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• Alleviation of impurities levels in Li for IFMIF-type neutron source are suggested. • No fatal problems are found by applying the alleviation of impurities levels. • Level of Be-7 in Li under process can be diminished by the alleviation. The alleviated limits of impurities of nitrogen and hydrogen isotopes in lithium are suggested to be 400wppm and 550appm, respectively for early realization of Advanced Fusion Neutron Source planned in Japan. The applicability of the suggestion is discussed by comparing to that of IFMIF. As the results, not less than 4-year buffer for R&D of nitrogen trap are brought in, which is thought as one of bottle neck. Additionally, the application of this suggestion would diminish the beryllium (a radioactive isotope will generate in A-FNS process) concentration in processing lithium, as well. On the other hand, it is indicated that there would be no fatal problems by application of the present suggestion. [ABSTRACT FROM AUTHOR]

Abstract: The tritium permeation behavior through F82H steel from approximately pure tritiated water vapor diluted with He was investigated and the as-received and the permeation-experimented sample surfaces were analyzed by means of SEM, EDX and XRD. Extremely slow but obvious tritium permeation through F82H steel could be observed. It is indicated from the surface analyses that an oxide layer composed of magnetite and hematite is formed during the permeation experiments. The permeabilities estimated using the partial pressure of molecular tritium under the extreme assumption that the redox reaction among tritiated water vapor, magnetite and hematite with forming molecular tritium reaches equilibrium indicate that apparent tritium permeation using tritiated water vapor for the inlet side would be proceeded by molecular tritium generated by the redox reaction, although it is expected that actual tritium permeation would be derived from dissociated tritium atom indicated in transition state during the redox reaction. [Copyright &y& Elsevier]

The annihilation behavior of irradiation defects induced in neutron-irradiated Li2TiO3 and Li2ZrO3 were investigated with the tritium release behavior. It was revealed that the common characteristics in both samples were that the annihilation process of irradiation defects consisted of two first-order processes and E’-center could act as tritium trapping site, and otherwise was the way how the E’-center annihilated. The difference was suggested to attribute to the mobility of M as M4+ (or M3+, etc.). [ABSTRACT FROM AUTHOR]

Abstract: Recently, we have investigated the annihilation behavior of the radiation defects and the tritium release behavior in/from some neutron-irradiated ceramic breeder materials. In the present study, correlation between the tritium release and the annihilation of irradiation defects were investigated for neutron-irradiated Li2ZrO3. The annihilation of irradiation defects consisted of two processes, the fast and the slow processes. The activation energies for these processes were determined to be 0.32±0.01 and 0.80±0.01eV, respectively. The E′-center of irradiation defects was found to act as a tritium trapping site by comparison between the annihilation of irradiation defects and tritium release. From the present and the previous works, the annihilation process of oxygen vacancies such as E′-center and F+-center could play an important role in tritium behavior. The results of the present and previous studies suggest that there appears to be similar correlation between annihilation of irradiation defects and tritium release for several ceramic breeder materials. [Copyright &y& Elsevier]

Abstract: A PACVD apparatus was designed and fabricated at Shizuoka University in order to prepare high-pure boron coating films. In the present study, some parameters, especially feeding gas concentration, substrate temperature and CVD input power, have been optimized to prepare pure boron coating films. It was found that the purity of boron coating film was controlled by the decaborane concentration of feeding gas and substrate temperature during the PACVD process, and each optimized values were 0.4 and 473K, respectively. The atomic composition of boron in the boron coating film under the optimized condition has been achieved to be 0.94. [Copyright &y& Elsevier]

Abstract: Helium irradiation effects on the retention of energetic deuterium implanted into the boron coating film were investigated by means of X-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS). It was found, by XPS, that the B 1s peak was shifted to lower binding energy side by He+ ion irradiation and the FWHM was extended. These facts show that the some defects were introduced into the boron coating film by He+ ion irradiation. From TDS experiment, the deuterium retention, especially the amount of B–D terminal bond, increased by the pre-He+ ion irradiation. However, it decreased by the post-He+ ion irradiation. These experimental results indicate that the B–D terminal bond was mainly influenced by the He+ ion irradiation because the two neighbor B–D bonds have to dissociate simultaneously for the B–D–B bridge bond. [Copyright &y& Elsevier]

