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Diagnostic integration solutions in the ITER first wall
Fusion Engineering and Design 98–99 (2015) 1548–1551
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Diagnostic integration solutions in the ITER first wall Gonzalo Martínez b,∗ , Alex Martin a , Christopher Watts a , Evgeny Veshchev a , Roger Reichle a , Pavel Shigin a,e , Flavien Sabourin c , Stefan Gicquel a , Raphael Mitteau a , Jorge González d a
ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex, France Technical University of Catalonia (UPC), Barcelona-Tech, Barcelona, Spain ABMI-Groupe, Parc du Relais BatD 201 Route de SEDS, 13127 Vitrolles, France d RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat, Spain e National Research Nuclear University (MEPhI), Kashirskoe shosse, 115409 Moscow, Russian Federation b c
h i g h l i g h t s • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets.
• An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW.
a r t i c l e
i n f o
Article history: Received 24 September 2014 Received in revised form 9 June 2015 Accepted 10 June 2015 Available online 3 July 2015 Keywords: Integration Blanket first wall (FW) FW diagnostics
a b s t r a c t ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems. © 2015 Elsevier B.V. All rights reserved.
1. Introduction Diagnostics play an essential role for the successful operation of the ITER machine. As a key project requirement, ITER shall not operate without a working diagnostic which provides measurements for machine protection, basic control and advanced plasma control [1]. Some diagnostics require a plasma facing element attached to the first wall (FW), commonly known as a FW diagnostic, which contribute to measure a wide range of plasma parameters. Therefore, in order to guarantee a proper integration in the machine their design must comply not only with their functional requirements but also with the design of the blankets.
∗ Corresponding author. Tel.: +34 650842075. E-mail address: [email protected] (G. Martínez). http://dx.doi.org/10.1016/j.fusengdes.2015.06.046 0920-3796/© 2015 Elsevier B.V. All rights reserved.
This is particularly challenging because not all systems have reached the same level of design maturity. While most of the internal components (blankets and divertor) have passed the final design review, the majority of diagnostic systems are at a conceptual design level. 2. Blanket system The blanket system consists of 440 blanket shield modules (BSMs). Each module is composed of a first wall (FW) and a shield block (SB). The plasma-facing FW is composed of beryllium armour components attached to copper substrates mounted on a watercooled stainless steel support structure called central beam. The FW structure is mounted on the SB through a central bolt. The hydraulic connection between the FW and the SB is provided by flexible pipes on the back of the beam. The mechanical, water connection and electrical interfaces between FW and SB are accessible
G. Martínez et al. / Fusion Engineering and Design 98–99 (2015) 1548–1551
Fig. 1. EHF (upper) and NHF (lower) technologies used on the FWs.
through the front face of the FW on a 60 mm width recessed central slot, where FW diagnostics will be mounted through a dedicate interface described in Section 5. There are two versions of the FWs depending on the heat flux level they are exposed to – the enhanced heat flux (EHF) and the normal heat flux (NHF) panels, capable of accommodating 4.7 MW/m2 and 1–2 MW/m2 respectively [2]. Two different manufacturing routes are currently anticipated for NHF and EHF; hot isostatic pressing (HIP) by Europe, beryllium tile brazing by Russia and beryllium tile HIP by China [3] (Fig. 1). 3. FW diagnostics The concept FW diagnostic is introduced to group diagnostics that require a plasma facing component installed in the central slot of a FW panel. These are: The visible infrared (vis/IR) TV system has the primary role of the measuring FW surface temperature, luminance and temperature during edge-localized modes (ELMs) [4]. Its main priorities are machine protection and basic machine control measurements [1]. The vis/IR TV system requires black bodies (BB) for their calibration. A number of BBs are located on the central beam of the dedicated FWs [5]. The primary aim of the edge and core Thomson scattering systems is to measure the electron temperature and density profiles at the periphery and core of the plasma for physics studies and advanced plasma control [1]. Installation of beam dumps for the high-power lasers will require unique cut-outs on the FW central beam and a reduction of the FW central slot insert. To simplify manufacturing