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ITER diagnostic port plug engineering design analysis in the EU
Fusion Engineering and Design 82 (2007) 1231–1237
ITER diagnostic port plug engineering design analysis in the EU E. Ciattaglia a,∗ , L.C. Ingesson a , D. Campbell a , G. Saibene a , C. Walker b , L. Doceul c , P. Dirken d , L. Petrizzi e , R. Heidinger f a
European Fusion Development Agreement (EFDA), Close Support Unit (CSU), Garching, Boltzmannstr. 2, D-85748 Garching bei M¨unchen, Germany b ITER International Team, Cadarache, France c CEA, Association Euratom-CEA, Commissariat a ` l’´energie Atomique, Cadarache, France d UKAEA/EURATOM Fusion Association, Culham Science Centre, Culham, UK e ENEA, Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, Frascati, Italy f FZK, Forschungszentrum Karlsruhe, Association FZK-EURATOM, Karlsruhe, Germany
Abstract Engineering analysis has been carried out on a representative equatorial diagnostic port plug of ITER, outcome of which is given here. A preliminary overview of the work for a prototypical diagnostic upper port plug is also reported. To ensure that the port plug structure satisfies the requirements of the ITER environment, the following analyses have been performed: finite element (FE) analyses of static and dynamic behaviour under electromagnetic loads, FE vibration analysis, FE thermal-structural analysis and nuclear heating and radiation dose studies. The main outcomes from these analyses and consequent design developments are a revision of the stainless steel/water ratio in the plug neutron shielding modules and the incorporation of improved neutron shielding of the port duct, a modification of the port plug top plate arrangement, to increase torsional stiffness of the structure under disruption loads, and improved solutions for cooling arrangements. Manufacturing studies have also been performed, involving the assessment of methods for the production of plug flanges, the analysis of welding techniques for parts assembly and methods for the introduction of cooling features in plug components. Reference design solutions and possible modifications will be presented and discussed. © 2007 Published by Elsevier B.V. Keywords: ITER; Diagnostics; Engineering; Design; Analysis
1. Introduction
∗
Corresponding author. Tel.: +49 89 3299 4420; fax: +49 89 3299 4312. E-mail address: [email protected] (E. Ciattaglia). 0920-3796/$ – see front matter © 2007 Published by Elsevier B.V. doi:10.1016/j.fusengdes.2007.08.018
ITER will have a comprehensive plasma measurement system consisting of about 45 individual diagnostics, involving a wide range of technologies. The diagnostics provide measurements of a wide range
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E. Ciattaglia et al. / Fusion Engineering and Design 82 (2007) 1231–1237
of plasma parameters for use in machine protection, feedback loops for basic and advanced plasma control, and performance evaluation and physics analysis [1]. A large part of the diagnostic systems will be placed inside port plugs, large water cooled St-St structures located in the tokamak vacuum vessel ports, inside the primary vacuum, at the equatorial and upper levels. The EU will supply to ITER a number of diagnostic procurement packages, which include diagnostics and port plugs. In the past year, the EU has contributed significantly to activities launched within ITER to review and advance the work carried out by the ITER international team (IT) on the engineering design of the ITER diagnostic port plugs. The objective of these activities, undertaken as a co-ordinated activity and involving all ITER Participant Teams and the ITER IT, is to develop recommendations for common guidelines and procedures on design and manufacturing of the port structures. Within this framework, a series of performance analyses has been undertaken by the European Associations on the port plug outline design to ensure that the structures satisfy the requirements of the ITER environment. Two representative ports were chosen for this activity: one at the equatorial level and one at the upper level. The work presented here focuses on the engineering analysis of the representative equatorial port plug, although an overview is given on preliminary work for some prototypical upper port plug designs.