Characterization of the structure and the chemical state of boron in boron coating film were performed by XRD and TDS measurements. The structure of the boron coating film was shown to be amorphous by XRD measurement. By changing the total pressure, the discharge power, and the substrate temperature, the growth rate of the boron coating film and the hydrogen retention were evaluated. It was found that the growth rate decreased as the total pressure increased, and increased with an increase in the discharge power. For the substrate temperature dependence, the hydrogen retention decreased as the substrate temperature increased. However, the growth rate remained constant, although the hydrogen retention decreased as the substrate temperature increased. This indicates that the growth rates are not controlled by the chemical form of B–H bond, but by that of B-B bond. In the TDS analysis, hydrogen atoms released from B–H–B and B–H bonds were observed. [Copyright &y& Elsevier]

Abstract: Retention of hydrogen isotopes in the carbon (isotropic graphite) first wall tiles of JT-60U was studied by secondary ion mass spectrometry and thermal desorption spectroscopy. The surface morphology and erosion/deposition profiles of the tiles were characterized using scanning electron microscope and X-ray photoelectron spectroscopy. The upper area is mainly eroded, while the bottom area of the inboard wall is dominated by deposition. In contrast to the divertor area, hydrogen isotope retention in the eroded wall area was generally larger than that in the deposition dominated area. Measured near surface concentrations of hydrogen isotopes in the wall tiles, as well as the D/H ratios, were a little higher than those in the divertor area. This indicates direct implantation of high-energy D from NBI into the first wall. The lower temperature of the first wall relative to the divertor tiles would reduce desorption and/or replacement of implanted D by subsequent D or H impingement. [Copyright &y& Elsevier]

Abstract: Ion fluence dependence on chemical behavior of energetic deuterium implanted into oxygen-contained boron and pure boron films was investigated by XPS and TDS. It was found that large amount of D implanted into the oxygen-contained sample was quickly desorbed as D2O by chemical sputtering during implantation. The D retention for both samples increased as the fluence increased. The implanted D was preferentially trapped by O with forming O–D bond, and thereafter by B with forming B–D bond. These facts indicate that the stability of D bound to O is higher than that bound to B. It was concluded that the chemical sputtering of oxygen by energetic particles and the chemical stability of O–D bond should be taken into consideration for the evaluation of tritium inventory in the oxygen-contained boron film. [Copyright &y& Elsevier]

Abstract: TDS analyses have been performed in order to examine helium irradiation effects for deuterium retention in boron coating films prepared by PACVD. It is clarified that the deuterium retention is reduced due to the irradiation prior to deuterium. There is no correlation between the retention variations of deuterium and helium, and thus the helium occupation seems hardly to contribute to the reduction. The third peaks in the TDS spectra shift on the high temperature side with the irradiation fluence. It is presumed that helium irradiation forms certain layers by knocking on boron atoms, in which deuterium retention is partly restricted, and trapped deuterium needs higher energy for the desorption. [Copyright &y& Elsevier]

The behavior of hydrogen isotopes in boron was studied by XPS and TDS. The samples used were boron polycrystals and boron thin films produced by plasma CVD using decaborane (B10H14) diluted by He gas. The 1.0 keV deuterium ions were implanted into the samples up to fluence of 5.4 × 1021 D m-2 at various temperatures. It was found that deuterium retention at 573 K was decreased less than 35% compared to that at room temperature. The activation energies for B–D–B and B–D bonds were estimated to be 1.11 ± 0.21 and 2.17 ± 0.36 eV for the boron thin film, and 1.36 ± 0.20 and 2.27 ± 0.45 eV for the boron polycrystal, respectively. The B–H distance and the activation energy for the B–H bond were also calculated by the Gaussian 03 code and these were consistent with the experimental values. [Copyright &y& Elsevier]

