Japanese activities in ITER transitional arrangements

Fusion Engineering and Design 81 (2006) 69–77

Japanese activities in ITER transitional arrangements M. Mori ∗ , ITER Japanese Participant Team Japan Atomic Energy Research Institute, 801-1 Mukouyama, Naka, Ibaraki 311-0193, Japan Received 12 April 2005; received in revised form 22 September 2005; accepted 22 September 2005 Available online 18 January 2006

Abstract The ITER transitional arrangements (ITA) is conducted by the International Team (IT) with supports from participant teams (PTs). The Japanese Participant Team (JA-PT) has contributed to the ITA by sharing a lot of technical tasks to verify feasibilities of fabrication and quality control method in procuring ITER equipments and facilities. For examples, trial fabrications of Nb3 Sn strands have been performed by the JA-PT with four potential suppliers, and it has already been confirmed that one of the strands produced by a certain supplier meet the ITER requirements. All strands produced by other suppliers will be fully qualified by the end of 2005. The trial fabrications of jackets for the central solenoid (CS), structural material for toroidal field coil, the partial mock-ups of the vacuum vessel (VV) and the shield blanket module are also ongoing to verify the feasibility of fabrications and quality inspections. Furthermore, the JA-PT has made several technical developments on neutral beam (NB) system to improve reliability in high voltage (1 MV) insulation and in stable high current density operations over a long period. Design of equatorial EC launcher and some techniques towards reliable steady-state EC operation have been developed. The paper also describes other activities to confirm validity of the present design. © 2005 Elsevier B.V. All rights reserved. Keywords: Fusion; ITER; Tokamak; Magnet; Vacuum vessel; Blanket

1. Introduction The ITER negotiation for preparing the ITER Joint Implementation Agreements began in 2001 after completion of the ITER Engineering Design Activities (EDA). The Coordinated Technical Activities (CTA) were jointly carried out by Japan, Canada, the European Union (EU) and the Russian Federation (Rus∗ Corresponding author. Tel.: +89 29 270 7363; fax: +89 29 270 7468. E-mail address: [email protected].

0920-3796/$ – see front matter © 2005 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2005.09.075

sia) to make technical supports for the negotiation from July 2001 to December 2002. Now, the six parties participating the ITER project, i.e. The People’s Republic of China, EU, Japan, the Republic of Korea, Russia and the United States of America, are jointly making technical preparations for efficient start of ITER joint construction since January 2003 under the ITER transitional arrangements (ITA). The technical activities in the ITA include preparation for procurement documents of ITER facilities and equipments, site-specific design adaptation for a preferred site, preparation for licensing of ITER including neces-

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sary safety analyses, and they are conducted by the International Team (IT) with supports from participant teams (PTs). The IT is organized by scientists and engineers sent from the parties, and each party establishes its own PT. Japan sends ∼20 staffs to the IT in the ITA. The JAERI, which is designated as the Japanese implementing institute for the ITA by Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japanese Government, set up a Japanese Participant Team (JA-PT), and the JA-PT has contributed to the technical activities in the ITA mostly by undertaking technical tasks under the task agreements concluded between the IT and the PT. Up to the present, ITER site has not been decided yet. Taking the situation into account, the JA-PT has undertaken tasks for design improvements and optimization, providing technical basis for narrowing design options down, or qualification of manufacturing and quality control methods, giving priority to site-independently useful ones. Besides the framework of the ITA tasks, there are some ITER-related Japanese activities carried on by experts from many Japanese institutes and universities: developments of ITER test blanket modules, contribution to the International Tokamak Physics Activity promoted aiming at cooperation in development of the physics basis for burning tokamak plasmas including ITER, supports for licensing preparation of Japanese government, and so on. This paper presents the Japanese contributions made under the ITA task agreements.

