P. Calderoni

Comparison of deuterium retention for ion-irradiated and neutron-irradiated tungsten

The behavior of D retention for Fe2+-irradiated tungsten with a damage of 0.025–3 dpa was compared with that for neutron-irradiated tungsten with 0.025 dpa. The D2 thermal desorption spectroscopy (TDS) spectra for Fe2+-irradiated tungsten consisted of two desorption stages at 450 and 550 K, while that for neutron-irradiated tungsten was composed of three stages and an addition desorption stage was found at 750 K. The desorption rate of the major desorption stage at 550 K increased as the displacement damage increased due to Fe2+ irradiation increasing. In addition, the first desorption stage at 450 K was found only for damaged samples. Therefore, the second stage would be based on intrinsic defects or vacancy produced by Fe2+ irradiation, and the first stage should be the accumulation of D in mono-vacancy and the activation energy would be relatively reduced, where the dislocation loop and vacancy is produced. The third one was found only for neutron irradiation, showing the D trapping by a void or vacancy cluster, and the diffusion effect is also contributed to by the high full-width at half-maximum of the TDS spectrum. Therefore, it can be said that the D2 TDS spectra for Fe2+-irradiated tungsten cannot represent that for the neutron-irradiated one, indicating that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten is different from that for the ion-irradiated one.

The deuterium depth profile in neutron-irradiated tungsten exposed to plasma

The effect of radiation damage has been mainly simulated using high-energy ion bombardment. The ions, however, are limited in range to only a few microns into the surface. Hence, some uncertainty remains about the increase of trapping at radiation damage produced by 14 MeV fusion neutrons, which penetrate much farther into the bulk material. With the Japan-US joint research project: Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), the tungsten samples (99.99 % pure from A.L.M.T., 6mm in diameter, 0.2mm in thickness) were irradiated to high flux neutrons at 50 C and to 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). Subsequently, the neutron-irradiated tungsten samples were exposed to a high-flux deuterium plasma (ion flux: 1021-1022 m-2s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment (TPE) at the Idaho National Laboratory (INL). First results of deuterium retention in neutron-irradiated tungsten exposed in TPE have been reported previously. This paper presents the latest results in our on-going work of deuterium depth profiling in neutron-irradiated tungsten via nuclear reaction analysis. The experimental data is compared with the result from non neutron-irradiated tungsten, and is analyzed with the Tritium Migration Analysis Program (TMAP) to elucidate the hydrogen isotope behavior such as retention and depth distribution in neutron-irradiated and non neutron-irradiated tungsten.

Characterization of surface morphology and retention in tungsten materials exposed to high fluxes of deuterium ions in the tritium plasma experiment

Under appropriate conditions, exposing tungsten to a high flux D plasma creates near-surface blisters and other changes in surface morphology. We have characterized the sizes of blisters formed at different temperatures (147 °C≤Tsurface≤704 °C) and performed a surface analysis to elucidate factors that influence blister formation. Tungsten targets that were exposed to low energy (70 eV) D ions at a flux of 1.1×1022 m−2 s−1 in the tritium plasma experiment (TPE) were considered. We used AES to analyze the surface for evidence of implanted impurities. Blister diameters and heights were quantified using SEM imagery and vertical scanning interferometry. Given the likelihood of D precipitation in blisters, we expect that the data obtained here could be incorporated into a computational model to better simulate the diffusion and desorption of D in W. With this in mind, we present an analysis of thermal desorption profiles showing the release of D from the surface.

Development of Monte Carlo Simulation Code to Model Behavior of Hydrogen Isotopes Loaded into Tungsten Containing Vacancies

Abstract The behavior of hydrogen isotopes implanted into tungsten containing vacancies was simulated using a Monte Carlo technique. The correlations between the distribution of implanted deuterium and fluence, trap density and trap distribution were evaluated. Throughout the present study, qualitatively understandable results were obtained. In order to improve the precision of the model and obtain quantitatively reliable results, it is necessary to deal with the following subjects: (1) how to balance long-time irradiation processes with a rapid diffusion process, (2) how to prevent unrealistic accumulation of hydrogen, and (3) how to model the release of hydrogen forcibly loaded into a region where hydrogen densely exist already.

Overview of Recent Tritium Experiments in TPE

Abstract Tritium retention in plasma-facing components influences the design, operation, and lifetime of fusion devices such as ITER. Most of the retention studies were carried out with the use of either hydrogen or deuterium. Tritium Plasma Experiment is a unique linear plasma device that can handle radioactive fusion fuel of tritium, toxic material of beryllium, and neutron-irradiated material. A tritium depth profiling method up to mm range was developed using a tritium imaging plate and a diamond wire saw. A series of tritium experiments (T2/D2 ratio: 0.2 and 0.5 %) was performed to investigate tritium depth profiling in bulk tungsten, and the results shows that tritium is migrated into bulk tungsten up to mm range.

Optimization of heat treatment and calibration procedures for high temperature irradiation resistant thermocouples

ABSTRACT A detailed description is provided for the optimization of heat treatment and calibrating procedure for Idaho National Laboratory’s high temperature irradiation resistant thermocouples (HTIR-TC). Also discussed is the implication of the procedures on the overall performance of the HTIR-TC finished product; in particular, for the case of long lead sections that are typical of sensors deployed in nuclear reactors. The effect on the measurement accuracy of fluctuations of the reference temperature and localized heating along the sensor lead is also investigated. A calibration graphical user interface (GUI) and accompanying script is presented for unifying the calibration process. A fifth-order polynomial function was found to be best for fitting HTIR-TC calibration data. The performance of the HTIR-TCs subject to the optimized heat treatment and calibration procedures is shown to be robust and consistent with other TC industry standards. Electromotive force curves versus length of the TC are presented to clarify positioning of TCs in general and how its effect on the calibrated temperatures.