Table 1 Main dimensions and weights of representative equatorial and upper port plug Upper
Equatorial
1.12 1.11 5.69 6.08 12.98
2.16 1.66 2.23 3.64 16.14
Diagnostic/shield module Height (all) (m) Width (all) (m) Length (typical) (m) Drained weight (typical) (t) Drained weight (max) (t)
0.91 0.44 1.00 1.99 4.00
1.88 1.44 0.78 5.42 6.41
BSM Height (m) Width (m) Length (m) Drained weight (t)
1.37 0.80 0.50 2.29
2.10 1.85 0.6 9.77
22.37
42.17
23.97
45.03
Port plug structure Height (m) Width (m) Length (m) Length with BSM (m) Drained weight (t)
Total drained weight (including diagnostic and shielding) (t) Total water filled weight (t)
structure is made of stainless steel plates and flanges, cooled by water channels providing cooling to the plug components, the diagnostic modules and the blanket shield module (BSM). The principal loads on the port plug structure and on its joint arrangement to the vacuum vessel port are: own weight, electromagnetic (EM) loads from plasma
2. Port plug engineering Port plugs are large structures located inside the vacuum vessel ports, supported at the rear flange through a bolted joint to the vacuum vessel ports flange (where the primary vacuum is closed). Main dimensions and weights for a typical equatorial and upper port plug are shown in Table 1. The port plugs host the diagnostic components and provide, through the rear vacuum closure plate, feedthroughs for the diagnostic optical lines and services [2] (Figs. 1 and 2). The port plug main functions are nuclear and heat shielding, support of the blanket shield modules and provision of the primary vacuum closure. In addition, they must support the diagnostic components, while providing the required working conditions for the diagnostic systems. Their
Fig. 1. Equatorial port plug 1.
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transients and thermal loads from nuclear heating. In order to check the strength of the port plug structure against these loads, several preliminary analyses have been performed: structural analysis, neutronic analysis, thermal and cooling arrangements analysis and manufacturing studies. All the analysis reported in the next section were performed on the representative equatorial port plug, equatorial number 1, and the features present in the details of the plug components are those relevant to the diagnostic components located in the equatorial port plug 1 (Figs. 1 and 2).
3. Design developments on equatorial port plug
Fig. 3. Displacement for representative equatorial port plug when EM loads are applied.
3.1. Structural analysis A preliminary design of the representative equatorial plug includes three diagnostic shielding modules (as shown in Fig. 2) and assumes that those are assembled from the top of the plug, bolted to individual top beams forming a top cover for the plug. An engineering study has been carried out to evaluate different design options to ensure the required structural stiffness of the plug structure when worse case EM loads are applied [3]. This is mainly dependent on the joint of the top
beams to the side, front and back walls of the plug and between themselves, all the other joins (e.g. between side walls and base plate) being welded joints. Bolted and welded arrangements were evaluated using finite element (FE) analysis and it was found that having a continuous top beam welded on the structure would reduce the plug structure displacement by more than a factor of two compared with the design using the individual beams bolted, to a value of about 2.2 mm (Fig. 3). This is achieved within acceptable stresses if the top
Fig. 2. Exploded model of equatorial port plug 1.
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plate is welded to the structure, while a bolted arrangement needs more careful evaluation for it to provide the appropriate strength. Electromagnetic loads employed in the first evaluation of the plug structural stiffness were characterised by a predominant radial torque of about 11 MNm, arising from a central disruption (CD) of 27 ms characteristic linear decay time. At a later stage, the loads were recalculated for an exponential 18 ms main disruption and considering a corrected map of the toroidal field versus the radius of the tokamak. The maximum radial torque was calculated to be 5.8 MNm. This new reduced EM loads make the design of a bolted joint of the top plate, a preferred option to ease assembly and disassembly of the structure, less challenging. Results of a preliminary vibration analysis of the equatorial plug show [3] that no resonance phenomena to the 27 ms CD event occur and that the dynamic amplification factor for stress and displacement is negligible. A seismic analysis was also carried out [3], which did not show any problems for the plug survivability and functionality. Studies of different design options for the BSM attachments are presently being carried out, following ITER guidelines, to check for required performance under the severe structural and thermal loadings. Ongoing analysis and future work will focus on a detailed analysis of the bolted arrangement of the plug top plate to the rest of the structure with the newly evaluated EM loads. 3.2. Neutronic analysis Neutronic calculations were performed using the Monte Carlo N-Particle Transport Code (MCNP version 4C2) [5], supplied with FENDL2.0 [6] nuclear data library. The model for the representative equatorial plug, including for simplicity only the components for the radial neutron camera (RNC), was used to calculate the dose rate after shut down at the back of the port plug, the effect on this of the diagnostic apertures, the effect in neutron streaming of the 20 mm assembly gap between port and plug, as well as an optimisation of the stainless steel to water ratio of the plug, and the nuclear heating on different regions of the plug. The dose rate due to activation in the interspace behind the port plug was found below the prescribed
Fig. 4. Nuclear heating at various radial distances from the front of the plug.