2. Superconducting magnet

Table 1 ITER requirements on TF Nb3 Sn strand Diameter Cu area/non-Cu area Cr plate

0.82 mm 1.0 2.0 + 1, −0 ␮m

Jc (at 12 T, 4.2 K)

>700 A/mm2 , bronze >800 A/mm2 , internal tin

Hysteresis loss

<1000 mJ/cm3 for ±3 T cycle

bronze process and the other supplier produces them by using an internal tin process. The specification of critical current density (Jc ) for the strand produced by bronze process has increased from the value in the EDA design, 650 A/mm2 , by taking account of both recent progress in strand performance and the evaluation of ITER Model Coil results [1,2]. Effects of three major parameters, Sn contents in bronze, filament diameter and barrier material, on superconducting performance were studied by using strands produced from small billets. Sn content in bronze was increased up to 15–16% from 14.5% in the Model Coil strands and a filament diameter was reduced to about 3 ␮m from 4 ␮m. A barrier material was also changed from Ta to Nb or combination of Nb and Ta to improve production yield, although hysteresis loss tended to increase with introducing Nb. Table 2 shows the measurement results of superconducting performance of six candidate strands for mass production produced by one of the suppliers. In case that filament diameter was 3.4 ␮m (No. 6), all measured vales meet the ITER requirements even with Nb barrier. Consequently, strand layout of No. 6 shown in Fig. 1 was adopted for a mass production strand, and a demonstration of productivity of the strand with larger

Activities to study and demonstrate fabrication feasibility at industrial level are in progress on the trial fabrication of strand, structure for toroidal filed (TF) coil and conductor jacket sections for the central solenoid (CS). 2.1. Trial fabrication of ITER TF Nb3 Sn strand In order to prepare for mass production of Nb3 Sn strands which meet ITER requirements shown in Table 1, trial fabrications of Nb3 Sn strands have been performed by JAERI with four potential suppliers. Three suppliers produce Nb3 Sn strands by using a

Fig. 1. Cross-section of a new strand for mass production.

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Table 2 Measured superconducting performance of six candidate strands in one of the suppliers (bronze process) Number

Filament diameter (␮m)

Jc (A/mm2 )

Barrier

Hysteresis loss (mJ/cm3 )

1 2 3 4 5 6

2.3 2.8 3.4 2.3 2.8 3.4

760 768 726 743 754 750

Nb/Ta Nb/Ta Nb/Ta Nb Nb Nb

434 398 310 1398 1028 870

billet size (>50 kg) is in progress. The other suppliers have also performed similar tests and demonstrations for mass production by using 50–100 kg billets. All strands produced in the trial fabrication will be fully qualified by the end of 2005 and we will be ready for procurement of TF Nb3 Sn strand with a reasonable cost. 2.2. Trial fabrication of materials for TF structure High strength and toughness stainless steels, JJ1 [3] and nitrogen strengthened 316LN (referred to as ST316LN), will be used in TF structure. Especially, a large forging block over 350 mm thick is necessary in the inboard region of TF coil case. Trial fabrication of JJ1 and ST316LN large forgings having the same crosssection as an actual TF coil case has been carried out to demonstrate manufacturing process and high cryogenic mechanical properties of large forgings. Large ingots of JJ1 and ST316LN (about 20 and 30 tonnes, respectively) were produced and have been forged to rectangular blocks (∼3.4 m (length) × ∼0.95 m (width) × ∼0.4 m (thickness) for JJ1 and ∼6.3 m (length) × ∼0.95 m (width) × ∼0.4 m (thickness) for ST316LN). Fig. 2 shows an overview

Fig. 2. Overview of a ST316LN block at an intermediate stage of forging processes.