an identical beam dump interface will be used for both diagnostics [5]. The poloidal polarimeter (PoPola) is used as the primary diagnostic to measure the current profile. It is used for advanced plasma control physics studies. Major components of the system are located in upper port 10 and equatorial port 10. In addition, 12 corner cube retroreflectors (CCR) are required to be installed at the central slot of the inner wall blankets and two in the divertor for reflection of the probing laser beams launched into the vacuum vessel [5]. The primary role for FW samples is to measure erosion. Supplementary measurements are dust accumulation, deposition effects and tritium retention. FW samples will be used for machine protection basic machine control measurements and there may be a “safety function” in the future. The samples are to be removed by remote handling by the multipurpose deployer during each
1549
shutdown during the hydrogen phase, deuterium phase and D–T phase. The samples distribution in the machine includes areas close to gas injectors and glow discharge cleaning (GDC) anodes. Access to the samples is assumed to be approximately on an annual basis [6]. The primary role of H-alpha spectroscopy (+visible spectroscopy) is to measure line emission of hydrogen isotopes and impurities from ITER FW [7]. This diagnostic provides essential information for plasma control, machine protection and physics study. H-alpha spectroscopy consist of four optical channels located in 3 port plugs: two poloidal wide field of view (FOV) channels in EPP#11 which cover the inner and upper part of the wall, one tangential wide-FOV channel in EPP#12 (covering the outer wall), and one wide-FOV channel in UPP#2 covering divertor and BM #18. The majority of H-alpha beam dumps are allocated in the central slots of blanket modules in front of EPP#11 covering the inner and upper wall [5,8]. The primary roles of the In Vessel Viewing System (IVVS) are in vessel inspection for viewing and metrology purposes and to monitor operational parameters of the main plasma FW for physics studies, advanced plasma control and safety limits (1000 kg dust limit in vessel) [9]. The IVVS will be installed in 6 lower ports (3, 5, 9, 11, 15, 17) and consists out of a mechanical deployer and a scanning head in each position. On the blankets, fiducials (specific reference points) are foreseen to improve the precision of the measurements [6]. 4. Main requirements and constraints There are high levels of potentially damaging fluxes at the FW. Thus, the FW diagnostic designs have to cope with nuclear heating, surface heat and electromagnetic radiation fluxes. Moreover, limited space and a compacted geometry of the blanket, force the FW diagnostics to be passively cooled through the FW beam structure which is actively cooled. Blanket constraints: • FW mounted diagnostics can be implemented only in the central slot by a dedicated cut-out subject to a minimum clearance between existing holes and cooling channels. • Re-arrangement of cooling channels or modification of the water box is only possible if the water flow imbalance between wings is less than 10%. • Hot spots at the beam structured shall not create excessive stresses. • Cut-outs and diagnostic attachments should be compatible with FW manufacturing route. • Under no circumstance can the FW diagnostics jeopardize the Blanket system requirements. FW diagnostics requirements: • FW diagnostics should be visually accessible or properly aligned to the focal points located in ports. • It must be protected from damaging heat and redisposition fluxes. Proper diagnostic recession in the central slot is necessary. Also, it is fundamental to have good passive cooling to ensure survivability and performance. • No remote handling class 1 is foreseen for the majority of FW diagnostics except for FW samples. 5. Design of interfaces Accordingly several unique FW diagnostic attachments which meet the required measurements capabilities and respect the
1550
G. Martínez et al. / Fusion Engineering and Design 98–99 (2015) 1548–1551
Fig. 3. Toroidal cross section of the CSI showing the interface solution for EHF panels.
Fig. 2. Poloidal cross section of the FW beam showing the interface solution for NHF panels.
blanket design constraints have been designed. To facilitate interfaces FW diagnostics have been grouped into two categories: – Standard attachments which will use the standard FW design modifications, already implemented to support such interfaces for Group 1 diagnostics and – Variant attachment which require variants on the FW design for Group 2 diagnostics.
Fig. 4. Detail of the Thomson scattering beam dump interface.
