, are briefly described. This report includes a summary of the presentations made by the meeting participants and a review of the accomplishments of the Co-ordinated Research Project (CRP). In addition, short summaries from the participants are included indicating the specific research completed in support of this CRP. (author)

  15. Heat loads on poloidal and toroidal edges of castellated plasma-facing components in COMPASS

    Science.gov (United States)

    Dejarnac, R.; Corre, Y.; Vondracek, P.; Gaspar, J.; Gauthier, E.; Gunn, J. P.; Komm, M.; Gardarein, J.-L.; Horacek, J.; Hron, M.; Matejicek, J.; Pitts, R. A.; Panek, R.

    2018-06-01

    Dedicated experiments have been performed in the COMPASS tokamak to thoroughly study the power deposition processes occurring on poloidal and toroidal edges of castellated plasma-facing components in tokamaks during steady-state L-mode conditions. Surface temperatures measured by a high resolution infra-red camera are compared with reconstructed synthetic data from a 2D thermal model using heat flux profiles derived from both the optical approximation and 2D particle-in-cell (PIC) simulations. In the case of poloidal leading edges, when the contribution from local radiation is taken into account, the parallel heat flux deduced from unperturbed, upstream measurements is fully consistent with the observed temperature increase at the leading edges of various heights, respecting power balance assuming simple projection of the parallel flux density. Smoothing of the heat flux deposition profile due to finite ion Larmor radius predicted by the PIC simulations is found to be weak and the power deposition on misaligned poloidal edges is better described by the optical approximation. This is consistent with an electron-dominated regime associated with a non-ambipolar parallel current flow. In the case of toroidal gap edges, the different contributions of the total incoming flux along the gap have been observed experimentally for the first time. They confirm the results of recent numerical studies performed for ITER showing that in specific cases the heat deposition does not necessarily follow the optical approximation. Indeed, ions can spiral onto the magnetically shadowed toroidal edge. Particle-in-cell simulations emphasize again the role played by local non-ambipolarity in the deposition pattern.

  16. Design and operation results of nitrogen gas baking system for KSTAR plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang-Tae [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Kim, Young-Jin, E-mail: k43689@nfri.re.kr [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Joung, Nam-Yong; Im, Dong-Seok; Kim, Kang-Pyo; Kim, Kyung-Min; Bang, Eun-Nam; Kim, Yaung-Soo [National Fusion Research Institute, 113 Gwahang-ro, Yuseong-gu, Daejeon 305-806 (Korea, Republic of); Yoo, Seong-Yeon [Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 305-764 (Korea, Republic of)

    2013-11-15

    Highlights: • Vacuum pressure in a vacuum vessel arrived at 7.24 × 10{sup −8} mbar. • PFC temperature was reached maximum 250 °C by gas temperature at 300 °C. • PFC inlet gas temperature was changed 5 °C per hour during rising and falling. • PFC gas balancing was made temperature difference among them below 8.3 °C. • System has a pre-cooler and a three-way valve to save operation energy. -- Abstract: A baking system for the Korea Superconducting Tokamak Advanced Research (KSTAR) plasma facing components (PFCs) is designed and operated to achieve vacuum pressure below 5 × 10{sup −7} mbar in vacuum vessel with removing impurities. The purpose of this research is to prevent the fracture of PFC because of thermal stress during baking the PFC, and to accomplish stable operation of the baking system with the minimum life cycle cost. The uniformity of PFC temperature in each sector was investigated, when the supply gas temperature was varied by 5 °C per hour using a heater and the three-way valve at the outlet of a compressor. The alternative of the pipe expansion owing to hot gas and the cage configuration of the three-way valve were also studied. During the fourth campaign of the KSTAR in 2011, nitrogen gas temperature rose up to 300 °C, PFC temperature reached at 250 °C, the temperature difference among PFCs was maintained at below 8.3 °C, and vacuum pressure of up to 7.24 × 10{sup −8} mbar was achieved inside the vacuum vessel.

  17. Low cycle thermal fatigue testing of beryllium grades for ITER plasma facing components

    International Nuclear Information System (INIS)

    Watson, R.D.; Youchison, D.L.; Dombrowski, D.E.; Guiniatouline, R.N.; Kupriynov, I.B.

    1996-01-01

    A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium, which is proposed as plasma facing armor for fusion reactor first wall, limiter, and divertor components. The 30 kW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature ''spike'' of 750 degree C for each pass of the beam. Large thermal stresses in excess of the yield strength are generated due to very high spot heat flux, 250 MW/m 2 . Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S- 65H, S-200F, S-200F-H, SR-200, I-400, extruded high purity, HIP'd spherical powder, porous beryllium (94% and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe 12 . Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation, and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT, and TShG-56. Rolled sheet Be (SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis

  18. Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

    International Nuclear Information System (INIS)

    Neu, R.; Riesch, J.; Coenen, J.W.; Brinkmann, J.; Calvo, A.; Elgeti, S.; García-Rosales, C.; Greuner, H.; Hoeschen, T.; Holzner, G.; Klein, F.; Koch, F.

    2016-01-01

    Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W_f/W) has been developed incorporating extrinsic toughening mechanisms. Small W_f/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO_3 compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

  19. Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Neu, R., E-mail: Rudolf.Neu@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Fakultät für Maschinenbau, Technische Universität München, D-85748 Garching (Germany); Riesch, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Coenen, J.W. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Brinkmann, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Calvo, A. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Elgeti, S. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); García-Rosales, C. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Greuner, H.; Hoeschen, T.; Holzner, G. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Klein, F. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Koch, F. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); and others

    2016-11-01

    Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W{sub f}/W) has been developed incorporating extrinsic toughening mechanisms. Small W{sub f}/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO{sub 3} compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

  20. Data for Erosion and Tritium Retention in Beryllium Plasma-Facing Materials. Summary Report of the First Research Coordination Meeting

    International Nuclear Information System (INIS)

    Braams, B.J.

    2013-04-01

    Nine experts in the field of plasma-wall interaction on beryllium surfaces together with IAEA staff met at IAEA Headquarters 26-28 September 2012 for the First Research Coordination Meeting of an IAEA Coordinated Research Project on data for erosion and tritium retention in beryllium plasma-facing materials. They described their on-going research, reviewed the main data needs and made plans for coordinated research during the remaining years of the project. The proceedings of the meeting are summarized in this report. (author)

  1. Induced charge of spherical dust particle on plasma-facing wall in non-uniform electric field

    International Nuclear Information System (INIS)

    Tomita, Y.; Smirnov, R.; Zhu, S.

    2005-01-01

    Induced charge of a spherical dust particle on a plasma-facing wall is investigated analytically, where non-uniform electric field is applied externally. The one-dimensional non-uniform electrostatic potential is approximated by the polynomial of the normal coordinate toward the wall. The bipolar coordinate is introduced to solve the Laplace equation of the induced electrostatic potential. The boundary condition at the dust surface determines the unknown coefficients of the general solution of the Laplace equation for the induced potential. From the obtained potential the surface induced charge can be calculated. This result allows estimating the effect of the surrounding plasma, which shields the induced charge. (author)

  2. Strongly emissive plasma-facing material under space-charge limited regime: Application to emissive probes

    Czech Academy of Sciences Publication Activity Database

    Cavalier, Jordan; Lemoine, N.; Bousselin, G.; Plihon, N.; Ledig, J.

    2017-01-01

    Roč. 24, č. 1 (2017), č. článku 013506. ISSN 1070-664X Institutional support: RVO:61389021 Keywords : plasma * tokamak * emissive probes Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.115, year: 2016 http://dx.doi.org/10.1063/1.4973557

  3. Cryo-cooled high-power window for high-frequency plasma heating

    International Nuclear Information System (INIS)

    Norajitra, P.; Bojarsky, E.; Reiser, H.; Haefner, H.E.

    1991-09-01

    Within the framework of gyrotron window development, IMF pursues the concept of a single-disk window with edge cooling. Compared to a double-disk window with surface cooling, this concept offers a number of advantages in terms of safety, reliability, and ease of design and fabrication. The coolants which can be used for this purpose are liquid nitrogen and helium at very low temperature. (orig.) [de

  4. Ion acoustic solitons and supersolitons in a magnetized plasma with nonthermal hot electrons and Boltzmann cool electrons

    Energy Technology Data Exchange (ETDEWEB)

    Rufai, O. R., E-mail: rajirufai@gmail.com; Bharuthram, R., E-mail: rbharuthram@uwc.ac.za [University of the Western Cape, Belville (South Africa); Singh, S. V., E-mail: satyavir@iigs.iigm.res.in; Lakhina, G. S., E-mail: lakhina@iigs.iigm.res.in [Indian Institute of Geomagnetism, New Panvel (W), Navi Mumbai (India)

    2014-08-15

    Arbitrary amplitude, ion acoustic solitons, and supersolitons are studied in a magnetized plasma with two distinct groups of electrons at different temperatures. The plasma consists of a cold ion fluid, cool Boltzmann electrons, and nonthermal energetic hot electrons. Using the Sagdeev pseudo-potential technique, the effect of nonthermal hot electrons on soliton structures with other plasma parameters is studied. Our numerical computation shows that negative potential ion-acoustic solitons and double layers can exist both in the subsonic and supersonic Mach number regimes, unlike the case of an unmagnetized plasma where they can only exist in the supersonic Mach number regime. For the first time, it is reported here that in addition to solitions and double layers, the ion-acoustic supersoliton solutions are also obtained for certain range of parameters in a magnetized three-component plasma model. The results show good agreement with Viking satellite observations of the solitary structures with density depletions in the auroral region of the Earth's magnetosphere.

  5. Development of high current density neutral beam injector with a low energy for interaction of plasma facing materials

    International Nuclear Information System (INIS)

    Nishikawa, Masahiro; Ueda, Yoshio; Goto, Seiichi

    1991-01-01

    A high current density neutral beam injector with a low energy has been developed to investigate interactions with plasma facing materials and propagation processes of damages. The high current density neutral beam has been produced by geometrical focusing method employing a spherical electrode system. The hydrogen beam with the current density of 140 mA/cm 2 has been obtained on the focal point in the case of the acceleration energy of 8 keV. (orig.)

  6. Plasma-wall interactions data compendium-1. ''Hydrogen retention property, diffusion and recombination coefficients database for selected plasma-facing materials''

    Energy Technology Data Exchange (ETDEWEB)

    Iwakiri, Hirotomo [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics; Matsuhiro, Kenjirou [Osaka Univ., Osaka (Japan); Hirooka, Yoshi [National Inst. for Fusion Science, Toki, Gifu (Japan); Yamamura, Yasunori [Okayama Univ. of Scinece, Okayama (Japan)

    2002-05-01

    A summary on the recent activities of the plasma-wall interactions database task group at the National Institute for Fusion Science is presented in this report. These activities are focused on the compilation of literature data on the key parameters related to wall recycling characteristics that affect dynamic particle balance during plasma discharges and also on-site tritium inventory. More specifically, in this task group a universal fitting formula has been proposed and successfully applied to help compile hydrogen implantation-induced retention data. Also, presented here are the data on hydrogen diffusion and surface recombination coefficients, both critical in modeling dynamic wall recycling behavior. Data compilation has been conducted on beryllium, carbon, tungsten and molybdenum, all currently used for plasma-facing components in magnetic fusion experiments. (author)

  7. Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.; Chiocchio, S.; Esser, B.; Dietz, J.; Igitkhanov, Y.; Janeschitz, G.

    1995-01-01

    This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC's) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m 2 ) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM's) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects the target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC's clad with different PFM's are discussed

  8. Binary-collision-approximation simulation for noble gas irradiation onto plasma facing materials

    International Nuclear Information System (INIS)

    Saito, Seiki; Nakamura, Hiroaki; Takayama, Arimichi; Ito, Atsushi M

    2014-01-01

    A number of experiments show that helium plasma constructs filament (fuzz) structures whose diameter is in nanometer-scale on the tungsten material under the suitable experimental condition. In this paper, binary-collision-approximation-based simulation is performed to reveal the mechanism and the conditions of fuzz formation of tungsten material under plasma irradiation. The irradiation of the plasma of hydrogen, deuterium, and tritium, and also the plasma of noble gas such as helium, neon, and argon atoms are investigated. The possibility of fuzz formation is discussed on the simulation result of penetration depth of the incident atoms

  9. Thermal conductivity reduction of tungsten plasma facing material due to helium plasma irradiation in PISCES using the improved 3-omega method

    Energy Technology Data Exchange (ETDEWEB)

    Cui, Shuang [Department of Mechanical and Aerospace Engineering, University of California, San Diego, La Jolla, CA 92093 (United States); Simmonds, Michael [Department of Physics, University of California, San Diego, La Jolla, CA 92093 (United States); Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States); Qin, Wenjing; Ren, Feng [School of Physics and Technology, Wuhan University, Wuhan, Hubei 430072 (China); Tynan, George R. [Department of Mechanical and Aerospace Engineering, University of California, San Diego, La Jolla, CA 92093 (United States); Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States); Doerner, Russell P. [Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States); Chen, Renkun, E-mail: rkchen@ucsd.edu [Department of Mechanical and Aerospace Engineering, University of California, San Diego, La Jolla, CA 92093 (United States); Center for Energy Research, University of California, San Diego, La Jolla, CA 92093 (United States)

    2017-04-01

    The near-surface region of plasma facing material (PFM) plays an important role in thermal management of fusion reactors. In this work, we measured thermal conductivity of tungsten (W) surface layers damaged by He plasma in PISCES at UCSD. We studied the damage effect on both bulk, and thin film, W. We observed that the surface morphology of both bulk and thin film was altered after exposure to He plasma with the fluence of 1 × 10{sup 26} m{sup −2} (bulk) and 2 × 10{sup 24} m{sup −2} (thin film). Transmission electron microscopy (TEM) analysis reveals that the depth of the irradiation damaged layer was approximately 20 nm on the bulk W exposed to He plasma at 773 K for 2000 s. In order to measure the thermal conductivity of this exceedingly thin damaged layer in the bulk W, we adopted the well-established ‘3-omega’ method and employed novel nanofabrication techniques to improve the measurement sensitivity. For the damaged W thin film sample, we measured the reduction in electrical conductivity and used the Wiedemann-Franz (W-F) law to extract the thermal conductivity. Results from both measurements show that thermal conductivity in the damaged layers was reduced by at least ∼80% compared to that of undamaged W. This large reduction in thermal conductivity can be attributed to the scattering of electrons, the dominant heat carriers in W, caused by defects introduced by He plasma irradiation.

  10. Progress in the engineering design and assessment of the European DEMO first wall and divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Barrett, Thomas R., E-mail: tom.barrett@ukaea.uk [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Ellwood, G.; Pérez, G.; Kovari, M.; Fursdon, M.; Domptail, F.; Kirk, S.; McIntosh, S.C.; Roberts, S.; Zheng, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Boccaccini, L.V. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); You, J.-H. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Bachmann, C. [EUROfusion, PPPT, Boltzmann Str. 2, 85748 Garching (Germany); Reiser, J.; Rieth, M. [KIT, IAM, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Visca, E.; Mazzone, G. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Arbeiter, F. [KIT, INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Domalapally, P.K. [Research Center Rez, Hlavní 130, 250 68 Husinec – Řež (Czech Republic)

    2016-11-01

    Highlights: • The engineering of the plasma facing components for DEMO is an extreme challenge. • PFC overall requirements, methods for assessment and designs status are described. • Viable divertor concepts for 10 MW/m{sup 2} surface heat flux appear to be within reach. • The first wall PFC concept will need to vary poloidally around the wall. • First wall coolant, structural material and PFC topology are open design choices. - Abstract: The European DEMO power reactor is currently under conceptual design within the EUROfusion Consortium. One of the most critical activities is the engineering of the plasma-facing components (PFCs) covering the plasma chamber wall, which must operate reliably in an extreme environment of neutron irradiation and surface heat and particle flux, while also allowing sufficient neutron transmission to the tritium breeding blankets. A systems approach using advanced numerical analysis is vital to realising viable solutions for these first wall and divertor PFCs. Here, we present the system requirements and describe bespoke thermo-mechanical and thermo-hydraulic assessment procedures which have been used as tools for design. The current first wall and divertor designs are overviewed along with supporting analyses. The PFC solutions employed will necessarily vary around the wall, depending on local conditions, and must be designed in an integrated manner by analysis and physical testing.

  11. Heat loads on JET plasma facing components from ICRF and LH wave absorption in the SOL

    Czech Academy of Sciences Publication Activity Database

    Jacquet, P.; Colas, L.; Mayoral, M.-L.; Arnoux, G.; Bobkov, V.; Brix, M.; Coad, P.; Czarnecka, A.; Dodt, D.; Durodie, F.; Ekedahl, A.; Frigione, D.; Fursdon, M.; Gauthier, E.; Goniche, M.; Graham, M.; Joffrin, E.; Korotkov, A.; Lerche, E.; Mailloux, J.; Monakhov, I.; Noble, C.; Ongena, J.; Petržílka, Václav; Portafaix, C.; Rimini, F.; Sirinelli, A.; Riccardo, V.; Vizvary, Z.; Widdowson, A.; Zastrow, K.-D.

    2011-01-01

    Roč. 51, č. 10 (2011), s. 103018-103018 ISSN 0029-5515 R&D Projects: GA ČR GA202/07/0044 Institutional research plan: CEZ:AV0Z20430508 Keywords : LH wave * plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.090, year: 2011 http://iopscience.iop.org/0029-5515/51/10/103018/pdf/0029-5515_51_10_103018.pdf

  12. 1st IAEA research coordination meeting on tritium retention in fusion reactor plasma facing components. October 5-6, 1995, Vienna, Austria. Summary report

    International Nuclear Information System (INIS)

    Langley, R.A.

    1995-12-01

    The proceedings and results of the 1st IAEA research Coordination Meeting on ''Tritium Retention in Fusion Reactor Plasma Facing Components'' held on October 5 and 6, 1995 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of a data survey and needs assessment for the retention, release and removal of tritium from plasma facing components, a summary of data evaluation, and recommendations regarding future work. (author). 4 tabs

  13. Development of non-destructive examination techniques for CFC-metal joints in annular geometry and their application to the manufacturing of plasma-facing components

    International Nuclear Information System (INIS)

    Di Pietro, E.; Visca, E.; Orsini, A.; Sacchetti, M.; Borruto, T.M.R.; Varone, P.; Vesprini, R.

    1995-01-01

    The design of plasma-facing components for ITER, as for any of the envisaged next-step machines, relies heavily on the use of brazed junctions to couple armour materials to the heat sink and cooling tubes. Moreover, the typical number of brazed components and the envisaged effects of local overheating due to failure in a single brazed junction stress the importance of having a set of NDE techniques developed that can ensure the flawless quality of the joint. The qualification and application of two NDE techniques (ultrasonic and thermographic analysis) for inspection of CFC-to-metal joints is described with particular regard to the annular geometry typical of macroblock/monoblock solutions for divertor high-heat-flux components. The results of the eddy current inspection are not reported. The development has been focused specifically on the joint between carbon-fiber composite and TZM molybdenum alloy; techniques for the production of reference defect samples have been devised and a set of reference defect samples produced. The comparative results of the NDE inspections are reported and discussed, also on the basis of the destructive examination of the samples. The nature and size of relevant and detectable defects are discussed together with hints for a possible NDE strategy for divertor high-heat-flux components

  14. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Manly, W.D. [Oak Ridge National Lab., TN (United States); Dombrowski, D.E. [Brush Wellman, Inc., Cleveland, OH (United States)] [and others

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers.

  15. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers

  16. A dynamic monitoring approach for the surface morphology evolution measurement of plasma facing components by means of speckle interferometry

    Science.gov (United States)

    Wang, Hongbei; Cui, Xiaoqian; Feng, Chunlei; Li, Yuanbo; Zhao, Mengge; Luo, Guangnan; Ding, Hongbin

    2017-11-01

    Plasma Facing Components (PFCs) in a magnetically confined fusion plasma device will be exposed to high heat load and particle fluxes, and it would cause PFCs' surface morphology to change due to material erosion and redeposition from plasma wall interactions. The state of PFCs' surface condition will seriously affect the performance of long-pulse or steady state plasma discharge in a tokamak; it will even constitute an enormous threat to the operation and the safety of fusion plasma devices. The PFCs' surface morphology evolution measurement could provide important information about PFCs' real-time status or damage situation and it would help to a better understanding of the plasma wall interaction process and mechanism. Meanwhile through monitoring the distribution of dust deposition in a tokamak and providing an upper limit on the amount of loose dust, the PFCs' surface morphology measurement could indirectly contribute to keep fusion operational limits and fusion device safety. Aiming at in situ dynamic monitoring PFCs' surface morphology evolution, a laboratory experimental platform DUT-SIEP (Dalian University of Technology-speckle interferometry experimental platform) based on the speckle interferometry technique has been constructed at Dalian University of Technology (DUT) in China. With directional specific designing and focusing on the real detection condition of EAST (Experimental Advanced Superconducting Tokamak), the DUT-SIEP could realize a variable measurement range, widely increased from 0.1 μm to 300 μm, with high spatial resolution (<1 mm) and ultra-high time resolution (<2 s for EAST measuring conditions). Three main components of the DUT-SIEP are all integrated and synchronized by a time schedule control and data acquisition terminal and coupled with a three-dimensional phase unwrapping algorithm, the surface morphology information of target samples can be obtained and reconstructed in real-time. A local surface morphology of the real divertor

  17. POD analysis of flow over a backward-facing step forced by right-angle-shaped plasma actuator.

    Science.gov (United States)

    Wang, Bin; Li, Huaxing

    2016-01-01

    This study aims to present flow control over the backward-facing step with specially designed right-angle-shaped plasma actuator and analyzed the influence of various scales of flow structures on the Reynolds stress through snapshot proper orthogonal decomposition (POD). 2D particle image velocimetry measurements were conducted on region (x/h = 0-2.25) and reattachment zone in the x-y plane over the backward-facing step at a Reynolds number of Re h  = 27,766 (based on step height [Formula: see text] and free stream velocity [Formula: see text]. The separated shear layer was excited by specially designed right-angle-shaped plasma actuator under the normalized excitation frequency St h  ≈ 0.345 along the 45° direction. The spatial distribution of each Reynolds stress component was reconstructed using an increasing number of POD modes. The POD analysis indicated that the flow dynamic downstream of the step was dominated by large-scale flow structures, which contributed to streamwise Reynolds stress and Reynolds shear stress. The intense Reynolds stress localized to a narrow strip within the shear layer was mainly affected by small-scale flow structures, which were responsible for the recovery of the Reynolds stress peak. With plasma excitation, a significant increase was obtained in the vertical Reynolds stress peak. Under the dimensionless frequencies St h  ≈ 0.345 and [Formula: see text] which are based on the step height and momentum thickness, the effectiveness of the flow control forced by the plasma actuator along the 45° direction was ordinary. Only the vertical Reynolds stress was significantly affected.

