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Technological challenges at ITER plasma facing components production in Russia
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Technological challenges at ITER plasma facing components production in Russia I.V. Mazul a,∗ , V.A. Belyakov a , A.A. Gervash a , R.N. Giniyatulin a , T.М. Guryeva a , V.E. Kuznetsov a , A.N. Makhankov a , A.A. Okunev a , O.N. Sevryukov b a b
Efremov Institute, 196641 St. Petersburg, Russia MEPhI, 115409 Moscow, Russia
h i g h l i g h t s • 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.
a r t i c l e
i n f o
Article history: Received 27 August 2015 Received in revised form 13 January 2016 Accepted 19 January 2016 Available online xxx Keywords: First wall Divertor Multilayered structure Joining technologies NDT Mockups testing
a b s t r a c t 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. © 2016 Elsevier B.V. All rights reserved.
1. Introduction In the frames of the contracts signed with the ITER Organization (IO) Russia has to supply, in particular, 179 first wall (FW) panels (40% of full FW coverage), 58 divertor dome (DD) assemblies (100% of these components) and provide high heat flux (HHF) testing of various divertor components supplied by all Parties [1]. By now, manufacturing and testing of medium-sized mockups (semi-prototypes) of the divertor and FW have been, in general, completed, thus the design solutions and manufacturing technologies selected by RF have been qualified and proved [2]. Preparation of working models of full-size prototypes is nearly completed and after manufacturing and testing of these prototypes
∗ Corresponding author. Tel.: +7 9217710056. E-mail address: [email protected] (I.V. Mazul).
in 2018 decision on serial production will be made. So, the range of technologies to be used during manufacturing of ITER plasma facing components (PFC) in Russia has been, in general, defined and their complexity, originality and difficulty are described below. 2. Design features and specific requirements DD and FW are water (∼100 ◦ C, 4 MPa) cooled components subjected by plasma to high surface heat loads (up to 5 MW/m2 for 103 –104 cycles), significant electromagnetic loads and neutron irradiation. Above-mentioned components consist of (see Fig. 1) multilayered heat-sink elements (HSEs), the plasma facing wall of which is made from bronze CuCrZr [3]. This wall is armoured on the plasma facing side by protective tiles made of tungsten W (for the divertor) or beryllium Be (for FW). The rear part of HSE is made from steel SS316L(N)-IG and is thick enough to provide strength and geometrical stability of the structure under electromagnetic
http://dx.doi.org/10.1016/j.fusengdes.2016.01.030 0920-3796/© 2016 Elsevier B.V. All rights reserved.
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Fig. 1. Design of DD assembly (upper left) and FW panel (upper right); FW HSE and its cross-section (bottom left) and parts (bottom right) of FW HSE: steel rear part and bronze wall (overturned).
and thermal loads. Multiple HSE (in the average, ∼40 elements for FW panel and ∼34 elements for the DD assembly) are arranged on larger units (the so-called panels) based on massive steel supports, which serve also as water manifolds. Support structures have units (lugs for divertor pinning and bolts for FW pre-stressed pulling) for attachment to other reactor structures (to divertor cassettes or blanket modules). Panels segmentation on multiple HSE is forced by the necessity to solve the following issues: reduction of electromagnetic loads (decrease in area of induced current loops); decrease in thermal stresses (by minimizing sizes of regions with thermal gradients); reduction of cost/risk of whole panel rejection during manufacturing (a panel consists of a number of pre-tested HSE). Due to deficiency of coolant mass flow rate for FW panels the use of more conventional tubular seamless cooling channels was not possible. For this reason HSE is a bimetallic (CuCrZr–SS) box-like structure and has a complicated profile of plasma facing cooling channel (hypervapotron type), which is able to provide adequate cooling at low flow velocities. Manufacturing of such channel by seamless technology is not practical. As a result of technological optimization each unit length of this channel contains double length of bimetallic SS–CuCrZr joint and double length of steel weld. Totally for all ∼9000 HSE the length of joints achieves about 25 km. For 239 supplied PFC assemblies, each having significant sizes (∼1.5 m) and unit weight (∼1 t) there are about 40 different design options. This means that, in fact, instead of serial production of similar parts we have a vast number of parts different in size and shape. The total number of armour tiles is as many as beryllium (toxic, fragile) and tungsten (very hard and fragile). Moreover, ∼100 000 parts from steel, bronze and bimetal CuCrZr–SS have to be machined with an accuracy better than 50 m. For nuclear machine the nondestructive testing (NDT) of manufactured parts and their joining/welding is strongly required. In particular, we have to provide 100% control (by X-rays or ultrasonically) of vacuum tight SS–SS welds and SS–CuCrZr joints. About 300 m2 of brazing joints (Be–CuCrZr and Wu–Cu–CuCrZr) has to be subjected to ultrasonic and HHF testing. Despite that pressurized water (4 MPa) is used for PFC cooling the tightness of various joints and welds has to
