Progress on Design and R&D for ITER Diagnostic Systems in Japan Domestic Agency

Transcription

1 1 ITR/P5-35 Progress on Design and R&D for ITER Diagnostic Systems in Japan Domestic Agency Y. Kawano 1, Y. Kusama 1, T. Kondoh 1, T. Hatae 1, K. Sato 1, H. Ogawa 1, M. Ishikawa 1, T. Sugie 1, E. Yatsuka 1, R. Imazawa 1, M. Takeuchi 1, T. Shimada 1, T. Hayashi 1, T. Yamamoto 1, T. Ono 1, Y. Seki 1, S. Suzuki 1, L. Bertalot 2, G. Vayakis 2, C. Watts 2, R. Barnsley 2, P. Andrew 2, R. Reichle 2, A. Encheva 2, C. Walker 2, S. Pitcher 2, V.S. Udintsev 2, M. Walsh 2 1 Japan Atomic Energy Agency, Mukoyama, Naka, Ibaraki, , Japan 2 ITER Organization, Route de Vinon sur Verdon, Saint Paul Lez Durance, France contact of main author: kawano.yasunori@jaea.go.jp Abstract. Japan Domestic Agency has been conducting the design and R&D for six ITER diagnostic systems that Japan Domestic Agency is responsible for. In summary: For the Microfission Chambers, the prototyping of the vacuum tight triaxial connector has indicated that it could be used in the ITER environment, and a neutron transport analysis has shown that the cooling water pipe in the blanket module should be filled with water for in-situ calibration of the diagnostic. For the Edge Thomson Scattering System, the target performance of the prototype YAG laser system, i.e. 5 J of output energy and 100 Hz repetition rate, has been successfully achieved with world highest performance of 7.66 J and 100 Hz. Stray light level can be reduced by a factor of 10 by a new beam dump design with chevron shaped fins. A novel in-situ calibration method has been proposed based on the detection of bremsstrahlung emissions. For the Poloidal Polarimeter, a new concept of robust mirror modules in the port plugs has been developed, in which the plasma facing mirrors and mirror support structure are unified. Optical analyses of the retro reflectors, which are deformed by nuclear and radiation loads, have indicated that about 50%~70% of the laser power can be returned to the diagnostic hall. An optical analysis has shown that the change in polarization state of the probing laser beam caused by thermal deformation of the first mirrors and retro reflector are smaller than target values. For the Divertor Impurity Influx Monitor, the detected signal would increase by a factor of 10 using a new design of equatorial port optics with components which are tolerant to gamma-ray irradiation. The bidirectional reflectivity distribution function of the tungsten block has been measured for the first time to study the surface reflection effect. Light images obtained by optical simulation have qualitatively agreed with experiment. For the Divertor IR-Thermography, design of optical systems at an equatorial port plug has progressed. Finally, for the Divertor Thermocouples for Outer Target, an R&D result indicated that a metal foil could be bonded to the divertor target as an attachment point for the thermocouples. 1. Introduction Japan Domestic Agency (JADA) has been steadily conducting the design and R&D for six ITER diagnostic systems [1,2] as followings; 1) Microfission chambers, 2) Edge Thomson Scattering System, 3) Poloidal Polarimeter, 4) Divertor Impurity Influx Monitor, 5) Divertor IR Thermography, and 6) Divertor Themocouples for Outer Target. Recent highlighted results from the design and R&D activities are presented. 2. Progress on Design and R&D for ITER Diagnostic Systems in JADA 2.1. Microfission Chambers The Microfission Chambers (MFCs), pencil size gas counters containing the fissile material (uranium-235, 235 U), will measure the total neutron source strength for direct evaluation of the fusion power of ITER [3,4]. They will be installed behind Blanket Modules (BMs) in the