  18. Response of plasma facing components in Tokamaks due to intense energy deposition using Particle-In-Cell (PIC) methods

    Science.gov (United States)

    Genco, Filippo

    Damage to plasma-facing components (PFC) due to various plasma instabilities is still a major concern for the successful development of fusion energy and represents a significant research obstacle in the community. It is of great importance to fully understand the behavior and lifetime expectancy of PFC under both low energy cycles during normal events and highly energetic events as disruptions, Edge-Localized Modes (ELM), Vertical Displacement Events (VDE), and Run-away electron (RE). The consequences of these high energetic dumps with energy fluxes ranging from 10 MJ/m2 up to 200 MJ/m 2 applied in very short periods (0.1 to 5 ms) can be catastrophic both for safety and economic reasons. Those phenomena can cause a) large temperature increase in the target material b) consequent melting, evaporation and erosion losses due to the extremely high heat fluxes c) possible structural damage and permanent degradation of the entire bulk material with probable burnout of the coolant tubes; d) plasma contamination, transport of target material into the chamber far from where it was originally picked. The modeling of off-normal events such as Disruptions and ELMs requires the simultaneous solution of three main problems along time: a) the heat transfer in the plasma facing component b) the interaction of the produced vapor from the surface with the incoming plasma particles c) the transport of the radiation produced in the vapor-plasma cloud. In addition the moving boundaries problem has to be considered and solved at the material surface. Considering the carbon divertor as target, the moving boundaries are two since for the given conditions, carbon doesn't melt: the plasma front and the moving eroded material surface. The current solution methods for this problem use finite differences and moving coordinates system based on the Crank-Nicholson method and Alternating Directions Implicit Method (ADI). Currently Particle-In-Cell (PIC) methods are widely used for solving

  19. Tritium distribution on plasma facing graphite tiles of JT-60U

    International Nuclear Information System (INIS)

    Tanabe, T.; Sugiyama, K.; Masaki, K.; Gotoh, Y.; Tobita, K.; Miya, N.

    2003-01-01

    Tritium distributions on the graphite divertor tiles, the dome units and the baffle plates of JT-60U were successfully measured. Poloidally, the highest tritium level was found at the dome top tiles and the outer baffle plates, where the plasma did not hit directly. On the other hand, although the toroidal tritium profiles on each tiles appeared uniform, detailed profiles in full toroidal direction clearly showed a periodic variation corresponding to the position of the magnetic field coils, indicating the ripple loss of high energy tritons as suggested by the OFMC code. Finally, the temperature increase owing to the plasma heat load was found to release the once retained tritium. (author)

  20. Thinning of N-face GaN (0001) samples by inductively coupled plasma etching and chemomechanical polishing

    International Nuclear Information System (INIS)

    Rizzi, F.; Gu, E.; Dawson, M. D.; Watson, I. M.; Martin, R. W.; Kang, X. N.; Zhang, G. Y.

    2007-01-01

    The processing of N-polar GaN (0001) samples has been studied, motivated by applications in which extensive back side thinning of freestanding GaN (FS-GaN) substrates is required. Experiments were conducted on FS-GaN from two commercial sources, in addition to epitaxial GaN with the N-face exposed by a laser lift-off process. The different types of samples produced equivalent results. Surface morphologies were examined over relatively large areas, using scanning electron microscopy and stylus profiling. The main focus of this study was on inductively coupled plasma (ICP) etch processes, employing Cl 2 /Ar or Cl 2 /BCl 3 Ar gas mixtures. Application of a standard etch recipe, optimized for feature etching of Ga-polar GaN (0001) surfaces, caused severe roughening of N-polar samples and confirmed the necessity for specific optimization of etch conditions for N-face material. A series of recipes with a reduced physical (sputter-based) contribution to etching allowed average surface roughness values to be consistently reduced to below 3 nm. Maximum N-face etch rates of 370-390 nm/min have been obtained in recipes examined to date. These are typically faster than etch rates obtained on Ga-face samples under the same conditions and adequate for the process flows of interest. Mechanistic aspects of the ICP etch process and possible factors contributing to residual surface roughness are discussed. This study also included work on chemomechanical polishing (CMP). The optimized CMP process had stock removal rates of ∼500 nm/h on the GaN N face. This was much slower than the ICP etching but showed the important capability of recovering smooth surfaces on samples roughened in previous processing. In one example, a surface roughened by nonoptimized ICP etching was smoothed to give an average surface roughness of ∼2 nm

  1. Development of unidirectional C/C composite with high thermal conductivity and its application to plasma facing materials

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Onozuka, Masanori; Ikeda, Takeshi; Akiba, Masato.

    1994-01-01

    Unidirectional C/C composite named 'MFC-1' with high conductivity was developed, and full-scale armor tiles were fabricated. The thermal conductivity in the direction perpendicular to the plasma-side surface is more than 300-500 W/m·degC, which is higher than those of other C/C composites ever made, even superior to that of pyrolytic carbon. It was shown by high heat load tests done using an electron beam test facility that the unidirectional C/C composite was very resistant against both surface erosion as well as severe thermal shock. The 'MFC-1' was successfully brazed to copper substrate, and its high thermal shock resistance was observed in heat load tests (20 MW/m 2 , 3s, not cooled). A functionally gradient material has been also developed as compliant layer for the MFC-1 bonded to copper. (author)

  2. Development of unidirectional C/C composite with high thermal conductivity and its application to plasma facing materials

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, Kimihiro (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Onozuka, Masanori; Ikeda, Takeshi; Akiba, Masato

    1994-03-01

    Unidirectional C/C composite named 'MFC-1' with high conductivity was developed, and full-scale armor tiles were fabricated. The thermal conductivity in the direction perpendicular to the plasma-side surface is more than 300-500 W/m[center dot]degC, which is higher than those of other C/C composites ever made, even superior to that of pyrolytic carbon. It was shown by high heat load tests done using an electron beam test facility that the unidirectional C/C composite was very resistant against both surface erosion as well as severe thermal shock. The 'MFC-1' was successfully brazed to copper substrate, and its high thermal shock resistance was observed in heat load tests (20 MW/m[sup 2], 3s, not cooled). A functionally gradient material has been also developed as compliant layer for the MFC-1 bonded to copper. (author).

  3. Preparation of erosion and deposition investigations on plasma facing components in Wendelstein 7-X

    Science.gov (United States)

    Dhard, C. P.; Balden, M.; Braeuer, T.; Brezinsek, S.; Coenen, J. W.; Dudek, A.; Ehrke, G.; Hathiramani, D.; Klose, S.; König, R.; Laux, M.; Linsmeier, Ch; Manhard, A.; Masuzaki, S.; Mayer, M.; Motojima, G.; Naujoks, D.; Neu, R.; Neubauer, O.; Rack, M.; Ruset, C.; Schwarz-Selinger, T.; Pedersen, T. Sunn; Tokitani, M.; Unterberg, B.; Yajima, M.; W7-X Team1, The

    2017-12-01

    In the Wendelstein 7-X stellarator with its twisted magnetic geometry the investigation of plasma wall interaction processes in 3D plasma configurations is an important research subject. For the upcoming operation phase i.e. OP1.2, three different types of material probes have been installed within the plasma vessel for the erosion/deposition investigations in selected areas with largely different expected heat load levels, namely, ≤10 MW m-2 at the test divertor units (TDU), ≤500 kW m-2 at the baffles, heat shields and toroidal closures and ≤100 kW m-2 at the stainless steel wall panels. These include 18 exchangeable target elements at TDU, about 30 000 screw heads at graphite tiles and 44 wafer probes on wall panels, coated with marker layers. The layer thicknesses, surface morphologies and the impurity contents were pre-characterized by different techniques and subjected to various qualification tests. The positions of these probes were fixed based on the strike line locations on the divertor predicted by field line diffusion and EMC3/EIRENE modeling calculations for the OP1.2 plasma configurations and availability of locations on panels in direct view of the plasma. After the first half of the operation phase i.e. OP1.2a the probes will be removed to determine the erosion/deposition pattern by post-mortem analysis and replaced by a new set for the second half of the operation phase, OP1.2b.

  4. Radiative cooling of a cilindrical Z-pinch in the stage of plasma shett motion

    International Nuclear Information System (INIS)

    Gerusov, A.V.; Imshennik, V.S.

    1982-01-01

    Dinamics of a cylindrically symmetric Z-pinch in deiterium-neon mixture on the motion stage with volume radiation losses of energy of neon ions is considered. A two-temperature MHD model of Z-pinch is numerically calculated with various percentage of gases in mixture. It is found that radiation losses are essential in the dynamics of the discharqe. They decrease distribution of magnetic field and curreqt density. The temoerature of anode surface heated by absorbtion of radiation from the plasma sheath which forms in the discharge is obtained. The value of anode surface temperature prior to the arrival of the shock front is insufficient for operating of previously proposed mechanism of X-ray plasma focus regime. Another mechanism leading to X-ray regime being due to radiative cooiang as aiso suggested

  5. Hot radial pressing: An alternative technique for the manufacturing of plasma-facing components

    International Nuclear Information System (INIS)

    Visca, E.; Libera, S.; Mancini, A.; Mazzone, G.; Pizzuto, A.; Testani, C.

    2005-01-01

    The Hot radial pressing (HRP) manufacturing technique is based on the radial diffusion bonding principle performed between the cooling tube and the armour tile. The bonding is achieved by pressurizing the cooling tube while the joining interface is kept at the vacuum and temperature conditions. This technique has been used for the manufacturing of relevant mock-ups of the ITER divertor vertical target. Tungsten monoblock mock-ups were successfully tested to high heat flux thermal fatigue (20 MW/m 2 of absorbed heat flux for 1000 cycles). After these good results the activity is now focused on the developing of a manufacturing process suitable also for the CFC monoblock mock-ups. A FE calculation was performed to investigate the stress involved in the CFC tiles during the process and to avoid the CFC fracture. The results obtained by the FE calculation and by the test performed in air simulating a HRP manufacturing process for a CFC monoblock mock-ups is reported in the paper

  6. Tungsten nitride coatings obtained by HiPIMS as plasma facing materials for fusion applications

    Czech Academy of Sciences Publication Activity Database

    Tiron, V.; Velicu, I. L.; Porosnicu, C.; Burducea, I.; Dinca, P.; Malinský, Petr

    Roč. 416, SEP (2017), s. 878-884 ISSN 0169-4332 R&D Projects: GA ČR(CZ) GBP108/12/G108; GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : Tugensten nitride layers * m-HIPIMS * deuterium retention * deuterium plasma jet * thermal desorption spectrometry Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 3.387, year: 2016

  7. Effect of heat loads on the plasma facing components of demo

    Energy Technology Data Exchange (ETDEWEB)

    Igitkhanov, Yu., E-mail: juri.igitkhanov@partner.kit.edu [ITEP, Karlsruhe Institute of Technology (Germany); Fetzer, R. [IHM, Karlsruhe Institute of Technology (Germany); Bazylev, B. [INR, Karlsruhe Institute of Technology (Germany)

    2016-11-01

    Highlights: • Under the DEMO1 stationary operation the nominal power fluxes along the magnetic field at the FW blanket modules is expected about 50 MW/m{sup 2}. • In the current design and averaged incident angle about 3–4.5° (similar to ITER) the engineering power load to the FW is expected within 2.5÷3.9 MW/m{sup 2}. • In the case of the unmitigated Type I ELMs unavoidable in the higher confinement H-mode of operation energy load per ELM is about 20 MJ/m{sup 2} along the field line, arriving at a frequency of 0.8 Hz with deposition time of 0.6 ms per each ELM. - Abstract: In this paper we analyse a thermo-hydraulic performance of the first wall blanket module during the stationary DEMO operation with the edge localized mode (ELM). Heat loads are estimated based on scaling arguments and predictions from the peeling-ballooning ELM model. Effect of parallel heat fluxes intersecting with the first wall panels and avoidance of overheating by inclination of the panels are considered. The material temperatures of the W/EUROFER sandwich type module with water cooling stainless steel tube and Cu alloy compliance embedded into EUROFER is calculated by using the MEMOS code. The calculations were carried out indicating the required geometric parameters as well as the cooling conditions which allow keeping materials temperatures within allowable engineering limits. Effect of inclination of the first wall plates to avoid the misalignment problems is considered.

  8. TPX heating and cooling system

    International Nuclear Information System (INIS)

    Kungl, D.J.; Knutson, D.S.; Costello, J.; Stoenescu, S.; Yemin, L.

    1995-01-01

    TPX, while having primarily super-conducting coils that do not require water cooling, still has very significant water cooling requirements for the plasma heating systems, vacuum vessel, plasma facing components, diagnostics, and ancillary equipment. This is accentuated by the 1000-second pulse requirement. Two major design changes, which have significantly affected the TPX Heating and Cooling System, have been made since the conceptual design review in March of 1993. This paper will discuss these changes and review the current status of the conceptual design. The first change involves replacing the vacuum vessel neutron shielding configuration of lead/glass composite tile by a much simpler and more reliable borated water shield. The second change reduces the operating temperature of the vacuum vessel from 150 C to ≥50 C. With this temperature reduction, all in-vessel components and the vessel will be supplied by coolant at a common ≥50 C inlet temperature. In all, six different heating and cooling supply requirements (temperature, pressure, water quality) for the various TPX components must be met. This paper will detail these requirements and provide an overview of the Heating and Cooling System design while focusing on the ramifications of the TPX changes described above

  9. Steady-state operation of magnetic fusion devices: Plasma control and plasma facing components. Report on the IAEA technical committee meeting held at Fukuoka, 25-29 October 1999

    International Nuclear Information System (INIS)

    Engelmann, F.

    2000-01-01

    An IAEA Technical Committee Meeting on Steady-State Operation of Magnetic Fusion Devices - Plasma Control and Plasma Facing Components was held at Fukuoka, Japan, from 25 to 29 October 1999. The meeting was the second IAEA Techical Committee Meeting on the subject, following the one held at Hefei, China, a year earlier. The meeting was attended by over 150 researchers from 10 countries

  10. Plasma cooling by metal snow. I. Origin of the effect and elimination procedures

    Energy Technology Data Exchange (ETDEWEB)

    Khan, I A; Dietz, K J; Waelbroeck, F; Wienhold, P [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Plasmaphysik

    1978-06-01

    In many plasma devices, steel or inconel walls are exposed to large flux densities phi/sub 1/ of atomic hydrogen particles which, soon after the reduction of the surface oxides and carbides has started, penetrate into the lattice. The stationary hydrogen concentration c/sub 1/ in the lattice is expressed as function of phi/sub 1/, of the wall temperature Tsub(w) and of the surface roughness factor sigma. It is found to be much larger than in an H/sub 2/ surrounding. Dissolved atoms recombining on internal surfaces (i.e. at grain boundaries) within the solid can then build up a considerable pressure Psup(p)sub(H/sub 2/) within resulting gas pockets; Psup(p)sub(H/sub 2/) should depend strongly on Tsub(w). Near room temperature, the computed values are such that surface-near pockets should crack open, releasing locally high pressure H/sub 2/ gas and some metal dust (impurity source) and increasing sigma. The expected distribution of csub(H) within the sponge-like structure which is expected to result from a prolonged exposure to phi/sub 1/ at low Tsub(w) is derived. Means of avoiding the plasma contamination by dust release are pointed out.

  11. NIFS joint research meeting on plasma facing components, PSI, and heat/particle control

    International Nuclear Information System (INIS)

    Yamashina, T.

    1997-10-01

    The LHD collaboration has been started in 1996. Particle and heat control is one of the categories for the collaboration, and a few programs have been nominated in these two years. A joint research meeting on PFC, PSI, heat and particle meeting was held at NIFS on June 27, 1997, in which present status of these programs were reported. This is a collection of the notes and view graphs presented in this meeting. Brief reviews and research plan of each program are included in relation to divertor erosion and sputtering, impurity generation, hydrogen recycling, edge plasma structure, edge transport and its control, heat removal, particle exhaust, wall conditioning etc. (author)

  12. Development of high conductive C/C composite tiles for plasma facing armor

    International Nuclear Information System (INIS)

    Ioki, K.; Namiki, K.; Tsujimura, S.; Toyoda, M.; Seki, M.; Takatsu, H.

    1991-01-01

    C/C composites with high thermal conductivity were developed in unidirectional, two-dimensional and felt types, and were fabricated as full-scale armor tile. Their thermal conductivity in the direction perpendicular to the plasma-side surface is 250∝550 W/mdeg C, that is comparable to that of pyrolytic graphite. It was shown by heat load tests that the C/C composites have low surface erosion characteristics and high thermal shock resistance. Various kinds of C/C composites were successfully bonded to metal substrate, and their mechanical strength and thermal shock resistance were tested. (orig.)

  13. On fractal properties of equipotentials over a real rough surface faced to plasma in fusion devices

    International Nuclear Information System (INIS)

    Budaev, V.P.; Yakovlev, M.

    2008-01-01

    We consider a sheath region bounded by a corrugated surface of material conductor and a flat boundary held to a constant voltage bias. The real profile of the film deposited from plasma on a limiter in a fusion device was used in numerical solving of the Poisson's equation to find a profile of electrostatic potential. The rough surface influences the equipotential lines over the surface. We characterized a shape of equipotential lines by a fractal dimension. The long-range correlation in the potential field is imposed by the non-trivial fractal structure of the surface. Dust particles bounced in such irregular potential field can accelerate due to the Fermi acceleration. (author)

  14. Effect of Three Days Storage of Coated Spermatozoa at Cooling and Adding Seminal Plasma on Ram Fertility

    Directory of Open Access Journals (Sweden)

    Alireza Vaferi

    2016-08-01

    Full Text Available Introduction Artificial insemination (AI has only been used as a supplement to natural mating. AI, when used in conjunction with accurate progeny testing schemes, can substantially increase the rate of genetic progress compared with that of natural service. Moreover, the use of AI causes the limitation of the transmitted diseases. Cervical insemination with frozen-thawed ram semen has not been widely adopted, probably because of the relative poor fertility obtained. Thus using fresh and diluted semen is only approach for performing AI. AI is currently limited by the poor fertility achieved after cervical insemination with the storage of liquid semen at sub-ambient temperature. The success of this procedure in sheep is restricted by the short length of time that ram sperm can be stored in a liquid state. Moreover, the effect of cooling on sperm differs depending on species. It is also well known that ram spermatozoa are more sensitive to cold-shock stress than those of other species. Seminal plasma, as physiological secretion, is a complex mixture of secretions originating from testis, epididymis and accessory sex glands which is mixed with epididymal sperm at ejaculation; it serves as the carrier of sperm to the female genital tract. This mixture contains numerous factors such as organic and nonorganic material which play an important role in the final maturation of the spermatozoa through hormonal, enzymatic and surface-modifying events. During natural mating, a mechanism may be activated to separate spermatozoa from seminal plasma. After being ejaculated into the vagina, sperm swim through cervical mucus and enter the uterus within minutes (>30 min; cervical mucus acts as a barrier for seminal plasma. In the artificial insemination industry, seminal plasma with all the useful and harmful components is not removed from semen and is in contact with sperm throughout cooling, freezing and storage. On the other hand, it was demonstrated that the

  15. Alternative electro-chemically based processing routes for joining of plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, Wolfgang, E-mail: wolfgang.krauss@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Materials Research III, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen (Germany); Lorenz, Julia; Holstein, Nils; Konys, Juergen [Karlsruhe Institute of Technology (KIT), Institute for Materials Research III, Hermann-von-Helmholtz-Platz 1 76344 Eggenstein-Leopoldshafen (Germany)

    2011-10-15

    Tungsten is considered in fusion technology as functional and structural material in the area of blanket and divertor for future application in DEMO. The KIT design of a He-cooled divertor includes joints between W and W-alloys as well as of W with Eurofer-steel. The main challenges range from expansion mismatch problem for tungsten/steel joints over metallurgical reactions with brittle phase formation to crack stopping ability and excellent surface wetting. These requirements were only met partly and insufficiently in the past e.g. by direct Cu-casting of tungsten onto steel. Both, the joining needs and the observed failure scenarios of conventionally joined components initiated the development of improved joining technologies based on electro-chemical processing routes. As electrolytes aqueous and aprotic, water free, system are integrated into this development line. In the first step principle requirements are presented to guarantee a reproducible and adherent deposition of scales based on Ni and Cu acting as inter layers and filler, respectively, to generate a real metallurgical bonding as demonstrate by 1100 deg. C joining tests. The development field aprotic systems based on ionic liquids is discussed with respect to enable development of refractory metal based fillers with focus high temperature W-W brazing.

  16. Chemical vapor deposition of SiC on C-C composites as plasma facing materials for fusion application

    International Nuclear Information System (INIS)

    Kim, W. J.; Lee, M. Y.; Park, J. Y.; Hong, G. W.; Kim, J. I.; Choi, D. J.

    2000-01-01

    Because of the low activation and excellent mechanical properties at elevated temperatures, carbon-fiber reinforced carbon(C-C) composites have received much attention for plasma facing materials for fusion reactor and high-temperature structural applications such as aircrafts and space vehicles. These proposed applications have been frustrated by the lack of resistance to hydrogen erosion and oxidation on exposure to ambient oxidizing conditions at high temperature. Although Silicon Carbide (SiC) has shown excellent properties as an effective erosion-and oxidation-protection coating, many cracks are developed during fabrication and thermal cycles in use due to the Coefficients of Thermal Expansion(CTE) mismatch between SiC and C-C composite. In this study, we adopted a pyrolitic carbon as an interlayer between SiC and C-C substrate in order to minimize the CTE mismatch. The oxidation-protection performance of this composite was investigated as well

  17. Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

    International Nuclear Information System (INIS)

    Singh, K.P.; Pandya, Santosh P.; Khirwadkar, S.S.; Patel, Alpesh; Patil, Y.; Buch, J.J.U.; Khan, M.S.; Tripathi, Sudhir; Pandya, Shwetang; Govindrajan, J.; Jaman, P.M.; Rathore, Devendra; Rangaraj, L.; Divakar, C.

    2011-01-01

    Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R and D area in fusion research. Pre-qualification tests for brazed joints between W-CuCrZr and C-CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m 2 uniform heat flux. Details about experimental and computational work are presented here.

  18. Pre-qualification of brazed plasma facing components of divertor target elements for ITER like tokamak application

    Energy Technology Data Exchange (ETDEWEB)

    Singh, K.P., E-mail: kpsingh@ipr.res.in [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat (India); Pandya, Santosh P.; Khirwadkar, S.S.; Patel, Alpesh; Patil, Y.; Buch, J.J.U.; Khan, M.S.; Tripathi, Sudhir; Pandya, Shwetang; Govindrajan, J. [Institute for Plasma Research, Bhat, Gandhinagar, Gujarat (India); Jaman, P.M.; Rathore, Devendra; Rangaraj, L.; Divakar, C. [Materials Science Division, National Aerospace Laboratories, CSIR, Bangalore, Karnataka (India)

    2011-10-15

    Qualification of tungsten (W) and graphite (C) based brazed plasma facing components (PFCs) is an important R and D area in fusion research. Pre-qualification tests for brazed joints between W-CuCrZr and C-CuCrZr using NDT (IR thermography and ultrasonic test) and thermal fatigue test are attempted. Mockups having good quality brazed joints of W and C based PFCs were identified using NDT. Subsequently, thermal fatigue test was performed on the identified mockups. All brazed tiles of W based PFC mockups could withstand thermal fatigue test, however, few tiles of C based PFC mockup were found detached. Thermal analyses of mockups are performed using finite element analysis (ANSYS) software to simulate the thermal hydraulic condition with 10 MW/m{sup 2} uniform heat flux. Details about experimental and computational work are presented here.