provide the water leakage into the reactor vacuum vessel less than 10−8 Pa m3 /s. 3. Status of supplier qualification Technological capabilities of RF industry had to be demonstrated before signing the Procurement Agreement between RF and IO. Manufacturing of representative medium-scale (0.5–0.6 m long and 0.2–0.3 m wide) mockups and their testing under operational condition are typical methods used to check selected design solution and to qualify the potential supplier. Fig. 2 shows the divertor and FW panel semi-prototypes, designed and manufactured at the Efremov Institute. The most important operational parameters influencing the design, reliability and life time of PFC are surface heat load and number of thermal cycles. These operational parameters were simulated on the Tsefey and IDTF electron beam facilities [4]. The divertor semi-prototype successfully withstood 1000 cycles at 3 MW/m2 , then 1000 cycles at 5 MW/m2 and was damaged only after ∼500 cycles at a double operational load of 10 MW/m2 . Preliminary testing of the FW panel semi-prototype at operational loads of 4.7 MW/m2 for 100 cycles demonstrates good quality of all joints. Further more comprehensive testing of this semi-prototype is under way. 4. Machining of parts and required tolerances Machining of DD elements involves the following works: machining of HSE basements and bronze walls, cutting of W/Cu tiles, bending of tubes, machining of the manifolds and steel support structure. The most complicated technological problems are machining of deep slots of the manifolds and stubs of the steel support structure (Fig. 3). The highest accuracy requirements are imposed on the surfaces under laser welding to ensure the required quality of welded joints and elements fixing the divertor to cassette. Manufacturing precision of the above elements is within ±0.05 mm. Advanced 5-axis
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Fig. 2. Qualification semi-prototypes of the divertor (left) and FW panel (right).
machines with a positioning accuracy of less than 0.005 mm are required to perform the most complicated work.
5. Joining of dissimilar materials 5.1. CuCrZr–SS316 joining When developing fabrication process for ITER PFC, special attention shall be paid to the production of reliable joints between heat sink materials made of CuCrZr bronze and support structure made of austenitic steel. Two different SS/CuCrZr joining techniques have been proposed and tried in Russia: hot isostatic pressing (HIP) and explosion bonding [5]. Both of mentioned joining techniques were comparatively studied to emphasize their advantage or disadvantage. Mechanical properties of SS/CuCrZr joint, its vacuum tightness and structural parameters (uniformity of elements distribution, grain size of copper alloy and hardness of joint area) were selected for the joint characterization. Studies of the SS/CuCrZr diffusion zone after HIP (980 ◦ C, 2 h, 150 MPa) have shown that both techniques provide a continuous, defect-free zone structure. Analysis of the width of the diffusion zone shows that the overall width of the diffusion zone is about 110 m for HIP joint and order of 25 m for exploded one. For both technologies, the grain size of bronze is not greater than 200 m, which meets the ITER requirements (Fig. 4). Investigations of the SS/CuCrZr mechanical properties after manufacturing
cycle showed that both methods of forming steel/bronze joint allow achieving the ultimate strength more than 350 MPa at room temperature. To assess a vacuum tightness of the joint the number of samples of each technology were manufactured and successfully passed through the repeated internal pressure test (100 cycles at 7 MPa, 30 min). To validate SS/CuCrZr heat sink performance the number of actively cooled SS/CuCrZr mock-ups were subjected to high heat flux expected for such components. The results confirm the efficiency of both joint technologies with the expected values of the heat flux. Summarizing the results of comparative study one can conclude that both of considered joint methods give satisfied properties for HHF components application.