2 2 ITR/P5-35 vacuum vessel (VV) at upper and lower outboard positions with two toroidal locations (four positions in total). Figure 1 shows the configuration of a MFC [3]. As shown in FIG.1(a), uranium oxide ( 235 UO 2, a total amount of 235 U is 10 mg) is coated to cover the outer cylindrical electrode of the MFC with active length of 76 mm. The inside of MFC is filled with argon (Ar) gas (ionization gas) at a pressure of 14.6 atm. Figure 1(b) shows a drawing of the MFC designed to satisfy the requirements of the VV environment of ITER. There, the MFC is enclosed in a case made of stainless steel (SS) so that Ar gas flow into VV can be avoided in case of a gas leakage. An exhaust pipe is used to detect a gas leakage and to vent the gas to outside the VV. From the MFC, a triaxial mineral insulated (MI) cable will carry signals outside the VV. Figure 1(c) shows the structure of the MI cable which consists of the core, inner sheath, outer sheath and electric insulator (silica, SiO 2 ). Inside of the inner sheath is filled by Ar gas to avoid discharge between the core and the inner sheath. (A) (a) (B)(b) (C)(c) Outer skin (Cu) Insulator (SiO 2 ) Ionizing gas (Argo ) Inner sheath (SUS) (SS) Inner sheath (Cu) Core(Cu) Insulator (SiO 2 ) + Argon gas FIG.1. Schematic views of; (a) typical MFC, (2) MFC designed to avoid Ar gas flow into VV, and (3) triaxial MI cable. (a) (b) Triaxial MI cable Connector Based on the machine assembling planning of ITER, MFCs will be installed at the VV after the first plasma and MI cables will be installed before the first plasma. This requires a new method of in situ connection of two MI cables in a limited space while keeping enough vacuum tightness. For this purpose, a prototype of a vacuum tight triaxial connector of the MI cable has been designed and fabricated as shown in FIG.2 [3]. A process for FIG.2. Schematic views of the connector for triaxial MI cable; (a) external envelope, (b) inner structure, (c) inner structure around connection point. connecting MI cables are 1) two connecting pipes for inner sheath and outer sheath are put through by one MI cable, 2) the core is connected by resistance spot welding using nickel (Ni) foil, 3) ceramic shields are inserted between the core and the inner sheath, 4) inner connecting pipe is slid to the connection point and jointed, 5) the ceramic shield are inserted between the inner sheath and the outer sheath, 6) the outer connecting pipe is slid to the connection point and jointed. Various performance tests of the prototype connector have been conducted. Results from a signal transmission test indicate that the attenuation of the input signals was low as only ~3% with negligible signal reflection. Helium leak rate was less than the detection limit of 1x10-10 Pa m 3 /s. It was confirmed that the stand-off voltage is more than 1 kv, which is more than five times higher than the operation voltage of 200 V. Accordingly, it has been confirmed that the newly developed connector is feasible to ITER. An effect of cooling water in BM on the in-situ calibration of MFCs has been evaluated by the neutron transport analysis using the MCNP code [5] as shown in FIG.3. In this analysis, neutron flux at the installation position of MFC and those neutron response are compared under the following two conditions of BM; 1) SS316 70% + water 30%, 2) SS316 70% + void 30%. As a result, the neutron response of MFCs for a neutron source in the case where the cooling water pipe in the BM is not filled with water is about twice as high as the case where (c) Ni foil

3 3 ITR/P5-35 the cooling water pipe is filled with water. This indicates that the cooling water pipe in the BM should be filled with water for the in-situ calibration as during ITER operations Edge Thomson Scattering System The Edge Thomson Scattering System will measure the profiles of electron density and temperature in the plasma edge region especially for edge pedestal study and control of H-mode [6-9]. For the precise and accurate measurement, a high energy (5 J) and high repetition rate (100 Hz) pulsed Nd:YAG laser system (laser wavelength λ=1064 nm) is required. In JADA, development of a prototype YAG laser system has progressed remarkably based on the design of that for Thomson scattering system in JT-60U (7.46 J, 50 Hz) [10]. In order to increase the repetition rate from 50 Hz to 100 Hz, input energy to flash lamps (corresponding to the pumping energy to amplifier rods) of each laser pulse should be reduced to be a half since averaged pumping power is limited to avoid severe thermal effects in the amplifier. Hence, higher amplification gain at a laser amplifier is needed. To obtain higher amplification gain, harmful effects that cause a loss of gain, i.e. ASE (amplified spontaneous emission at λ=1064 nm) and parasitic oscillations, should be suppressed as much as possible. For this purpose, samarium (Sm) doped glass was adopted for the flow tubes which accommodate amplifier rods and flash lamps. Figure 4 shows the small signal gain (SSG) of an amplifier as a function of input energy to the flash lamps. For the case of 50 Hz repetition rate, it was demonstrated that SSG with Sm doped tubes is substantially increased as ~2.8 times higher than those for without Sm doped tubes and for the case of JT-60U. At 100 Hz repetition rate with reduce pumping energy, the nearly same level of SSG to that for 50 Hz repetition rate was obtained. Figure 5 shows a photograph of the developed YAG laser system equipped with the higher gain amplifier. Consequently, 7.66 J of output energy and 100 Hz repetition rate (766 W of averaged output power) have been successfully achieved [7]. This performance shows that ITER target values have been fulfilled and the world highest performance of the laser system for diagnostics has been established as shown in FIG.6. neutron response (a.u.) with water without water neutron energy (MeV) FIG.3. Energy dependence of neutron response of the upper MFC. Opera&on)condi&on)for) ITER)laser:)50)J,)100)Hz Opera&on)condi&on)for) JT<60)laser:)94)J,)50)Hz 2.8 FIG.4. Small signal gain of the prototype YAG laser amplifier as a function of input energy to flash lamps. FIG.5. A photograph of the prototype YAG laser system.