  19. Beryllium processing technology review for applications in plasma-facing components

    International Nuclear Information System (INIS)

    Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

    1993-07-01

    Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included

  20. Beryllium processing technology review for applications in plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

    1993-07-01

    Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included.

  1. FOREWORD: 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications

    Science.gov (United States)

    Kreter, Arkadi; Linke, Jochen; Rubel, Marek

    2009-12-01

    The 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications (PFMC-12) was held in Forschungszentrum Jülich (FZJ) in Germany in May 2009. This symposium is the successor to the International Workshop on Carbon Materials for Fusion Applications series. Between 1985 and 2003, 10 'Carbon Workshops' were organized in Jülich, Stockholm and Hohenkammer. After this time, the scope of the symposium was redefined to reflect the new requirements of ITER and the ongoing evolution of the field. The workshop was first organized under its new name in 2006 in Greifswald, Germany. The main objective of this conference series is to provide a discussion forum for experts from research institutions and industry dealing with materials for plasma-facing components in present and future controlled fusion devices. The operation of ASDEX-Upgrade with tungsten-coated wall, the fast progress of the ITER-Like Wall Project at JET, the plans for the EAST tokamak to install tungsten, the start of ITER construction and a discussion about the wall material for DEMO all emphasize the importance of plasma-wall interactions and component behaviour, and give much momentum to the field. In this context, the properties and behaviour of beryllium, carbon and tungsten under plasma impact are research topics of foremost relevance and importance. Our community realizes both the enormous advantages and serious drawbacks of all the candidate materials. As a result, discussion is in progress as to whether to use carbon in ITER during the initial phase of operation or to abandon this element and use only metal components from the start. There is broad knowledge about carbon, both in terms of its excellent power-handling capabilities and the drawbacks related to chemical reactivity with fuel species and, as a consequence, about problems arising from fuel inventory and dust formation. We are learning continuously about beryllium and tungsten under fusion conditions, but our

  2. Platelet-rich plasma and hyaluronic acid - an efficient biostimulation method for face rejuvenation.

    Science.gov (United States)

    Ulusal, Betul Gozel

    2017-03-01

    Cosmetic applications of platelet-rich plasma (PRP) are new, and reports are scarce and dispersed in the literature. There are a variety of commercially available kits and injection techniques, and the number and intervals of injections vary. New investigations should focus on developing a standardized procedure for PRP preparation and application methods to augment its efficacy and potency. In this report, we aim to provide data and commentary to assist and add to current guidelines. A series of 94 female patients with varying degrees of facial aging signs were treated with PRP and hyaluronic acid (HA). Mean age was 53.0 ± 5.6. The mean injection number was 3.6 ± 2.0. Platelet-poor and platelet- rich plasma parts were mixed with 0.5 cc %3.5 hyaluronic acid and 0.5 cc procaine and injected with a 30G, 13-mm needle into deep dermis and hypodermis. Patients were asked to rate their personal satisfaction with their skin texture, pigmentation, and sagging. In addition, the overall results were rated by three independent physicians and the patients themselves. The outcomes were peer-reviewed, and correlations between the degree of the aesthetic scores and the number of injections were explored. There was a statistically significant difference in general appearance, skin firmness-sagging and skin texture according to the patients' before and after applications of PRP. A statistically significant correlation was found between the number of injections and overall satisfaction. Compared to the baseline, the PRP and HA injections provided clinically visible and statistically significant improvement on facial skin. The improvements were more remarkable as the injection numbers increased. © 2016 Wiley Periodicals, Inc.

  3. Progress of ITER full tungsten divertor technology qualification in Japan: Manufacturing full-scale plasma-facing unit prototypes

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Yamada, Hirokazu; Hirayama, Tomoyuki; Yokoyama, Kenji; Escourbiac, Frederic; Hirai, Takeshi

    2016-01-01

    Highlights: • JADA has demonstrated the feasibility of manufacturing the full-W plasma-facing units (W-PFU). • The surface profiles of the W monoblocks of the W-PFU prototypes on the test frame to mimic the support structure of the ITER OVT were examined by using an optical three-dimensional measurement system. The results show the most W monoblock surface in the target part locates within + 0.25 mm from the CAD data. • The strict profile control with the profile tolerance of ±0.3 mm is imposed on the OVT to prevent the leading edges of the W monoblocks from over-heating. • The present full-scale prototyping demonstrates to satisfy this requirement on the surface profile. • It can be concluded that the technical maturities of JADA and its suppliers are as high as to start series manufacturing the ITER divertor components. - Abstract: Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m 2 for 10 s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146 W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.

  4. Anchoring plant metallothioneins to the inner face of the plasma membrane of Saccharomyces cerevisiae cells leads to heavy metal accumulation.

    Directory of Open Access Journals (Sweden)

    Lavinia Liliana Ruta

    Full Text Available In this study we engineered yeast cells armed for heavy metal accumulation by targeting plant metallothioneins to the inner face of the yeast plasma membrane. Metallothioneins (MTs are cysteine-rich proteins involved in the buffering of excess metal ions, especially Cu(I, Zn(II or Cd(II. The cDNAs of seven Arabidopsis thaliana MTs (AtMT1a, AtMT1c, AtMT2a, AtMT2b, AtMT3, AtMT4a and AtMT4b and four Noccaea caerulescens MTs (NcMT1, NcMT2a, NcMT2b and NcMT3 were each translationally fused to the C-terminus of a myristoylation green fluorescent protein variant (myrGFP and expressed in Saccharomyces cerevisiae cells. The myrGFP cassette introduced a yeast myristoylation sequence which allowed directional targeting to the cytosolic face of the plasma membrane along with direct monitoring of the intracellular localization of the recombinant protein by fluorescence microscopy. The yeast strains expressing plant MTs were investigated against an array of heavy metals in order to identify strains which exhibit the (hyperaccumulation phenotype without developing toxicity symptoms. Among the transgenic strains which could accumulate Cu(II, Zn(II or Cd(II, but also non-canonical metal ions, such as Co(II, Mn(II or Ni(II, myrGFP-NcMT3 qualified as the best candidate for bioremediation applications, thanks to the robust growth accompanied by significant accumulative capacity.

  5. Anchoring plant metallothioneins to the inner face of the plasma membrane of Saccharomyces cerevisiae cells leads to heavy metal accumulation.

    Science.gov (United States)

    Ruta, Lavinia Liliana; Lin, Ya-Fen; Kissen, Ralph; Nicolau, Ioana; Neagoe, Aurora Daniela; Ghenea, Simona; Bones, Atle M; Farcasanu, Ileana Cornelia

    2017-01-01

    In this study we engineered yeast cells armed for heavy metal accumulation by targeting plant metallothioneins to the inner face of the yeast plasma membrane. Metallothioneins (MTs) are cysteine-rich proteins involved in the buffering of excess metal ions, especially Cu(I), Zn(II) or Cd(II). The cDNAs of seven Arabidopsis thaliana MTs (AtMT1a, AtMT1c, AtMT2a, AtMT2b, AtMT3, AtMT4a and AtMT4b) and four Noccaea caerulescens MTs (NcMT1, NcMT2a, NcMT2b and NcMT3) were each translationally fused to the C-terminus of a myristoylation green fluorescent protein variant (myrGFP) and expressed in Saccharomyces cerevisiae cells. The myrGFP cassette introduced a yeast myristoylation sequence which allowed directional targeting to the cytosolic face of the plasma membrane along with direct monitoring of the intracellular localization of the recombinant protein by fluorescence microscopy. The yeast strains expressing plant MTs were investigated against an array of heavy metals in order to identify strains which exhibit the (hyper)accumulation phenotype without developing toxicity symptoms. Among the transgenic strains which could accumulate Cu(II), Zn(II) or Cd(II), but also non-canonical metal ions, such as Co(II), Mn(II) or Ni(II), myrGFP-NcMT3 qualified as the best candidate for bioremediation applications, thanks to the robust growth accompanied by significant accumulative capacity.

  6. Progress of ITER full tungsten divertor technology qualification in Japan: Manufacturing full-scale plasma-facing unit prototypes

    Energy Technology Data Exchange (ETDEWEB)

    Ezato, Koichiro, E-mail: ezato.koichiro@jaea.go.jp [Department of ITER Project, Naka Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency (Japan); Suzuki, Satoshi; Seki, Yohji; Yamada, Hirokazu; Hirayama, Tomoyuki; Yokoyama, Kenji [Department of ITER Project, Naka Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency (Japan); Escourbiac, Frederic; Hirai, Takeshi [ITER Organization, route de vinon sur Verdon, 13067 St Paul lez Durance (France)

    2016-11-01

    Highlights: • JADA has demonstrated the feasibility of manufacturing the full-W plasma-facing units (W-PFU). • The surface profiles of the W monoblocks of the W-PFU prototypes on the test frame to mimic the support structure of the ITER OVT were examined by using an optical three-dimensional measurement system. The results show the most W monoblock surface in the target part locates within + 0.25 mm from the CAD data. • The strict profile control with the profile tolerance of ±0.3 mm is imposed on the OVT to prevent the leading edges of the W monoblocks from over-heating. • The present full-scale prototyping demonstrates to satisfy this requirement on the surface profile. • It can be concluded that the technical maturities of JADA and its suppliers are as high as to start series manufacturing the ITER divertor components. - Abstract: Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20 MW/m{sup 2} for 10 s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146 W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.

  7. Influence of boronization on operation with high-Z plasma facing components in Alcator C-Mod

    International Nuclear Information System (INIS)

    Lipschultz, B.; Lin, Y.; Marmar, E.S.; Whyte, D.G.; Wukitch, S.; Hutchinson, I.H.; Irby, J.; LaBombard, B.; Reinke, M.L.; Terry, J.L.; Wright, G.

    2007-01-01

    We report the results of operation of Alcator C-Mod with all high-Z molybdenum plasma facing component (PFC) surfaces. Without boron-coated PFCs energy confinement was poor (H ITER,89 ∼ 1) due to high core molybdenum (n Mo /n e ≤ 0.1%) and radiation. After applying boron coatings, n Mo /n e was reduced by a factor of 10-20 with H ITER,89 approaching 2. Results of between-discharge boronization, localized at various major radii, point towards important molybdenum source regions being small, outside the divertor, and due to RF-sheath-rectification. Boronization also has a significant effect on the plasma startup phase lowering Z eff , radiation, and lowering the runaway electron damage. The requirement of low-Z coatings over at least a fraction of the Mo PFCs in C-Mod for best performance together with the larger than expected D retention in Mo, give impetus for further high-Z PFC investigations to better predict the performance of un-coated tungsten surfaces in ITER and beyond

  8. The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices

    Science.gov (United States)

    Rubel, Marek; Petersson, Per; Alves, Eduardo; Brezinsek, Sebastijan; Coad, Joseph Paul; Heinola, Kalle; Mayer, Matej; Widdowson, Anna

    2016-03-01

    First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma-wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1-10), high sensitivity and combination of several methods in a single run. The role of 3He-based NRA, RBS (standard and micro-size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. 15N) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed.

  9. Tritium retention on the surface of stainless steel samples fixed on the plasma-facing wall in LHD

    International Nuclear Information System (INIS)

    Matsuyama, Masao; Abe, Shinsuke; Nishimura, Kiyohiko; Ashikawa, Naoko; Sagara, Akio; Oya, Yasuhisa; Okuno, Kenji; Yamauchi, Yuji; Nobuta, Yuji

    2014-01-01

    Effects of pre-heating for retention and distribution of tritium have been studied using samples fixed on the wall of the Large Helical Device during a plasma campaign. The samples were fixed at four different locations. The plasma-facing surface of the samples was covered with deposition layers of different thickness in each sample. Retention behavior in deposition layers was observed using β-ray-induced X-ray spectrometry and imaging plate technique. Pre-heating of the samples in vacuum was changed in a temperature range from 300 to 623 K, and subsequent tritium exposure was carried out at 300 K in every runs. Non-uniformity of tritium distribution clearly appeared even in the as-received samples which was not pre-heated. It is considered, therefore, that non-uniform adsorption sites of tritium have been produced during a formation process of deposition layers. In addition, it was seen that the amount of tritium retention increased with an increase in the pre-heating temperature, indicating that adsorption sites of tritium were newly formed in the deposition layers by heating in vacuum. (author)

  10. Crystal orientation effects on helium ion depth distributions and adatom formation processes in plasma-facing tungsten

    International Nuclear Information System (INIS)

    Hammond, Karl D.; Wirth, Brian D.

    2014-01-01

    We present atomistic simulations that show the effect of surface orientation on helium depth distributions and surface feature formation as a result of low-energy helium plasma exposure. We find a pronounced effect of surface orientation on the initial depth of implanted helium ions, as well as a difference in reflection and helium retention across different surface orientations. Our results indicate that single helium interstitials are sufficient to induce the formation of adatom/substitutional helium pairs under certain highly corrugated tungsten surfaces, such as (1 1 1)-orientations, leading to the formation of a relatively concentrated layer of immobile helium immediately below the surface. The energies involved for helium-induced adatom formation on (1 1 1) and (2 1 1) surfaces are exoergic for even a single adatom very close to the surface, while (0 0 1) and (0 1 1) surfaces require two or even three helium atoms in a cluster before a substitutional helium cluster and adatom will form with reasonable probability. This phenomenon results in much higher initial helium retention during helium plasma exposure to (1 1 1) and (2 1 1) tungsten surfaces than is observed for (0 0 1) or (0 1 1) surfaces and is much higher than can be attributed to differences in the initial depth distributions alone. The layer thus formed may serve as nucleation sites for further bubble formation and growth or as a source of material embrittlement or fatigue, which may have implications for the formation of tungsten “fuzz” in plasma-facing divertors for magnetic-confinement nuclear fusion reactors and/or the lifetime of such divertors.

  11. Research status and issues of tungsten plasma facing materials for ITER and beyond

    International Nuclear Information System (INIS)

    Ueda, Y.; Coenen, J.W.; De Temmerman, G.; Doerner, R.P.; Linke, J.; Philipps, V.; Tsitrone, E.

    2014-01-01

    This review summarizes surface morphology changes of tungsten caused by heat and particle loadings from edge plasmas, and their effects on enhanced erosion and material lifetime in ITER and beyond. Pulsed heat loadings by transients (disruption and ELM) are the largest concerns due to surface melting, cracking, and dust formation. Hydrogen induced blistering is unlikely to be an issue of ITER. Helium bombardment would cause surface morphology changes such as W fuzz, He holes, and nanometric bubble layers, which could lead to enhanced erosion (e.g. unipolar arcing of W fuzz). Particle loadings could enhance pulsed heat effects (cracking and erosion) due to surface layer embrittlement by nanometric bubbles and solute atoms. But pulsed heat loadings alleviate surfaces morphology changes in some cases (He holes by ELM-like heat pulses). Effects of extremely high fluence (∼10 30 m −2 ), mixed materials, and neutron irradiation are important issues to be pursued for ITER and beyond. In addition, surface refurbishment to prolong material lifetime is also an important issue

  12. Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    International Nuclear Information System (INIS)

    Bourham, Mohamed A.; Gilligan, John G.

    1999-01-01

    Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m 2 over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER

  13. Overview of advanced techniques for fabrication and testing of ITER multilayer plasma facing walls

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.-A.F. [Commissariat a l`Energie Atomique, Saclay, Gif-sur-Yvette (France)

    1998-09-01

    The design of the ITER primary first wall incorporates a multi-layered structure consisting of a layer of beryllium bonded to a layer of copper alloy with embedded stainless steel tubes which in turn is bonded to a stainless steel structure. In this configuration, the stainless steel provides structural support, the copper alloy improved resistance to high heat loads, and the beryllium layer a low Z metal interface with plasma. Fabrication, testing and control of this multi-layered structure, and indeed the entire blanket shield module, calls for advanced methods. Several associations in the four home teams and their industrial partners have been involved in various fabrication and joining tasks now grouped under L4 blanket project. In this paper, an overview of the work done so far for joining stainless steel to stainless steel, stainless steel to copper alloy, copper alloy to copper alloy, and copper alloy to beryllium is presented. Specialised papers dealing with most of the topics treated here are scheduled in this symposium. The fabrication and joining methods presented here, other than the conventional welding and brazing, follow four main routes. Two of them make extensive use of hot-isostatic pressing (HIP); (a) solid to solid; (b) solid or powder to powder, with or without a prior cold or hot isostatic pressing of one of the products. The third combines advantages of casting and HIPping for fabricating large and complex parts. The fourth investigates the possibility of using explosive welding for joining copper alloys to stainless steel. Other methods, including friction welding, are investigated for specific parts. (orig.) 34 refs.

  14. Effects of an evaporative cooling system on plasma cortisol, IGF-I, and milk production in dairy cows in a tropical environment

    Science.gov (United States)

    Titto, Cristiane Gonçalves; Negrão, João Alberto; Titto, Evaldo Antonio Lencioni; Canaes, Taissa de Souza; Titto, Rafael Martins; Pereira, Alfredo Manuel Franco

    2013-03-01

    Access to an evaporative cooling system can increase production in dairy cows because of improved thermal comfort. This study aimed to evaluate the impact of ambient temperature on thermoregulation, plasma cortisol, insulin-like growth factor 1 (IGF-I), and productive status, and to determine the efficiency of an evaporative cooling system on physiological responses under different weather patterns. A total of 28 Holstein cows were divided into two groups, one with and the other without access to a cooling system with fans and mist in the free stall. The parameters were analyzed during morning (0700 hours) and afternoon milking (1430 hours) under five different weather patterns throughout the year (fall, winter, spring, dry summer, and rainy summer). Rectal temperature (RT), body surface temperature (BS), base of tail temperature (TT), and respiratory frequency (RF) were lower in the morning ( P milk production during spring and summer ( P < 0.01).

  15. FOREWORD: 13th International Workshop on Plasma-Facing Materials and Components for Fusion Applications/1st International Conference on Fusion Energy Materials Science 13th International Workshop on Plasma-Facing Materials and Components for Fusion Applications/1st International Conference on Fusion Energy Materials Science

    Science.gov (United States)

    Jacob, Wolfgang; Linsmeier, Christian; Rubel, Marek

    2011-12-01

    subjected to extremely high heat loads and very high particle and neutron fluxes. They must have high thermal conductivity for efficient heat transport, high cohesive energy for low erosion by particle bombardment and low atomic number to minimize plasma cooling. These contradictory requirements make the development of plasma-facing materials one of the greatest challenges ever faced by materials scientists. The erosion of plasma-facing materials is one of the main factors influencing the operational schedule of experimental fusion reactors and future power plants. A number of materials selected for current designs cannot withstand the presently foreseen plasma scenarios of a power plant for a commercially viable period of time. Therefore, further coordinated development of plasma scenarios and materials is essential for the realization of fusion as an energy source. The design and development of plasma-facing materials requires a detailed understanding of the processes that occur when a material surface is bombarded with an intense flux of heat, particles and neutrons simultaneously. These materials-related topics are the focus of this series of workshops which has established itself as a discussion forum for experts from research institutions and industry dealing with materials for plasma-facing components in present and future thermonuclear fusion devices. During the joint conference PFMC-13/FEMaS-1 recent developments and research results in the following fields were addressed: carbon, beryllium, and tungsten based materials mixed materials erosion and redeposition high heat flux component development benchmarking of radiation damage modelling synchrotron and neutron based characterization techniques application of advanced transmission electron microscopy and micro-/nano-mechanical testing. With the approaching technical realization of ITER, the ITER-related PFMC topics are naturally the main focus of research. In this respect the start of the ITER-like wall experiment

  16. Development of bonding techniques between tungsten and copper alloy for plasma facing components by HIP method. 1. Bonding between tungsten and oxygen free copper

    International Nuclear Information System (INIS)

    Saito, Shigeru; Fukaya, Kiyoshi; Ishiyama, Shintaro; Eto, Motokuni; Akiba, Masato

    1999-08-01

    In recent years, it has been considered that W (tungsten) is one of candidate materials for armor tiles of plasma facing components, like first wall or divertor, of fusion reactor. On the other hand, oxygen free high thermal conductivity (OFHC)-copper is proposed as heat sink materials behind the plasma facing materials because of its high thermal conductivity. However, plasma facing components are exposed to cyclic high heat load and heavily irradiated by 14 MeV neutron. Under these conditions, many unfavorable effects, for instance, thermal stresses of bonding interface, irradiation damage and He atom production by nuclear transmutation, will be decreased bonding strength between W and Cu alloys. Therefore, it is necessary to develop a reliable bonding techniques in order to make plasma facing components which can resist them. Then, we started the bonding technology development by hot isostatic press (HIP) method to bond W with Cu alloys. In this experiments, to optimize HIP bonding conditions, four point bending were performed for each bonded conditions at temperature from R.T. to 873 K and we could get the best HIP bonding conditions for W and OFHC-Cu as 1273 K x 2 hours x 147 MPa. To evaluate bonding strength of the specimen bonded at these conditions, tensile tests were also performed at same temperature range. The tensile strength was similar with OFHC-Cu which were treated at same conditions. (author)

  17. Experimental results of near real-time protection system for plasma facing components in Wendelstein 7-X at GLADIS

    Science.gov (United States)

    Ali, A.; Jakubowski, M.; Greuner, H.; Böswirth, B.; Moncada, V.; Sitjes, A. Puig; Neu, R.; Pedersen, T. S.; the W7-X Team

    2017-12-01

    One of the aims of stellarator Wendelstein 7-X (W7-X), is to investigate steady state operation, for which power exhaust is an important issue. The predominant fraction of the energy lost from the confined plasma region will be absorbed by an island divertors, which is designed for 10 {{MWm}}-2 steady state operation. In order to protect the divertor targets from overheating, 10 state-of-the-art infrared endoscopes will be installed at W7-X. In this work, we present the experimental results obtained at the high heat flux test facility GLADIS (Garching LArge DIvertor Sample test facility in IPP Garching) [1] during tests of a new plasma facing components (PFCs) protection algorithm designed for W7-X. The GLADIS device is equipped with two ion beams that can generate a heat load in the range from 3 MWm-2 to 55 MWm-2. The algorithms developed at W7-X to detect defects and hot spots are based on the analysis of surface temperature evolution and are adapted to work in near real-time. The aim of this work was to test the near real-time algorithms in conditions close to those expected in W7-X. The experiments were performed on W7-X pre-series tiles to detect CFC/Cu delaminations. For detection of surface layers, carbon fiber composite (CFC) blocks from the divertor of the Wendelstein 7-AS stellarator were used to observe temporal behavior of fully developed surface layers. These layers of re-deposited materials, like carbon, boron, oxygen and iron, were formed during the W7-AS operation. A detailed analysis of the composition and their thermal response to high heat fluxes (HHF) are described in [2]. The experiments indicate that the automatic detection of critical events works according to W7-X PFC protection requirements.