of a ST316LN block at an intermediate stage of several forging processes. These blocks will be machined to the shape of the inner leg of TF coil case after ultrasonic testing by the middle of 2005. It is also indispensable to establish database of cryogenic mechanical properties of structural materials to assure mechanical integrity of the ITER magnet system. JAERI has developed the design fatigue S–N curve (stress versus number of cycles) at 4 K through tension–compression fatigue tests, and we have so far demonstrated that JJ1 has enough fatigue properties to withstand planned cyclic loads [4]. JA-PT activities for establishing the database will be continued by using specimens taken form the large forgings mentioned above. 2.3. Trial fabrication of CS jacket A stainless steel having a low coefficient of thermal expansion is required as a CS jacket material to obtain compressive force on winding packs by the cooldown (form room temperature to 4K). From this point of view, a new material, JK2LB [5] was selected for the CS jacket. Mechanical properties of several JK2LB materials having slightly different chemical compositions were evaluated as a function of main elements, such as C, N and B. The compositions of N (0.17–0.23%), C (<0.03%) and B (0.001–0.004%) were chosen from this study to achieve yield strength of more than 1000 MPa and fracture toughness of more than 130 MPa m0.5 at 4 K after Nb3 Sn reaction heat treatment (650 ◦ C × 240 h). Trial fabrication of a circle-in-square JK2LB tube with the processes of hot extrusion followed by cold drawing was also performed to confirm achievable accuracy of dimension and available unit length of the jacket, in parallel with the optimization of chemical

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Fig. 3. CS circle-in-square jacket produced in the trial fabrication.

compositions in JK2LB. The jacket has an outer dimension of 51.3 mm and an inner diameter of 35.1 mm with target tolerance of ±0.25 mm. The produced jacket is shown in Fig. 3. It was confirmed that target tolerance of ±0.25 mm and unit length of more than 7 m could be guaranteed as specifications. At the next step, new JK2LB billets with optimized chemical compositions having a diameter of 163 mm and total weight of 1.6 tonnes, were produced by using conventional electric furnace followed by electro-slug remelting (ESR) process. A prototype of the CS jacket using this billet will be produced by the end of March 2005.

3. Vacuum vessel 3.1. Structural analysis of VV support structure with multiple flexible plates ITER vacuum vessel (VV) is a safety-related component containing radioactive materials, such as tritium and activated dust. Since the VV supports was directly connected to the TF coils in the EDA design [6], structural integrity under anticipated conditions must be assured in consideration of the strong linkage of the VV and the TF coils. JA-PT proposed an alternative VV support concept to remove the unfavorable direct linkage as shown in Fig. 4, where the VV support system with multiple flexible plates located at the bottom of VV lower port is connected to the cryostat ring instead of the TF coil. The proposed concept has two advantages comparing to that of the EDA design:

Fig. 4. Alternative design of support structure for ITER vacuum vessel.

(1) the TF coil can be categorized as a non-safetyrelated component because the TF coils are mechanically separated from the VV supports and no direct contact of the coils with the VV would be expected under anticipated loads; (2) the thermal load of TF coil due to heat conduction from VV is reduced. The following results obtained in stress analyses of the proposed VV support structure shows the feasibility of the proposed structure as an improved VV support concept [7]. • The maximum value of the relative displacement between VV and TF coils is found to be 15 mm, which is much less than 100 mm shown in the EDA concept. • The maximum primary stress of the flexible plates is estimated to be 200 MPa and lower than the allowable value of 264 MPa defined in ASME Section III NF, under the off-normal load condition. The maximum stress of the connection bolt was also estimated to be lower than the allowable value.

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whether distortions caused by welding are within tolerances. The basic data necessary for preparing the technical specifications of the VV procurement, such as weld distortions, minimum space required for welding and inspection, welding and inspection accessibilities to the weld joints in the present design and applicability of weld inspections, will be obtained through the fabrication of the partial mockups.

4. Blanket and divertor 4.1. Shielding blanket module

Fig. 5. Partial VV mock-ups.