5.2. Be–CuCrZr brazing Proposing beryllium as plasma facing armour for the ITER FW and fast brazing as Be/CuCrZr joining technique [6], three options of fast heating were considered in Russia: electron beam heating, a direct ohmic heating and induction heating. When brazing Be tiles onto CuCrZr heat-sink it is necessary to prevent forming of brittle intermetallic Be/CuCrZr layer and to preserve the strength of bronze at the required level (of the order of 300 MPa at room temperature). Thus it is necessary to minimize the time taken at temperatures above 450–500 ◦ C.
Fig. 3. Machining of deep slots (width—25 mm, depth—285 mm) of the manifolds (upper line) and machining of stubs of the steel support structure (bottom line).
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Fig. 4. Microstructure of SS–CuCrZr joint produced by explosion bonding (upper left) and HIP (upper right) and grain size of CuCrZr after explosion bonding (bottom left) and HIP (bottom right) manufacturing cycles.
Fig. 5. Inductive (55 kHz, 20 kW) heating/brazing of Be tiles to 0.6 m long FW HSE. (Several thermocouples T1–T8 on HSE lateral walls in several positions along HSE length demonstrate sufficient uniformity of heating).
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Fig. 6. Example of welding of two cover plates (in the centre) 11 mm thick and metallographic cross-section of welds cut in several control points (at the periphery).
Fig. 7. Check of inner walls of box-like HSE by videoscope (left). Inner view: with spatters after conventional welding (centre) and after of proper technology (right).
Heating by electron beam leads to significant overheating of beryllium tiles. Ohmic heating requires extended bronze parts to be cut after brazing. The inductive heating method does not have the above disadvantage (Fig. 5). Inductive heating allows one to fulfil the fast heating requirements: pre-heating to 450 ◦ C for 60–90 min, fast heating from 450 to 710 ◦ C for 3–5 min. The temperature uniformity through the brazing zone is ±20 ◦ C. Brazing is
performed in a vacuum chamber at a residual pressure not exceeding 4 × 10−5 mbar. 5.3. W/Cu–CuCrZr brazing Joining technologies for W–Cu–CuCrZr composition are well described elsewhere [7,8]. We use vacuum casting of pure copper
Fig. 8. Welding of DD triangular tubular support structure (clock-wise numbering starting from upper left figure): (1) ring holding tool for distortion and residual stress control; (2) synchronized orbital automatic welding of two parallel tubes; (3) and (4) high-quality inner and outer surfaces of 11 mm-thick weld; (5) use of laser tracker for distortion control at each of 60 welding passes.
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Fig. 9. Flow chart of divertor manufacturing operations.
on tungsten to produce W–Cu joint. Then W–Cu tiles are brazed to the CuCrZr heat sink using STEMET® 1108 brazing alloy.
- high tolerance welding technology for welding of massive triangular divertor support structure with minimal residual stresses and distortions has been developed (Fig. 8).