  18. Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Brooks, A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Lopes-Cardozo, N. [TU/Eindhoven, Eindhoven (Netherlands); Menard, J.; Ono, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Rindt, P. [TU/Eindhoven, Eindhoven (Netherlands); Tresemer, K. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

    2016-11-15

    Highlights: • An upgrade path for the NSTX-U tokamak is proposed that maintains scientific productivity while enabling exploration of novel, liquid metal PFC. • Pre-filled liquid metal divertor targets are proposed as an intermediate step that mitigates technical and scientific risks associated with liquid metal PFC. • Analysis of leading edge features show a strong link between engineering design considerations and expected performance as a PFC. • A method for optimizing porous liquid metal targets restrained by capillary forces is provided indicating pore-sizes well within current technical capabilities. - Abstract: Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. Two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

  19. High Confinement and High Density with Stationary Plasma Energy and Strong Edge Radiation Cooling in Textor-94

    Science.gov (United States)

    Messiaen, A. M.

    1996-11-01

    A new discharge regime has been observed on the pumped limiter tokamak TEXTOR-94 in the presence of strong radiation cooling and for different scenarii of additional hearing. The radiated power fraction (up to 90%) is feedback controlled by the amount of Ne seeded in the edge. This regime meets many of the necessary conditions for a future fusion reactor. Energy confinement increases with increasing densities (reminiscent of the Z-mode obtained at ISX-B) and as good as ELM-free H-mode confinement (enhancement factor verus ITERH93-P up to 1.2) is obtained at high densities (up to 1.2 times the Greenwald limit) with peaked density profiles showing a peaking factor of about 2 and central density values around 10^14cm-3. In experiments where the energy content of the discharges is kept constant with an energy feedback loop acting on the amount of ICRH power, stable and stationary discharges are obtained for intervals of more than 5s, i.e. 100 times the energy confinement time or about equal to the skin resistive time, even with the cylindrical q_α as low as 2.8 β-values up to the β-limits of TEXTOR-94 are achieved (i.e. β n ≈ 2 of and β p ≈ 1.5) and the figure of merit for ignition margin f_Hqa in these discharges can be as high as 0.7. No detrimental effects of the seeded impurity on the reactivity of the plasma are observed. He removal in these discharges has also been investigated. [1] Laboratoire de Physique des Plasmas-Laboratorium voor Plasmafysica, Association "EURATOM-Belgian State", Ecole Royale Militaire-Koninklijke Militaire School, Brussels, Belgium [2] Institut für Plasmaphysik, Forschungszentrum Jülich, GmbH, Association "EURATOM-KFA", Jülich, Germany [3] Fusion Energy Research Program, Mechanical Engineering Division, University of California at San Diego, La Jolla, USA [4] FOM Institüt voor Plasmafysica Rijnhuizen, Associatie "FOM-EURATOM", Nieuwegein, The Netherlands [*] Researcher at NFSR, Belgium itemize

  20. Heat transfer characteristics of rectangular coolant channels with various aspect ratios in the plasma-facing components under fully developed MHD laminar flow

    International Nuclear Information System (INIS)

    Takase, K.; Hasan, M.Z.

    1995-01-01

    Convective heat transfer in MHD laminar flow through rectangular channels in the plasma-facing components of a fusion reactor has been analyzed numerically to investigate the effects of channel aspect ratio, defined as the ratio of the lengths of the plasma-facing side to the other side. The adverse effect of the nonuniformity of surface heat flus on Nusselt number (Nu) at the plasma-facing side can be alleviated by increasing the aspect ratio of a rectangular duct. At the center and corner of the plasma-facing side of a square duct, the Nu of non-MHD flow are 6.8 and 2.2, respectively, for uniform surface heat flux. In the presence of a strong magnetic field, Nu at the center and corner increases to 22 and 3.6, respectively. However, when the heat flux is highly nonuniform, as in the plasma-facing components, Nu decreases from 22 to 3.1 at the center and from 3.6 to 3.1 at the corner. When the aspect ratio is increased to 4, Nu at the center and corner increase to 5 and 4.7. Along the circumference of a rectangular channel, there are locations where the wall temperature is equal to or less than the bulk coolant temperature, thus making the Nu with conventional definition infinity or negative. The ratio between Nu of MHD flow and Nu of non-MHD flow for various aspect ratios is constant in the region of Hartmann number of more than 200 at least. On the other hand, its ratio increases monotonously with increasing the aspect ratio

  1. Modification of adhered dust on plasma-facing surfaces due to exposure to ELMy H-mode plasma in DIII-D

    Directory of Open Access Journals (Sweden)

    I. Bykov

    2017-08-01

    Full Text Available Transient heat load tests have been conducted in the lower divertor of DIII-D using DiMES manipulator in order to study the behavior of dust on tungsten Plasma Facing Components (PFCs during ELMy H-mode discharges. Samples with pre-adhered, pre-characterized dust have been exposed at the outer strike point (OSP in a series of discharges with varied intra-(inter- ELM heat fluxes. We used C dust because of its high sublimation temperature and non-metal properties. Al dust as a surrogate for Be and W dust were employed as relevant to that in the ITER divertor. The poor initial thermal contact between the substrate and the particles led to overheating, sublimation and shrinking of the carbon dust, and wetting induced coagulation of Al dust. Little modification of the W dust was observed. An enhanced surface adhesion and improvement of the thermal contact of C and Al dust were the result of exposure. A post mortem “adhesive tape” sampling showed that 70% of Al, <5% of W and C particles could not be removed from the surface owing to the improved adhesion. Al and C but not W particles that could be lifted had W inclusions indicating damage to the substrate. This suggests that non destructive methods may be inefficient for removal of dust in ITER.

  2. A study of plasma facing tungsten components with electrical discharge machined surface exposed to cyclic thermal loads

    Energy Technology Data Exchange (ETDEWEB)

    Seki, Yohji, E-mail: seki.yohji@jaea.go.jp; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Yamada, Hirokazu; Hirayama, Tomoyuki

    2016-11-01

    Through R&D for a plasma facing units (PFUs) in an outer vertical target of an ITER full-tungsten (W) divertor, Japan Atomic Energy Agency succeeded in demonstrating the durability of the W divertor shaped by an electrical discharge machining (EDM). To prevent melting of W armors in the PFUs, an adequate technology to meet requirements of a geometrical shape and a tolerance is one of the most important key issues in a manufacturing process. From the necessity, the EDM has been evaluated to control the final shape of the W armor. Though the EDM was known to be advantages such as an easy workability, a potential disadvantage of presence of micro-cracks on the W surface appeared. In order to examine a potential effect of the micro-crack on a heat removal durability, a high heat flux testing was carried out for the W divertor mock-up with the polish and the EDM. As the result, all of the W armors endured the repetitive heat load of 1000 cycles at an absorbed heat flux of more than 20 MW/m{sup 2}, which strongly encourages the realization of the PFUs of the ITER full-W divertor with the various geometrical shape and the high accuracy tolerance.

  3. The role and application of ion beam analysis for studies of plasma-facing components in controlled fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Rubel, Marek, E-mail: Marek.Rubel@ee.kth.se [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Fusion Plasma Physics, Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden); Petersson, Per [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Fusion Plasma Physics, Royal Institute of Technology (KTH), 100 44 Stockholm (Sweden); Alves, Eduardo [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisbon (Portugal); Brezinsek, Sebastijan [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Institut für Klima- und Energieforschung, Forschungszentrum Jülich, D-52425 Jülich (Germany); Coad, Joseph Paul [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Heinola, Kalle [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); University of Helsinki, 00014 Helsinki (Finland); Mayer, Matej [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Max-Planck-Institut für Plasmaphysik, 85478 Garching (Germany); Widdowson, Anna [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2016-03-15

    First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma–wall interactions. Detailed information is required for the assessment of material lifetime and accumulation of hydrogen isotopes in wall materials. The intention of this work is to give a concise overview of key issues in the characterization of plasma-facing materials and components in tokamaks, especially in JET with an ITER-Like Wall. IBA techniques play a particularly prominent role here because of their isotope selectivity in the low-Z range (1–10), high sensitivity and combination of several methods in a single run. The role of {sup 3}He-based NRA, RBS (standard and micro-size beam) and HIERDA in fuel retention and material migration studies is presented. The use of tracer techniques with rare isotopes (e.g. {sup 15}N) or marker layers on wall diagnostic components is described. Special instrumentation, development of equipment to enhance research capabilities and issues in handling of contaminated materials are addressed.

  4. Measurement of thickness of film deposited on the plasma-facing wall in the QUEST tokamak by colorimetry.

    Science.gov (United States)

    Wang, Z; Hanada, K; Yoshida, N; Shimoji, T; Miyamoto, M; Oya, Y; Zushi, H; Idei, H; Nakamura, K; Fujisawa, A; Nagashima, Y; Hasegawa, M; Kawasaki, S; Higashijima, A; Nakashima, H; Nagata, T; Kawaguchi, A; Fujiwara, T; Araki, K; Mitarai, O; Fukuyama, A; Takase, Y; Matsumoto, K

    2017-09-01

    After several experimental campaigns in the Kyushu University Experiment with Steady-state Spherical Tokamak (QUEST), the originally stainless steel plasma-facing wall (PFW) becomes completely covered with a deposited film composed of mixture materials, such as iron, chromium, carbon, and tungsten. In this work, an innovative colorimetry-based method was developed to measure the thickness of the deposited film on the actual QUEST wall. Because the optical constants of the deposited film on the PFW were position-dependent and the extinction coefficient k 1 was about 1.0-2.0, which made the probing light not penetrate through some thick deposited films, the colorimetry method developed can only provide a rough value range of thickness of the metal-containing film deposited on the actual PFW in QUEST. However, the use of colorimetry is of great benefit to large-area inspections and to radioactive materials in future fusion devices that will be strictly prohibited from being taken out of the limited area.

  5. A study of plasma facing tungsten components with electrical discharge machined surface exposed to cyclic thermal loads

    International Nuclear Information System (INIS)

    Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Yamada, Hirokazu; Hirayama, Tomoyuki

    2016-01-01

    Through R&D for a plasma facing units (PFUs) in an outer vertical target of an ITER full-tungsten (W) divertor, Japan Atomic Energy Agency succeeded in demonstrating the durability of the W divertor shaped by an electrical discharge machining (EDM). To prevent melting of W armors in the PFUs, an adequate technology to meet requirements of a geometrical shape and a tolerance is one of the most important key issues in a manufacturing process. From the necessity, the EDM has been evaluated to control the final shape of the W armor. Though the EDM was known to be advantages such as an easy workability, a potential disadvantage of presence of micro-cracks on the W surface appeared. In order to examine a potential effect of the micro-crack on a heat removal durability, a high heat flux testing was carried out for the W divertor mock-up with the polish and the EDM. As the result, all of the W armors endured the repetitive heat load of 1000 cycles at an absorbed heat flux of more than 20 MW/m"2, which strongly encourages the realization of the PFUs of the ITER full-W divertor with the various geometrical shape and the high accuracy tolerance.

  6. Steady state and transient thermal-hydraulic characterization of full-scale ITER divertor plasma facing components

    International Nuclear Information System (INIS)

    Tincani, A.; Malavasi, A.; Ricapito, I.; Riccardi, B.; Di Maio, P.A.; Vella, G.

    2007-01-01

    In the frame of the activities related to ITER divertor R and D, ENEA CR Brasimone was charged by EFDA (European Fusion Design Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. Inner Vertical Target, Dome Liner and Outer Vertical Target, both in steady state and during draining and drying transient. More in detail, for each PFC, the first phase of the work is the steady state hydraulic characterization which consists of: - measurements of pressure drops at different temperatures; - determination of the velocity distribution in the internal channels; - check the possible insurgence of cavitation. The subsequent phase of the thermal-hydraulic characterization foresees a testing campaign of draining and drying procedure by means of a suitable gas flow. The objective of this experimental procedure is to eliminate in the most efficient way the residual amount of water after gravity discharge. In order to accomplish this experimental campaign a significant modification of CEF1 loop has been designed and realized. This paper presents, first of all, the experimental set-up, the agreed test matrix and the achieved results for both steady state and transient tests. Moreover, the level of the implementation of a predictive hydraulic model, based on RELAP 5 code, as well as its results are described, discussed and compared with the experimental ones. (orig.)

  7. Structural safety assessment of a tokamak-type fusion facility for a through crack to cause cooling water leakage and plasma disruption

    International Nuclear Information System (INIS)

    Nakahira, Masataka

    2004-01-01

    A tokamak-type fusion machine has inherent safety associated with plasma shutdown. A small water leak can cause a plasma disruption although there is another possibility to terminate plasma without disruption. This plasma disruption will induce electromagnetic (EM) forces acting in the vacuum vessel (VV). From a radiological safety viewpoint, the VV is designed to form a physical barrier that encloses tritium and activated dust. If the VV can sustain an unstable fracture by EM forces from a through crack to cause the small leak, the structural safety will be assured and the inherent safety will be demonstrated. Therefore, a systematic approach to assure the structural safety is developed. A new analytical model to evaluate the through crack and leak rate of cooling water is proposed, with verification by experimental leak measurements. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is concluded that EM forces induced by the small leak to terminate plasma will not cause unstable fracture of the VV; thus the inherent safety is demonstrated. (author)

  8. LOFA analyses for the water and helium cooled SEAFP reactors

    International Nuclear Information System (INIS)

    Sponton, L.; Sjoeberg, A.; Nordlinder, S.

    2001-01-01

    This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

  9. Development of bonding techniques of W and Cu-alloys for plasma facing components of fusion reactor with HIP method

    International Nuclear Information System (INIS)

    Saito, S.; Fukaya, K.; Ishiyama, S.; Eto, M.; Sato, K.; Akiba, M.

    1998-01-01

    W (tungsten) and Cu (copper)-alloys, like oxygen free high thermal conductivity (OFHC)-copper or dispersion strengthened (DS)-copper, are candidate materials for plasma facing components(PFC) of TOKAMAK type fusion reactor as armor tile and heat sink, respectively. However, PFC are exposed to cyclic high heat load and heavy irradiation by 14 MeV neutrons. Under these conditions, thermal stresses at bonding interface and irradiation damage will decrease the bonding strength between W and Cu alloys. Therefore, it is necessary to develop a reliable bonding techniques in order to make PFC with enough integrity. We have applied the hot isostatic press (HIP) method to bond W with Cu-alloys. In this experiments, to optimize HIP bonding conditions, four point bending tests were performed for different bonding conditions at temperatures from R.T. to 873 K and we obtained an optimum HIP bonding condition for W and OFHC-Cu as 1273 SK x 2 hours x 98 ∼ 147 MPa. Tensile tests were also performed at the same temperature range. The tensile strength of the bonded W / Cu was almost equal to that of OFHC Cu which was HIPed at the same conditions. Tensile specimens were broken at the bonding interface or OFHC-Cu side. Bonding tests of W and DS-Cu showed that HIP was not successful because tungsten oxide was produced at the bonding interface and residual stresses were not relaxed. Therefore, it was concluded that some insert materials will be needed to bond W and DS-Cu. (author)

  10. General directions and recently test modelling results of lithium capillary-pore systems as plasma facing components for tokamak-reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.; Azizov, E.A.; Mirnov, S.V.; Lazaret, V.B.; Safronov, V.M.

    2003-01-01

    Full text: At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing material. A solid CPS filled with liquid lithium will have high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stress induced cracking in steady state and during plasma transitions (disruptions, ELMs, VDEs, runaways) to provide the normal operation of divertor target plates and first wall protection elements. These materials would not be the sources of impurities inducing the raise of Z eff and they will not be collected as dust in the divertor area and in ducts. The key directions of experimental investigation of lithium CPS behaviour in first wall and divertor operation simulating conditions are considered. Experiments with lithium CPS in plasma disruption simulation conditions on the hydrogen plasma accelerator MK-200UG (∼10-15 MJ/m 2 , ∼50 μs) have been performed. Shielding lithium plasma layer formation and high stability of these systems have been shown. The new lithium limiter with a thermal regulation system tests on up graded T-11M tokamak (plasma current up to 100 kA, pulse length ∼0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, limiter deposited power are investigated in this tests

  11. U.S. Assessment of advanced limiter-divertor plasma-facing systems (ALPS) design, analysis, and R and D needs

    International Nuclear Information System (INIS)

    Mattas, R. F.

    1998-01-01

    The purpose of the ALPS program is to identify and evaluate advanced limiter/diverter systems that will enhance the attractiveness of fusion power. The highest priority goals at present are achieving high power density, up to 50 MW/m 2 , and showing compatibility of plasma-facing surfaces with plasma operation. Personnel representing a wide range of disciplines from a number of institutions are engaged in the program, where an evaluation phase of the program is planned for three years. Successful identification of promising concepts in the evaluation phase should lead to an R and D phase that includes proof-of-principle experiments

  12. INFLUENCE OF COOLING CONDITIONS AND THE SIZE OF STOCK MATERIAL DURING CASTING OF TOOL STEEL ON CAPABILITY TO THE SUBSEQUENT THERMAL HARD-FACING

    Directory of Open Access Journals (Sweden)

    V. N. Fedulov

    2016-01-01

    Full Text Available Influence of cooling conditions and the size of stock material of instrumental steel 4H5MF1S on сability to surface hardening after high-temperature tempering at 500–650 °C is investigated. Comparison with hardening of forgings is given.

  13. Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Guilhem, D.; Adjeroud, B.; Balorin, C.; Buravand, Y.; Bertrand, B.; Bondil, J.L.; Desgranges, C.; Gauthier, E.; Lipa, M.; Messina, P.; Missirlian, M.; Mitteau, R.; Moulin, D.; Pocheau, C.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.; Vallet, S

    2004-07-01

    Tore-Supra has a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T proportional to {radical}(t)) during the discharge until radiation cooling is equal to the incoming heat flux (T > 1800 K). Such an environment is found in most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra infrared thermography system is composed of 7 infrared endoscopes, this system is described in details in the paper, the new JET infrared thermography system is presented and some insights of the ITER set of visible/infrared endoscope is given. (authors)

  14. Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

    International Nuclear Information System (INIS)

    Guilhem, D.; Adjeroud, B.; Balorin, C.; Buravand, Y.; Bertrand, B.; Bondil, J.L.; Desgranges, C.; Gauthier, E.; Lipa, M.; Messina, P.; Missirlian, M.; Mitteau, R.; Moulin, D.; Pocheau, C.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.; Vallet, S.

    2004-01-01

    Tore-Supra has a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T proportional to √(t)) during the discharge until radiation cooling is equal to the incoming heat flux (T > 1800 K). Such an environment is found in most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra infrared thermography system is composed of 7 infrared endoscopes, this system is described in details in the paper, the new JET infrared thermography system is presented and some insights of the ITER set of visible/infrared endoscope is given. (authors)

  15. Cooling techniques

    International Nuclear Information System (INIS)

    Moeller, S.P.

    1994-01-01

    After an introduction to the general concepts of cooling of charged particle beams, some specific cooling methods are discussed, namely stochastic, electron and laser cooling. The treatment concentrates on the physical ideas of the cooling methods and only very crude derivations of cooling times are given. At the end three other proposed cooling schemes are briefly discussed. (orig.)

  16. Investigations on in situ diagnostics by an infrared camera to distinguish between the plasma facing tiles with carbonaceous surface layer and defect in the underneath junction

    International Nuclear Information System (INIS)

    Cai, Laizhong; Gauthier, Eric; Corre, Yann; Liu, Jian

    2013-01-01

    Both a deposition surface layer and a delamination underneath junction existing on plasma facing components (PFCs) can result in abnormal high surface temperature under normal heating conditions. The tile with delamination has to be replaced to prevent from a critical failure (complete delamination) during plasma operation while the carbon deposit can be removed without any repairing. Therefore, distinguishing in situ deposited tiles and junction defect tiles is crucial to avoid the critical failure without unwanted shutdown. In this paper, the thermal behaviors of junction defect tiles and carbon deposit tiles are simulated numerically. A modified time constant method is then introduced to analyze the thermal behaviors of deposited tiles and junction defect tiles. The feasibility of discrimination by analyzing the thermal behaviors of tiles is discussed and the requirements of this method for discrimination are described. Finally, the time resolution requirement of IR cameras to do the discrimination is mentioned

  17. 2nd IAEA research coordination meeting on collection and evaluation of reference data for thermo-mechanical properties of fusion reactor plasma facing materials. Summary report

    International Nuclear Information System (INIS)

    Langley, R.A.

    1996-08-01

    The proceedings and results of the 2nd IAEA Research Coordination Meeting on ''Collection and Evaluation of Reference Data for Thermo-mechanical Properties of Fusion Reactor Plasma Facing Materials'' held on March 25, 26 and 27, 1996 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of discussions amongst the participants regarding the status of data, publication of a multi-author review paper and recommendations regarding future work. (author). 1 tab

  18. Expected energy fluxes onto ITER Plasma Facing Components during disruption thermal quenches from multi-machine data comparisons

    International Nuclear Information System (INIS)

    Loarte, A.; Andrew, P.; Matthews, G.F.; Paley, J.; Riccardo, V.; Counsell, G.; Eich, T.; Fuchs, C.; Gruber, O.; Herrmann, A.; Pautasso, G.; Federici, G.; Finken, K.H.; Maddaluno, G.; Whyte, D.

    2005-01-01

    A comparison of the power flux characteristics during the thermal quench of plasma disruptions among various tokamak experiments has been carried out and conclusions for ITER have been drawn. It is generally observed that the energy of the plasma at the thermal quench is much smaller than that of a full performance plasma. The timescales for power fluxes onto PFCs during the thermal quench, as determined by IR measurements, are found to scale with device size but not to correlate with pre-disruptive plasma characteristics. The profiles of the thermal quench power fluxes are very broad for diverted discharges, typically a factor of 5-10 broader than that measured during 'normal' plasma operation, while for limiter discharges this broadening is absent. The combination of all the above factors is used to derive the expected range of power fluxes on the ITER divertor target during the thermal quench. The new extrapolation derived in this paper indicates that the average disruption in ITER will deposit an energy flux approximately one order of magnitude lower than previously thought. The evaluation of the ITER divertor lifetime with these revised specifications is carried out. (author)

  19. Three-dimensional particle-in-cell simulations of gap crossings in castellated plasma-facing components in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Komm, Michael; Dejarnac, Renaud; Gunn, J. P.; Pekarek, Z.

    2013-01-01

    Roč. 55, č. 2 (2013), 025006-025006 ISSN 0741-3335 R&D Projects: GA ČR GA202/09/1467; GA ČR GAP205/11/2341; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * PIC divertor * castellation gaps Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.386, year: 2013 http://iopscience.iop.org/0741-3335/55/2/025006/pdf/0741-3335_55_2_025006.pdf

  20. Numerical model for swirl flow cooling in high-heat-flux particle beam targets and the design of a swirl-flow-based plasma limiter

    International Nuclear Information System (INIS)

    Milora, S.L.; Combs, S.K.; Foster, C.A.