4 4 ITR/P5-35 output energy (J) total stray light Chevron (3 Sheets) Chevron (2 Sheets) Wedge repetition rate (Hz) FIG.6. Comparison of YAG lasers for Thomson scattering index of surface roughness σ FIG.7. Total stray light of beam dump as a function of the index of surface roughness σ. A YAG laser beam will be launched into the VV along nearly radial direction from an equatorial port. A laser beam dump at an inboard BM is a significantly important component to terminate the powerful laser beam with low level of stray light. JADA has proposed a new beam dump design with chevron shaped fins as shown in FIG.7. Numerical simulations indicate that the new beam dump could significantly reduce the stray light level, by up to a factor of 10, as compared to a conventional beam dump with wedge type fins. Accordingly, this new design is promising for ITER [11]. Considering the possible degradation of transmission optical components during an ITER operation, a novel in-situ calibration method has been proposed utilizing the background bremsstrahlung emission [12]. Numerical simulations show that calibration coefficients for all viewing channels are obtained. In addition to that, a profile of effective electric charge (Z eff ) of the plasma can be also deduced by data from the Edge Thomson Scattering System, indicating the extension of its measurement capability Poloidal Polarimeter The Poloidal Polarimeter will measure the current profile based on the detection of change in polarization state (orientation angle θ and ellipticity angle ε) of far infrared (FIR) laser beams with the wavelength of ~119 µm due to so called Faraday rotation effect and Cotton-Mouton effect [13-16]. Figure 8 shows overview of this diagnostic system in the inter%space port%cell gallery diagnos0c%hall upper port plug equatorial port plug retro reflectors in blanket modules and a divertor cassette steering mirrors for beam alignment laser launching mirrors into port plugs diagnostic hall at upper port level beam convertor (focusing mirror) FIG. 8. Overview of the Poloidal Polarimeter in the tokamak building.

5 5 ITR/P5-35 FIG.9. Design of the mirror module of poloidal polarimeter for the upper port plug. tokamak building. Since the FIR laser beams must be stably injected into the plasma through an upper port and an equatorial port, a new concept for a robust mirror module in the port plugs has been developed, where the plasma FIG.10. Heat loads to the first mirrors (top) and temperature distribution of the first mirror block obtained by thermal analysis (bottom). facing mirrors and mirror support structure are unified. Figure 9 shows the design of mirror module for the upper port. This mirror module consists of three mirror blocks, i.e. the first, second and third mirror blocks, which are made of SS. Mirrors are formed by direct machining of surfaces of those blocks. Figure 10 shows heat loads to the first mirror block and temperature distribution of the first mirror block obtained by a thermal analysis. Figure 11 shows surface deformation of the first mirrors obtained by a structural analysis. The maximal deformation within 90 % of mirror surface is less than 10 µm. The laser beams that pass through the plasma are reflected by retro reflectors (RRs) in the BMs and divertor cassette. In addition to the conventional corner cube RR, a new concept of terrace retro-reflector array (TERRA) is proposed to reflect obliquely injected laser beams [17]. Optical analyses have investigated the characteristics of a laser beam reflected by a thermally deformed corner cube RR and TERRA. These analyses indicate that about 70% and 50%, respectively, of injected laser power is returned to the diagnostic hall [18]. An optical analysis was carried out for evaluation of polarization state of laser beams returned to the diagnostic hall as shown in FIG.12, where surface deformation of the first mirror and the deformation of the corner cube RR were taken into account. It is confirmed that the orientation angle θ (deg) ellipticity angle ε (deg) FIG.11. Thermal deformation of the first mirror surfaces of mirror block of poloidal polarimeter. FIG.12. Distribution of polarization state of a laser beam returned to the diagnostic hall; orientation angle θ (left) and ellipticity angle ε (right). Deformations of the first mirror surface and the corner cube RR were taken into account.