6. Welding
7. Non-destructive testing
The main volume of welding is related to welding of closed box-like HSE parts. As an example, welding of bimetallic hypervapotron wall to the box-like steel basement (welding 5 mm in depth) may be considered. Welding of various multiple cover plates in steel support structures (welding 5–11 mm in depth) is another example. To decrease residual stresses (especially in the bimetallic joint, which is only 5 mm far from a weld) and welding deformations (to minimize further machining and to provide tolerances better than 50 m) laser welding was selected, which produces the minimal heat affected zone in the structure. For the first cooling loop of a nuclear facility the above-mentioned welds have to meet such requirements as absence of spattered droplets, full penetration of weld and so on (see also EN ISO 13919.1). Application of laser welding for such loops is rather a new task, and the following results have been achieved by solving this task [9]:
Severe operational conditions, standards of nuclear facilities and requirement of high reliability of PFC necessitate wide application of NDT at every manufacturing step. Ultrasonic (US) control with immersion baths, robotic manipulators for automatic scanning of large-in-area joints and modern technique (phase arrays, for example) are used for detection of joint defects for Be–CuCrZr, CuCrZr–SS and W–Cu–CuCrZr compositions. US control is also planned to be used after welding of relatively thin (∼10 mm) cover plates to massive support structures with specific thicknesses in the range of 100–600 mm. Such “contrast” structures are difficult to control by X-ray methods. The traditional X-ray control along with visual inspection (including endoscopes) and use of dye penetrants will be applied to control HSE welds, which do not have so “contrast” composition. HHF testing similar to that described in the third paragraph (but with a heat load less than the operational ones) may be used to control the quality of joining of armour tiles to HSE. In case of defects we can observe (by IR thermography) hot spots on the surface of heated components and thus reject the most defective HSE before their assembly into large and unrepairable panels. According to the Procurement Agreement the supplier has to perform selective test up to 30% (in the average) of HSE with a higher selection rate at the beginning of manufacturing and a
- full-penetration vacuum tight welds (5–11 mm thick) without pores, undercuts and excess penetrations have been obtained (Fig. 6); - effective prevention/removal of spatters inside cooling channel has been developed (Fig. 7);
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lower rate in later phases. This testing for the divertor components supplied by all Parties will be carried out at the Efremov Institute. 8. General manufacturing flow chart and evaluation of productive capacity
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- technological issues of PFC manufacturing have been, in general, solved, but their further optimization to decrease the manufacturing cost will be continued; - placing of several procurement contracts in one place (i.e. the Efremov Institute) has a positive synergetic effect. References
As an example, Fig. 9 is a flow chart of DD manufacturing operations that are similar to the process of the FW manufacturing. It is seen, that machining is not a major operation, which accounts for about 20% of the total labour input of manufacturing products. The most frequent and critical operations are welding, non-destructive testing, geometry control and ultrasonic cleaning. It is assumed that the bottleneck of the process chain will be hydraulic and vacuum leak tests as the most prolonged and poorly expanded operation. Productive capacity of series production is supposed at the level of one assembly per month for the divertor and three panels per month for the FW. 9. Conclusion The presented review of technological challenges related to manufacturing of ITER PFC in Russia shows the following: - Russian industry has successfully undergone the preliminary qualification procedure and demonstrated satisfactory readiness for supply of PFC for ITER;
[1] M. Merola, F. Escorbiac, R. Raffray, P. Chappuis, T. Hirai, A. Martin, Overview and status of ITER internal components, Fusion Eng. Des. 89 (2014) 890–895. [2] I.V. Mazul, V.A. Belyakov, R.N. Giniyatulin, et al., Preparation to manufacturing of ITER plasma facing components in Russia, Fusion Eng. Des. 86 (2011) 576–579. [3] I. Mazul, A. Alekseev, V. Belyakov, D. Bondarchuk, R. Eaton, F. Escourbiac, A. Gervash, D. Glazunov, V. Kuznetsov, M. Merola, A. Labusov, I. Ovchinnikov, R. Raffray, R. Rulev, Russian development of enhanced heat flux technologies for ITER first wall, Fusion Eng. Des. 87 (2012) 437–442. [4] V. Kuznetsov, A. Gorbenko, V. Davydov, et al., Status of the IDTF high heat flux test facility, Fusion Eng. Des. 89 (2014) 955–959. [5] A. Gervash to be published in Fusion Eng. Des. in 2016 as a proceeding of ICFRM-17 held in AAchen in October 2015. [6] A. Gervash, R. Giniyatulin, I. Mazul, et al., Comparative thermal cyclic testing and strength investigation of different Be/Cu joints, Fusion Eng. Des. 39–40 (1998) 543–549. [7] N. Litunovsky, E. Alekseenko, A. Makhankov, I. Mazul, Development of the armouring technique for ITER Divertor Dome, Fusion Eng. Des. 86 (2011) 1749–1752. [8] N. Litunovsky, E. Alekseenko, V. Kuznetsov, et al., Repair of manufacturing defects in the armour of plasma facing units of the ITER Divertor Dome, Fusion Eng. Des. 88 (2013) 1739–1743. [9] S. Mazaev, T. Gurieva, A. Lapin, et al., Laser welding of plasma facing units for ITER divertor Dome manufacturing, in: IEEE/NPSS Proceedings of 24th Symposium on Fusion Engineering (SOFE), June, 2011, 2011, pp. 1–5, http://dx. doi.org/10.1109/SOFE.2011.6052257.