Based on the results of the IFMIF/EVEDA project in the broader approach (BA) activities, the conceptual design of the Advanced Fusion Neutron Source (A-FNS) to obtain irradiation data of DEMO blanket materials is under consideration in Japan. This paper describes the progress of the A-FNS conceptual design, which consists of a 40 MeV, 125 mA deuteron beam, a liquid lithium target system, and a test and post-irradiation test facilities. the main objectives of the A-FNS are the acquisition of RAFM irradiation data up to 2035 and the tritium recovery test on the blanket. In addition, it is expected to be used for various neutron researches. The technical design activities of the A-FNS are focused on improvement of the lithium target design, review of impurity criteria, remote handling, and irradiation application plan. [ABSTRACT FROM AUTHOR]

Abstract Operation of the JET tokamak with beryllium and tungsten ITER-like wall provides unique opportunity for detailed studies on dust generation: quantity, morphology, location, etc. The programme carried out in response to ITER needs for safety assessment comprises: (i) remotely controlled vacuum cleaning of the divertor; (ii) local sampling of loosely bound matter from plasma-facing components (PFC); (iii) collection of mobilized dust on various erosion-deposition probes located in the divertor and in the main chamber. Results of comprehensive analyses performed by a number of complementary techniques, e.g. a range of microscopy methods, electron and ion spectroscopy, liquid scintillography and thermal desorption, are summarized by following points: (a) Total amount of dust collected by vacuum cleaning after three campaigns is about 1–1.4 g per campaign (19.1–23.5 h plasma operation), i.e. over 100 times smaller than in JET operated with carbon walls (i.e. in JET-C). (b) Two major categories of Be dust are identified: flakes of co-deposits formed on PFC and droplets (2–10 μm in diameter). Small quantifies, below 1 g, of Be droplets and splashes are associated mainly with melting of beryllium limiters. (c) Tungsten dust occurs mainly as partly molten flakes originating from the W-coated tiles. [ABSTRACT FROM AUTHOR]

Abstract Micro/nanoscopic observations on the surface of the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with ITER-like Wall (JET ILW) have been carried out by means of several material analysis techniques. Previous results from the inner divertor were reported for a single poloidal section of the tile numbers 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. The formation of the thick stratified mixed-material deposition layer on tiles 1 and 4, and erosion on tile 3 were identified. This study is mostly focused on the outer divertor: tiles 6, 7 and 8. In contrast to the inner tile, remarkable surface modifications have not been observed on the vertical target (tiles 7 and 8) where sputtering erosion and impurity deposition would have been almost balanced. Only a specific part of tile 6 (horizontal target) located near the exhaust channel was covered with a stratified ("geological-like") mixed-material deposition layer which mainly included Be and Ni with the thickness of ∼2 μm. Special feature of this mixed layer was that a certain amount of nitrogen (N) was clearly detected in the layer. Since the concentration of N varied with the depth position, it could be depended on the amount of that gas puffed for plasma edge cooling during the JET experimental campaign. In addition to the outer divertor tiles, a very interesting feature of the local erosion and deposition effects is reported in this paper. [ABSTRACT FROM AUTHOR]

Abstract Targets of Reduced Activation Ferritic Martensitic (RAFM) steel F82H were exposed to low-energy (200 eV), high flux (about 1022 D/m2s) deuterium (D) plasma at 623–773 K to various D fluences in the range from 1 × 1025 to 2.5 × 1026 D/m2. The surface morphology of the plasma-exposed targets was examined with a field-emission scanning electron microscope. Cross-sectional observations of nano-structures formed on the F82H target surfaces were performed using a transmission electron microscope equipped with an energy dispersive X-ray spectrometer. It has been shown that nano-sized fiber-like layers are formed on the target surfaces under D plasma exposure. Micro-sized surface morphology pattern depends on the D fluence. As the D fluence increases, clusters of the fiber-like layers begin to be formed and organized into ordered structure. Highlights • F82H steel targets were exposed to 200 eV D plasma at elevated temperatures to various D fluences. • Nano-sized fiber-like layers were formed on the target surfaces. • As the D fluence increased, clusters of the fiber-like layers began to be formed and organized into ordered structure. [ABSTRACT FROM AUTHOR]