    1984-11-01

    An unsteady, two-dimensional heat conduction code has been used to study the performance of swirl-flow-based neutral particle beam targets. The model includes the effects of two-phase heat transfer and asymmetric heating of tubular elements. The calorimeter installed in the Medium Energy Test Facility, which has been subjected to 30-s neutral beam pulses with incident heat flux intensities of greater than or equal to 5 kW/cm 2 , has been modeled. The numerical results indicate that local heat fluxes in excess of 7 kW/cm 2 occur at the water-cooled surface on the side exposed to the beam. This exceeds critical heat flux limits for uniformly heated tubes wih straight flow by approximately a factor of 5. The design of a plasma limiter based on swirl flow heat transfer is presented

  1. Development of bonding techniques between W and Cu-alloys for plasma facing components by HIP method (3). Bonding tests with Au-foil insert

    International Nuclear Information System (INIS)

    Saito, Shigeru

    2002-07-01

    In recent years, it has been considered that W (tungsten) is one of candidate materials for armor tiles of plasma a facing components (PFC), like first wall or divertor, of fusion reactor. On the other hand, Cu-alloys, like OFHC-Cu or DS-Cu, are proposed as heat sink materials behind the plasma facing materials because of its high thermal conductivity. It is necessary to develop a reliable bonding techniques in order to fabricate PFC. JAERI has developed the hot isostatic press (HIP) bonding process to bond W with Cu-alloys. In this experiments, bonding tests with Au-foil insert were performed. We could get the best HIP bonding conditions for W and Cu-alloys with Au-foil as 1123K x 2hours x 147MPa. It was shown that the HIP temperature was 150K lower than that of without Au-foil. Furthermore, the tensile strength was similar to that of with without Au-foil. (author)

  2. The Effects of Temperature and Oxidation on Deuterium Retention in Solid and Liquid Lithium Films on Molybdenum Plasma-Facing Components

    Science.gov (United States)

    Capece, Angela

    2014-10-01

    Liquid metal plasma-facing components (PFCs) enable in-situ renewal of the surface, thereby offering a solution to neutron damage, erosion, and thermal fatigue experienced by solid PFCs. Lithium in particular has a high chemical affinity for hydrogen, which has resulted in reduced recycling and enhanced plasma performance on many fusion devices including TFTR, T11-M, FTU, CDX-U, LTX, TJ-II, and NSTX. A key component to the improvement in plasma performance is deuterium retention in Li; however, this process is not well understood in the complex tokamak environment. Recent surface science experiments conducted at the Princeton Plasma Physics Laboratory have used electron spectroscopy and temperature programmed desorption to understand the mechanisms for D retention in Li coatings on Mo substrates. The experiments were designed to give monolayer-control of Li films and were conducted in ultrahigh vacuum under controlled environments. An electron cyclotron resonance plasma source was used to deliver a beam of deuterium ions to the surface over a range of ion energies. Our work shows that D is retained as LiD in metallic Li films. However, when oxygen is present in the film, either by diffusion from the subsurface at high temperature or as a contaminant during the deposition process, Li oxides are formed that retain D as LiOD. Experiments indicate that LiD is more thermally stable than LiOD, which decomposes to liberate D2 gas and D2O at temperatures 100 K lower than the LiD decomposition temperature. Other experiments show how D retention varies with substrate temperature to provide insight into the differences between solid and liquid lithium films. This work was supported by DOE Contract No. DE AC02-09CH11466.

  3. R and D activities on manufacturing plasma-facing unit for prototype of ITER divertor outer target in JADA

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Nishi, Hiroshi; Mohri, Kensuke; Enoeda, Mikio

    2012-01-01

    Japan Domestic Agency (JADA) carried out R and Ds activities to improve joining CFC monoblocks onto a CuCrZr cooling tube in PFUs to boost the success rate of joint and to confirm load carrying capability of the monoblock attachments to Steel Support Structure (SSS) against tensile force simulating electromagnetic load to pull PFUs from SSS. In joining the CFC monoblocks to the cooling tube, JADA has adopted brazing by using noble-metal-free filler with the following improvements; (1) metalizing joint surface of CFC using Ti-coating with accurate thickness controlling, (2) Changing buffer layer material from soft pure copper to Cu–W alloy. By using the present improved joint, JADA has manufactured three mock-ups with 5 CFC monoblocks and tested against repetitive high heat loads more than 20 MW/m 2 . All of CFC monoblocks of each mockup can survive the high heat loads throughout 1000 cycles with no degradation of heat removal capability. Regarding the load carrying capability of monoblock attachments to SSS, tensile experiments were carried out using the same geometries of CFC and tungsten monoblocks in PFUs and the results show that both geometries and joints meet the ITER requirements, that is, 3 kN and 8 kN, respectively.

  4. R and D activities on manufacturing plasma-facing unit for prototype of ITER divertor outer target in JADA

    Energy Technology Data Exchange (ETDEWEB)

    Ezato, Koichiro, E-mail: ezatok.koichiro@jaea.go.jp [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka, Ibaraki (Japan); Suzuki, Satoshi; Seki, Yohji; Nishi, Hiroshi; Mohri, Kensuke; Enoeda, Mikio [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka, Ibaraki (Japan)

    2012-08-15

    Japan Domestic Agency (JADA) carried out R and Ds activities to improve joining CFC monoblocks onto a CuCrZr cooling tube in PFUs to boost the success rate of joint and to confirm load carrying capability of the monoblock attachments to Steel Support Structure (SSS) against tensile force simulating electromagnetic load to pull PFUs from SSS. In joining the CFC monoblocks to the cooling tube, JADA has adopted brazing by using noble-metal-free filler with the following improvements; (1) metalizing joint surface of CFC using Ti-coating with accurate thickness controlling, (2) Changing buffer layer material from soft pure copper to Cu-W alloy. By using the present improved joint, JADA has manufactured three mock-ups with 5 CFC monoblocks and tested against repetitive high heat loads more than 20 MW/m{sup 2}. All of CFC monoblocks of each mockup can survive the high heat loads throughout 1000 cycles with no degradation of heat removal capability. Regarding the load carrying capability of monoblock attachments to SSS, tensile experiments were carried out using the same geometries of CFC and tungsten monoblocks in PFUs and the results show that both geometries and joints meet the ITER requirements, that is, 3 kN and 8 kN, respectively.

  5. Development of bonding techniques between tungsten and copper alloy for plasma facing components by HIP method (2). Bonding between tungsten and DS-copper

    International Nuclear Information System (INIS)

    Saito, Shigeru; Fukaya, Kiyoshi; Eto, Motokuni; Ishiyama, Shintaro; Akiba, Masato

    2000-02-01

    Recently, W (tungsten)-alloys are considered as plasma facing material (PFM) for ITER because of these many favorable properties such as high melting point (3655 K), relatively high thermal conductivity and higher resistivity for plasma sputtering. On the other hand, Cu-alloys, especially DS (dispersion strengthened)-Cu, are proposed as heat sink materials because of its high thermal conductivity and good mechanical properties at high temperature. Plasma facing components (PFC) are designed as the duplex structure where W armor tiles are bonded with Cu-alloy heat sink. Then, we started the bonding technology development by hot isostatic press (HIP) method to bond W with Cu-alloys because of its many advantages. Until now, it was reported that we could get the best HIP bonding conditions for W and OFHC-Cu and the tensile strength was similar with HIP treated OFHC-Cu. In this experiments, bonding tests of W and DS-Cu with insert material were performed. As insert material, OFHC-Cu was used with different thickness. Bonding conditions were selected as 1273 K x 2 hours x 147 MPa. Bonding tests with 0.3 to 1.8 mm thickness OFHC-Cu were successfully bonded but with 0.1 mm thickness was not bonded. From the results of tensile tests, the tensile strength of the specimens with 0.3 and 0.5 mm thickness were decreased at elevated temperature. It was shown that over 1.0 mm thickness OFHC-Cu insert may be needed and the tensile strength were a little higher than that of HIP treated OFHC-Cu. (author)

  6. Design and Test of Wendelstein 7-X Water-Cooled Divertor Scraper

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J. [Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Garching, Germany; Greuner, Henri [Max Planck Institute for Plasma Physics, Garching, Germany; Ehrke, Gunnar [Max Planck Institute of Plasma Physics, Greifswald, Germany; Boeswirth, Bernd [Max Planck Institute for Plasma Physics, Garching, Germany; Wang, Zhongwei [Max Planck Institute for Plasma Physics, Garching, Germany; Clark, Emily [The University of Tennessee, Knoxville; Lumsdaine, Arnold [ORNL; Tretter, Jorg [Max Planck Institute for Plasma Physics, Garching, Germany; Junghanns, Patrick [Max Planck Institute for Plasma Physics, Garching, Germany; Stadler, Reinhold [Max Planck Institute for Plasma Physics, Garching, Germany; McGinnis, William Dean [ORNL; Lore, Jeremy D. [ORNL; Team, W7-X [Max-Planck-Institut fur Plasmaphysik, Griefswald, Germany

    2018-04-01

    Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipes of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper.

  7. Final Report for Project DE-SC0006958: "An Investigation of the Effects of magnetic Fields and Collisionality on Shock Formation in Radiatively Cooled Plasma Flows"

    Energy Technology Data Exchange (ETDEWEB)

    Bott-Suzuki, Simon

    2014-11-05

    We have developed a new experimental platform to study bow-shock formation in plasma flows generated using an inverse wire array z-pinch. We have made significant progress on the analysis of both hydrodynamic and magnetized shocks using this system. The hydrodynamic experiments show formation of a well-defined Mach cone, and highly localized shock strong associated with radiative losses and rapidly cooling over the shock. Magnetized shocks show that the balance of magnetic and ram pressures dominate the evolution of the shock region, generating a low plasma beta void around the target. Manuscripts are in preparation for publication on both these topics. We have also published the development of a novel diagnostic method which allow recovery of interferometry and self-emission data along the same line of sight. Finally, we have carried out work to integrate a kinetic routine with the 3D MHD code Gorgon, however it remains to complete this process. Both undergraduate and graduate students have been involved in both the experimental work and publications.

  8. Hygrothermal conditions in cold, north facing attic spaces under the eaves with vapour-open roofing underlay in a cool, temperate climate

    DEFF Research Database (Denmark)

    Bjarløv, Søren Peter; Johnston, C.J.; Hansen, M.H.

    2016-01-01

    compliance with the current Danish Building Regulations (BR10) for airtightness (moisture levels in attics with vapour-open roofing underlays. North facing cold attic spaces under the eaves constitute a worst case scenario. Following best...... to allow an influx of 3.3 l/s of conditioned indoor air 20 °C and 60% RH at a pressure difference of 50 Pa) and ventilation (singled-sided, passive ventilation) contained more moisture and had significantly higher levels of mould growth than the non-ventilated attics. Under the same physical conditions...... the ‘pressure equalized’ attic rooms were found to have moisture levels in between those observed in the ventilated and non-ventilated attic rooms. Likewise, the observed levels of mould growth were in between those observed in the cases of the ventilated and non-ventilated attic rooms. Attics with reduced...

  9. A dislocation-based crystal viscoplasticity model with application to micro-engineered plasma-facing materials

    Energy Technology Data Exchange (ETDEWEB)

    Rivera, David; Huang, Yue; Po, Giacomo; Ghoniem, Nasr M., E-mail: ghoniem@ucla.edu

    2017-03-15

    Materials developed with special surface architecture are shown here to be more resilient to the transient thermomechanical environments imposed by intermittent exposures to high heat flux thermal loading typical of long-pulse plasma transients. In an accompanying article, we present experimental results that show the relaxation of residual thermal stresses in micro-engineered W surfaces. A dislocation-based model is extended here within the framework of large deformation crystal plasticity. The model is applied to the deformation of single crystals, polycrystals, and micro-engineered surfaces composed of a uniform density of micro-pillars. The model is utilized to design tapered surface micro-pillar architecture, composed of a Re core and W coatings. Residual stresses generated by cyclic thermomechanical loading of these architectures show that the surface can be in a compressive stress state, following a short shakedown plasma exposure, thus mitigating surface fracture. - • Materials developed with special surface architecture are shown to be more resilient to the transient thermomechanical plasma transients. • A dislocation-based model is extended within the framework of large deformation crystal plasticity. • The model is applied to the deformation of single crystals, polycrystals, and micro-engineered surfaces. • The model is utilized to design tapered surface micro-pillar architecture, composed of a Re core and W coatings. • Residual stresses generated by cyclic thermomechanical loading show that the surface can be in a compressive stress state, thus mitigating surface fracture.

  10. A dislocation-based crystal viscoplasticity model with application to micro-engineered plasma-facing materials

    International Nuclear Information System (INIS)

    Rivera, David; Huang, Yue; Po, Giacomo; Ghoniem, Nasr M.

    2017-01-01

    Materials developed with special surface architecture are shown here to be more resilient to the transient thermomechanical environments imposed by intermittent exposures to high heat flux thermal loading typical of long-pulse plasma transients. In an accompanying article, we present experimental results that show the relaxation of residual thermal stresses in micro-engineered W surfaces. A dislocation-based model is extended here within the framework of large deformation crystal plasticity. The model is applied to the deformation of single crystals, polycrystals, and micro-engineered surfaces composed of a uniform density of micro-pillars. The model is utilized to design tapered surface micro-pillar architecture, composed of a Re core and W coatings. Residual stresses generated by cyclic thermomechanical loading of these architectures show that the surface can be in a compressive stress state, following a short shakedown plasma exposure, thus mitigating surface fracture. - • Materials developed with special surface architecture are shown to be more resilient to the transient thermomechanical plasma transients. • A dislocation-based model is extended within the framework of large deformation crystal plasticity. • The model is applied to the deformation of single crystals, polycrystals, and micro-engineered surfaces. • The model is utilized to design tapered surface micro-pillar architecture, composed of a Re core and W coatings. • Residual stresses generated by cyclic thermomechanical loading show that the surface can be in a compressive stress state, thus mitigating surface fracture.

  11. Experimental study of divertor plasma-facing components damage under a combination of pulsed and quasi-stationary heat loads relevant to expected transient events at ITER

    International Nuclear Information System (INIS)

    Klimov, N S; Podkovyrov, V L; Kovalenko, D V; Zhitlukhin, A M; Barsuk, V A; Mazul, I V; Giniyatulin, R N; Kuznetsov, V Ye; Riccardi, B; Loarte, A; Merola, M; Koidan, V S; Linke, J; Landman, I S; Pestchanyi, S E; Bazylev, B N

    2011-01-01

    This paper concerns the experimental study of damage of ITER divertor plasma-facing components (PFCs) under a combination of pulsed plasma heat loads (representative of controlled ITER type I edge-localized modes (ELMs)) and quasi-stationary heat loads (representative of the high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events). The PFC's tungsten armor damage under pulsed plasma exposure was driven by (i) the melt layer motion, which leads to bridges formation between neighboring tiles and (ii) the W brittle failure giving rise to a stable crack pattern on the exposed surface. The crack width reaches a saturation value that does not exceed some tens of micrometers after several hundreds of ELM-like pulses. HHF thermal fatigue tests have shown (i) a peeling-off of the re-solidified material due to its brittle failure and (ii) a significant widening (up to 10 times) of the cracks and the formation of additional cracks.

  12. Laser-induced breakdown spectroscopy for the analysis of plasma facing components of tokamaks: parametric study and calibration-free measurements

    International Nuclear Information System (INIS)

    Mercadier, L.

    2011-09-01

    During the operation of a nuclear fusion device like the future reactor ITER, a fraction of tritium is trapped in the plasma facing components and has to be measured in order to fulfill nuclear safety requirements. Laser-induced breakdown spectroscopy (LIBS) is proposed to achieve this measurement. The laser plasma produced on carbon fibre composite tiles from the Tore Supra reactor is analyzed via a parametric study: it has to have a temperature over 10000 K and an electron density over 10 17 cm -3 to optimize the application. A calibration-free procedure that takes into account self-absorption is proposed to determine the relative concentration of hydrogen from the experimental spectra. The time- and space-resolved spectral emission of the plasma plume is investigated and reveals the presence of a temperature gradient from the core towards the periphery. This gradient is taken into account and the H/C concentration ratio is deduced. The accuracy of the results is evaluated and discussed. The study of the D/H isotopic ratio under low pressure argon reveals the presence of plume segregation that leads to an error of about 50%, error that can partially be reduced. Tungsten materials are investigated and difficulties related to spectroscopic databases are discussed. Finally, the feasibility of LIBS analysis with depth resolution is validated for multilayered metallic samples. (author)

  13. Study of the hydrogen behavior in amorphous hydrogenated materials of type a - C:H and a - SiC:H facing fusion reactor plasma

    International Nuclear Information System (INIS)

    Barbier, G.

    1997-01-01

    Plasma facing components of controlled fusion test devices (tokamaks) are submitted to several constraints (irradiation, high temperatures). The erosion (physical sputtering and chemical erosion) and the hydrogen recycling (retention and desorption) of these materials influence many plasma parameters and thus affect drastically the tokamak running. First, we will describe the different plasma-material interactions. It will be pointed out, how erosion and hydrogen recycling are strongly related to both chemical and physical properties of the material. In order to reduce these interactions, we have selected two amorphous hydrogenated materials (a-C:H and a-SiC:H), which are known for their good thermal and chemical qualities. Some samples have been then implanted with lithium ions at different fluences. Our materials have been then irradiated with deuterium ions at low energy. From our results, it is shown that both the lithium implantation and the use of an a - SiC:H substrate can be beneficial in enhancing the hydrogen retention. These results were completed with thermal desorption studies of these materials. It was evidenced that the hydrogen fixation was more efficient in a-SiC:H than in a-C:H substrate. Results in good agreement with those described above have been obtained by exposing a - C:H and a - SiC:H samples to the scrape off layer of the tokamak of Varennes (TdeV, Canada). A modelling of hydrogen diffusion under irradiation has been also proposed. (author)

  14. Hydrogen behaviour study in plasma facing a-C:H and a-SiC:H hydrogenated amorphous materials for fusion reactors

    International Nuclear Information System (INIS)

    Barbier, Gauzelin

    1997-01-01

    Plasma facing components of controlled fusion test devices (tokamaks) are submitted to several constraints (irradiation, high temperatures). The erosion (physical sputtering and chemical erosion) and the hydrogen recycling (retention and desorption) of these materials influence many plasma parameters and thus affect drastically the tokamak running. Firstly, we will describe the different plasma-material interactions. It will be pointed out, how erosion and hydrogen recycling are strongly related to both chemical and physical properties of the material. In order to reduce this interactions, we have selected two amorphous hydrogenated materials (a-C:H and a-SiC:H), which are known for their good thermal and chemical qualities. Some samples have been then implanted with lithium ions at different fluences. Our materials have been then irradiated with deuterium ions at low energy. From our results, it is shown that both the lithium implantation and the use of an a-SiC:H substrate can be benefit in enhancing the hydrogen retention. These results were completed with thermal desorption studies of these materials. It was evidenced that the hydrogen fixation was more efficient in a -SiC:H than in a-C:H substrate. Results in good agreement with those described above have been obtained by exposing a-C:H and a-SiC:H samples to the scrape off layer of the tokamak of Varennes (TdeV, Canada). A modeling of hydrogen diffusion under irradiation has been also proposed. (author)

  15. Effects of the plasma-facing materials on the negative ion H ‑ density in an ECR (2.45 GHz) plasma

    Science.gov (United States)

    Bentounes, J.; Béchu, S.; Biggins, F.; Michau, A.; Gavilan, L.; Menu, J.; Bonny, L.; Fombaron, D.; Bès, A.; Lebedev, Yu A.; Shakhatov, V. A.; Svarnas, P.; Hassaine, T.; Lemaire, J. L.; Lacoste, A.

    2018-05-01

    Within the framework of fundamental research, the present work focuses on the role of surface material in the production of H ‑ negative ion, with a potential application of designing cesium-free H ‑ negative ion sources oriented to fusion application. It is widely accepted that the main reaction leading to H ‑ production, in the plasma volume, is the dissociative attachment of low-energy electrons (T e ≤ 1 eV) on highly ro-vibrationally excited hydrogen molecules. In parallel with other mechanisms, the density of these excited molecules may be enhanced by means of the recombinative desorption, i.e. the interaction between surface absorbed atoms with other atoms (surface adsorbed or not) through the path {H}{{ads}}+{H}{{gas}/{{ads}}}\\to {H}2{(v,J)}{{gas}}+{{Δ }}E. Accordingly, a systematic study on the role played by the surface in this reaction, with respect to the production of H ‑ ion in the plasma volume, is here performed. Thus, tantalum and tungsten (already known as H ‑ enhancers) and quartz (inert surface) materials are employed as inner surfaces of a test bench chamber. The plasma inside the chamber is produced by electron cyclotron resonance (ECR) driving and it is characterized with conventional electrostatic probes, laser photodetachment, and emission and absorption spectroscopy. Two different positions (close to and away from the ECR driving zone) are investigated under various conditions of pressure and power. The experimental results are supported by numerical data generated by a 1D model. The latter couples continuity and electron energy balance equations in the presence of magnetic field, and incorporates vibrational kinetics, H2 molecular reactions, H electronically excited states and ground-state species kinetics. In the light of this study, recombinative desorption has been evidenced as the most probable mechanism, among others, responsible for an enhancement by a factor of about 3.4, at 1.6 Pa and 175 W of microwave power, in the

  16. Effect of Seminal Plasma Removal on Cell Membrane, Acrosomal Integrity and Mitochondrial Activity of Cooled Stallion Semen

    Directory of Open Access Journals (Sweden)

    Dhafer M. Aziz

    2012-07-01

    Full Text Available Fresh semen samples were collected from 11 warm blood stallions, each ejaculate was distributed into three equal parts. The first part was diluted in a skim milk-glucose diluent (SMG, the second part was diluted in a skim milk-glucose supplemented with Tyrode's medium (SMG-T, the third part was centrifuged to remove the seminal plasma, then the sperm was resuspended in the second diluent (SMG-T-C. The diluted semen were evaluated immediately after dilution (0 hour and at 24, 48, 72, and 96 hours of storage at 5°C. Flow cytometry was performed to determine sperm viability, mitochondrial activity and acrosomal integrity. Immediately after dilution the tested parameters of sperms that diluted in SMG-T was significantly (P<0.001 higher than those diluted with SMG and SMG-T-C, and with SMG-T-C were higher significantly (P<0.05 than those diluted with SMG. The decreasing rate in tested sperm parameter was greater significantly (P<0.001 in semen samples which were diluted with SMG than those diluted with SMG-T and SMG-T-C. In conclusion, the present study indicated that viability, acrosomal integrity, and mitochondrial activity of stallion sperms were better preserved in SMG-T in comparison with SMG, also centrifugation and removal of the seminal plasma have an adverse effect on these three sperm parameters.

  17. Particle-in-cell studies of fast-ion slowing-down rates in cool tenuous magnetized plasma

    Science.gov (United States)

    Evans, Eugene S.; Cohen, Samuel A.; Welch, Dale R.