maximum dose rate 10 days after machine shutdown of 100 Sv/h. It was concluded that almost half of the dose rate (of a total of up to 70 Sv/h) originates from the port to plug gap, and the remainder from the RNC collimator. Improvements in the design of the port gap (dog legs at plug front and shielding elements at the back) are being implemented as the result of this analysis. The composition (steel to water ratio) of the port plug diagnostic shielding modules has been varied to assess the scope of a reduction of dry weight (as required for handling) from the reference composition of 80% steel and 20% water without significant increase in neutron streaming. A composition of 50% steel and 50% water seems acceptable, which has prompted a change of the reference design to 30, 60 and 70% water in the first, second and third diagnostic module respectively, bringing the dry weight below the limit value of 50 t. A radial profile of the nuclear heating in the plug is shown in Fig. 4. The total heat to be removed from the plug is 95 kW. This heating is only weakly dependent on the port gap, but is significantly higher around diagnostic apertures. The calculated values were used as input for the port plug thermal analysis (see Section 3.3). A preliminary analysis using the 3D discrete ordinates code Attila [7] was also carried out to prove, as a first step, the capability of the code to read CAD models and the higher speed in performing the calculations. Further future work in neutronic studies will include: (1) analysis including all diagnostics allocated to equatorial port plug 1, (2) evaluation of the St-St to water ratio proposed as reference by the ITER IT, (3) more parametric studies of neutron streaming through diagnostic labyrinths and (4) benchmark of Attila for streaming calculations.
E. Ciattaglia et al. / Fusion Engineering and Design 82 (2007) 1231–1237
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Fig. 5. Schematic of equatorial port plug cooling circuit.
3.3. Cooling and thermal analysis A reference layout for the cooling of the port plug is shown in Fig. 5. Preliminary hydraulic analysis and thermal calculations (temperature distribution and stress) were carried out to study the effectiveness of the cooling, given the input heating loads from the neutronic analysis (Fig. 4). The main results are (Fig. 6): (1) the front of the plug sees the highest heat input and requires the present cooling arrangements to be improved, (2) as the heat input decreases rapidly going towards the rear of the plug, some of the back com-
ponents (e.g. back flange) may be left un-cooled, (3) to satisfy baking temperature requirements and reduce thermal stresses, some form of heat input would be needed for the un-cooled parts of the plug. Thermal analysis was performed to check the temperature and stress rise in the plug structure, diagnostic modules and front flange (to which the BSM is attached) [3]. The parts at the front of the plug, receiving the higher nuclear heating, are the critical ones and present detailed analyses are investigating optimisation of the cooling channels around the different diagnostic apertures in the front frame and first diagnostic module. Present and future studies also include an overall revision of the cooling circuit to incorporate the design change to a plug continuous top beam and to identify critical features, also in view of plug draining requirements, and the manufacturing study of remote handling (RH) water weld connections. 3.4. Manufacturing studies
Fig. 6. Temperature distribution (◦ C) on equatorial port plug when nuclear heating is applied.