Micro-/nano-characterization of the surface structures on the divertor tiles used in the first campaign (2011–2012) of the JET tokamak with the ITER-like wall (JET ILW) were studied. The analyzed tiles were a single poloidal section of the tile numbers of 1, 3 and 4, i.e., upper, vertical and horizontal targets, respectively. A sample from the apron of Tile 1 was deposition-dominated. Stratified mixed-material layers composed of Be, W, Ni, O and C were deposited on the original W-coating. Their total thickness was ∼1.5 μm. By means of transmission electron microscopy, nano-size bubble-like structures with a size of more than 100 nm were identified in that layer. They could be related to deuterium retention in the layer dominated by Be. The surface microstructure of the sample from Tile 4 also showed deposition: a stratified mixed-material layer with the total thickness of 200–300 nm. The electron diffraction pattern obtained with transmission electron microscope indicated Be was included in the layer. No bubble-like structures have been identified. The surface of Tile 3, originally coated by Mo, was identified as the erosion zone. This is consistent with the fact that the strike point was often located on that tile during the plasma operation. The study revealed the micro- and nano-scale modification of the inner tile surface of the JET ILW. In particular, a complex mixed-material deposition layer could affect hydrogen isotope retention and dust formation. [ABSTRACT FROM AUTHOR]

Surface morphology changes and deuterium (D) retention in Toughened, Fine-Grained Recrystallized Tungsten (TFGR W) with TaC dispersoids (W-TaC) and pure tungsten exposed to D plasmas to a fluence of 10 26 D/m 2 s were studied as a function of the D ion flux (10 22 –10 24 D/m 2 s). As the flux increased from 10 22 D/m 2 s to 10 24 D/m 2 s, the numbers of blisters increased for both materials. However, smaller blisters were observed on W-TaC compared to pure W. In W-TaC, cracks beneath the surface along grain boundaries were observed, which were comparable to the blister sizes. The reason for the smaller blister sizes may arise from smaller grain sizes of W-TaC. In addition, reduction of the D retention in W-TaC was observed for higher flux exposures. D depth profiles indicate this reduction arises due to decrease in trapping in the bulk. [ABSTRACT FROM AUTHOR]

Surface topography and deuterium retention in polycrystalline hot-rolled W and W–5%Re have been examined after exposure to a low-energy (76 eV), high flux (around 10 22 D/m 2 s) deuterium plasma to an ion fluence of 10 26 D/m 2 at various temperatures. The methods used were confocal laser scanning microscopy and the D( 3 He, p) 4 He nuclear reaction at 3 He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 348 to 673 K, small blisters of size in the range from about 1 to about 15 μm, depending on the exposure temperature, are formed on the W and W–5%Re surfaces. In the W–5%Re, the deuterium retention demonstrates its maximum at exposure temperature of 463 K, while in the W this maximum is shifted to 523 K. A difference in the temperature dependence of the D retention for the W and W–5%Re is explained, as a rough approximation, by temperature dependences of the ductility of these materials. [ABSTRACT FROM AUTHOR]

Abstract: Distribution of deuterium and hydrogen retained in graphite tiles placed in the divertor region of JT-60U with the dual side pumping geometry was investigated by thermal desorption spectroscopy. The retention of hydrogen isotopes was nearly proportional to the thickness of the carbon redeposited layers, though their concentration changes with the poloidal location and orientation of the surface tiles. The least concentration of (H+D)/C (∼0.02) was found in the redeposited layers on the inner divertor tile. Rather high hydrogen concentration was found in the redeposited layers on plasma shadowed area. In particular, the redeposited layers on the bottom side of the outer dome wing tile shadowed from plasma and facing to the pumping slot showed the highest concentration of 0.13 in (H+D)/C. In JT-60U, however, the carbon deposition in the shadowed area is very small, resulting in a smaller total retention in the divertor area compared to other large tokamaks. [Copyright &y& Elsevier]

Abstract: Dependence of implantation temperature on chemical behavior of energetic deuterium implanted into WC was investigated by TDS and XPS. 1.0keV ions were implanted into WC samples at the implantation temperature range of 323–873K. It was found that the deuterium retention decreased as the implantation temperature increased. Above 573K, most of the retained deuterium was bound to C, which was less than 20% of the total D retention after implantation at 323K. Above 673K, C was segregated on the WC surface and some of the implanted deuterium was retained in the segregated carbon layer. Additionally, it can be said that the D retention in WC was much less than that in other carbon-related materials, such as graphite and SiC. Hydrogen isotope retention can be reduced significantly when WC is formed on a divertor surface as a redeposited layer. [Copyright &y& Elsevier]