    2018-04-01

    We report on 3D-3V particle-in-cell simulations of fast-ion energy-loss rates in a cold, weakly-magnetized, weakly-coupled plasma where the electron gyroradius, ρe, is comparable to or less than the Debye length, λDe, and the fast-ion velocity exceeds the electron thermal velocity, a regime in which the electron response may be impeded. These simulations use explicit algorithms, spatially resolve ρe and λDe, and temporally resolve the electron cyclotron and plasma frequencies. For mono-energetic dilute fast ions with isotropic velocity distributions, these scaling studies of the slowing-down time, τs, versus fast-ion charge are in agreement with unmagnetized slowing-down theory; with an applied magnetic field, no consistent anisotropy between τs in the cross-field and field-parallel directions could be resolved. Scaling the fast-ion charge is confirmed as a viable way to reduce the required computational time for each simulation. The implications of these slowing down processes are described for one magnetic-confinement fusion concept, the small, advanced-fuel, field-reversed configuration device.

  18. Near-Infrared Spectroscopy for Zeeman Spectra of Ti I in Plasma Using a Facing Target Sputtering System

    Science.gov (United States)

    Kobayashi, Shinji; Nishimiya, Nobuo; Suzuki, Masao

    2017-10-01

    The saturated absorption lines of neutral titanium were measured in the region of 9950-14380 cm-1 using a Ti:sapphire ring laser. A facing target sputtering system was used to obtain the gaseous state of a Ti I atom. The Zeeman splitting of 38 transitions was observed under the condition that the electric field component of a linearly polarized laser beam was parallel to the magnetic field. The gJ factors of the odd parity states were determined for 28 states belonging to 3d24s4p and 3d34p using those of the even parity states reported by Stachowska in 1997. The gJ factors of z5P1,2,3 levels were newly determined. gJ of y3F2, y3D2, z3P2, and z5S2 levels were refined.

  19. Adipose-derived stromal vascular fraction cells and platelet-rich plasma: basic and clinical evaluation for cell-based therapies in patients with scars on the face.

    Science.gov (United States)

    Gentile, Pietro; De Angelis, Barbara; Pasin, Methap; Cervelli, Giulio; Curcio, Cristiano B; Floris, Micol; Di Pasquali, Camilla; Bocchini, Ilaria; Balzani, Alberto; Nicoli, Fabio; Insalaco, Chiara; Tati, Eleonora; Lucarini, Lucilla; Palla, Ludovico; Pascali, Michele; De Logu, Pamela; Di Segni, Chiara; Bottini, Davide J; Cervelli, Valerio

    2014-01-01

    Actually, autologous fat grafts have many clinical applications in breast surgery, facial rejuvenation, buttock augmentation, and Romberg syndrome as well as a treatment of liposuction sequelae. The aim of this article was to describe the preparation and isolation procedures for stromal vascular fraction (SVF), the preparation of platelet-rich plasma (PRP), and the clinical application in the treatment of the scar on the face. Ten patients with burns sequelae (n = 6) and post-traumatic scars (n = 4) were treated with SVF-enhanced autologous fat grafts obtained by the Celution System. Another 10 patients with burns sequelae (n = 5) and post-traumatic scars (n = 5) were treated with fat grafting based on the Coleman technique mixed with 0.5 mL of PRP.To assess the effects of their treatment, the authors compared their results with those of a control group consisting of 10 patients treated with centrifuged fat. In the patients treated with SVF-enhanced autologous fat grafts, we observed a 63% maintenance of contour restoring after 1 year compared with only 39% of the control group (n = 10) treated with centrifuged fat graft (P < 0.0001). In the patients treated with fat grafting and PRP, we observed a 69% maintenance of contour restoring after 1 year compared with that of the control group (n = 10). Autologous fat grafting is a good method for the correction of scars on the face instead of the traditional scar surgical excision.

  20. Effects of modified surfaces produced at plasma-facing surface on hydrogen release behavior in the LHD

    Directory of Open Access Journals (Sweden)

    Y. Nobuta

    2017-08-01

    Full Text Available In the present study, an additional deuterium (D ion irradiation was performed against long-term samples mounted on the helical coil can and in the outer private region in the LHD during the 17th experimental campaign. Based on the release behavior of the D and hydrogen (H retained during the experimental campaign, the difference of release behavior at the top surface and in bulk of modified surfaces is discussed. Almost all samples on the helical coil can were erosion-dominant and some samples were covered with boron or carbon, while a very thick carbon films were formed in the outer private region. In the erosion-dominant area, the D desorbed at much lower temperatures compared to that of H retained during the LHD plasma operation. For the samples covered with boron, the D tended to desorb at lower temperatures compared to H. For the carbon deposition samples, the D desorbed at much higher temperatures compared to no deposition and boron-covered samples, which was very similar to that of H. The D retention capabilities at the top surface of carbon and boron films were 2–3 times higher than no deposition area. The results indicate that the retention and release behavior at the top surface of the modified layer can be different from that of bulk substrate material.

  1. Development of real time system imaging software for the protection of plasma facing components(PFCs) in Wendelstein 7-X

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Adnan; Jakubowski, Marcin; Sunn Pedersen, Thomas; Rodatos, Alexander [Max-Planck-Institute for Plasma Physics, Greifswald (Germany); Greuner, Henri [Max-Planck-Institute for Plasma Physics, Garching (Germany)

    2016-07-01

    One of the main aims of Wendelstein 7-X, an advanced stellarator in Greifswald, is the investigation of quasi-steady state operation of magnetic fusion devices, for which power exhaust is a very important issue. The predominant fraction of the energy lost from the confined plasma region will be removed by 10 so-called island divertors, which can sustain up to 10 MW/Sq-m. In order to protect the divertor elements from overheating and to monitor power deposition onto the divertor elements, 10 state-of-the-art infrared endoscopes will be installed at W7-X and software is under development for real-time analysis of automatic detection of the hot spots and other abnormal events. The pre-defined algorithms designed for early detection of defects e.g. hotspots, surface layers and delaminations during the discharge are being implemented into the software acquiring the images from the infrared cameras and broadcast them to the main Discharge Control System(DCS). This allows for automatic control of the scenario of the discharge in order to assure safe operation of W7-X. The first online tests of the software will soon be performed at GLADIS in Garching.

  2. Efficacy of autologous platelet-rich plasma combined with fractional ablative carbon dioxide resurfacing laser in treatment of facial atrophic acne scars: A split-face randomized clinical trial

    OpenAIRE

    Gita Faghihi; Shima Keyvan; Ali Asilian; Saeid Nouraei; Shadi Behfar; Mohamad Ali Nilforoushzadeh

    2016-01-01

    Background: Autologous platelet-rich plasma has recently attracted significant attention throughout the medical field for its wound-healing ability. Aims: This study was conducted to investigate the potential of platelet-rich plasma combined with fractional laser therapy in the treatment of acne scarring. Methods: Sixteen patients (12 women and 4 men) who underwent split-face therapy were analyzed in this study. They received ablative fractional carbon dioxide laser combined with intradermal ...

  3. Study of heat fluxes on plasma facing components in a tokamak from measurements of temperature by infrared thermography

    International Nuclear Information System (INIS)

    Daviot, R.

    2010-05-01

    The goal of this thesis is the development of a method of computation of those heat loads from measurements of temperature by infrared thermography. The research was conducted on three issues arising in current tokamaks but also future ones like ITER: the measurement of temperature on reflecting walls, the determination of thermal properties for deposits observed on the surface of tokamak components and the development of a three-dimensional, non-linear computation of heat loads. A comparison of several means of pyrometry, monochromatic, bi-chromatic and photothermal, is performed on an experiment of temperature measurement. We show that this measurement is sensitive to temperature gradients on the observed area. Layers resulting from carbon deposition by the plasma on the surface of components are modeled through a field of equivalent thermal resistance, without thermal inertia. The field of this resistance is determined, for each measurement points, from a comparison of surface temperature from infrared thermographs with the result of a simulation, which is based on a mono-dimensional linear model of components. The spatial distribution of the deposit on the component surface is obtained. Finally, a three-dimensional and non-linear computation of fields of heat fluxes, based on a finite element method, is developed here. Exact geometries of the component are used. The sensitivity of the computed heat fluxes is discussed regarding the accuracy of the temperature measurements. This computation is applied to two-dimensional temperature measurements of the JET tokamak. Several components of this tokamak are modeled, such as tiles of the divertor, upper limiter and inner and outer poloidal limiters. The distribution of heat fluxes on the surface of these components is computed and studied along the two main tokamak directions, poloidal and toroidal. Toroidal symmetry of the heat loads from one tile to another is shown. The influence of measurements spatial resolution

  4. Activities of HIP joining of plasma-facing armors in the blanket first-wall in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang-Il, E-mail: yijung@kaeri.re.kr [Korea Atomic Energy Research Institute, Daedeok-daero, Daejeon 34057 (Korea, Republic of); Park, Jeong-Yong; Choi, Byoung-Kwon; Lee, Jung-Suk; Kim, Hyun-Gil; Park, Dong-Jun; Park, Jung-Hwan; Kim, Suk-Kwon; Lee, Dong-Won [Korea Atomic Energy Research Institute, Daedeok-daero, Daejeon 34057 (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Gwahak-ro, Yuseong, Daejeon 34133 (Korea, Republic of)

    2016-11-01

    Highlights: • HIP joints of Be/CuCrZr, Be/FMS, W/FMS were demonstrated. • The process conditions for HIP joining were developed. • For the joining of Be, coating interlayers as well as thick diffusion barrier was developed. • For the joining of W, double-staged HIP was applied for the joint integrity. • No significant defects nor a brittle failure were observed along the joint interface. - Abstract: Joining technology for dissimilar materials was developed for the fabrication of an ITER blanket first-wall, which consisted of Be, CuCrZr, and stainless steel (SS). The Be/CuCrZr/SS joint was fabricated using a hot isostatic pressing (HIP) method. Beryllium armor was joined to the CuCrZr/SS block at 580 °C under 100 MPa. The optimal interlayer coatings of Cr/Cu and Ti/Cr/Cu were developed using an ion-beam assisted physical vapor deposition. Beryllium is also a candidate armor material for the TBM first-wall. Successful joining of Be to ferritic-martensitic steel (FMS) was accomplished using an HIP method by introducing the thick diffusion barrier. A thick diffusion barrier of a Cu foil(10 μm) limited the excessive diffusion and prevented the formation of brittle phases at the Be/FMS interface. Be and FMS were bonded at 650–850 °C; however, a temperature of lower than 750 °C was recommended to avoid material degradation of FMS. In addition, the joining of W to FMS has been developed. Tungsten is another armor material applicable to more severe plasma conditions. The large difference in the thermal expansion between W and FMS was resolved by introducing the Ti interlayer and Mo separator. Moreover, the double-staged HIP (the first stage at 900 °C and 100 MPa and the second stage at 750 °C and 70 MPa) was applied to suppress the edge delamination of W/FMS joints during thermal history.

  5. Thermal radiation characteristics and direct evidence of tungsten cooling on the way to nanostructure formation on its surface

    Energy Technology Data Exchange (ETDEWEB)

    Takamura, S., E-mail: takamura@aitech.ac.jp [Faculty of Engineering, Aichi Institute of Technology, Yakusa-cho, Toyota 470-0392 (Japan); Miyamoto, T. [Faculty of Engineering, Aichi Institute of Technology, Yakusa-cho, Toyota 470-0392 (Japan); Ohno, N. [Graduate School of Engineering, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan)

    2013-07-15

    The physical properties of tungsten with nanostructure on its surface are investigated focusing on the thermal radiation and cooling characteristics. First, direct evidence of substantial W surface cooling has been clearly shown with use of a very thin thermocouple inserted into W target, which solves an uncertainty associated with a radiation thermometer. Second, the above measurements of W surface temperature make it possible to estimate quantitatively the total emissivity from which we may evaluate the radiative power through the Stefan–Boltzmann equation, which is very important for mitigation evaluation of a serious plasma heat load to the plasma-facing component.

  6. Thermal radiation characteristics and direct evidence of tungsten cooling on the way to nanostructure formation on its surface

    International Nuclear Information System (INIS)

    Takamura, S.; Miyamoto, T.; Ohno, N.

    2013-01-01

    The physical properties of tungsten with nanostructure on its surface are investigated focusing on the thermal radiation and cooling characteristics. First, direct evidence of substantial W surface cooling has been clearly shown with use of a very thin thermocouple inserted into W target, which solves an uncertainty associated with a radiation thermometer. Second, the above measurements of W surface temperature make it possible to estimate quantitatively the total emissivity from which we may evaluate the radiative power through the Stefan–Boltzmann equation, which is very important for mitigation evaluation of a serious plasma heat load to the plasma-facing component

  7. Face to Face

    OpenAIRE

    Robert Leckey

    2013-01-01

    This paper uses Queer theory, specifically literature on Bowers v. Hardwick, to analyze debates over legislation proposed in Quebec regarding covered faces. Queer theory sheds light on legal responses to the veil. Parliamentary debates in Quebec reconstitute the polity, notably as secular and united. The paper highlights the contradictory and unstable character of four binaries: legislative text versus social practice, act versus status, majority versus minority, and knowable versus unknowabl...

  8. Cooling towers

    International Nuclear Information System (INIS)

    Boernke, F.

    1975-01-01

    The need for the use of cooling systems in power plant engineering is dealt with from the point of view of a non-polluting form of energy production. The various cooling system concepts up to the modern natural-draught cooling towers are illustrated by examples. (TK/AK) [de

  9. The Plasma-Facing Components Transporter (PFCT) : a Prototype System for PFC Replacement on the new ITER 2001 Cassette Mock-up

    International Nuclear Information System (INIS)

    Micciche, G.; Lorenzelli, L.; Muro, L.; Irving, M.

    2006-01-01

    The remote maintainability of the early ITER divertor cassette (based on the ITER 1998 design) was successfully proved during test campaigns carried out in the Divertor Refurbishment Platform (DRP) at the ENEA research centre at Brasimone over the period 1999-2003. Due to subsequent major modifications in the ITER divertor cassette design, the main focus over the past few years has been on the design and manufacture of the various components, devices and tools needed for refurbishment of the new ITER 2001 Divertor Cassette. The design of this new cassette differs substantially from the earlier version: in particular the shape, weight and attachment system of the Plasma Facing Components (PFC's) has been completely revised, and this also entailed a review of the procedures adopted for its refurbishment. One of the major requirements of the cassette refurbishment process is removal and replacement of the three PFC's. In the old cassette concept, target replacement was performed by means of a purpose-built '' C '' frame slung from a standard bridge crane. The 2001 cassette design precludes such handling methods for a number of reasons, notably because of the extremely tight inter-PFC clearances, and the need for controlled inclination of the target in addition to normal translational movements, both impossible with a simple Cartesian crane. To demonstrate the refurbishment feasibility operations for the new ITER Divertor 2001 cassettes, an experimental machine known as the Plasma-Facing Component Transporter (PFCT) has been designed, fabricated and commissioned in the years 2004-5. This full six degree-of-freedom system has been designed to handle payloads of up to 5 tonnes with good positional accuracy, and axes capable of very low joint velocities, including inclination of the PFC's over the range of ± 10 o in both horizontal axes, and controlled rotation about the vertical axis. Preliminary trials carried out during the commissioning phase have proved its

  10. Study of the hydrogen behavior in amorphous hydrogenated materials of type a - C:H and a - SiC:H facing fusion reactor plasma; Etude du comportament de l`hydrogene dans des materiaux amorphes hydrogenes de type a - C:H et a - SiC:H devant faire face au plasma des reacteurs a fusion

    Energy Technology Data Exchange (ETDEWEB)

    Barbier, G. [Lyon-1 Univ., 69 - Villeurbanne (France). Inst. de Physique Nucleaire

    1997-04-10

    Plasma facing components of controlled fusion test devices (tokamaks) are submitted to several constraints (irradiation, high temperatures). The erosion (physical sputtering and chemical erosion) and the hydrogen recycling (retention and desorption) of these materials influence many plasma parameters and thus affect drastically the tokamak running. First, we will describe the different plasma-material interactions. It will be pointed out, how erosion and hydrogen recycling are strongly related to both chemical and physical properties of the material. In order to reduce these interactions, we have selected two amorphous hydrogenated materials (a-C:H and a-SiC:H), which are known for their good thermal and chemical qualities. Some samples have been then implanted with lithium ions at different fluences. Our materials have been then irradiated with deuterium ions at low energy. From our results, it is shown that both the lithium implantation and the use of an a - SiC:H substrate can be beneficial in enhancing the hydrogen retention. These results were completed with thermal desorption studies of these materials. It was evidenced that the hydrogen fixation was more efficient in a-SiC:H than in a-C:H substrate. Results in good agreement with those described above have been obtained by exposing a - C:H and a - SiC:H samples to the scrape off layer of the tokamak of Varennes (TdeV, Canada). A modelling of hydrogen diffusion under irradiation has been also proposed. (author) 176 refs.

  11. Processing of W-Cu functionally graded materials (FGM) through the powder metallurgy route: application as plasma facing components for ITER-like thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Raharijaona, J.J.

    2009-11-01

    The aim of this study was to study and optimize the sintering of W-Cu graded composition materials, for first wall of ITER-like thermonuclear reactor application. The graded composition in the material generates graded functional properties (Functionally Graded Materials - FGM). Rough thermomechanical calculations have shown the interest of W-Cu FGM to improve the lifetime of Plasma Facing Components (PFC). To process W-Cu FGM, powder metallurgy route was analyzed and optimized from W-CuO powder mixtures. The influence of oxide reduction on the sintering of powder mixtures was highlighted. An optimal heating treatment under He/H 2 atmosphere was determined. The sintering mechanisms were deduced from the analysis of the effect of the Cu-content. Sintering of W-Cu materials with a graded composition and grain size has revealed two liquid migration steps: i) capillary migration, after the Cu-melting and, ii) expulsion of liquid, at the end of sintering, from the dense part to the porous part, due to the continuation of W-skeleton sintering. These two steps were confirmed by a model based on capillary pressure calculation. In addition, thermal conductivity measurements were conducted on sintered parts and showed values which gradually increase with the Cu-content. Hardness tests on a polished cross-section in the bulk are consistent with the composition profiles obtained and the differential grain size. (author)

  12. Microneedling combined with platelet-rich plasma or trichloroacetic acid peeling for management of acne scarring: A split-face clinical and histologic comparison.

    Science.gov (United States)

    El-Domyati, Moetaz; Abdel-Wahab, Hossam; Hossam, Aliaa

    2018-02-01

    Minimally invasive procedures provide effective, safe, relatively long-lasting, and natural results without large damage to the skin. A combination treatment is considered an approach that includes at least 2 different and unrelated modalities. This study aims to evaluate the use and effectiveness of some combined minimally invasive procedures for management of acne scarring. Twenty-four volunteers with postacne atrophic scars were randomly divided into 3 equal groups according to performed procedure on each side of the face (microneedling by dermaroller alone or combined with platelet-rich plasma [PRP] or trichloroacetic acid [TCA] 15% peeling) and received 6 bi-weekly sessions of treatment. Photography and punch biopsies were taken before and after 3 months of treatment for clinical, histological, and histometrical evaluation. Combined treatment of dermaroller and PRP or dermaroller and TCA 15% showed significant improvement when compared with dermaroller alone (P = .015 and .011 respectively). Epidermal thickness showed statistically significant increase in studied groups, mainly after dermaroller and TCA 15%. Moreover, the 3 studied groups showed more organized collagen bundles and newly formed collagen formation and markedly decreased abnormal elastic fibers. Based on the clinical, histometrical, and histochemical assessment, inspite that most volunteers showed significant improvement after treatment, however, the combined use of dermaroller and TCA 15% was more effective in postacne atrophic scars than the use of dermaroller and PRP or dermaroller only. © 2017 Wiley Periodicals, Inc.

  13. Plasma shutdown device

    International Nuclear Information System (INIS)

    Hosogane, Nobuyuki; Nakayama, Takahide.

    1985-01-01

    Purpose: To prevent concentration of plasma currents to the plasma center upon plasma shutdown in a torus type thermonuclear device by the injection of fuels to the plasma center thereby prevent plasma disruption at the plasma center. Constitution: The plasma shutdown device comprises a plasma current measuring device that measures the current distribution of plasmas confined within a vacuum vessel and outputs a control signal for cooling the plasma center when the plasma currents concentrate to the plasma center and a fuel supply device that supplies fuels to the plasma center for cooling the center. The fuels are injected in the form of pellets into the plasmas. The direction and the velocity of the injection are set such that the pellets are ionized at the center of the plasmas. (Horiuchi, T.)

  14. Heteroepitaxial growth of In-face InN on GaN (0001) by plasma-assisted molecular-beam epitaxy

    International Nuclear Information System (INIS)

    Dimakis, E.; Iliopoulos, E.; Tsagaraki, K.; Kehagias, Th.; Komninou, Ph.; Georgakilas, A.

    2005-01-01

    The thermodynamic aspects of indium-face InN growth by radio frequency plasma-assisted molecular-beam epitaxy (rf-MBE) and the nucleation of InN on gallium-face GaN (0001) surface were investigated. The rates of InN decomposition and indium desorption from the surface were measured in situ using reflected high-energy electron diffraction and the rf-MBE 'growth window' of In-face InN (0001) was identified. It is shown that sustainable growth can be achieved only when the arrival rate of active nitrogen species on the surface is higher than the arrival rate of indium atoms. The maximum substrate temperature permitting InN growth as a function of the active nitrogen flux was determined. The growth mode of InN on Ga-face GaN (0001) surface was investigated by reflected high-energy electron diffraction and atomic force microscopy. It was found to be of the Volmer-Weber-type for substrate temperatures less than 350 deg. C and of the Stranski-Krastanov for substrate temperatures between 350 and 520 deg. C. The number of monolayers of initial two-dimensional growth, in the case of Stranski-Krastanov mode, varies monotonically with substrate temperature, from 2 ML at 400 deg. C to about 12 ML at 500 deg. C. The evolution and coalescence of nucleated islands were also investigated as a function of substrate temperature. It was found that at higher temperature their coalescence is inhibited leading to porous-columnar InN thin films, which exhibit growth rates higher than the nominal value. Therefore, in order to achieve continuous InN layers on GaN (0001) a two-step growth approach is introduced. In that approach, InN is nucleated at low temperatures on GaN and the growth continues until full coalescence of the nucleated islands. Subsequently, this nucleation layer is overgrown at higher substrate temperature in order to achieve high-quality continuous films. The InN films grown by the two-step method were investigated by x-ray diffraction, Hall-effect measurements, and

  15. Spray cooling

    International Nuclear Information System (INIS)

    Rollin, Philippe.