This first group includes the following diagnostics: Vis/IR TV system, FW samples and IVVS fiducial markers. The main constraints for developing the attachment design are geometrical and heat loads. A through hole of 55 mm in the case of NHF and a conical depression on the central slot insert of EHF panels have been provided to accommodate them. Although the requirements for each system may differ, a suitable solution for Group 1 diagnostics has been found. Depending on the FW technology used (NHF/EHF) one of two possible attachments solutions will be used. In the case of NHF a CuCrZr body has been added as a heat sink/adaptor in order to improve the thermal conductivity between the FW sample and the SS beam structure. A pressing fit sleeve will provide enough pressure to be cooled down passively through the beam structure walls. FW samples are required to be remote hand-able (RHC1), thus they will be screwed into position and locked by a specific feature (Fig. 2). In the case of EHF panels, the FW diagnostic is directly mounted on the central slot insert (CSI) which provides protection from heat fluxes and a good thermal conductance through a conical hole and a threaded part (Fig. 3).
need to be placed close to the burning plasma even though they are particularly susceptible to re-deposition and how this impacts the reflectivity, with beryllium as the dominant constituent. Consequently, CCRs will be recessed deeper in the central beam structure and properly aligned with the corresponding line of sight. This adds complexity to the compacted geometry of the blanket which must now cope with bigger cut-outs. Design details of this interface are still to be agreed. H-alpha dumps are mainly a cavity which works as a black body. They are vital for suppressing strong divertor stray light which can be up to few hundreds the signal from the FW. Dumps will enable the stray light to be separated from the true signal. They shall provide at least 10 fold reduction of stray light [7]. The solution proposed to integrate dumps is to use the already existing hole in the central slot such as the ones provided for PoPola’s CCRs and Thomson scattering beam dumps. Applying a surface treatment in the cavities, the roughening required to collect more light and improve dumps performance can be achieved. Edge and core Thomson scattering beam dumps are used for the purpose of safely terminating the laser beam and minimizing the generation of stray light. Their interfaces are placed on the top of the central slot inside the FW central beam. A cut out of 100 mm × 60 mm × 77 mm on the central beam will host the beam dumps and a bolted interface of 4×M5 will fix it in position. The cooling of the BD is passive, based on the contact with the FW (Fig. 4).
5.2. Variant attachment
6. Ongoing analysis
This second group includes: poloidal polarimeter, H-alpha spectroscopy and edge and core Thomson scattering. The main constraints for developing their attachment are not only geometrical and head loads, but also re-deposition and light reflection. Due to the harsh ITER environment, in vessel optics have to address important issues such as survivability and thermal distortion. PoPola systems are no exception. The CCR and mirrors
The feasibility of the designs is in process of being validated through a Thermo-mechanical and hydraulic analysis. The loads applied are based on the Blanket System Load Specification [10] and Nuclear Heating and Radiation Damage in BM04 with Detailed Manifold and IM Key Configuration [11]. Maximal temperature found on samples under photonic and neutronic heat loads for EHF panel is 400 ◦ C. While the hydraulic
5.1. Standard attachment
G. Martínez et al. / Fusion Engineering and Design 98–99 (2015) 1548–1551
1551
Disclaimer The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. Acknowledgements
Fig. 5. (Left) Max. temperature on EHF standard attachment. (Right) Pressure drop map on NHF panel.
analyses shows that the impact of additional diagnostic barrels on the finger flow distribution is minor, the finger flow balance is below 10%. These first results are investigating the feasibility of implementing standard attachments. Nevertheless, only part of the thermal and hydraulic analysis has been done (Fig. 5). A global thermomechanical analysis in order to fully validate the interfaces design solutions remains to be done. 7. Summary An iterative communication with both parties (blanket and diagnostic groups) has resulted in a unique integration design that incorporates FW diagnostics into the machine and achieves the project requirements. An integrated design concept has been developed and provides a design that respects the requirements of each of the systems. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW.
