Performances of Helium, Neon and Argon Glow Discharges for Reduction of Fuel Hydrogen Retention in Tungsten, Stainless Steel and Graphite

Transcription

1 1 FTP/P7-10 Performances of Helium, Neon and Argon Glow Discharges for Reduction of Fuel Hydrogen Retention in Tungsten, Stainless Steel and Graphite T. Hino 1), Y. Yamauchi1), Y. Kimura1), A. Matsumoto1), K. Nishimura2), Y. Ueda3) 1) Hokkaido University, Sapporo, Japan 2) National Institute for Fusion Science, Toki-shi, Japan 3) Osaka University, Suita-shi, Osaka, Japan address of main author: Abstract. It is quite important to investigate the performance of glow discharge conditionings for controls of in-vessel tritium (T) inventory and hydrogen recycling. For this purpose, first, the deuterium (D) retentions in tungsten (W), graphite (C) and stainless steel (SS) were measured. The retention in W was not small as expected, several times larger than that of SS, although the retention in SS was one order smaller than that of C. Such the large retention in W is owing to the growth of rough surface structure produced by plasma irradiations. For reduction of deuterium retention in W, SS and C, second, inert gas (He, Ne, Ar) glow discharges were conducted under the same condition, and these performances were compared. The removal ratio of deuterium retention was highest in He discharge, and lowest in Ar discharge. These values are well explained by the numerical analyses using SRIM code. The removal ratios for SS and C were significantly large, but quite small for W. This reason is again owing to the rough surface structure in W. For W, thirdly, the hydrogen isotope exchange and the wall baking experiments were conducted. It is found that the wall baking with a temperature higher than 700 K can well reduce the retention, and the hydrogen isotope exchange using deuterium glow discharge is also useful to reduce the tritium retention in the wall. The present results significantly contribute to control the fuel hydrogen retention and to reduce the in-vessel tritium inventory in fusion reactors. 1. Introduction The inventory of the tritium in a fusion reactor plant has to be lower than several kilograms. The inventory inside of the vacuum chamber, called as in-vessel tritium inventory, has to be lower than several hundred grams. In the case of International Thermonuclear Experimental Reactor, ITER, tungsten may be employed as the plasma facing wall in the divertor region from the beginning of HH operation. In future fusion reactors, one of the candidate plasma facing materials is tungsten, owing to low sputtering yield, high melting point, low activation for fusion neutron and low amount of retained tritium. The other candidates are graphite and ferritic steel. However, the fuel hydrogen

2 2 FTP/P7-10 retention and tritium retention/inventory in these materials has not been systematically investigated yet. Thus, it is required to evaluate the amount of retained tritium, i.e. tritium retention/inventory for these materials. In addition, the scheme to reduce the tritium inventory has to be developed. It is known that the fuel hydrogen recycling deteriorates the energy confinement of fusion plasma. The fuel hydrogen recycling often becomes dominant if the fuel hydrogen retention, i.e. amount of (D+T) retained in the wall, is large. From this view point, the fuel hydrogen retention also has to be controlled. In the present study, the fuel hydrogen retention and the tritium inventory for tungsten, graphite and stainless steel were evaluated based upon the deuterium glow discharge. It is believed that the retention and desorption behaviour of tritium is the same as the other hydrogen isotope, deuterium or hydrogen (H). Thus, in this study the deuterium plasma was employed to evaluate the fuel hydrogen retention and the tritium inventory. In fusion reactors, the tritium accumulates in the wall as increase of a plasma operation time. The glow discharge with hydrogen isotope gas or inert gas may be employed in the intervals of the main discharge shots if the in-vessel tritium inventory exceeds an allowable certain amount. The glow discharge with inert gas may be also used to reduce the fuel hydrogen retention, i.e. (D+T) retention in the wall [1-5]. In this study, the glow discharges using inert gases such as helium (He), neon (Ne) and argon (Ar) gases were conducted under the same condition to reduce the deuterium retention, in order to evaluate the reduction of the tritium retention by the inert gas discharges. In addition, the hydrogen discharge followed by the deuterium discharge was conducted to reduce the hydrogen retention through the isotope exchange from H to D, in order to apply this method for the reduction of the tritium inventory. The thermal deuterium behaviour was also investigated, in order to clarify the effect of wall baking for reduction of the tritium retention. Finally, the methods suitable to reduce the tritium inventory and the fuel hydrogen retention are suggested. 2. Experiments and Results The glow discharge apparatus as shown in FIG.1 [1-5] was used to implant the deuterium into the W, SS or C liner. The liner temperature was room temperature (RT).The W, SS and C liners were made by sheets of polycrystalline tungsten (Nilaco), 316L SS and IG-430U (Toyo Tanso), respectively. Before the discharge, the deuterium gas was supplied to the chamber with a constant flow rate. After the beginning of the discharge, the partial pressure of D 2 rapidly dropped and the partial pressures of HD and H 2 gradually increased. Gas retention was obtained based upon a residual gas analysis, RGA. FIG. 2 shows the temporal changes of these partial pressures for W. The net reduction of the partial