Abstract: Deuterium (D) and Hydrogen (H) retention in carbon tiles exposed to DD and HH discharges in the JT-60U W-shaped divertor has been investigated by means of thermal desorption spectroscopy (TDS) and secondary ion mass spectrometry (SIMS). In the thick re-deposited layers on the outer divertor, hydrogen (H+D) was retained rather homogeneously with its concentration of 0.03 in (D+H)/C atomic ratio, which is attributed to the surface temperature rise above 900K. Owing to the HH discharges performed after the DD discharges for tritium removal, H-dominated re-deposited layers were added on D-dominated re-deposited layers. In addition D retained in near surface layers were replaced by H both for re-deposited and eroded area. Accordingly deuterium concentration peaked at a little inside behind the H dominated surface layers for all tiles. The temperature increase during the discharge very likely causes the saturation of hydrogen retention even in the re-deposited layers. Accordingly the global wall saturation was observed in JT-60U. [Copyright &y& Elsevier]

Abstract: Depth profile and retention properties of hydrogen in the graphite armor tiles of JT-60 open-divertor used for HH discharge period were investigated by secondary ion mass spectroscopy and thermal desorption spectroscopy. Two tiles were investigated, one at the inner divertor area covered with the re-deposited layers of the largest thickness (70μm) and the other at the private flux region near the outer divertor striking point hardly covered with the re-deposited layers (only 1–2μm). In the former, hydrogen was retained uniformly in the carbon re-deposited layers with H/C of 0.03, which was significantly smaller than H/C observed in the re-deposited layers of other tokamaks. Such small hydrogen retention in the re-deposited layers is attributed to the temperature rise of the re-deposited layers during discharges due to their poor thermal contact to the substrate. The latter retained rather large amount of hydrogen only in the shallow surface layers and its depth profile decayed sharply to 1.3×10−4 in H/C of the bulk retention as expected from the very thin re-deposited layers and observed temperature rise above 900K. [Copyright &y& Elsevier]

Abstract: The chemical behavior of energetic deuterium implanted into tungsten carbide (WC) samples has been investigated by XPS and TDS. The TDS experiments showed that almost of all deuterium retained in WC was desorbed in the temperature region from 300 to 700K and the residue at around 1000K. The deuterium (D2) desorption peak at around 1000K was found to originate from the desorption of deuterium atom (D) trapped by carbon (C). Its retention was considerably small and was saturated at low deuterium ion (D2 +) fluence. The area of the large D2 desorption peaks, which corresponds to the D retention in WC, increased as D2 + fluence increased. The D2 desorption in the temperature range of 300–700K takes place in three stages, the lower two peaks were attributed to the desorption of D retained in two interstitial sites, and the highest peak corresponds to D trapped by carbon vacancies induced by chemical sputtering of C by energetic D2 +. From these experimental results, the D trapping and detrapping behaviors are discussed and it is concluded that the hydrogen isotope retention is governed by the amount of interstitial sites and carbon vacancies in WC. The hydrogen isotope retention could be lowered if the surface temperature of WC were above 700K. [Copyright &y& Elsevier]

Abstract: Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature. [Copyright &y& Elsevier]

Abstract: In the present study, the thermal desorption behavior of deuterium implanted into diamond was investigated by means of TDS and XPS. From the results of TDS experiments, the most of all implanted deuterium was desorbed by forming D2 mainly, and two desorption peaks were observed in the temperature range from 700 to 1100K and from 1100 to 1300K. The lower temperature deuterium desorption from the diamond was similar to that from HOPG. However, the D/C value of the diamond was approximately 0.2, which was almost half the value of the HOPG. The C1s XPS spectra of diamond were shifted towards lower energy with increasing heating temperature above 773K. In addition, the π–π∗ transition peak appeared above this temperature. These meant that the disordered diamond structure were reconstructed through the desorption of deuterium, and stabilized by formation of a graphite network. [Copyright &y& Elsevier]