Diagnostic value of texture analysis of apparent diffusion coefficient maps for differentiating fat-poor angiomyolipoma from non-clear-cell renal cell carcinoma

Purpose: To investigate the feasibility of texture analysis of apparent diffusion coefficient (ADC) maps for differentiating fat-poor angiomyolipomas (fpAMLs) from non-clear-cell renal cell carcinomas (non-ccRCCs). Methods: In this bi-institutional study, we included two consecutive cohorts from different institutions with pathologically confirmed solid renal masses: 67 patients (fpAML = 46; non-ccRCC = 21) for model development and 39 (fpAML = 24; non-ccRCC = 15) for validation. Patients underwent preoperative magnetic resonance imaging (MRI), including diffusion-weighted imaging. We extracted 45 texture features using a software with volumes of interest on ADC maps. Receiver operating characteristic curve analysis was performed to compare the diagnostic performance between the random forest (RF) model (derived from extracted texture features) and conventional subjective evaluation using computed tomography and MRI by radiologists. Results: RF analysis revealed that grey-level zone length matrix long-zone high grey-level emphasis was the dominant texture feature for diagnosing fpAML. The area under the curve (AUC) of the RF model to distinguish fpAMLs from non-ccRCCs was not significantly different between the validation and development cohorts (p = .19). In the validation cohort, the AUC of the RF model was similar to that of board-certified radiologists (p = .46) and significantly higher than that of radiology residents (p = .03). Conclusions: Texture analysis of ADC maps demonstrated similar diagnostic performance to that of board-certified radiologists for discriminating between fpAMLs and non-ccRCCs. Diagnostic performances in the development and validation cohorts were comparable despite using data from different imaging device manufacturers and institutions

Influence of dynamic annealing of irradiation defects on the deuterium retention behaviors in tungsten irradiated with neutron

Tungsten (W) samples were damaged by neutron and 6.4 MeV Fe-ion irradiation above 1000 K simulating the divertor operation temperature. Deuterium (D) retention properties were examined by decorating the damaged W with D and subsequent thermal desorption spectroscopy (TDS) measurements. Vacancy clusters were the major D trapping site in the W irradiated with Fe-ion at 873 K, although D retention by vacancy clusters decreased in the W irradiated with Fe-ion at 1173 K due to dynamic annealing. The D de-trapping activation energy from vacancy clusters was found to be 1.85 eV. D retention in neutron damage W was larger than that damaged by Fe-ion due to the uniform distribution of irradiation defects. The D desorption behaviors from neutron damaged W was simulated well by assuming the D de-trapping activation energy to be 1.52 eV

Investigation of remaining tritium in the LHD vacuum vessel after the first deuterium experimental campaign

Remaining tritium in the vacuum vessel after the first deuterium plasma experimental campaign conducted over four months was investigated in the large helical device (LHD) for the first time in stellarator/heliotron devices by using the tritium imaging plate technique. In-vessel components such as divertor tiles and first wall panels, and long-term material probes retrieved from the vacuum vessel were analyzed. The in-vessel component in which tritium remained most densely is the baffle part of divertor tiles made of graphite retrieved from the inboard-side divertor. Asymmetric tritium retention is observed on divertor tiles located at magnetically symmetric positions, and can be attributed to the toroidal field direction dependence of the asymmetric loss of energetic tritons generated by deuterium–deuterium nuclear fusion reactions. On the first wall, tritium remained in a deposited layer, which mainly consists of carbon

Current Status of Large Helical Device and Its Prospect for Deuterium Experiment

Achievement of reactor relevant plasma condition in Helical type magnetic devices and exploration in its related plasma physics and fusion engineering are the aim of the Large Helical Device (LHD) project. In the recent experiments on LHD, we have achieved ion-temperature of 8.1 keV at 1 × 1019 m−3 by the optimization of wall conditioning using long pulse discharge by Ion Cyclotron Heating (ICH). The electron temperature of 10 keV at 1.6 × 1019 m−3 was also achieved by the optimization of Electron Cyclotron Heating (ECH). For further improvement in plasma performance, the upgrade of the Large Helical Device (LHD), including the deuterium experiment, is planned. In this paper, the recent achievements on LHD and the upgrade of LHD are described

Development of the Tritium Transport Model for Pebbles of Li2TiO3

Tritium transport code in Li2TiO3 pebbles were developed in this study. In the simulation code, tritium transports in a grain, a closed-pore and an opened-pore were modeled and combined. The tritium transport in grain was modeled by the tritium diffusion, trapping/detrapping by defects, and the annihilation process of defects. The adsorption/desorption equilibrium of diffusing tritium gas molecule was modeled for the tritium transports in opened-pores. A vacant-core and shell model was used to model the tritium transport in closed-pores. The results by simulation code suggested that the tritium trapping by closed-pores would result in the shift of tritium release toward higher temperature side as dissociation of tritium gas molecule in closed-pores requires higher activation energy

Effect of neutron dose on the tritium release behavior of Li2TiO3–0.5Li4SiO4 biphasic ceramic

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Deuterium Permeation Behavior in Fe Ion Damaged Tungsten Studied by Gas-Driven Permeation Method

The deuterium (D) permeation behavior for 1 displacement per atom Fe2+ damaged tungsten (W) was studied by the gas-driven permeation method and compared with undamaged W. The results of thermal desorption spectroscopy showed that dislocation loops and voids were formed in damaged W. It was found that the D permeation behavior in W was affected by irradiation defects. The effective diffusivity and permeability in the damaged W were lower than that in undamaged W. However, the difference in effective diffusivity and permeability between the undamaged sample and the damaged sample was reduced with increasing the heating temperature. Under 965 K, which was enough for D detrapping from voids, the permeability for damaged W was consistent with that for undamaged W

Microstructure change and deuterium permeation behavior of ceramic-metal multi-layer coatings after immersion in liquid lithium-lead alloy

For the establishment of liquid tritium breeding concepts, static lithium-lead corrosion tests for single-layer erbium oxide coatings and erbium oxide-metal multi-layer coatings were carried out, followed by deuterium permeation measurements. Grain boundary corrosion of erbium oxide coatings was confirmed by the static immersion tests at 550 and 600 °C. An erbium oxide-iron two-layer coating sustained its structure after lithium-lead immersion at 600 °C for up to 3000 h. The results of gas-driven deuterium permeation measurements for multi-layer coatings immersed at 550 and 600 °C indicated that crystallization and grain growth of erbium oxide would sufficiently occur during the immersion at 600 °C, and then the sample showed lower permeabilities in the first measurements at lower temperature. On the other hand, permeation reduction factors of the sample immersed at 550 °C were estimated to be 100‒200 in the temperature range of 400‒550 °C. A corrosion layer formed on the coating surface might work as an additional diffusion barrier after the permeation test at 600 °C. Keywords: Lithium-lead, Tritium permeation barrier, Corrosion, Erbium oxid