The authors are thankful to all ITER IO members, external contractors and colleagues in the Domestic Agencies for their continuous support and help in the design development of the integration of FW diagnostic. References [1] S. Chiocchio, Project Requirements, ITER Cadarache, May 2010 (Private communication). [2] R. Mitteau, B. Calcagno, R. Chappuis, The design of the ITER first wall panels, Fusion Eng. Des. 88 (October (6–8)) (2013) 568–570. [3] M. Merola, F. Escourbiac, R. Raffray, et al., Overview and Status of ITER Internal Components, ISFNT, 2013. [4] S. Salasca, Development of equatorial visible/infrared wide angle viewing system and radial neutron camera for ITER, Fusion Eng. Des. 84 (June) (2009) 1689–1696. [5] A. Martin, In Vessel Component Variants V1.15, ITER Cadarache, 2014 (Private communication). [6] R. Reichle, Review of the ITER Diagnostics Suite for Erosion, Deposition, Dust and Tritium Measurements, PSI, 2014. [7] A.B. Kukushkin, Divertor Stray Light Analysis in JET-ILW and Implications for the H-alpha Diagnostic in ITER, AIP, 2014. [8] A.J.H. Donné, Diagnostics, Nucl. Fusion 47 (2007) S337–S384 (Chapter 7). [9] C. Neri, ITER in vessel viewing system design and assessment activities, Fusion Eng. Des. 86 (October) (2011) 1954–1957. [10] B. Calcagno, Blanket System Load Specification, ITER Cadarache, 2013 (Private communication). [11] M. Sawan, Nuclear Heating and Radiation Damage in BM04 with Detailed Manifold and IM Key Configuration, ITER Cadarache, 2012 (Private communication).
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Diagnostic integration solutions in the ITER first wall Gonzalo Martínez b,∗ , Alex Martin a , Christopher Watts a , Evgeny Veshchev a , Roger Reichle a , Pavel Shigin a,e , Flavien Sabourin c , Stefan Gicquel a , Raphael Mitteau a , Jorge González d a
ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex, France Technical University of Catalonia (UPC), Barcelona-Tech, Barcelona, Spain ABMI-Groupe, Parc du Relais BatD 201 Route de SEDS, 13127 Vitrolles, France d RÜECKER LYPSA, Carretera del Prat, 65, Cornellá de Llobregat, Spain e National Research Nuclear University (MEPhI), Kashirskoe shosse, 115409 Moscow, Russian Federation b c
h i g h l i g h t s • This paper describes the current status of the integration efforts to implement diagnostics in the ITER first wall (FW). • Some diagnostics require a plasma facing element attached to the FW, commonly known as a FW diagnostic. Their design must comply not only with their functional requirements but also with the design of the blankets.
• An integrated design concept has been developed. It provides a design that respects the requirements of each system. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW.
a r t i c l e
i n f o
Article history: Received 24 September 2014 Received in revised form 9 June 2015 Accepted 10 June 2015 Available online 3 July 2015 Keywords: Integration Blanket first wall (FW) FW diagnostics
a b s t r a c t ITER will have about 50 diagnostic systems for machine protection, plasma control and optimization, and understanding the physics of burning plasma. The implementation in the ITER machine is challenging, particularly for the in-vessel diagnostics, region defined between the vacuum vessel and first wall (FW) contours, where space is constrained by the high number of systems. This paper describes the current status of design integration efforts to implement diagnostics in the ITER first wall. These approaches are the basis for detailed optimization and improvement of conceptual interfaces designs between systems. © 2015 Elsevier B.V. All rights reserved.
1. Introduction Diagnostics play an essential role for the successful operation of the ITER machine. As a key project requirement, ITER shall not operate without a working diagnostic which provides measurements for machine protection, basic control and advanced plasma control [1]. Some diagnostics require a plasma facing element attached to the first wall (FW), commonly known as a FW diagnostic, which contribute to measure a wide range of plasma parameters. Therefore, in order to guarantee a proper integration in the machine their design must comply not only with their functional requirements but also with the design of the blankets.