3 3 FTP/P7-10 pressure corresponds to the implanted deuterium amount. The implanted amount saturated within the discharge time of approximately 60 min. The saturated value for the amount of retained D was ~5 x10 16 D/cm 2. In order to investigate the hydrogen isotope effect, the H 2 glow discharge was similarly conducted under the same condition and the amount of retained H was measured. The amount of retained H was ~5.9 x10 16 H/cm 2, very close to the amount of retained D. No significant difference between H and D was observed. Thus, the amount of retained T is also regarded as ~5 x T/cm 2. The deuterium retention in SS and C was similarly measured under the same discharge and substrate conditions. It was seen that the deuterium retention in W was about double of that for SS. The large retention in W is owing to the growth of surface roughness explained later. The retention in C was one order of magnitude larger than that in SS, owing to the production of active carbon dangling bond (C- bond) D During glow discharge HD H Time[s] 0.0 FIG. 1 Glow discharge apparatus. FIG.2 Changes of partial pressures during D 2 glow discharge for W liner. It is believed that the W wall has a low tritium inventory. If the present result (~5 x10 16 T/cm 2 ) is applied, the tritium inventory in a fusion reactor with a full W wall becomes only a few grams. It is known that the tritium retention becomes significantly large if the carbon impurities deposits on the tungsten wall. In the fusion device, the wall is always contaminated by the impurities such as carbon and oxygen. The thickness of carbon impurity layer on the plasma facing wall is often larger than several hundred monolayers, i.e. several ten nm. In such the case, the tritium inventory exceeds several hundred grams. Then, the purging for tritium inventory has to be often conducted even if the W wall is used. However, there is little suggestion on the useful method for purging tritium inventory.

4 4 FTP/P7-10 In order to reduce the tritium inventory and control the hydrogen recycling, the performance of glow discharge conditionings has to be investigated. For reduction of deuterium retention in W, SS and C, inert gas (He, Ne, Ar) glow discharges were conducted after the deuterium glow discharge, and these performances were compared. FIG. 3 shows the changes of partial pressures during the He, Ne and Ar glow discharges for W. During the inert gas discharge, the increase of deuterium partial pressure corresponds to the desorbed amount of deuterium. The ratios of the removed deuterium amount by He, Ne and Ar discharges for W were 4%, 3% and 2%, respectively. This ranking is proportional to the implantation depth of inert gas ion calculated by SRIM code. The ratios for SS by He, Ne and Ar discharges were 45%, 30% and 15%, respectively. The ratios for C were about half of the values for SS. TABLE 1 shows the removal ratio of fuel hydrogen amount in W, SS and C by He, Ne and Ar discharges. It is found that the removed ratio for W is extremely small. The atomic force microscope (AFM) photographs after the deuterium discharge shows that the rough surface with many holes with a size of ~100 nm was produced, as shown in FIG. 4. In addition, Auger electron spectroscopy (AES) analysis showed that the carbon content was not reduced even after the deuterium discharge. The present results show that the glow discharge conditionings for SS and C is suitable but not for W D 2 During He glow discharge HD He D 2 During Ne glow discharge HD Ne HD During Ar glow discharge D 2 Ar H Time[s] H Time[s] H Time[s] 0.00 FIG.3 Changes of partial pressured during He, Ne and Ar glow discharges for W liner. The removal ratio can be explained by both surface erosion and ion impact desorption. We now consider the case of C. FIG. 5 shows the depth profiles of D, He, Ne and Ar ions in C. The inert gas ion colloids with the implanted deuterium, so that the deuterium retention is partly reduced. If the range of the inert gas ion is larger, the reduction of the deuterium retention becomes larger. The erosion depths in C are shown in FIG. 6. Most of deuterium may be removed by the erosion. Thus, the removal ratio estimated for C becomes TABLE 2. The removal ratio in the case of He is very higher than those in the cases of Ne and Ar. The estimated values are roughly the same as the experimental values shown in TABLE 1. For SS and W, similar results were obtained.