Please cite this article in press as: I.V. Mazul, et al., Technological challenges at ITER plasma facing components production in Russia, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.030
ARTICLE IN PRESS
FUSION-8470; No. of Pages 7
Fusion Engineering and Design xxx (2016) xxx–xxx
Contents lists available at ScienceDirect
Fusion Engineering and Design journal homepage: www.elsevier.com/locate/fusengdes
Technological challenges at ITER plasma facing components production in Russia I.V. Mazul a,∗ , V.A. Belyakov a , A.A. Gervash a , R.N. Giniyatulin a , T.М. Guryeva a , V.E. Kuznetsov a , A.N. Makhankov a , A.A. Okunev a , O.N. Sevryukov b a b
Efremov Institute, 196641 St. Petersburg, Russia MEPhI, 115409 Moscow, Russia
h i g h l i g h t s • 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.
a r t i c l e
i n f o
Article history: Received 27 August 2015 Received in revised form 13 January 2016 Accepted 19 January 2016 Available online xxx Keywords: First wall Divertor Multilayered structure Joining technologies NDT Mockups testing
a b s t r a c t 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. © 2016 Elsevier B.V. All rights reserved.
1. Introduction In the frames of the contracts signed with the ITER Organization (IO) Russia has to supply, in particular, 179 first wall (FW) panels (40% of full FW coverage), 58 divertor dome (DD) assemblies (100% of these components) and provide high heat flux (HHF) testing of various divertor components supplied by all Parties [1]. By now, manufacturing and testing of medium-sized mockups (semi-prototypes) of the divertor and FW have been, in general, completed, thus the design solutions and manufacturing technologies selected by RF have been qualified and proved [2]. Preparation of working models of full-size prototypes is nearly completed and after manufacturing and testing of these prototypes
∗ Corresponding author. Tel.: +7 9217710056. E-mail address: [email protected] (I.V. Mazul).
in 2018 decision on serial production will be made. So, the range of technologies to be used during manufacturing of ITER plasma facing components (PFC) in Russia has been, in general, defined and their complexity, originality and difficulty are described below. 2. Design features and specific requirements DD and FW are water (∼100 ◦ C, 4 MPa) cooled components subjected by plasma to high surface heat loads (up to 5 MW/m2 for 103 –104 cycles), significant electromagnetic loads and neutron irradiation. Above-mentioned components consist of (see Fig. 1) multilayered heat-sink elements (HSEs), the plasma facing wall of which is made from bronze CuCrZr [3]. This wall is armoured on the plasma facing side by protective tiles made of tungsten W (for the divertor) or beryllium Be (for FW). The rear part of HSE is made from steel SS316L(N)-IG and is thick enough to provide strength and geometrical stability of the structure under electromagnetic
http://dx.doi.org/10.1016/j.fusengdes.2016.01.030 0920-3796/© 2016 Elsevier B.V. All rights reserved.