∗ Corresponding author. Tel.: +34 650842075. E-mail address: [email protected] (G. Martínez). http://dx.doi.org/10.1016/j.fusengdes.2015.06.046 0920-3796/© 2015 Elsevier B.V. All rights reserved.
This is particularly challenging because not all systems have reached the same level of design maturity. While most of the internal components (blankets and divertor) have passed the final design review, the majority of diagnostic systems are at a conceptual design level. 2. Blanket system The blanket system consists of 440 blanket shield modules (BSMs). Each module is composed of a first wall (FW) and a shield block (SB). The plasma-facing FW is composed of beryllium armour components attached to copper substrates mounted on a watercooled stainless steel support structure called central beam. The FW structure is mounted on the SB through a central bolt. The hydraulic connection between the FW and the SB is provided by flexible pipes on the back of the beam. The mechanical, water connection and electrical interfaces between FW and SB are accessible
G. Martínez et al. / Fusion Engineering and Design 98–99 (2015) 1548–1551
Fig. 1. EHF (upper) and NHF (lower) technologies used on the FWs.
through the front face of the FW on a 60 mm width recessed central slot, where FW diagnostics will be mounted through a dedicate interface described in Section 5. There are two versions of the FWs depending on the heat flux level they are exposed to – the enhanced heat flux (EHF) and the normal heat flux (NHF) panels, capable of accommodating 4.7 MW/m2 and 1–2 MW/m2 respectively [2]. Two different manufacturing routes are currently anticipated for NHF and EHF; hot isostatic pressing (HIP) by Europe, beryllium tile brazing by Russia and beryllium tile HIP by China [3] (Fig. 1). 3. FW diagnostics The concept FW diagnostic is introduced to group diagnostics that require a plasma facing component installed in the central slot of a FW panel. These are: The visible infrared (vis/IR) TV system has the primary role of the measuring FW surface temperature, luminance and temperature during edge-localized modes (ELMs) [4]. Its main priorities are machine protection and basic machine control measurements [1]. The vis/IR TV system requires black bodies (BB) for their calibration. A number of BBs are located on the central beam of the dedicated FWs [5]. The primary aim of the edge and core Thomson scattering systems is to measure the electron temperature and density profiles at the periphery and core of the plasma for physics studies and advanced plasma control [1]. Installation of beam dumps for the high-power lasers will require unique cut-outs on the FW central beam and a reduction of the FW central slot insert. To simplify manufacturing an identical beam dump interface will be used for both diagnostics [5]. The poloidal polarimeter (PoPola) is used as the primary diagnostic to measure the current profile. It is used for advanced plasma control physics studies. Major components of the system are located in upper port 10 and equatorial port 10. In addition, 12 corner cube retroreflectors (CCR) are required to be installed at the central slot of the inner wall blankets and two in the divertor for reflection of the probing laser beams launched into the vacuum vessel [5]. The primary role for FW samples is to measure erosion. Supplementary measurements are dust accumulation, deposition effects and tritium retention. FW samples will be used for machine protection basic machine control measurements and there may be a “safety function” in the future. The samples are to be removed by remote handling by the multipurpose deployer during each
1549
shutdown during the hydrogen phase, deuterium phase and D–T phase. The samples distribution in the machine includes areas close to gas injectors and glow discharge cleaning (GDC) anodes. Access to the samples is assumed to be approximately on an annual basis [6]. The primary role of H-alpha spectroscopy (+visible spectroscopy) is to measure line emission of hydrogen isotopes and impurities from ITER FW [7]. This diagnostic provides essential information for plasma control, machine protection and physics study. H-alpha spectroscopy consist of four optical channels located in 3 port plugs: two poloidal wide field of view (FOV) channels in EPP#11 which cover the inner and upper part of the wall, one tangential wide-FOV channel in EPP#12 (covering the outer wall), and one wide-FOV channel in UPP#2 covering divertor and BM #18. The majority of H-alpha beam dumps are allocated in the central slots of blanket modules in front of EPP#11 covering the inner and upper wall [5,8]. The primary roles of the In Vessel Viewing System (IVVS) are in vessel inspection for viewing and metrology purposes and to monitor operational parameters of the main plasma FW for physics studies, advanced plasma control and safety limits (1000 kg dust limit in vessel) [9]. The IVVS will be installed in 6 lower ports (3, 5, 9, 11, 15, 17) and consists out of a mechanical deployer and a scanning head in each position. On the blankets, fiducials (specific reference points) are foreseen to improve the precision of the measurements [6]. 