Manufacturing methods for the different parts of the plug structure have been investigated. For the plug side and bottom plates, casting versus forging methods were evaluated. Casting would need to be followed by hot isostatic pressure (HIP) treatment to reduce the number of defects introduced during the casting procedure. However, casting does not seem a viable option for various reasons, including: (1) plate dimensions
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are too large to be treated using the HIP method, (2) different grade of stainless steel (compared to ITER Grade) is required for casting, (3) vacuum compatibility and aqueous corrosion for ITER conditions need to be evaluated. Forging and machining of the front and back flanges should be achievable with standard tools and techniques. Gun-drilling can be used to introduce cooling holes in the plates, although the effects of the drilling ‘drift’ (about 1 mm every 1 m) need further evaluation. Possible techniques for the 130 mm weld between the base and sides were investigated, including narrow gap tungsten inert gas (TIG) welding (ITER reference technique), sub-merged arc welding, electroslag welding and electron beam (EB) welding. The first three techniques bring the highest heat load to the system and therefore the highest potential distortion. The weld area for EB welding is the smallest therefore distortion is expected to be minimal. In order to achieve the dimensional tolerances required (±1 mm generally, ±0.5 mm on high-tolerances surfaces), distortion during and after welding will need to be carefully controlled. Rigid jigging of the components during welding and stress relief by heat treatment afterwards is required. However, machining of the final components may still be needed to achieve the design tolerances. Testing of weld techniques, weld preparations, fixtures, heat treatments, post welding machining, etc. will be required to establish which technique is best suited to the manufacture of the port plug. A present task is looking at simulating welding distortions through software codes with the aim of comparing welding distortion of the sides-to-base joint for different welding techniques.
assembly of the system components. Similarly to the equatorial plug, the engineering challenge is presented by the joint between the top plate and the rest of the structure, which is required to have appropriate shear strength against the EM radial torque. For this reason, various design solutions are being evaluated for different dimensions of the aperture.
5. Conclusions The work carried out has given preliminary results on the port plug reference design, highlighting challenges in the engineering and manufacturability of the diagnostic port plug structures. With the enhanced level of preparations for ITER construction, the scope of this work is expanding to include: (a) Considerably more detailed analyses (b) Quality assurance aspects throughout the design stages (c) Integration of the port plugs with the diagnostic systems and of the latter with other systems in the same plug. An effort is being made to co-ordinate ongoing and future activities within EU and, thorough the ITER team, with the other ITER parties, in order to: (i) reduce the variety of technical solutions, (ii) avoid analysis duplications, (iii) develop common guidelines for integration and (iv) harmonise plans for integration. Acknowledgments
4. Design developments on upper port plug While in the past year the EU Associations have mainly been involved in the engineering of the representative equatorial port plug (as shown in the sections above) and work on the upper port plug was developed by other ITER parties, this year some studies have started in the EU to investigate the possibility of an upper port plug structure common for diagnostics and the four upper plugs hosting the electron cyclotron resonance heating (ECRH) systems [4]. The potential common design is based on that of a variant of the ECRH upper plug under study at present, where a top aperture is introduced to ease assembly and dis-
This work, supported by the European Communities under the contract of Association between EURATOM and the EU Fusion laboratories involved, was carried out within the framework of the European Fusion Development Agreement (EFDA). The views and opinions expressed herein do not necessarily reflect those of the European Commission. References [1] A.E. Costley, T. Sugie, G. Vayakis, et al., Technological challenges of ITER diagnostics, 23rd Symposium on Fusion Technology, September 20–24, 2004, Venice, Italy, Part A, pp. 109–119.
E. Ciattaglia et al. / Fusion Engineering and Design 82 (2007) 1231–1237 [2] C.I. Walker, A.Costley, K. Itami, T. Kondoh, T. Sugie, G. Vayakis, et al., ITER diagnostics: integration and engineering aspects, Review of Scientific Instruments, October 2004, vol. 75, No. 10, pp. 4243–4246. [3] L. Doceul, C. Walker, C. Ingesson, E. Ciattaglia, P. Chappuis, C. Portafaix, et al., CEA Engineering Studies and Integration of the ITER Diagnostic Port Plugs, in: 24rd Symposium on Fusion Technology, Warsaw, Poland, September 11–15, 2006. [4] P. Sp¨ah, R. Heidinger, G. Hailfinger, M. Henderson, K. Kleefeld, A. Serikov, et al., Design and Analysis of the Structural Components in the ITER ECH Upper Port Plug, in: 24rd Symposium on Fusion Technology, Warsaw, Poland, September 11-15, 2006.