Please cite this article in press as: I.V. Mazul, et al., Technological challenges at ITER plasma facing components production in Russia, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.030
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Fig. 1. Design of DD assembly (upper left) and FW panel (upper right); FW HSE and its cross-section (bottom left) and parts (bottom right) of FW HSE: steel rear part and bronze wall (overturned).
and thermal loads. Multiple HSE (in the average, ∼40 elements for FW panel and ∼34 elements for the DD assembly) are arranged on larger units (the so-called panels) based on massive steel supports, which serve also as water manifolds. Support structures have units (lugs for divertor pinning and bolts for FW pre-stressed pulling) for attachment to other reactor structures (to divertor cassettes or blanket modules). Panels segmentation on multiple HSE is forced by the necessity to solve the following issues: reduction of electromagnetic loads (decrease in area of induced current loops); decrease in thermal stresses (by minimizing sizes of regions with thermal gradients); reduction of cost/risk of whole panel rejection during manufacturing (a panel consists of a number of pre-tested HSE). Due to deficiency of coolant mass flow rate for FW panels the use of more conventional tubular seamless cooling channels was not possible. For this reason HSE is a bimetallic (CuCrZr–SS) box-like structure and has a complicated profile of plasma facing cooling channel (hypervapotron type), which is able to provide adequate cooling at low flow velocities. Manufacturing of such channel by seamless technology is not practical. As a result of technological optimization each unit length of this channel contains double length of bimetallic SS–CuCrZr joint and double length of steel weld. Totally for all ∼9000 HSE the length of joints achieves about 25 km. For 239 supplied PFC assemblies, each having significant sizes (∼1.5 m) and unit weight (∼1 t) there are about 40 different design options. This means that, in fact, instead of serial production of similar parts we have a vast number of parts different in size and shape. The total number of armour tiles is as many as beryllium (toxic, fragile) and tungsten (very hard and fragile). Moreover, ∼100 000 parts from steel, bronze and bimetal CuCrZr–SS have to be machined with an accuracy better than 50 m. For nuclear machine the nondestructive testing (NDT) of manufactured parts and their joining/welding is strongly required. In particular, we have to provide 100% control (by X-rays or ultrasonically) of vacuum tight SS–SS welds and SS–CuCrZr joints. About 300 m2 of brazing joints (Be–CuCrZr and Wu–Cu–CuCrZr) has to be subjected to ultrasonic and HHF testing. Despite that pressurized water (4 MPa) is used for PFC cooling the tightness of various joints and welds has to
provide the water leakage into the reactor vacuum vessel less than 10−8 Pa m3 /s. 3. Status of supplier qualification Technological capabilities of RF industry had to be demonstrated before signing the Procurement Agreement between RF and IO. Manufacturing of representative medium-scale (0.5–0.6 m long and 0.2–0.3 m wide) mockups and their testing under operational condition are typical methods used to check selected design solution and to qualify the potential supplier. Fig. 2 shows the divertor and FW panel semi-prototypes, designed and manufactured at the Efremov Institute. The most important operational parameters influencing the design, reliability and life time of PFC are surface heat load and number of thermal cycles. These operational parameters were simulated on the Tsefey and IDTF electron beam facilities [4]. The divertor semi-prototype successfully withstood 1000 cycles at 3 MW/m2 , then 1000 cycles at 5 MW/m2 and was damaged only after ∼500 cycles at a double operational load of 10 MW/m2 . Preliminary testing of the FW panel semi-prototype at operational loads of 4.7 MW/m2 for 100 cycles demonstrates good quality of all joints. Further more comprehensive testing of this semi-prototype is under way. 4. Machining of parts and required tolerances Machining of DD elements involves the following works: machining of HSE basements and bronze walls, cutting of W/Cu tiles, bending of tubes, machining of the manifolds and steel support structure. The most complicated technological problems are machining of deep slots of the manifolds and stubs of the steel support structure (Fig. 3). The highest accuracy requirements are imposed on the surfaces under laser welding to ensure the required quality of welded joints and elements fixing the divertor to cassette. Manufacturing precision of the above elements is within ±0.05 mm. Advanced 5-axis
Please cite this article in press as: I.V. Mazul, et al., Technological challenges at ITER plasma facing components production in Russia, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.030
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Fig. 2. Qualification semi-prototypes of the divertor (left) and FW panel (right).
machines with a positioning accuracy of less than 0.005 mm are required to perform the most complicated work.