5 5 FTP/P7-10 In the case of W, the removal ratio was extremely smaller compared with the cases of SS and C. Very rough structure on the W surface was developed by the ion irradiation, owing to dropping of crystal grains. The ion flow in the glow discharge concentrates to protrusive regions of the rough surface, so that the removal of deuterium is quite localized. Then, the deuterium retention is little reduced. This is a reason why the removal ratio becomes very small. The present tungsten was polycrystalline tungsten. For other tungsten materials such as laminar tungsten, the rough surface also grows by particle and heat flows. Thus, for reduction of fuel hydrogen retention, other methods have to be developed. The deuterium retention in W was several times larger than that in SS, as described above. In the case of SS, the rough surface was not developed and the surface carbon concentration was significantly reduced by the ion irradiation. In W, at the concave regions of the rough surface, the carbon impurity concentration little decreased by the ion irradiation. Thus, the deuterium retention becomes larger compared with the case of a flat surface since the carbon content well traps the deuterium. (a) (b) (c) FIG. 4 AFM photographs of W before irradiation (a), after D 2 plasma irradiation (b) and after D 2 plasma irradiation followed by Ar plasma irradiation. TABLE 1 Removal ratio of fuel hydrogen amount for SS, C and W by He, Ne and Ar discharges.

6 6 FTP/P7-10 TABLE 2 Estimated removal ratio of fuel hydrogen amount in the case of graphite (C). FIG. 5 Depth profile of implanted D, He, Ne and Ar ions in graphite (C). FIG. 6 Depth profile of implanted D and erosion depths of He, Ne and Ar ions in graphite (C). For W, the hydrogen isotope exchange and the wall baking experiments were conducted. After the implantation of hydrogen using the hydrogen glow discharge, the deuterium glow discharge was conducted to replace the hydrogen into the deuterium. FIG. 7 shows the temporal change of partial pressures during the hydrogen glow discharge and the deuterium glow discharge. Approximately 60% of implanted hydrogen was replaced by the deuterium. Such the high exchange ratio is owing to the effect of neutral radicals, namely deuterium atoms. In the cases SS and C, the removal ratio was also high (60%). The hydrogen isotope exchange ratio is significantly high, so that this method can be applied to reduce the in-vessel tritium inventory in fusion reactors. Namely, the tritium retention in the wall can be reduced by using deuterium glow discharge conditionings.

7 7 FTP/P7-10 (a) (b) FIG. 7 Change of partial pressure of H 2 during H 2 discharge (a), and changes of D 2, HD and H 2 during D 2 discharge conducted after H 2 discharge. In the glow discharge apparatus, the small W sample was placed on the sample holder and exposed to the deuterium plasma. This sample was extracted and transferred to the vacuum chamber of thermal desorption spectroscopy (TDS) device. The sample was heated in the TDS device, and the gas desorption spectra were measured by a quadruple mass spectrometer (QMS). For W, the thermal desorption spectra is shown in FIG. 8. The horizontal axis is the heating temperature. In the wall baking treatment for the walls in fusion devices, this temperature corresponds to the baking temperature. It is found that the wall baking with a temperature higher than 700 K can well reduce the retention. The baking with 700 K is relatively easy, so that baking treatment is also a very useful method. FIG.8 Thermal desorption spectra of W liner after D 2 glow discharge.

8 8 FTP/P Conclusion The deuterium retentions in tungsten (W), graphite (C) and stainless steel (SS) were measured under the same discharge and substrate conditions. The retention in W was not small as expected, several times larger than that of SS, although the retention in SS was one order smaller than that of C. Such the large retention in W is owing to the growth of rough surface structure produced by the plasma irradiation. For reduction of deuterium retention in W, SS and C, inert gas (He, Ne, Ar) glow discharges were conducted and these performances were compared. The removal ratio was highest in He discharge, and lowest in Ar discharge. These values are well explained by the numerical analyses using SRIM code. The removal ratios for SS and C were significantly large, so that the inert gas glow discharges are useful for the wall conditionings to the SS and graphite walls. However, the ratio is quite small for W. This reason is owing to the rough surface structure in W. For W, the hydrogen isotope exchange and the wall baking experiments were conducted. The wall baking with a temperature higher than 700 K can well reduce the retention, and the hydrogen isotope exchange using deuterium glow discharge is a useful tool to reduce the tritium retention in the wall. The present results significantly contribute to control the fuel hydrogen retention and to reduce the in-vessel tritium inventory in fusion reactors. Acknowledgements This work was supported by the collaboration study program of National Institute of Fusion Science, NIFS08KOBS013, and by the Grant-in-Aid for Scientific Research from the Ministry of Education, Science, Sports and Culture in Japan. The graphite sheets of isotropic graphite, IG-430U, were supplied by Toyo Tanso Ltd. References [1] HINO, T. et al, Fusion Eng. and Design 85 (2010) [2] YAMAUCHI, Y. et al, J. Nucl. Mater., 390 (2009) [3] KIMURA, Y. et al, J. Nucl. Mater., 417 (2011) [4] HINO, T. et al, Materials Science Forum 721 (2012) [5] HINO, T. et al, Fuel hydrogen retention of tungsten and the reduction by inert gas glow discharges, To appear in Fusion Eng and Design (2012).