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[5] MCNP 4B, Monte Carlo N-Particle Transport System, Los Alamos National Laboratory, Los Alamos, New Mexico, 1997. [6] A. Pashshenko, H. Wienke, FENDL/E-2.0, Evaluated Nuclear Data Library of Neutron Nuclear Interaction Cross Sections and Photon Production Cross Sections and Photon-Atom Interaction Cross Sections for Fusion Applications, Report IAEA-NDS-175, March 1997. [7] T.A. Wareing, J.M. McGhee, J.E. Morel, ATTILA, a three dimensional unstructured tetrahedral mesh discrete ordinales code, Trans. Am. Nucl. Soc. 75 (1996) 146–147.
ITER diagnostic port plug engineering design analysis in the EU E. Ciattaglia a,∗ , L.C. Ingesson a , D. Campbell a , G. Saibene a , C. Walker b , L. Doceul c , P. Dirken d , L. Petrizzi e , R. Heidinger f a
European Fusion Development Agreement (EFDA), Close Support Unit (CSU), Garching, Boltzmannstr. 2, D-85748 Garching bei M¨unchen, Germany b ITER International Team, Cadarache, France c CEA, Association Euratom-CEA, Commissariat a ` l’´energie Atomique, Cadarache, France d UKAEA/EURATOM Fusion Association, Culham Science Centre, Culham, UK e ENEA, Associazione EURATOM-ENEA sulla Fusione, ENEA Centro Ricerche, Frascati, Italy f FZK, Forschungszentrum Karlsruhe, Association FZK-EURATOM, Karlsruhe, Germany
Abstract Engineering analysis has been carried out on a representative equatorial diagnostic port plug of ITER, outcome of which is given here. A preliminary overview of the work for a prototypical diagnostic upper port plug is also reported. To ensure that the port plug structure satisfies the requirements of the ITER environment, the following analyses have been performed: finite element (FE) analyses of static and dynamic behaviour under electromagnetic loads, FE vibration analysis, FE thermal-structural analysis and nuclear heating and radiation dose studies. The main outcomes from these analyses and consequent design developments are a revision of the stainless steel/water ratio in the plug neutron shielding modules and the incorporation of improved neutron shielding of the port duct, a modification of the port plug top plate arrangement, to increase torsional stiffness of the structure under disruption loads, and improved solutions for cooling arrangements. Manufacturing studies have also been performed, involving the assessment of methods for the production of plug flanges, the analysis of welding techniques for parts assembly and methods for the introduction of cooling features in plug components. Reference design solutions and possible modifications will be presented and discussed. © 2007 Published by Elsevier B.V. Keywords: ITER; Diagnostics; Engineering; Design; Analysis
1. Introduction
∗
Corresponding author. Tel.: +49 89 3299 4420; fax: +49 89 3299 4312. E-mail address: [email protected] (E. Ciattaglia). 0920-3796/$ – see front matter © 2007 Published by Elsevier B.V. doi:10.1016/j.fusengdes.2007.08.018
ITER will have a comprehensive plasma measurement system consisting of about 45 individual diagnostics, involving a wide range of technologies. The diagnostics provide measurements of a wide range
1232
E. Ciattaglia et al. / Fusion Engineering and Design 82 (2007) 1231–1237
of plasma parameters for use in machine protection, feedback loops for basic and advanced plasma control, and performance evaluation and physics analysis [1]. A large part of the diagnostic systems will be placed inside port plugs, large water cooled St-St structures located in the tokamak vacuum vessel ports, inside the primary vacuum, at the equatorial and upper levels. The EU will supply to ITER a number of diagnostic procurement packages, which include diagnostics and port plugs. In the past year, the EU has contributed significantly to activities launched within ITER to review and advance the work carried out by the ITER international team (IT) on the engineering design of the ITER diagnostic port plugs. The objective of these activities, undertaken as a co-ordinated activity and involving all ITER Participant Teams and the ITER IT, is to develop recommendations for common guidelines and procedures on design and manufacturing of the port structures. Within this framework, a series of performance analyses has been undertaken by the European Associations on the port plug outline design to ensure that the structures satisfy the requirements of the ITER environment. Two representative ports were chosen for this activity: one at the equatorial level and one at the upper level. The work presented here focuses on the engineering analysis of the representative equatorial port plug, although an overview is given on preliminary work for some prototypical upper port plug designs.