4. Main requirements and constraints There are high levels of potentially damaging fluxes at the FW. Thus, the FW diagnostic designs have to cope with nuclear heating, surface heat and electromagnetic radiation fluxes. Moreover, limited space and a compacted geometry of the blanket, force the FW diagnostics to be passively cooled through the FW beam structure which is actively cooled. Blanket constraints: • FW mounted diagnostics can be implemented only in the central slot by a dedicated cut-out subject to a minimum clearance between existing holes and cooling channels. • Re-arrangement of cooling channels or modification of the water box is only possible if the water flow imbalance between wings is less than 10%. • Hot spots at the beam structured shall not create excessive stresses. • Cut-outs and diagnostic attachments should be compatible with FW manufacturing route. • Under no circumstance can the FW diagnostics jeopardize the Blanket system requirements. FW diagnostics requirements: • FW diagnostics should be visually accessible or properly aligned to the focal points located in ports. • It must be protected from damaging heat and redisposition fluxes. Proper diagnostic recession in the central slot is necessary. Also, it is fundamental to have good passive cooling to ensure survivability and performance. • No remote handling class 1 is foreseen for the majority of FW diagnostics except for FW samples. 5. Design of interfaces Accordingly several unique FW diagnostic attachments which meet the required measurements capabilities and respect the
1550
G. Martínez et al. / Fusion Engineering and Design 98–99 (2015) 1548–1551
Fig. 3. Toroidal cross section of the CSI showing the interface solution for EHF panels.
Fig. 2. Poloidal cross section of the FW beam showing the interface solution for NHF panels.
blanket design constraints have been designed. To facilitate interfaces FW diagnostics have been grouped into two categories: – Standard attachments which will use the standard FW design modifications, already implemented to support such interfaces for Group 1 diagnostics and – Variant attachment which require variants on the FW design for Group 2 diagnostics.
Fig. 4. Detail of the Thomson scattering beam dump interface.
This first group includes the following diagnostics: Vis/IR TV system, FW samples and IVVS fiducial markers. The main constraints for developing the attachment design are geometrical and heat loads. A through hole of 55 mm in the case of NHF and a conical depression on the central slot insert of EHF panels have been provided to accommodate them. Although the requirements for each system may differ, a suitable solution for Group 1 diagnostics has been found. Depending on the FW technology used (NHF/EHF) one of two possible attachments solutions will be used. In the case of NHF a CuCrZr body has been added as a heat sink/adaptor in order to improve the thermal conductivity between the FW sample and the SS beam structure. A pressing fit sleeve will provide enough pressure to be cooled down passively through the beam structure walls. FW samples are required to be remote hand-able (RHC1), thus they will be screwed into position and locked by a specific feature (Fig. 2). In the case of EHF panels, the FW diagnostic is directly mounted on the central slot insert (CSI) which provides protection from heat fluxes and a good thermal conductance through a conical hole and a threaded part (Fig. 3).
need to be placed close to the burning plasma even though they are particularly susceptible to re-deposition and how this impacts the reflectivity, with beryllium as the dominant constituent. Consequently, CCRs will be recessed deeper in the central beam structure and properly aligned with the corresponding line of sight. This adds complexity to the compacted geometry of the blanket which must now cope with bigger cut-outs. Design details of this interface are still to be agreed. H-alpha dumps are mainly a cavity which works as a black body. They are vital for suppressing strong divertor stray light which can be up to few hundreds the signal from the FW. Dumps will enable the stray light to be separated from the true signal. They shall provide at least 10 fold reduction of stray light [7]. The solution proposed to integrate dumps is to use the already existing hole in the central slot such as the ones provided for PoPola’s CCRs and Thomson scattering beam dumps. Applying a surface treatment in the cavities, the roughening required to collect more light and improve dumps performance can be achieved. Edge and core Thomson scattering beam dumps are used for the purpose of safely terminating the laser beam and minimizing the generation of stray light. Their interfaces are placed on the top of the central slot inside the FW central beam. A cut out of 100 mm × 60 mm × 77 mm on the central beam will host the beam dumps and a bolted interface of 4×M5 will fix it in position. The cooling of the BD is passive, based on the contact with the FW (Fig. 4).