6 6 ITR/P5-35 change in polarization state caused by above deformation are smaller than target measurement accuracies of θ and ε of (Δθ, Δε) (0.5, 3 ) [19] Divertor Impurity Influx Monitor The Divertor Impurity Influx Monitor will measure the parameters of impurities and isotopes of hydrogen (hydrogen, deuterium and tritium) in the divertor plasma by using spectroscopic techniques in the wavelength range of nm. In order to perform FIG.13. A new optical design for the equatorial measurement system of the Divertor Impurity Influx Monitor utilizing achromatic lenses. the two-dimensional measurement, this system observes the divertor region using five optical systems installed in the divertor cassette, the divertor port (two systems to view through the gap between the divertor cassettes), the equatorial port and the upper port [20, 21]. In order to increase the light intensity measured by the equatorial port optics, the diameter of the entrance pupil of front end optics was increased from 1.5 mm to 5 mm. Illumination analysis of the new design of the equatorial port optics with the above pupil size has confirmed that the image of the optical fiber (0.2 mm dia.) on the divertor region is less than the required value of 50 mm. However, a significantly larger size of optical mirrors is needed. This problem is resolved by adopting achromatic lenses made of the silica (SiO 2 ) and high purity calcium fluoride (CaF 2 ), which is tolerant to gamma-ray irradiation, as shown in FIG.13. Here, it was reported that internal transmittance of CaF 2 is not degraded after the gamma-ray irradiation of 50 kgy. According to this new design with the achromatic lenses, detected signal will be increased up to a factor of 10 (object space numerical aperture is increased from 0.03 to 0.12) as compared with that of the previous design for smaller pupil, together with keeping mirror sizes as small as those in the previous design [22]. Since light reflected by wall surface contaminates the signal as the stray lights, characteristics of wall reflections should be investigated to develop the methods to mitigate influences of the stray light. For this purpose, the bidirectional reflectivity distribution function (BRDF) of the tungsten (W) block, which will be used in the divertor dome and the baffle, has been measured for the first time [23]. Figure 14 shows the laser microscope data of surface roughness of the milled W block; the same type of W block in ITER. The maximum depth of the engraved lines is 2.8 µm and space between the engraved lines is 1 to 100 µm. Figure 15 shows the photographs of light reflected by the W block. For parallel light incidence, the light is reflected in a nearly straight-line shape. For perpendicular light incidence, the light is reflected in an arc shape. An optical simulation, where the surface model from the laser microscope measurement of the W block is adopted, has reproduced the above characteristics qualitatively well as shown in FIG.16. It is to note that sand blasted surface can diffuse the reflected light as shown in FIG.15(c), this result indicates the importance of surface finishing and monitoring of surface condition. FIG.14. Laser microscope data of surface roughness of the W block milled with a fraise.