3.2. Fabrication of VV partial mock-ups In the present design of the VV, a large number of interface structures, such as keys and housings to support the blanket modules on the VV, have been introduced as shown in Fig. 5 instead of the backplates in the 1998 ITER design, in order to reduce the procurement cost [6]. As a result, the number of weld joints in the VV has increased and the distance between weld joints becomes much shorter than that in the previous design, for which fabrication feasibility has already been demonstrated in the EDA R&D of VV [8]. The weld joints of housings for blanket support are very close to the reinforced ribs connected between inner and outer walls of VV especially in the inboard cylindrical region. Furthermore, accessibilities of welding and weld inspection are extremely restricted by the highly curved inner and outer walls in the top curved region. The critical issues in the present VV design are therefore to confirm feasibilities of fabrication and quality inspection even with the expected large welding distortion and the low accessibility for welding and weld inspection. JA-PT is now fabricating a partial VV mock-up (shown in Fig. 5), which consists of an inboard cylindrical region and a top curved region, in order to confirm the fabrication methods and Non-Destructive Testing (NDT) methods and in order to examine

The JA-PT has performed three-dimensional electro-magnetic (EM) analyses to evaluated EM force acting on the shielding blanket modules at plasma disruptions with linear current decay time of 40 ms, based on disruption scenarios renewed by the ITER International Team instead of previous scenarios in the EDA design with current decay time of 27 ms. In the preliminary analysis, the stress of some modules was shown to be almost the design allowable limit. Then, structure designs of such modules have been modified in order to keep the stress below the design allowable limit; slots of the inboard module has been deepened to reduce eddy current, and the key structure has been strengthened for the module located at the top of the vacuum vessel. It has been confirmed that all EM stress of the improved modules is kept below the design allowable limit. It has been also shown that maximum EM force appears at the modules located near the divertor region. Furthermore, the dynamic stress on the intermodular key and the stub key have been analyzed for blanket modules located near the divertor region taking into account of possible variation of initial positioning of the key and the key groove, damping effects and waveforms of EM load and thermal load. The results showed that the evaluated stress was below the design allowable limit. In addition to improvements of module design, the JA-PT also started qualification of the first wall panel fabrication and verification of inspection procedure. Fabrication of first wall panel mock-up made of Be, CuCrZr and SS316L with a real width is progressing.

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4.2. A large-scale divertor mock-up with an annular flow To accommodate reduction of the space assigned to the divertor in ITER, JAERI has proposed a divertor cooling structure with an annular flow concept. In this concept, the inlet and outlet of coolant are both located at the same side of the plasma facing component. At the other end, an end plug for hairpin return of the coolant is employed to eliminate the complex coolant manifold. In addition, using a larger armor tile (width = 33 mm), the number of the components can be reduced to twothirds of that in the reference design. To demonstrate the thermal and mechanical performance of the annular flow concept, the JA-PT has fabricated a divertor mockup and carried out a high heat flux test. The mock-up has successfully withstood 3000 thermal cycles at a heat flux of 10 MW/m2 and 1000 thermal cycles at 20 MW/m2 , which are corresponding to the steady state and transient heat load conditions of the ITER divertor. No degradation of the thermal performance of the mock-up has been observed in these thermal cycling tests. These results show that the divertor with the annular flow concept can be one of the promising options for the ITER divertor.

5. Heating and current drive systems 5.1. NB system A neutral beam (NB) system in ITER is designed to inject 33 MW of D0 beams at the beam energy of 1 MeV using two NB injectors. To fulfill the requirement, the ITER beam source (ion source and accelerator) has to generate 1 MeV, 40 A (ion current density: 200 A/m2 ) of D− ion beams. Since ion beams of such high energy and current have never been generated, JAPT has started R&D aiming at demonstration of 1 MeV H− ion beam acceleration up to around 1 A, which is limited by JAERI’s power source. Since the insulation gas used in conventional accelerators brings about radiation-induced conductivity (RIC) in the 1 MV ITER accelerator [9], vacuum insulation has been adopted in the end of EDA instead of gas insulation in a previous design. A vigorous R&D has been carried out after EDA to improve the reliability in the vacuum insulation technology of 1 MV

Fig. 6. The JAERI MeV accelerator.