5.2. Variant attachment
6. Ongoing analysis
This second group includes: poloidal polarimeter, H-alpha spectroscopy and edge and core Thomson scattering. The main constraints for developing their attachment are not only geometrical and head loads, but also re-deposition and light reflection. Due to the harsh ITER environment, in vessel optics have to address important issues such as survivability and thermal distortion. PoPola systems are no exception. The CCR and mirrors
The feasibility of the designs is in process of being validated through a Thermo-mechanical and hydraulic analysis. The loads applied are based on the Blanket System Load Specification [10] and Nuclear Heating and Radiation Damage in BM04 with Detailed Manifold and IM Key Configuration [11]. Maximal temperature found on samples under photonic and neutronic heat loads for EHF panel is 400 ◦ C. While the hydraulic
5.1. Standard attachment
G. Martínez et al. / Fusion Engineering and Design 98–99 (2015) 1548–1551
1551
Disclaimer The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. Acknowledgements
Fig. 5. (Left) Max. temperature on EHF standard attachment. (Right) Pressure drop map on NHF panel.
analyses shows that the impact of additional diagnostic barrels on the finger flow distribution is minor, the finger flow balance is below 10%. These first results are investigating the feasibility of implementing standard attachments. Nevertheless, only part of the thermal and hydraulic analysis has been done (Fig. 5). A global thermomechanical analysis in order to fully validate the interfaces design solutions remains to be done. 7. Summary An iterative communication with both parties (blanket and diagnostic groups) has resulted in a unique integration design that incorporates FW diagnostics into the machine and achieves the project requirements. An integrated design concept has been developed and provides a design that respects the requirements of each of the systems. Thermo-mechanical analyses are on-going to confirm that this configuration respects the heat loads limits on the blanket FW.
The authors are thankful to all ITER IO members, external contractors and colleagues in the Domestic Agencies for their continuous support and help in the design development of the integration of FW diagnostic. References [1] S. Chiocchio, Project Requirements, ITER Cadarache, May 2010 (Private communication). [2] R. Mitteau, B. Calcagno, R. Chappuis, The design of the ITER first wall panels, Fusion Eng. Des. 88 (October (6–8)) (2013) 568–570. [3] M. Merola, F. Escourbiac, R. Raffray, et al., Overview and Status of ITER Internal Components, ISFNT, 2013. [4] S. Salasca, Development of equatorial visible/infrared wide angle viewing system and radial neutron camera for ITER, Fusion Eng. Des. 84 (June) (2009) 1689–1696. [5] A. Martin, In Vessel Component Variants V1.15, ITER Cadarache, 2014 (Private communication). [6] R. Reichle, Review of the ITER Diagnostics Suite for Erosion, Deposition, Dust and Tritium Measurements, PSI, 2014. [7] A.B. Kukushkin, Divertor Stray Light Analysis in JET-ILW and Implications for the H-alpha Diagnostic in ITER, AIP, 2014. [8] A.J.H. Donné, Diagnostics, Nucl. Fusion 47 (2007) S337–S384 (Chapter 7). [9] C. Neri, ITER in vessel viewing system design and assessment activities, Fusion Eng. Des. 86 (October) (2011) 1954–1957. [10] B. Calcagno, Blanket System Load Specification, ITER Cadarache, 2013 (Private communication). [11] M. Sawan, Nuclear Heating and Radiation Damage in BM04 with Detailed Manifold and IM Key Configuration, ITER Cadarache, 2012 (Private communication).