5. Joining of dissimilar materials 5.1. CuCrZr–SS316 joining When developing fabrication process for ITER PFC, special attention shall be paid to the production of reliable joints between heat sink materials made of CuCrZr bronze and support structure made of austenitic steel. Two different SS/CuCrZr joining techniques have been proposed and tried in Russia: hot isostatic pressing (HIP) and explosion bonding [5]. Both of mentioned joining techniques were comparatively studied to emphasize their advantage or disadvantage. Mechanical properties of SS/CuCrZr joint, its vacuum tightness and structural parameters (uniformity of elements distribution, grain size of copper alloy and hardness of joint area) were selected for the joint characterization. Studies of the SS/CuCrZr diffusion zone after HIP (980 ◦ C, 2 h, 150 MPa) have shown that both techniques provide a continuous, defect-free zone structure. Analysis of the width of the diffusion zone shows that the overall width of the diffusion zone is about 110 m for HIP joint and order of 25 m for exploded one. For both technologies, the grain size of bronze is not greater than 200 m, which meets the ITER requirements (Fig. 4). Investigations of the SS/CuCrZr mechanical properties after manufacturing
cycle showed that both methods of forming steel/bronze joint allow achieving the ultimate strength more than 350 MPa at room temperature. To assess a vacuum tightness of the joint the number of samples of each technology were manufactured and successfully passed through the repeated internal pressure test (100 cycles at 7 MPa, 30 min). To validate SS/CuCrZr heat sink performance the number of actively cooled SS/CuCrZr mock-ups were subjected to high heat flux expected for such components. The results confirm the efficiency of both joint technologies with the expected values of the heat flux. Summarizing the results of comparative study one can conclude that both of considered joint methods give satisfied properties for HHF components application.
5.2. Be–CuCrZr brazing Proposing beryllium as plasma facing armour for the ITER FW and fast brazing as Be/CuCrZr joining technique [6], three options of fast heating were considered in Russia: electron beam heating, a direct ohmic heating and induction heating. When brazing Be tiles onto CuCrZr heat-sink it is necessary to prevent forming of brittle intermetallic Be/CuCrZr layer and to preserve the strength of bronze at the required level (of the order of 300 MPa at room temperature). Thus it is necessary to minimize the time taken at temperatures above 450–500 ◦ C.
Fig. 3. Machining of deep slots (width—25 mm, depth—285 mm) of the manifolds (upper line) and machining of stubs of the steel support structure (bottom line).
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Fig. 4. Microstructure of SS–CuCrZr joint produced by explosion bonding (upper left) and HIP (upper right) and grain size of CuCrZr after explosion bonding (bottom left) and HIP (bottom right) manufacturing cycles.
Fig. 5. Inductive (55 kHz, 20 kW) heating/brazing of Be tiles to 0.6 m long FW HSE. (Several thermocouples T1–T8 on HSE lateral walls in several positions along HSE length demonstrate sufficient uniformity of heating).
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Fig. 6. Example of welding of two cover plates (in the centre) 11 mm thick and metallographic cross-section of welds cut in several control points (at the periphery).
Fig. 7. Check of inner walls of box-like HSE by videoscope (left). Inner view: with spatters after conventional welding (centre) and after of proper technology (right).
Heating by electron beam leads to significant overheating of beryllium tiles. Ohmic heating requires extended bronze parts to be cut after brazing. The inductive heating method does not have the above disadvantage (Fig. 5). Inductive heating allows one to fulfil the fast heating requirements: pre-heating to 450 ◦ C for 60–90 min, fast heating from 450 to 710 ◦ C for 3–5 min. The temperature uniformity through the brazing zone is ±20 ◦ C. Brazing is
performed in a vacuum chamber at a residual pressure not exceeding 4 × 10−5 mbar. 5.3. W/Cu–CuCrZr brazing Joining technologies for W–Cu–CuCrZr composition are well described elsewhere [7,8]. We use vacuum casting of pure copper
Fig. 8. Welding of DD triangular tubular support structure (clock-wise numbering starting from upper left figure): (1) ring holding tool for distortion and residual stress control; (2) synchronized orbital automatic welding of two parallel tubes; (3) and (4) high-quality inner and outer surfaces of 11 mm-thick weld; (5) use of laser tracker for distortion control at each of 60 welding passes.