high voltage. The accelerator as shown in Fig. 6 has been developed at JAERI. Features of the accelerator include: (1) the ion source is directly mounted on the accelerator. And hence, pressure in the accelerator is in the range of 0.05–0.2 Pa during the operation [10]. (2) The fiber-reinforced plastic (FRP) insulator stack, as a vacuum boundary, surrounds the accelerator with a vacuum gap of 50 mm wide between the accelerator and the insulator stack. This vacuum gap allows direct line of sight from −1 MV to the ground (in distance ∼1.8 m). Hence the accelerator was designed so as to fulfill insulation criteria for both glow (gas) discharges and vacuum arc discharges [11]. (3) The electric field concentration at the triple junction was reduced to 1.2 kV/mm by a large stress ring [12] to avoid surface flashover along the insulator. It has been demonstrated that the accelerator stably withstands up to 1 MV without breakdown by using these vacuum insulation techniques mentioned above. Following the demonstration of the stable insulation at 1 MV, beam acceleration test has been intensively carried out to increase the beam current density. Closed circles in Fig. 7 shows progress in accelerated H− ion current density at the beam energy around 1 MeV. According to source tuning under Cs (cesium) seeded condition and high power operations, the beam current density increased step by step. Typical beam parameters [13] obtained for both with and without Cs are as

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Fig. 7. Progress of accelerated H− ion current density, as a function of the beam energy.

follows: • 1 MeV, 70 mA (18 A/m2 ) for 1.0 s (without Cs); • 900 keV, 110 mA (80 A/m2 ) for 0.5 s (with Cs); • 800 keV, 140 mA (100 A/m2 ) for 0.5 s (with Cs). At present, the beam power and the pulse length were limited by heat handling capability of the beam dump, which is to be replaced so as to receive high heat load of 160 MW/m2 (normal to the beam incident) for 1 MeV, 200 A/m2 H− ion beams. Since the ability of the KAMABOKO source itself has already been demonstrated in H− ion production of 300 A/m2 (at the beam energy of 50 keV) [14], the current density of ITER requirement (200 A/m2 ), together with ampere level H− ion beam current, is to be achieved soon by further tuning of source operation conditions, such as Cs seeding. Confirming simultaneous achievement of required beam current and beam energy will be developed during the ITER construction phase.

By integrating the following breakthroughs in high power gyrotron technology, i.e. energy recovery by a depressed collector [15], 1 MW oscillation with a higher order mode cavity and application of a synthetic diamond window [16], high power operation of 1.3 MW (1 ms) was demonstrated by JA-PT. However, the extension of the pulse duration was limited by outgassing from an inner component made of stainless steel due to the absorption of stray radiation. By Cu coating on the stainless steel component, the outgassing has been significantly reduced and a quasi-steady-state oscillation of 100 s has been demonstrated at the power of ∼0.5 MW [17], where the local temperatures of major components of the gyrotron have reached steady state. JA-PT is now improving a built-in mode converter to suppress the stray radiation. In addition, a constant current control of the electron beam will be applied to obtain a stable oscillation. With these improvements, the pulse extension test will be carried out aiming at 1MW/CW operation. 5.2.2. Equatorial EC launcher JA-PT has contributed to refining the ITER equatorial launcher design. The launcher consists of the front shield and the launcher plug as shown in Fig. 8. The former has three horizontal slots for RF beam injection and is segmented into 14 shield modules. The latter consists of three sets of steering mirrors, waveguides, miter bends for each waveguide-runs, diamond windows, the nuclear shields and so on. In order to confirm durability of the launcher components and improve the reliability of launcher design, JA-PT carried out several tests of the launcher components. Smooth rotation of irradiated bearings for the steering mirror was confirmed

5.2. RF heating system 5.2.1. Gyrotron ITER requires 170 GHz high power gyrotron systems with a total power of 24 MW, for electron cyclotron heating (ECH), electron cyclotron current drive (ECCD) and suppression of plasma instabilities. Intensive development of a 1 MW–170 GHz gyrotron is carried out to satisfy the requirements of the ITER.

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Fig. 8. ITER equatorial launcher.