    1975-01-01

    Spray cooling - using water spraying in air - is surveyed as a possible system for make-up (peak clipping in open circuit) or major cooling (in closed circuit) of the cooling water of the condensers in thermal power plants. Indications are given on the experiments made in France and the systems recently developed in USA, questions relating to performance, cost and environmental effects of spray devices are then dealt with [fr

  16. Face to Face

    Directory of Open Access Journals (Sweden)

    Robert Leckey

    2013-12-01

    Full Text Available This paper uses Queer theory, specifically literature on Bowers v. Hardwick, to analyze debates over legislation proposed in Quebec regarding covered faces. Queer theory sheds light on legal responses to the veil. Parliamentary debates in Quebec reconstitute the polity, notably as secular and united. The paper highlights the contradictory and unstable character of four binaries: legislative text versus social practice, act versus status, majority versus minority, and knowable versus unknowable. As with contradictory propositions about homosexuality, contradiction does not undermine discourse but makes it stronger and more agile. Este artículo utiliza la teoría Queer, más concretamente la literatura sobre Bowers vs. Hardwick, para analizar los debates sobre la legislación propuesta en Quebec en relación al velo. La teoría Queer arroja luz sobre las respuestas legales al velo. Los debates parlamentarios en Quebec reconstituyen la forma de gobierno, especialmente como secular y unido. El documento pone de relieve el carácter contradictorio e inestable de cuatro binarios: texto legislativo frente a las prácticas sociales; legislación frente a estado; mayoría versus minoría; y conocible frente a incognoscible. Al igual que con las proposiciones contradictorias acerca de la homosexualidad, la contradicción no socava el discurso, sino que lo hace más fuerte y más ágil.

  17. Characterization of the liquid Li-solid Mo (1 1 0) interface from classical molecular dynamics for plasma-facing applications

    Science.gov (United States)

    Vella, Joseph R.; Chen, Mohan; Fürstenberg, Sven; Stillinger, Frank H.; Carter, Emily A.; Debenedetti, Pablo G.; Panagiotopoulos, Athanassios Z.

    2017-11-01

    An understanding of the wetting properties and a characterization of the interface between liquid lithium (Li) and solid molybdenum (Mo) are relevant to assessing the efficacy of Li as a plasma-facing component in fusion reactors. In this work, a new second-nearest neighbor modified embedded-atom method (2NN MEAM) force field is parameterized to describe the interactions between Li and Mo. The new force field reproduces several benchmark properties obtained from first-principles quantum mechanics simulations, including binding curves for Li at three different adsorption sites and the corresponding forces on Li atoms adsorbed on the Mo (1 1 0) surface. This force field is then used to study the wetting of liquid Li on the (1 1 0) surface of Mo and to examine the Li-Mo interface using molecular dynamics simulations. From droplet simulations, we find that liquid Li tends to completely wet the perfect Mo (1 1 0) surface, in contradiction with previous experimental measurements that found non-zero contact angles for liquid Li on a Mo substrate. However, these experiments were not carried out under ultra-high vacuum conditions or with a perfect (1 1 0) Mo surface, suggesting that the presence of impurities, such as oxygen, and surface structure play a crucial role in this wetting process. From thin-film simulations, it is observed that the first layer of Li on the Mo (1 1 0) surface has many solid-like properties such as a low mobility and a larger degree of ordering when compared to layers further away from the surface, even at temperatures well above the bulk melting temperature of Li. These findings are consistent with temperature-programmed desorption experiments.

  18. Evaluation of helium cooling for fusion divertors

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1993-09-01

    The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m 2 at an average heat flux of 2 MW/m 2 . The divertors have a requirement of both minimum temperature (100 degrees C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m 2 . This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m 2 . The pumping power required was less than 1% of the power removed. These results verified the design prediction

  19. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state-of-the-art of ventil......This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state......-of-the-art of ventilative cooling potentials and limitations, its consideration in current energy performance regulations, available building components and control strategies and analysis methods and tools. In addition, the report provides twenty six examples of operational buildings using ventilative cooling ranging from...

  20. The efficacy of autologous platelet rich plasma combined with ablative carbon dioxide fractional resurfacing for acne scars: a simultaneous split-face trial.

    Science.gov (United States)

    Lee, Jin Woong; Kim, Beom Joon; Kim, Myeung Nam; Mun, Seog Kyun

    2011-07-01

    Ablative carbon dioxide (CO(2) ) fractional resurfacing is a promising therapeutic intervention for the treatment of acne scars, although this technique is associated with prolonged surgical site erythema and edema, which may affect the daily lives of patients. Autologous platelet-rich plasma (PRP) is known to enhance wound healing and has applications in many areas of medicine. To evaluate the synergistic effects of autologous PRP with CO(2) fractional resurfacing for acne scars. A split-face trial was conducted in 14 Korean participants with acne scars. All participants received one session of ablative CO(2) fractional resurfacing. Immediately after resurfacing, facial halves were randomly assigned to receive treatment with autologous PRP injections on one side (experimental side) and normal saline injections on the other side (control side). The participants were monitored for degree of recovery and resurfacing-associated adverse events, including prolonged erythema, edema, and other effects on days 0, 2, 4, 6, 8, 15, and 30. The intensity of erythema was objectively measured using a chromometer at the same time intervals. After one additional treatment session using the same protocol, two independent dermatologists evaluated clinical improvement using a quartile grading scale. All participants completed the study. Erythema on the experimental side improved faster than on the control side and was significantly less at day 4 (p=.01). This difference was confirmed using a chromometer (p=.049). Total duration of erythema was an average of 10.4±2.7 days on the control side and 8.6±2.0 days on the experimental side (p=.047). Edema also improved faster on the experimental side than on the control side. The total duration of edema was an average of 7.1±1.5 days on the control side and 6.1±1.1 days on the experimental side (p=.04). Participants were also assessed for duration of post-treatment crusting, with a mean of 6.8±1.0 days on the control side and 5.9±1

  1. Efficacy of autologous platelet-rich plasma combined with fractional ablative carbon dioxide resurfacing laser in treatment of facial atrophic acne scars: A split-face randomized clinical trial

    Directory of Open Access Journals (Sweden)

    Gita Faghihi

    2016-01-01

    Full Text Available Background: Autologous platelet-rich plasma has recently attracted significant attention throughout the medical field for its wound-healing ability. Aims: This study was conducted to investigate the potential of platelet-rich plasma combined with fractional laser therapy in the treatment of acne scarring. Methods: Sixteen patients (12 women and 4 men who underwent split-face therapy were analyzed in this study. They received ablative fractional carbon dioxide laser combined with intradermal platelet-rich plasma treatment on one half of their face and ablative fractional carbon dioxide laser with intradermal normal saline on the other half. The injections were administered immediately after laser therapy. The treatment sessions were repeated after an interval of one month. The clinical response was assessed based on patient satisfaction and the objective evaluation of serial photographs by two blinded dermatologists at baseline, 1 month after the first treatment session and 4 months after the second. The adverse effects including erythema and edema were scored by participants on days 0, 2, 4, 6, 8, 15 and 30 after each session. Results: Overall clinical improvement of acne scars was higher on the platelet-rich plasma-fractional carbon dioxide laser treated side but the difference was not statistically significant either 1 month after the first treatment session (P = 0.15 or 4 months after the second (P = 0.23. In addition, adverse effects (erythema and edema on the platelet-rich plasma-fractional carbon dioxide laser-treated side were more severe and of longer duration. Limitations: Small sample size, absence of all skin phototypes within the study group and lack of objective methods for the evaluation of response to treatment and adverse effects were the limitations. Conclusion: This study demonstrated that adding platelet-rich plasma to fractional carbon dioxide laser treatment did not produce any statistically significant synergistic effects

  2. Efficacy of autologous platelet-rich plasma combined with fractional ablative carbon dioxide resurfacing laser in treatment of facial atrophic acne scars: A split-face randomized clinical trial.

    Science.gov (United States)

    Faghihi, Gita; Keyvan, Shima; Asilian, Ali; Nouraei, Saeid; Behfar, Shadi; Nilforoushzadeh, Mohamad Ali

    2016-01-01

    Autologous platelet-rich plasma has recently attracted significant attention throughout the medical field for its wound-healing ability. This study was conducted to investigate the potential of platelet-rich plasma combined with fractional laser therapy in the treatment of acne scarring. Sixteen patients (12 women and 4 men) who underwent split-face therapy were analyzed in this study. They received ablative fractional carbon dioxide laser combined with intradermal platelet-rich plasma treatment on one half of their face and ablative fractional carbon dioxide laser with intradermal normal saline on the other half. The injections were administered immediately after laser therapy. The treatment sessions were repeated after an interval of one month. The clinical response was assessed based on patient satisfaction and the objective evaluation of serial photographs by two blinded dermatologists at baseline, 1 month after the first treatment session and 4 months after the second. The adverse effects including erythema and edema were scored by participants on days 0, 2, 4, 6, 8, 15 and 30 after each session. Overall clinical improvement of acne scars was higher on the platelet-rich plasma-fractional carbon dioxide laser treated side but the difference was not statistically significant either 1 month after the first treatment session (P = 0.15) or 4 months after the second (P = 0.23). In addition, adverse effects (erythema and edema) on the platelet-rich plasma-fractional carbon dioxide laser-treated side were more severe and of longer duration. Small sample size, absence of all skin phototypes within the study group and lack of objective methods for the evaluation of response to treatment and adverse effects were the limitations. This study demonstrated that adding platelet-rich plasma to fractional carbon dioxide laser treatment did not produce any statistically significant synergistic effects and also resulted in more severe side effects and longer downtime.

  3. Cooling towers

    International Nuclear Information System (INIS)

    Korik, L.; Burger, R.

    1992-01-01

    What is the effect of 0.6C (1F) temperature rise across turbines, compressors, or evaporators? Enthalpy charts indicate for every 0.6C (1F) hotter water off the cooling tower will require an additional 2 1/2% more energy cost. Therefore, running 2.2C (4F) warmer due to substandard cooling towers could result in a 10% penalty for overcoming high heads and temperatures. If it costs $1,250,000.00 a year to operate the system, $125,000.00 is the energy penalty for hotter water. This paper investigates extra fuel costs involved in maintaining design electric production with cooling water 0.6C (1F) to 3C (5.5F) hotter than design. If design KWH cannot be maintained, paper will calculate dollar loss of saleable electricity. The presentation will conclude with examining the main causes of deficient cold water production. State-of-the-art upgrading and methodology available to retrofit existing cooling towers to optimize lower cooling water temperatures will be discussed

  4. Passive containment cooling water distribution device

    Science.gov (United States)

    Conway, Lawrence E.; Fanto, Susan V.

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

  5. Pre-conceptual design activities for the materials plasma exposure experiment

    International Nuclear Information System (INIS)

    Lumsdaine, Arnold; Rapp, Juergen; Varma, Venugopal; Bjorholm, Thomas; Bradley, Craig; Caughman, John; Duckworth, Robert; Goulding, Richard; Graves, Van; Giuliano, Dominic; Lessard, Timothy; McGinnis, Dean; Meitner, Steven

    2016-01-01

    Highlights: • The development of long-pulse nuclear fusion devices requires testing plasma facing components at reactor relevant conditions. • The pre-conceptual design of a proposed linear plasma facility is presented. • Engineering considerations for multiple systems—plasma source and heating, magnet, vacuum, water cooling, and target, are presented. - Abstract: The development of next step fusion facilities such as DEMO or a Fusion Nuclear Science Facility (FNSF) requires first closing technology gaps in some critical areas. Understanding the material-plasma interface is necessary to enable the development of divertors for long-pulse plasma facilities. A pre-conceptual design for a proposed steady-state linear plasma device, the Materials Plasma Exposure Experiment (MPEX), is underway. A helicon plasma source along with ion cyclotron and electron Bernstein wave heating systems will produce ITER divertor relevant plasma conditions with steady-state parallel heat fluxes of up to 40 MW/m"2 with ion fluxes up to 10"2"4/m"2 s on target. Current plans are for the device to use superconducting magnets to produce 1–2 T fields. As a steady-state device, active cooling will be required for components that interact with the plasma (targets, limiters, etc.), as well as for other plasma facing components (transport regions, vacuum tanks, diagnostic ports). Design concepts for the vacuum system, the cooling system, and the plasma heating systems have been completed. The device will include the capability for handling samples that have been neutron irradiated in order to consider the multivariate effects of neutrons, plasma, and high heat-flux on the microstructure of divertor candidate materials. A vacuum cask, which can be disconnected from the high field environment in order to perform in-vacuo diagnosis of the surface evolution is also planned for the facility.

  6. Pre-conceptual design activities for the materials plasma exposure experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lumsdaine, Arnold, E-mail: lumsdainea@ornl.gov; Rapp, Juergen; Varma, Venugopal; Bjorholm, Thomas; Bradley, Craig; Caughman, John; Duckworth, Robert; Goulding, Richard; Graves, Van; Giuliano, Dominic; Lessard, Timothy; McGinnis, Dean; Meitner, Steven

    2016-11-01

    Highlights: • The development of long-pulse nuclear fusion devices requires testing plasma facing components at reactor relevant conditions. • The pre-conceptual design of a proposed linear plasma facility is presented. • Engineering considerations for multiple systems—plasma source and heating, magnet, vacuum, water cooling, and target, are presented. - Abstract: The development of next step fusion facilities such as DEMO or a Fusion Nuclear Science Facility (FNSF) requires first closing technology gaps in some critical areas. Understanding the material-plasma interface is necessary to enable the development of divertors for long-pulse plasma facilities. A pre-conceptual design for a proposed steady-state linear plasma device, the Materials Plasma Exposure Experiment (MPEX), is underway. A helicon plasma source along with ion cyclotron and electron Bernstein wave heating systems will produce ITER divertor relevant plasma conditions with steady-state parallel heat fluxes of up to 40 MW/m{sup 2} with ion fluxes up to 10{sup 24}/m{sup 2} s on target. Current plans are for the device to use superconducting magnets to produce 1–2 T fields. As a steady-state device, active cooling will be required for components that interact with the plasma (targets, limiters, etc.), as well as for other plasma facing components (transport regions, vacuum tanks, diagnostic ports). Design concepts for the vacuum system, the cooling system, and the plasma heating systems have been completed. The device will include the capability for handling samples that have been neutron irradiated in order to consider the multivariate effects of neutrons, plasma, and high heat-flux on the microstructure of divertor candidate materials. A vacuum cask, which can be disconnected from the high field environment in order to perform in-vacuo diagnosis of the surface evolution is also planned for the facility.

  7. Cooling tower

    Energy Technology Data Exchange (ETDEWEB)

    Norbaeck, P; Heneby, H

    1976-01-22

    Cooling towers to be transported on road vehicles as a unit are not allowed to exceed certain dimensions. In order to improve the efficiency of such a cooling tower (of cross-flow design and box-type body) with given dimensions, it is proposed to arrange at least one of the scrubbing bodies displaceable within a module or box. Then it can be moved out of the casing into working position, thereby increasing the front surface available for the inlet of air (and with it the efficiency) by nearly a factor of two.

  8. About Face

    Medline Plus

    Full Text Available ... Basics PTSD Treatment What is AboutFace? Resources for Professionals Get Help Home Watch Videos by Topic Videos ... Basics PTSD Treatment What is AboutFace? Resources for Professionals Get Help PTSD We've been there. After ...

  9. About Face

    Medline Plus

    Full Text Available ... Treatment What is AboutFace? Resources for Professionals Get Help Home Watch Videos by Topic Videos by Type ... Treatment What is AboutFace? Resources for Professionals Get Help PTSD We've been there. After a traumatic ...

  10. Development of conductively cooled first wall armor and actively cooled divertor structure for ITER/FER

    International Nuclear Information System (INIS)

    Ioki, K.; Yamada, M.; Sakata, S.; Okada, K.; Toyoda, M.; Shimizu, K.; Tsujimura, S.; Iimura, M.; Akiba, M.; Araki, M.; Seki, M.

    1991-01-01

    Based on the design requirements for the plasma facing components in ITER/FER, we have performed design studies on the conductively cooled first wall armor and the divertor plate with sliding supports. The full-scale armor tiles were fabricated for heat load tests, and good thermal performances were obtained in heat load tests of 0.2-0.4 MW/m 2 . It is shown by the thermomechanical analysis on the divertor plate that thermal stresses and bending deformation are reduced significantly by using the sliding supports. The divertor test module with the sliding supports has been fabricated to investigate its fabricability and to verify the functions of the sliding supports during a high heat load of about 10 MW/m 2 . (orig.)

  11. Theory of semiconductor laser cooling

    Science.gov (United States)

    Rupper, Greg

    Recently laser cooling of semiconductors has received renewed attention, with the hope that a semiconductor cooler might be able to achieve cryogenic temperatures. In order to study semiconductor laser cooling at cryogenic temperatures, it is crucial that the theory include both the effects of excitons and the electron-hole plasma. In this dissertation, I present a theoretical analysis of laser cooling of bulk GaAs based on a microscopic many-particle theory of absorption and luminescence of a partially ionized electron-hole plasma. This theory has been analyzed from a temperature 10K to 500K. It is shown that at high temperatures (above 300K), cooling can be modeled using older models with a few parameter changes. Below 200K, band filling effects dominate over Auger recombination. Below 30K excitonic effects are essential for laser cooling. In all cases, excitonic effects make cooling easier then predicted by a free carrier model. The initial cooling model is based on the assumption of a homogeneous undoped semiconductor. This model has been systematically modified to include effects that are present in real laser cooling experiments. The following modifications have been performed. (1) Propagation and polariton effects have been included. (2) The effect of p-doping has been included. (n-doping can be modeled in a similar fashion.) (3) In experiments, a passivation layer is required to minimize non-radiative recombination. The passivation results in a npn heterostructure. The effect of the npn heterostructure on cooling has been analyzed. (4) The effect of a Gaussian pump beam was analyzed and (5) Some of the parameters in the cooling model have a large uncertainty. The effect of modifying these parameters has been analyzed. Most of the extensions to the original theory have only had a modest effect on the overall results. However we find that the current passivation technique may not be sufficient to allow cooling. The passivation technique currently used appears

  12. The plasma-wall interaction region: a key low temperature plasma for controlled fusion

    International Nuclear Information System (INIS)

    Counsell, G F

    2002-01-01

    The plasma-wall interaction region of a fusion device provides the interface between the hot core plasma and the material surfaces. To obtain acceptably low levels of erosion from these surfaces requires most of the power leaving the core to be radiated. This is accomplished in existing devices by encouraging plasma detachment, in which the hot plasma arriving in the region is cooled by volume recombination and ion-neutral momentum transfer with a dense population of neutrals recycled from the surface. The result is a low temperature (1 eV e e >10 19 m -3 ) but weakly ionized (n 0 >10 20 m -3 , n e /n 0 <0.1) plasma found nowhere else in the fusion environment. This plasma provides many of the conditions found in industrial plasmas exploiting plasma chemistry and the presence of carbon in the region (in the form of carbon-fibre composite used in the plasma facing materials) can result in the formation of deposited hydrocarbon films. The plasma-wall interaction region is therefore among the most difficult in fusion to model, requiring an understanding of atomic, molecular and surface physics issues

  13. Plasma generator

    International Nuclear Information System (INIS)

    Omichi, Takeo; Yamanaka, Toshiyuki.

    1976-01-01

    Object: To recycle a coolant in a sealed hollow portion formed interiorly of a plasma limiter itself to thereby to cause direct contact between the coolant and the plasma limiter and increase of contact area therebetween to cool the plasma limiter. Structure: The heat resulting from plasma generated during operation and applied to the body of the plasma limiter is transmitted to the coolant, which recycles through an inlet and outlet pipe, an inlet and outlet nozzle and a hollow portion to hold the plasma limiter at a level less than a predetermined temperature. On the other hand, the heater wire is, at the time of emergency operation, energized to heat the plasma limiter, but this heat is transmitted to the limiter body to increase the temperature thereof. However, the coolant recycling the hollow portion comes into direct contact with the limiter body, and since the plasma limiter surround the hollow portion, the heat amount transmitted from the limiter body to the coolant increases to sufficiently cool the plasma limiter. (Yoshihara, H.)

  14. Test results from a helium gas-cooled porous metal heat exchanger

    International Nuclear Information System (INIS)

    North, M.T.; Rosenfeld, J.H.; Youchison, D.L.

    1996-01-01

    A helium-cooled porous metal heat exchanger was built and tested, which successfully absorbed heat fluxes exceeding all previously tested gas-cooled designs. Helium-cooled plasma-facing components are being evaluated for fusion applications. Helium is a favorable coolant for fusion devices because it is not a plasma contaminant, it is not easily activated, and it is easily removed from the device in the event of a leak. The main drawback of gas coolants is their relatively poor thermal transport properties. This limitation can be removed through use of a highly efficient heat exchanger design. A low flow resistance porous metal heat exchanger design was developed, based on the requirements for the Faraday shield for the International Thermonuclear Experimental Reactor (ITER) device. High heat flux tests were conducted on two representative test articles at the Plasma Materials Test Facility (PMTF) at Sandia National Laboratories. Absorbed heat fluxes as high as 40 MW/m 2 were successfully removed during these tests without failure of the devices. Commercial applications for electronics cooling and other high heat flux applications are being identified

  15. Stochastic cooling

    International Nuclear Information System (INIS)

    Bisognano, J.; Leemann, C.

    1982-03-01

    Stochastic cooling is the damping of betatron oscillations and momentum spread of a particle beam by a feedback system. In its simplest form, a pickup electrode detects the transverse positions or momenta of particles in a storage ring, and the signal produced is amplified and applied downstream to a kicker. The time delay of the cable and electronics is designed to match the transit time of particles along the arc of the storage ring between the pickup and kicker so that an individual particle receives the amplified version of the signal it produced at the pick-up. If there were only a single particle in the ring, it is obvious that betatron oscillations and momentum offset could be damped. However, in addition to its own signal, a particle receives signals from other beam particles. In the limit of an infinite number of particles, no damping could be achieved; we have Liouville's theorem with constant density of the phase space fluid. For a finite, albeit large number of particles, there remains a residue of the single particle damping which is of practical use in accumulating low phase space density beams of particles such as antiprotons. It was the realization of this fact that led to the invention of stochastic cooling by S. van der Meer in 1968. Since its conception, stochastic cooling has been the subject of much theoretical and experimental work. The earliest experiments were performed at the ISR in 1974, with the subsequent ICE studies firmly establishing the stochastic cooling technique. This work directly led to the design and construction of the Antiproton Accumulator at CERN and the beginnings of p anti p colliding beam physics at the SPS. Experiments in stochastic cooling have been performed at Fermilab in collaboration with LBL, and a design is currently under development for a anti p accumulator for the Tevatron

  16. Experience gained from high heat flux actively cooled PFCs in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.; Bayetti, P.; Brosset, C.; Bucalossi, J.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Ghendrih, Ph.; Guilhem, D.; Gunn, J.; Loarer, T.; Lipa, M.; Mitteau, R.; Pegourie, B.; Reichle, R.; Schlosser, J.; Tsitrone, E.; Vallet, J.C.

    2005-01-01

    The implementation of actively cooled high heat flux plasma facing components (PFCs) is one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW/m 2 of nominal convected heat flux. Technical information is drawn from the whole development up to the industrialisation and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non-destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final 'burial' of about half of the injected gas during long discharges

  17. Management of water leaks on Tore Supra actively cooled fusion device

    International Nuclear Information System (INIS)

    Hatchressian, J.C.; Gargiulo, L.; Samaille, F.; Soler, B.