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Fig. 9. Flow chart of divertor manufacturing operations.
on tungsten to produce W–Cu joint. Then W–Cu tiles are brazed to the CuCrZr heat sink using STEMET® 1108 brazing alloy.
- high tolerance welding technology for welding of massive triangular divertor support structure with minimal residual stresses and distortions has been developed (Fig. 8).
6. Welding
7. Non-destructive testing
The main volume of welding is related to welding of closed box-like HSE parts. As an example, welding of bimetallic hypervapotron wall to the box-like steel basement (welding 5 mm in depth) may be considered. Welding of various multiple cover plates in steel support structures (welding 5–11 mm in depth) is another example. To decrease residual stresses (especially in the bimetallic joint, which is only 5 mm far from a weld) and welding deformations (to minimize further machining and to provide tolerances better than 50 m) laser welding was selected, which produces the minimal heat affected zone in the structure. For the first cooling loop of a nuclear facility the above-mentioned welds have to meet such requirements as absence of spattered droplets, full penetration of weld and so on (see also EN ISO 13919.1). Application of laser welding for such loops is rather a new task, and the following results have been achieved by solving this task [9]:
Severe operational conditions, standards of nuclear facilities and requirement of high reliability of PFC necessitate wide application of NDT at every manufacturing step. Ultrasonic (US) control with immersion baths, robotic manipulators for automatic scanning of large-in-area joints and modern technique (phase arrays, for example) are used for detection of joint defects for Be–CuCrZr, CuCrZr–SS and W–Cu–CuCrZr compositions. US control is also planned to be used after welding of relatively thin (∼10 mm) cover plates to massive support structures with specific thicknesses in the range of 100–600 mm. Such “contrast” structures are difficult to control by X-ray methods. The traditional X-ray control along with visual inspection (including endoscopes) and use of dye penetrants will be applied to control HSE welds, which do not have so “contrast” composition. HHF testing similar to that described in the third paragraph (but with a heat load less than the operational ones) may be used to control the quality of joining of armour tiles to HSE. In case of defects we can observe (by IR thermography) hot spots on the surface of heated components and thus reject the most defective HSE before their assembly into large and unrepairable panels. According to the Procurement Agreement the supplier has to perform selective test up to 30% (in the average) of HSE with a higher selection rate at the beginning of manufacturing and a
- full-penetration vacuum tight welds (5–11 mm thick) without pores, undercuts and excess penetrations have been obtained (Fig. 6); - effective prevention/removal of spatters inside cooling channel has been developed (Fig. 7);
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lower rate in later phases. This testing for the divertor components supplied by all Parties will be carried out at the Efremov Institute. 8. General manufacturing flow chart and evaluation of productive capacity
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- technological issues of PFC manufacturing have been, in general, solved, but their further optimization to decrease the manufacturing cost will be continued; - placing of several procurement contracts in one place (i.e. the Efremov Institute) has a positive synergetic effect. References
As an example, Fig. 9 is a flow chart of DD manufacturing operations that are similar to the process of the FW manufacturing. It is seen, that machining is not a major operation, which accounts for about 20% of the total labour input of manufacturing products. The most frequent and critical operations are welding, non-destructive testing, geometry control and ultrasonic cleaning. It is assumed that the bottleneck of the process chain will be hydraulic and vacuum leak tests as the most prolonged and poorly expanded operation. Productive capacity of series production is supposed at the level of one assembly per month for the divertor and three panels per month for the FW. 9. Conclusion The presented review of technological challenges related to manufacturing of ITER PFC in Russia shows the following: - Russian industry has successfully undergone the preliminary qualification procedure and demonstrated satisfactory readiness for supply of PFC for ITER;
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Please cite this article in press as: I.V. Mazul, et al., Technological challenges at ITER plasma facing components production in Russia, Fusion Eng. Des. (2016), http://dx.doi.org/10.1016/j.fusengdes.2016.01.030