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in neutron irradiation tests with a total fast neutron fluence of 1024 n/m2 (neutron energy > 1.0 MeV), which corresponds to the 10 years neutron irradiation on the bearing position in ITER [18]. In neutron irradiation tests of diamond specimen for the windows, no degradation of thermal conductivity was observed with the neutron fluence up to 1020 n/m2 , which is estimated to be much higher than the ITER lifetime fluence at the window position, while it degraded gradually over 1020 n/m2 . No change on the dielectric properties and the bending strength were observed up to 1021 n/m2 [18]. In fatigue stress tests of the flexible cooling tubes for the mirror, no water leakage was observed with more than 106 fatigue cycles [19]. 6. Tritium systems and other analyses The JA-PT has taken data to confirm the performance of designed tritium system. (1) Oxidation efficiency in a detritiation system has been tested to confirm the sufficient efficiency even in unusual conditions such as fires, where CO and CO2 coexist with air. It was verified that the value of normal oxidation efficiency in the tested detritiation system, which is >99.99% and >99.9% for H2 (2%) and CH4 (2%) in the air, respectively, was maintained even at the concentration of CO up to 10% and CO2 up to 20% [20]. (2) Tritium permeation through the plasma facing components has been re-evaluated based on newly obtained experimental data [21]. Results indicate that tritium permeation is mainly attributed through the first wall region because of its large surface and tritium generation in the Be armor (9 Be(n,t)7 Li). The amount of tritium permeation to the secondary cooling water through the heat exchanger wall was estimated to be negligibly small (∼0.4 GBq for 20 years) [21]. (3) A series of irradiation tests has been carried out to confirm the durability of polymer electrolyte film (NafionN117) used in the electrolysis cell of ITER water detritiation system. The cell is designed to be usable for 2 years at a tritium concentration of 9.25 TBq/kg (530 kGy). No serious change was observed in tensile strength and ion exchange capacity of the film up to 850 kGy of gamma irradiation.

In order to evaluate the detailed heat deposition distribution of the ITER NBI duct wall, which will be applied to the input data of the thermo-mechanical design, three-dimensional neutron and photon transport calculations have been performed using Monte Carlo code MCNP-4C with Fusion Evaluated Nuclear Data Library FENDL-2. The estimated results on nuclear heating rate and surface heat flux have been delivered to the IT in order to confirm the feasibility of the present design from a thermo-mechanical view point. Optimized plasma start-up scenario was also developed to reduce the heat deposition to the blanket first wall, and has been proposed to the IT. The JA-PT has also made technical supports for Japanese proposal of Rokkasho as ITER site. For instance, the JA-PT has prepared technical basis of the seismic isolation with well-experienced rubber bearings, which was proposed in the Japanese site proposal. Sufficient margin to the elastic design limit has been confirmed even in an earthquake larger than the design basis, by testing full- and sub-scale ITER bearings.

7. Summary The JAERI, which is designated as the Japanese implementing institute for the ITA by MEXT, set up the JA-PT, and the JA-PT has contributed to the technical preparations for ITER joint construction and operation by undertaking technical tasks under the ITA. Giving priority to site-independently useful tasks before the ITER site decision, the JA-PT has undertaken tasks on: • verification tests of fabrications and quality control methods for superconducting magnet, vacuum vessel and shield blanket module; • technical developments on NB and EC systems to improve reliability in stable and long pulse operation; • establishing technical data to confirm validity of the design (for instance, on magnet, divertor cooling structure, heating and current drive systems, tritium system and heat deposition distribution on NB duct wall); • making operation scenario in detail; • developments of design in detail including supporting analysis and tests.

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Completing these technical preparations and also design adaptations to the selected ITER site will make it possible to finalize the specification documents for ITER procurements.

Acknowledgements The author should like to express grateful thanks to H. Nakajima, K. Hamada, Y. Nunoya, K. Shibanuma, M. Enoeda, S. Suzuki, K. Sakamoto, T. Inoue, K. Takahashi, T. Hayashi and S. Sato who assisted the author in writing this over view paper.

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