Table 1 Main dimensions and weights of representative equatorial and upper port plug Upper
Equatorial
1.12 1.11 5.69 6.08 12.98
2.16 1.66 2.23 3.64 16.14
Diagnostic/shield module Height (all) (m) Width (all) (m) Length (typical) (m) Drained weight (typical) (t) Drained weight (max) (t)
0.91 0.44 1.00 1.99 4.00
1.88 1.44 0.78 5.42 6.41
BSM Height (m) Width (m) Length (m) Drained weight (t)
1.37 0.80 0.50 2.29
2.10 1.85 0.6 9.77
22.37
42.17
23.97
45.03
Port plug structure Height (m) Width (m) Length (m) Length with BSM (m) Drained weight (t)
Total drained weight (including diagnostic and shielding) (t) Total water filled weight (t)
structure is made of stainless steel plates and flanges, cooled by water channels providing cooling to the plug components, the diagnostic modules and the blanket shield module (BSM). The principal loads on the port plug structure and on its joint arrangement to the vacuum vessel port are: own weight, electromagnetic (EM) loads from plasma
2. Port plug engineering Port plugs are large structures located inside the vacuum vessel ports, supported at the rear flange through a bolted joint to the vacuum vessel ports flange (where the primary vacuum is closed). Main dimensions and weights for a typical equatorial and upper port plug are shown in Table 1. The port plugs host the diagnostic components and provide, through the rear vacuum closure plate, feedthroughs for the diagnostic optical lines and services [2] (Figs. 1 and 2). The port plug main functions are nuclear and heat shielding, support of the blanket shield modules and provision of the primary vacuum closure. In addition, they must support the diagnostic components, while providing the required working conditions for the diagnostic systems. Their
Fig. 1. Equatorial port plug 1.
E. Ciattaglia et al. / Fusion Engineering and Design 82 (2007) 1231–1237
1233
transients and thermal loads from nuclear heating. In order to check the strength of the port plug structure against these loads, several preliminary analyses have been performed: structural analysis, neutronic analysis, thermal and cooling arrangements analysis and manufacturing studies. All the analysis reported in the next section were performed on the representative equatorial port plug, equatorial number 1, and the features present in the details of the plug components are those relevant to the diagnostic components located in the equatorial port plug 1 (Figs. 1 and 2).
3. Design developments on equatorial port plug
Fig. 3. Displacement for representative equatorial port plug when EM loads are applied.
3.1. Structural analysis A preliminary design of the representative equatorial plug includes three diagnostic shielding modules (as shown in Fig. 2) and assumes that those are assembled from the top of the plug, bolted to individual top beams forming a top cover for the plug. An engineering study has been carried out to evaluate different design options to ensure the required structural stiffness of the plug structure when worse case EM loads are applied [3]. This is mainly dependent on the joint of the top
beams to the side, front and back walls of the plug and between themselves, all the other joins (e.g. between side walls and base plate) being welded joints. Bolted and welded arrangements were evaluated using finite element (FE) analysis and it was found that having a continuous top beam welded on the structure would reduce the plug structure displacement by more than a factor of two compared with the design using the individual beams bolted, to a value of about 2.2 mm (Fig. 3). This is achieved within acceptable stresses if the top
Fig. 2. Exploded model of equatorial port plug 1.
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plate is welded to the structure, while a bolted arrangement needs more careful evaluation for it to provide the appropriate strength. Electromagnetic loads employed in the first evaluation of the plug structural stiffness were characterised by a predominant radial torque of about 11 MNm, arising from a central disruption (CD) of 27 ms characteristic linear decay time. At a later stage, the loads were recalculated for an exponential 18 ms main disruption and considering a corrected map of the toroidal field versus the radius of the tokamak. The maximum radial torque was calculated to be 5.8 MNm. This new reduced EM loads make the design of a bolted joint of the top plate, a preferred option to ease assembly and disassembly of the structure, less challenging. Results of a preliminary vibration analysis of the equatorial plug show [3] that no resonance phenomena to the 27 ms CD event occur and that the dynamic amplification factor for stress and displacement is negligible. A seismic analysis was also carried out [3], which did not show any problems for the plug survivability and functionality. Studies of different design options for the BSM attachments are presently being carried out, following ITER guidelines, to check for required performance under the severe structural and thermal loadings. Ongoing analysis and future work will focus on a detailed analysis of the bolted arrangement of the plug top plate to the rest of the structure with the newly evaluated EM loads. 3.2. Neutronic analysis Neutronic calculations were performed using the Monte Carlo N-Particle Transport Code (MCNP version 4C2) [5], supplied with FENDL2.0 [6] nuclear data library. The model for the representative equatorial plug, including for simplicity only the components for the radial neutron camera (RNC), was used to calculate the dose rate after shut down at the back of the port plug, the effect on this of the diagnostic apertures, the effect in neutron streaming of the 20 mm assembly gap between port and plug, as well as an optimisation of the stainless steel to water ratio of the plug, and the nuclear heating on different regions of the plug. The dose rate due to activation in the interspace behind the port plug was found below the prescribed
Fig. 4. Nuclear heating at various radial distances from the front of the plug.