7 7 ITR/P5-35 FIG.15. Photographs of light reflected by the W block. FIG.16. Results of optical simulation for light reflected by W block surface Divertor IR Thermography The Divertor IR Thermography will measure the temperature distribution and heat load of the inner and outer divertor targets and baffles by spectroscopic measurement of infrared (IR) light with the wavelength of 1-5 µm with high temporal and spatial resolution for physics study. For temporal resolution, 20 µs and 2 ms are required for the temperature ranges of C and C, respectively. For spatial resolution, 3 mm is required. Figure 17 shows the overview of the components at an equatorial port and observation area. As shown in the figure, the outer divertor target is observed obliquely down direction from the port plug. There are two different temperature detection systems based on the light intensity ratio of 3 µm and 5 µm, and spectroscopy (1-5 µm) for selected 100 spatial points. Figure 18 shows a design of spectroscope installed in port cell. Light of the wavelength of 1-5 µm can be resolved on the detector within an appropriate size of 17 mm [24]. Intensive optical analysis was carried out for change in shape and size of spot diagram on the detector. With a decrease in the angle of field of view, the spot diagram becomes smaller. The spot diagram significantly becomes larger as the wavelength becomes longer. These results obtained by optical analyses provide basis of the design optimization. Observa(on+area+ for+outer+divertor Front2end+op(cs+ for+inner+divertor Observa(on+area++ for+inner+divertor Front2end+op(cs+ for+outer+divertor+ Relay+op(cs FIG.17. Overview of the components at an equatorial port plug and observation area of the Divertor IR Thermography.

8 8 ITR/P Divertor Thermocouples at Outer Target This diagnostic system will measure the temperature distribution of armor tiles of the outer vertical target by use of thermocouples. Obtained data will be also used for calibration of the Divertor IR Thermography diagnostic. For fixation of the thermocouples, metallic coating of armor tiles is needed for spot welding of the thermocouples. JADA has experimentally confirmed that nickel (Ni) foil can be bonded by brazing to the CFC armor tile and spot welding of the thermocouples onto it can be implemented. Based on these results, a heating test to simulate the plasma operation conditions is planned. 3. Conclusion FIG.18. Design of spectroscope of the Divertor IR Thermography (top), evaluated spot diagram on detector (bottom). Recent highlighted results from the design and R&D activities in JADA have been presented, showing the steady progress towards the successful construction of ITER diagnostic systems. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. References [1] A.E. Costley et al., Proc. 22 nd IAEA Fusion Energy Conf., IT/P6-21 (2008). [2] Y. Kusama et al., 19 th Meeting of the ITPA TG on Diagnostics (Naka, Oct., 2010). [3] M. Ishikawa et al., Rev. Sci. Instrum. 81 (2010) 10D308. [4] T. Kondoh et al., Proc. Plasma Conference 2011, 22P136-P (Kanazawa, Nov., 2011). [5] M. Ishikawa et al., Proc. 38 th EPS conf. Plas. Phys., P2.058 (2011) [6] E. Yatsuka et al., Rev. Sci. Instrum. 81 (2010) 10D541. [7] T. Hatae et al., Rev. Sci. Instrum. 83 (2012) 10E344. [8] E. Yatsuka et al.,rev. Sci. Instrum. 83 (2012) 10E328. [9] E. Yatsuka et al., Nucl. Fusion 51 (2011) [10] T. Hatae et al., J. Plasma Fusion Res. SERIES, Vol. 9, 253 (2010). [11] E. Yatsuka et al., 9 th Fusion Energy Rengokoenkai, 29A137p (Kobe, June, 2012). [12] E. Yatsuka et al., Proc. Plasma Conference 2011, 25C08 (Kanazawa, Nov., 2011), submitted to Plasma Fusion Res. [13] Y. Kawano, et al., Proc. Plasma Conference 2011, 22P120-P (Kanazawa, Nov., 2011). [14] R. Imazawa et al., Nucl. Fusion 51 (2011) [15] R. Imazawa et al., Plasma Phys. Control. Fusion 54 (2012) [16] R. Imazawa et al., ibid ITR/P5-38. [17] R. Imazawa et al., Rev. Sci. Instrum. 82 (2011) [18] R. Imazawa et al., Plasma Fusion Res. 6 (2011) [19] R. Imazawa et al., 9 th Fusion Energy Rengokoenkai, 29A138p (Kobe, June, 2012). [20] H. Ogawa et al., Fusion Eng. Des. 83 (2008) [21] T. Sugie et al., AIP Conf. Proc. 988 (2008) 218. [22] H. Ogawa et al., Proc. Plasma Conference 2011, 22P137-P (Kanazawa, Nov., 2011). [23] A. Iwamae et al., JAEA-Research (2012). [24] M. Takeuchi et al., 9 th Fusion Energy Rengokoenkai, 29A-139P (Kobe, Jun., 2012).