    2005-01-01

    Up to now, Tore Supra is the only fusion device fully equipped with actively cooled Plasma Facing Components (PFCs). In case of abnormal events during a plasma discharge, the PFCs could be submitted to a transient high power density (run away electrons) or to a continuous phenomena as local thermal flux induced by trapped suprathermal electrons or ions). It could lead to a degradation of the PFC integrity and in the worst case to a water leak occurrence. Such water leak has important consequence on the tokamak operation that concerns PFCs themselves, monitoring equipment located in the vacuum vessel or connected to the ports as RF antennas, diagnostics or pumping systems. Following successive water leak events (the most important water leak, that occurred in September 2002, is described in the paper), a large feedback experience has been gained on Tore supra since more than 15 years that could be useful to actively cooled next devices as W7X and ITER. (authors)

  18. Quantified Faces

    DEFF Research Database (Denmark)

    Sørensen, Mette-Marie Zacher

    2016-01-01

    artist Marnix de Nijs' Physiognomic Scrutinizer is an interactive installation whereby the viewer's face is scanned and identified with historical figures. The American artist Zach Blas' project Fag Face Mask consists of three-dimensional portraits that blend biometric facial data from 30 gay men's faces...... and critically examine bias in surveillance technologies, as well as scientific investigations, regarding the stereotyping mode of the human gaze. The American artist Heather Dewey-Hagborg creates three-dimensional portraits of persons she has “identified” from their garbage. Her project from 2013 entitled...

  19. Reading faces and Facing words

    DEFF Research Database (Denmark)

    Robotham, Julia Emma; Lindegaard, Martin Weis; Delfi, Tzvetelina Shentova

    unilateral lesions, we found no patient with a selective deficit in either reading or face processing. Rather, the patients showing a deficit in processing either words or faces were also impaired with the other category. One patient performed within the normal range on all tasks. In addition, all patients......It has long been argued that perceptual processing of faces and words is largely independent, highly specialised and strongly lateralised. Studies of patients with either pure alexia or prosopagnosia have strongly contributed to this view. The aim of our study was to investigate how visual...... perception of faces and words is affected by unilateral posterior stroke. Two patients with lesions in their dominant hemisphere and two with lesions in their non-dominant hemisphere were tested on sensitive tests of face and word perception during the stable phase of recovery. Despite all patients having...

  20. About Face

    Medline Plus

    Full Text Available Skip to Content Menu Closed (Tap to Open) Home Interviews Our Stories Search All Videos PTSD Basics PTSD Treatment What is AboutFace? Resources for Professionals Get Help Home Watch Interviews Our ...

  1. About Face

    Medline Plus

    Full Text Available ... not feeling better, you may have PTSD (posttraumatic stress disorder). Watch the intro This is AboutFace In these videos, Veterans, family members, and clinicians share their experiences with PTSD ...

  2. About Face

    Medline Plus

    Full Text Available Skip to Content Menu Closed (Tap to Open) Home Videos by Topic Videos by Type Search All ... What is AboutFace? Resources for Professionals Get Help Home Watch Videos by Topic Videos by Type Search ...

  3. About Face

    Medline Plus

    Full Text Available Skip to Content Menu Closed (Tap to Open) Home Interviews Our Stories Search All Videos PTSD Basics ... What is AboutFace? Resources for Professionals Get Help Home Watch Interviews Our Stories Search All Videos Learn ...

  4. A very cool cooling system

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The NA62 Gigatracker is a jewel of technology: its sensor, which delivers the time of the crossing particles with a precision of less than 200 picoseconds (better than similar LHC detectors), has a cooling system that might become the precursor to a completely new detector technique.   The 115 metre long vacuum tank of the NA62 experiment. The NA62 Gigatracker (GTK) is composed of a set of three innovative silicon pixel detectors, whose job is to measure the arrival time and the position of the incoming beam particles. Installed in the heart of the NA62 detector, the silicon sensors are cooled down (to about -20 degrees Celsius) by a microfluidic silicon device. “The cooling system is needed to remove the heat produced by the readout chips the silicon sensor is bonded to,” explains Alessandro Mapelli, microsystems engineer working in the Physics department. “For the NA62 Gigatracker we have designed a cooling plate on top of which both the silicon sensor and the...

  5. Cooling pancakes

    International Nuclear Information System (INIS)

    Bond, J.R.; Wilson, J.R.

    1984-01-01

    In theories of galaxy formation with a damping cut-off in the density fluctuation spectrum, the first non-linear structures to form are Zeldovich pancakes in which dissipation separates gas from any collisionless dark matter then present. One-dimensional numerical simulations of the collapse, shock heating, and subsequent thermal evolution of pancakes are described. Neutrinos (or any other cool collisionless particles) are followed by direct N-body methods and the gas by Eulerian hydrodynamics with conduction as well as cooling included. It is found that the pressure is relatively uniform within the shocked region and approximately equals the instantaneous ram pressure acting at the shock front. An analytic theory based upon this result accurately describes the numerical calculations. (author)

  6. Cool Sportswear

    Science.gov (United States)

    1982-01-01

    New athletic wear design based on the circulating liquid cooling system used in the astronaut's space suits, allows athletes to perform more strenuous activity without becoming overheated. Techni-Clothes gear incorporates packets containing a heat-absorbing gel that slips into an insulated pocket of the athletic garment and is positioned near parts of the body where heat transfer is most efficient. A gel packet is good for about one hour. Easily replaced from a supply of spares in an insulated container worn on the belt. The products, targeted primarily for runners and joggers and any other athlete whose performance may be affected by hot weather, include cooling headbands, wrist bands and running shorts with gel-pack pockets.

  7. Cooling systems

    International Nuclear Information System (INIS)

    Coutant, C.C.

    1978-01-01

    Progress on the thermal effects project is reported with regard to physiology and distribution of Corbicula; power plant effects studies on burrowing mayfly populations; comparative thermal responses of largemouth bass from northern and southern populations; temperature selection by striped bass in Cherokee Reservoir; fish population studies; and predictive thermoregulation by fishes. Progress is also reported on the following; cause and ecological ramifications of threadfin shad impingement; entrainment project; aquaculture project; pathogenic amoeba project; and cooling tower drift project

  8. Hydrogen retention in lithium on metallic walls from “in vacuo” analysis in LTX and implications for high-Z plasma-facing components in NSTX-U

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Lucia, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Allain, J.P.; Bedoya, F. [Department of Nuclear, Plasma, & Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL (United States); Bell, R.; Boyle, D. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Capece, A. [Department of Physics, The College of New Jersey, Ewing, NJ (United States); Jaworski, M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Koel, B.E. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Majeski, R. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Roszell, J. [Department of Chemical & Biological Engineering, Princeton University, Princeton, NJ (United States); Schmitt, J. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ (United States); Soukhanovskii, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States)

    2017-04-15

    The application of lithium to plasma-facing components (PFCs) has long been used as a technique for wall conditioning in magnetic confinement devices to improve plasma performance. Determining the characteristics of PFCs at the time of exposure to the plasma, however, is difficult because they can only be analyzed after venting the vacuum vessel and removing them at the end of an operational period. The Materials Analysis and Particle Probe (MAPP) addresses this problem by enabling PFC samples to be exposed to plasmas, and then withdrawn into an analysis chamber without breaking vacuum. The MAPP system was used to introduce samples that matched the metallic PFCs of the Lithium Tokamak Experiment (LTX). Lithium that was subsequently evaporated onto the walls also covered the MAPP samples, which were then subject to LTX discharges. In vacuo extraction and analysis of the samples indicated that lithium oxide formed on the PFCs, but improved plasma performance persisted in LTX. The reduced recycling this suggests is consistent with separate surface science experiments that demonstrated deuterium retention in the presence of lithium oxide films. Since oxygen decreases the thermal stability of the deuterium in the film, the release of deuterium was observed below the lithium deuteride dissociation temperature. This may explain what occurred when lithium was applied to the surface of the NSTX Liquid Lithium Divertor (LLD). The LLD had segments with individual heaters, and the deuterium-alpha emission was clearly lower in the cooler regions. The plan for NSTX-U is to replace the graphite tiles with high-Z PFCs, and apply lithium to their surfaces with lithium evaporation. Experiments with lithium coatings on such PFCs suggest that deuterium could still be retained if lithium compounds form, but limiting their surface temperatures may be necessary.

  9. About Face

    Medline Plus

    Full Text Available ... at first. But if it's been months or years since the trauma and you're not feeling better, you may have PTSD (posttraumatic stress disorder). Watch the intro This is AboutFace In these videos, Veterans, family members, ...

  10. About Face

    Medline Plus

    Full Text Available ... What is AboutFace? Resources for Professionals Get Help PTSD We've been there. After a traumatic event — ... you're not feeling better, you may have PTSD (posttraumatic stress disorder). Watch the intro This is ...

  11. Ultrafine tungsten as a plasma-facing component in fusion devices: effect of high flux, high fluence low energy helium irradiation

    International Nuclear Information System (INIS)

    El-Atwani, O.; Gonderman, Sean; Allain, J.P.; Efe, Mert; Klenosky, Daniel; Qiu, Tian; De Temmerman, Gregory; Morgan, Thomas; Bystrov, Kirill

    2014-01-01

    This work discusses the response of ultrafine-grained tungsten materials to high-flux, high-fluence, low energy pure He irradiation. Ultrafine-grained tungsten samples were exposed in the Pilot-PSI (Westerhout et al 2007 Phys. Scr. T128 18) linear plasma device at the Dutch Institute for Fundamental Energy Research (DIFFER) in Nieuwegein, the Netherlands. The He flux on the tungsten samples ranged from 1.0 × 10 23 –2.0 × 10 24  ions m −2  s −1 , the sample bias ranged from a negative (20–65) V, and the sample temperatures ranged from 600–1500 °C. SEM analysis of the exposed samples clearly shows that ultrafine-grained tungsten materials have a greater fluence threshold to the formation of fuzz by an order or magnitude or more, supporting the conjecture that grain boundaries play a major role in the mechanisms of radiation damage. Pre-fuzz damage analysis is addressed, as in the role of grain orientation on structure formation. Grains of (1 1 0) and (1 1 1) orientation showed only pore formation, while (0 0 1) oriented grains showed ripples (higher structures) decorated with pores. Blistering at the grain boundaries is also observed in this case. In situ TEM analysis during irradiation revealed facetted bubble formation at the grain boundaries likely responsible for blistering at this location. The results could have significant implications for future plasma-burning fusion devices given the He-induced damage could lead to macroscopic dust emission into the fusion plasma. (paper)

  12. Cool snacks

    DEFF Research Database (Denmark)

    Grunert, Klaus G; Brock, Steen; Brunsø, Karen

    2016-01-01

    Young people snack and their snacking habits are not always healthy. We address the questions whether it is possible to develop a new snack product that adolescents will find attractive, even though it is based on ingredients as healthy as fruits and vegetables, and we argue that developing...... such a product requires an interdisciplinary effort where researchers with backgrounds in psychology, anthropology, media science, philosophy, sensory science and food science join forces. We present the COOL SNACKS project, where such a blend of competences was used first to obtain thorough insight into young...... people's snacking behaviour and then to develop and test new, healthier snacking solutions. These new snacking solutions were tested and found to be favourably accepted by young people. The paper therefore provides a proof of principle that the development of snacks that are both healthy and attractive...

  13. Cool visitors

    CERN Multimedia

    2006-01-01

    Pictured, from left to right: Tim Izo (saxophone, flute, guitar), Bobby Grant (tour manager), George Pajon (guitar). What do the LHC and a world-famous hip-hop group have in common? They are cool! On Saturday, 1st July, before their appearance at the Montreux Jazz Festival, three members of the 'Black Eyed Peas' came on a surprise visit to CERN, inspired by Dan Brown's Angels and Demons. At short notice, Connie Potter (Head of the ATLAS secretariat) organized a guided tour of ATLAS and the AD 'antimatter factory'. Still curious, lead vocalist Will.I.Am met CERN physicist Rolf Landua after the concert to ask many more questions on particles, CERN, and the origin of the Universe.

  14. Transpiration cooling assisted ablative thermal protection of aerospace substructures

    International Nuclear Information System (INIS)

    Khan, M.B.; Iqbal, N.; Haider, Z.

    2009-01-01

    Ablatives are heat-shielding materials used to protect aerospace substructures. These materials are sacrificial in nature and provide protection primarily through the large endothermic transformation during exposure to hyper thermal environment such as encountered in re-entry modules. The performance of certain ablatives was reported in terms of their TGA/DTA in Advanced Materials-97 (pp 57-65). The focus of this earlier research resided in the consolidation of interface between the refractory inclusion and the host polymeric matrix to improve thermal resistance. In the present work we explore the scope of transpiration cooling in ablative performance through flash evaporation of liquid incorporated in the host EPDM (Ethylene Propylene Diene Monomer) matrix. The compression-molded specimens were exposed separately to plasma flame (15000 C) and oxyacetylene torch (3000 C) and the back face transient temperature is recorded in situ employing a thermocouple/data logger system. Both head on impingement (HOI) and parallel flow (PF) through a central cavity in the ablator were used. It is observed that transpiration cooling is effective and yields (a) rapid thermal equilibrium in the specimen, (b) lower back face temperature and (c) lower ablation rate, compared to conventional ablatives. SEM/EDS analysis is presented to amplify the point. (author)

  15. Numerical simulation of heat transfer and fluid flow of an impinging round jet of plasma into confined walls

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Shimizu, Akihiko; Kunugi, Tomoaki.

    1995-01-01

    Numerical simulations are presented on the flow and heat transfer characteristics of an impinging round jet of argon plasma with atmospheric pressure. The target slab with finite thickness upon which plasma jet impinges is assumed to be as SiC which is a candidate material for plasma facing material of fusion reactor. The plasma jet is treated by use of a magnetohydrodynamics model that takes its two-temperature non-equilibrium state into account. The rear side of the target slab is assumed to be cooled by a gas-solid suspension impinging round jet. The result shows that the plasma is in non-equilibrium state in which the electron temperature is higher than the heavy particle in the outer region of plasma jet core and that the heat flux to the target slab is over 8 MW/m 2 in the region of the plasma jet core contacts. (author)

  16. Scientific feedback from high heat flux actively cooled PFCs development, realization and first results in Tore Supra

    International Nuclear Information System (INIS)

    Grosman, A.; Bayetti, P.; Brosset, C.; Bucalossi, J.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Ghendrih, Ph.; Guilhem, D.; Gunn, J.; Loarer, T.; Lipa, M.; Mitteau, R.; Pegourie, B.; Reichle, R.; Schlosser, J.; Tsitrone, E.; Vallet, J.C.

    2004-01-01

    The implementation of actively cooled high heat flux plasma facing components (PFCs) are one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m -2 of nominal convected heat flux. A technical feedback is given from the whole development up to the industrialization and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final 'burial' of about the injected gas during long discharges. (authors)

  17. Scientific feedback from high heat flux actively cooled PFCs development, realization and first results in Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Grosman, A.; Bayetti, P.; Brosset, C.; Bucalossi, J.; Cordier, J.J.; Durocher, A.; Escourbiac, F.; Ghendrih, Ph.; Guilhem, D.; Gunn, J.; Loarer, T.; Lipa, M.; Mitteau, R.; Pegourie, B.; Reichle, R.; Schlosser, J.; Tsitrone, E.; Vallet, J.C

    2004-07-01

    The implementation of actively cooled high heat flux plasma facing components (PFCs) are one of the major ingredients required for operating the Tore Supra tokamak with very long pulses. A pioneering activity has been developed in this field from the very beginning of the device operation that is today culminating with the routine operation of an actively cooled toroidal pumped limiter (TPL) capable to sustain up to 10 MW.m{sup -2} of nominal convected heat flux. A technical feedback is given from the whole development up to the industrialization and focuses on a number of critical issues, such as bonding technology analysis, manufacture processes, repair processes, destructive and non destructive testing. The actual experience in Tore Supra allows to address the question of D retention on carbon walls. Redeposition on surfaces without plasma flux is suspected to cause the final 'burial' of about the injected gas during long discharges. (authors)

  18. Divertor cooling device

    International Nuclear Information System (INIS)

    Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

    1993-01-01

    Cooling water for a divertor cooling system cools the divertor, thereafter, passes through pipelines connecting the exit pipelines of the divertor cooling system and the inlet pipelines of a blanket cooling system and is introduced to the blanket cooling system in a vacuum vessel. It undergoes emission of neutrons, and cooling water in the divertor cooling system containing a great amount of N-16 which is generated by radioactivation of O-16 is introduced to the blanket cooling system in the vacuum vessel by way of pipelines, and after cooling, passes through exit pipelines of the blanket cooling system and is introduced to the outside of the vacuum vessel. Radiation of N-16 in the cooling water is decayed sufficiently with passage of time during cooling of the blanket, thereby enabling to decrease the amount of shielding materials such as facilities and pipelines, and ensure spaces. (N.H.)

  19. Development of a plasma driven permeation experiment for TPE

    Energy Technology Data Exchange (ETDEWEB)

    Buchenauer, Dean, E-mail: dabuche@sandia.gov [Sandia National Laboratories, Livermore, CA (United States); Kolasinski, Robert [Sandia National Laboratories, Livermore, CA (United States); Shimada, Masa [Idaho National Laboratory, Idaho Falls, ID (United States); Donovan, David [Sandia National Laboratories, Livermore, CA (United States); Youchison, Dennis [Sandia National Laboratories, Albuquerque, NM (United States); Merrill, Brad [Idaho National Laboratory, Idaho Falls, ID (United States)

    2014-10-15

    Highlights: • We have designed and fabricated a novel tritium permeation membrane holder for use in the Tritium Plasma Experiment (TPE). • The membrane temperature is controlled by varying the cooling flow rate and proximity of a spiral cooling channel. • Sealing tests have demonstrated adequate helium leak rates up to temperatures of 1000 °C. • Flow modeling indicates a minimal helium pressure drop across the membrane holder (<700 Pa). • Thermal modeling shows good heat removal and minimal membrane temperature variation (±2%) even up to peak TPE ion fluxes. - Abstract: Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.

  20. Development of a plasma driven permeation experiment for TPE

    International Nuclear Information System (INIS)

    Buchenauer, Dean; Kolasinski, Robert; Shimada, Masa; Donovan, David; Youchison, Dennis; Merrill, Brad

    2014-01-01

    Highlights: • We have designed and fabricated a novel tritium permeation membrane holder for use in the Tritium Plasma Experiment (TPE). • The membrane temperature is controlled by varying the cooling flow rate and proximity of a spiral cooling channel. • Sealing tests have demonstrated adequate helium leak rates up to temperatures of 1000 °C. • Flow modeling indicates a minimal helium pressure drop across the membrane holder (<700 Pa). • Thermal modeling shows good heat removal and minimal membrane temperature variation (±2%) even up to peak TPE ion fluxes. - Abstract: Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes

  1. WORKSHOP: Beam cooling

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Cooling - the control of unruly particles to provide well-behaved beams - has become a major new tool in accelerator physics. The main approaches of electron cooling pioneered by Gersh Budker at Novosibirsk and stochastic cooling by Simon van der Meer at CERN, are now complemented by additional ideas, such as laser cooling of ions and ionization cooling of muons

  2. Innovative technologies for Faraday shield cooling

    International Nuclear Information System (INIS)

    Rosenfeld, J.H.; Lindemuth, J.E.; North, M.T.; Goulding, R.H.

    1995-01-01

    Alternative advanced technologies are being evaluated for use in cooling the Faraday shields used for protection of ion cyclotron range of frequencies (ICR) antennae in Tokamaks. Two approaches currently under evaluation include heat pipe cooling and gas cooling. A Monel/water heat pipe cooled Faraday shield has been successfully demonstrated. Heat pipe cooling offers the advantage of reducing the amount of water discharged into the Tokamak in the event of a tube weld failure. The device was recently tested on an antenna at Oak Ridge National Laboratory. The heat pipe design uses inclined water heat pipes with warm water condensers located outside of the plasma chamber. This approach can passively remove absorbed heat fluxes in excess of 200 W/cm 2 ;. Helium-cooled Faraday shields are also being evaluated. This approach offers the advantage of no liquid discharge into the Tokamak in the event of a tube failure. Innovative internal cooling structures based on porous metal cooling are being used to develop a helium-cooled Faraday shield structure. This approach can dissipate the high heat fluxes typical of Faraday shield applications while minimizing the required helium blower power. Preliminary analysis shows that nominal helium flow and pressure drop can sufficiently cool a Faraday shield in typical applications. Plans are in progress to fabricate and test prototype hardware based on this approach

  3. Spatial structure of radio frequency ring-shaped magnetized discharge sputtering plasma using two facing ZnO/Al2O3 cylindrical targets for Al-doped ZnO thin film preparation

    Directory of Open Access Journals (Sweden)

    Takashi Sumiyama

    2017-05-01

    Full Text Available Spatial structure of high-density radio frequency ring-shaped magnetized discharge plasma sputtering with two facing ZnO/Al2O3 cylindrical targets mounted in ring-shaped hollow cathode has been measured and Al-doped ZnO (AZO thin film is deposited without substrate heating. The plasma density has a peak at ring-shaped hollow trench near the cathode. The radial profile becomes uniform with increasing the distance from the target cathode. A low ion current flowing to the substrate of 0.19 mA/cm2 is attained. Large area AZO films with a resistivity of 4.1 – 6.7×10-4 Ω cm can be prepared at a substrate room temperature. The transmittance is 84.5 % in a visible region. The surface roughnesses of AZO films are 0.86, 0.68, 0.64, 1.7 nm at radial positions of r = 0, 15, 30, 40 mm, respectively, while diffraction peak of AZO films is 34.26°. The grains exhibit a preferential orientation along (002 axis.

  4. Renewable Heating And Cooling

    Science.gov (United States)

    Renewable heating and cooling is a set of alternative resources and technologies that can be used in place of conventional heating and cooling technologies for common applications such as water heating, space heating, space cooling and process heat.

  5. Restaurant food cooling practices.

    Science.gov (United States)

    Brown, Laura Green; Ripley, Danny; Blade, Henry; Reimann, Dave; Everstine, Karen; Nicholas, Dave; Egan, Jessica; Koktavy, Nicole; Quilliam, Daniela N

    2012-12-01

    Improper food cooling practices are a significant cause of foodborne illness, yet little is known about restaurant food cooling practices. This study was conducted to examine food cooling practices in restaurants. Specifically, the study assesses the frequency with which restaurants meet U.S. Food and Drug Administration (FDA) recommendations aimed at reducing pathogen proliferation during food cooling. Members of the Centers for Disease Control and Prevention's Environmental Health Specialists Network collected data on food cooling practices in 420 restaurants. The data collected indicate that many restaurants are not meeting FDA recommendations concerning cooling. Although most restaurant kitchen managers report that they have formal cooling processes (86%) and provide training to food workers on proper cooling (91%), many managers said that they do not have tested and verified cooling processes (39%), do not monitor time or temperature during cooling processes (41%), or do not calibrate thermometers used for monitoring temperatures (15%). Indeed, 86% of managers reported