maximum dose rate 10 days after machine shutdown of 100 Sv/h. It was concluded that almost half of the dose rate (of a total of up to 70 Sv/h) originates from the port to plug gap, and the remainder from the RNC collimator. Improvements in the design of the port gap (dog legs at plug front and shielding elements at the back) are being implemented as the result of this analysis. The composition (steel to water ratio) of the port plug diagnostic shielding modules has been varied to assess the scope of a reduction of dry weight (as required for handling) from the reference composition of 80% steel and 20% water without significant increase in neutron streaming. A composition of 50% steel and 50% water seems acceptable, which has prompted a change of the reference design to 30, 60 and 70% water in the first, second and third diagnostic module respectively, bringing the dry weight below the limit value of 50 t. A radial profile of the nuclear heating in the plug is shown in Fig. 4. The total heat to be removed from the plug is 95 kW. This heating is only weakly dependent on the port gap, but is significantly higher around diagnostic apertures. The calculated values were used as input for the port plug thermal analysis (see Section 3.3). A preliminary analysis using the 3D discrete ordinates code Attila [7] was also carried out to prove, as a first step, the capability of the code to read CAD models and the higher speed in performing the calculations. Further future work in neutronic studies will include: (1) analysis including all diagnostics allocated to equatorial port plug 1, (2) evaluation of the St-St to water ratio proposed as reference by the ITER IT, (3) more parametric studies of neutron streaming through diagnostic labyrinths and (4) benchmark of Attila for streaming calculations.
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Fig. 5. Schematic of equatorial port plug cooling circuit.
3.3. Cooling and thermal analysis A reference layout for the cooling of the port plug is shown in Fig. 5. Preliminary hydraulic analysis and thermal calculations (temperature distribution and stress) were carried out to study the effectiveness of the cooling, given the input heating loads from the neutronic analysis (Fig. 4). The main results are (Fig. 6): (1) the front of the plug sees the highest heat input and requires the present cooling arrangements to be improved, (2) as the heat input decreases rapidly going towards the rear of the plug, some of the back com-
ponents (e.g. back flange) may be left un-cooled, (3) to satisfy baking temperature requirements and reduce thermal stresses, some form of heat input would be needed for the un-cooled parts of the plug. Thermal analysis was performed to check the temperature and stress rise in the plug structure, diagnostic modules and front flange (to which the BSM is attached) [3]. The parts at the front of the plug, receiving the higher nuclear heating, are the critical ones and present detailed analyses are investigating optimisation of the cooling channels around the different diagnostic apertures in the front frame and first diagnostic module. Present and future studies also include an overall revision of the cooling circuit to incorporate the design change to a plug continuous top beam and to identify critical features, also in view of plug draining requirements, and the manufacturing study of remote handling (RH) water weld connections. 3.4. Manufacturing studies
Fig. 6. Temperature distribution (◦ C) on equatorial port plug when nuclear heating is applied.
Manufacturing methods for the different parts of the plug structure have been investigated. For the plug side and bottom plates, casting versus forging methods were evaluated. Casting would need to be followed by hot isostatic pressure (HIP) treatment to reduce the number of defects introduced during the casting procedure. However, casting does not seem a viable option for various reasons, including: (1) plate dimensions
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are too large to be treated using the HIP method, (2) different grade of stainless steel (compared to ITER Grade) is required for casting, (3) vacuum compatibility and aqueous corrosion for ITER conditions need to be evaluated. Forging and machining of the front and back flanges should be achievable with standard tools and techniques. Gun-drilling can be used to introduce cooling holes in the plates, although the effects of the drilling ‘drift’ (about 1 mm every 1 m) need further evaluation. Possible techniques for the 130 mm weld between the base and sides were investigated, including narrow gap tungsten inert gas (TIG) welding (ITER reference technique), sub-merged arc welding, electroslag welding and electron beam (EB) welding. The first three techniques bring the highest heat load to the system and therefore the highest potential distortion. The weld area for EB welding is the smallest therefore distortion is expected to be minimal. In order to achieve the dimensional tolerances required (±1 mm generally, ±0.5 mm on high-tolerances surfaces), distortion during and after welding will need to be carefully controlled. Rigid jigging of the components during welding and stress relief by heat treatment afterwards is required. However, machining of the final components may still be needed to achieve the design tolerances. Testing of weld techniques, weld preparations, fixtures, heat treatments, post welding machining, etc. will be required to establish which technique is best suited to the manufacture of the port plug. A present task is looking at simulating welding distortions through software codes with the aim of comparing welding distortion of the sides-to-base joint for different welding techniques.
assembly of the system components. Similarly to the equatorial plug, the engineering challenge is presented by the joint between the top plate and the rest of the structure, which is required to have appropriate shear strength against the EM radial torque. For this reason, various design solutions are being evaluated for different dimensions of the aperture.
5. Conclusions The work carried out has given preliminary results on the port plug reference design, highlighting challenges in the engineering and manufacturability of the diagnostic port plug structures. With the enhanced level of preparations for ITER construction, the scope of this work is expanding to include: (a) Considerably more detailed analyses (b) Quality assurance aspects throughout the design stages (c) Integration of the port plugs with the diagnostic systems and of the latter with other systems in the same plug. An effort is being made to co-ordinate ongoing and future activities within EU and, thorough the ITER team, with the other ITER parties, in order to: (i) reduce the variety of technical solutions, (ii) avoid analysis duplications, (iii) develop common guidelines for integration and (iv) harmonise plans for integration. Acknowledgments
4. Design developments on upper port plug While in the past year the EU Associations have mainly been involved in the engineering of the representative equatorial port plug (as shown in the sections above) and work on the upper port plug was developed by other ITER parties, this year some studies have started in the EU to investigate the possibility of an upper port plug structure common for diagnostics and the four upper plugs hosting the electron cyclotron resonance heating (ECRH) systems [4]. The potential common design is based on that of a variant of the ECRH upper plug under study at present, where a top aperture is introduced to ease assembly and dis-
This work, supported by the European Communities under the contract of Association between EURATOM and the EU Fusion laboratories involved, was carried out within the framework of the European Fusion Development Agreement (EFDA). The views and opinions expressed herein do not necessarily reflect those of the European Commission. References [1] A.E. Costley, T. Sugie, G. Vayakis, et al., Technological challenges of ITER diagnostics, 23rd Symposium on Fusion Technology, September 20–24, 2004, Venice, Italy, Part A, pp. 109–119.
E. Ciattaglia et al. / Fusion Engineering and Design 82 (2007) 1231–1237 [2] C.I. Walker, A.Costley, K. Itami, T. Kondoh, T. Sugie, G. Vayakis, et al., ITER diagnostics: integration and engineering aspects, Review of Scientific Instruments, October 2004, vol. 75, No. 10, pp. 4243–4246. [3] L. Doceul, C. Walker, C. Ingesson, E. Ciattaglia, P. Chappuis, C. Portafaix, et al., CEA Engineering Studies and Integration of the ITER Diagnostic Port Plugs, in: 24rd Symposium on Fusion Technology, Warsaw, Poland, September 11–15, 2006. [4] P. Sp¨ah, R. Heidinger, G. Hailfinger, M. Henderson, K. Kleefeld, A. Serikov, et al., Design and Analysis of the Structural Components in the ITER ECH Upper Port Plug, in: 24rd Symposium on Fusion Technology, Warsaw, Poland, September 11-15, 2006.
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