Surface Tritium Research Articles

Theoretical design of a solid scintillation detector for tritium surface contamination measurement based on CaF2(Eu) sheet and self-coincidence technology of MCP-PMT

Measurement of tritium surface contamination is important for tritium related facilities, especially for fusion devices. A novel detector has been designed for tritium surface contamination measurements based on CaF2(Eu) sheet and Microchannel plate photomultiplier tube (MCP-PMT). Self-coincidence technology has been introduced to obtain lower detection limit by diminishing the background noise caused by γ-rays and hot electrons. Performance of the detector was optimized by specifying the key parameters including the distance of the scintillator and sample, the thickness of light guide, the number of annular electrodes et al., using Monte Carlo method. Results indicate that detection efficiency of 49.4 % and the response deviations at different positions less than 0.5 % could be achieved. The detection limit of tritium under the simulation conditions is 0.09 Bq/cm2 and 0.03 Bq/cm2 when the counting time is 10 s and 60 s, respectively.

Quantitative estimation of tritium amount in surface layer and bulk of tungsten sample using BIXS

Estimation of both the surface tritium and tritium in bulk in key materials of fusion reactors is of great importance for tritium safety and the management of tritium-contaminated material. For the further quantitative analysis of tritium in solids, the elaborate BIXS spectra were calculated based on Monte Carlo simulation. Four types of tritium depth profile were considered to evaluate the quantitative estimation method. It is found that the attenuation of X-rays in tungsten depends on both the energy of X-rays and the depth of X-rays. The evaluation of tritium amount in surface layer indicated that the intensity of Ar(Kα) peak could be used to evaluate the surface tritium within 400 nm from the surface in most cases and the deviations were less than 9% in the calculation. The intensity of both W(Lα) X-rays and the high energy X-rays can be employed to roughly estimate the total tritium amount. For linearly decreasing and exponentially decreasing distribution, the maximum calculation deviations were 24.9% and 28.8%, respectively. While for Gaussian distribution, the maximum deviations were 146% and 53%, respectively. And it can also be used for tritium estimation in other materials.

Hyaluronic acid adsorption on nanodiamonds: Quantitative characteristics and mechanism

Formation of adsorption complex between hyaluronic acid and detonation nanodiamonds was studied. Nanodiamonds that possess negative and positive zeta-potential in an aqueous suspension were used. To determine the direct amount of hyaluronic acid on the nanodiamond surface tritium label was introduced into the polymer by means of tritium thermal activation method. It was found that hyaluronic acid irreversibly adsorbs on both positively and negatively charged nanodiamonds, but in different quantities. Adsorption of hyaluronic acid on nanodiamonds of positive zeta potential leads to the isoelectric point at the surface coverage close to 100 mg/g, while further increase in surface amount of hyaluronic acid results in the formation of negatively charged particles and the increase of its species diameter comparing with the initial material. In the case of negatively charged nanodiamonds adsorption of hyaluronic acid is also observed, but it was as much as 50 times lower than in the case of positively charged nanodiamonds. Even small amounts of hyaluronic acid adsorbed on nanodiamond surface results in the disaggregation of clusters of nanodiamonds that are formed in an aqueous suspension, while zeta potential was from − 50 to − 40 mV that is around of the value for non-modified material. On the basis of the experimental data, we can suggest that the adsorption mechanism of complex formation between hyaluronic acid and nanodiamonds includes both electrostatic interaction and formation of hydrogen bonds between surface functional groups and adsorbed water molecules.

A planar-type ionization chamber for tritium surface contamination measurements in fusion facility

Gamma irradiation from nuclides produced by neutron activation places significant influence on tritium surface contamination measurements after DT discharge in both Inertial Confinement Fusion (ICF) and Magnetic Confinement Fusion (MCF) facilities. A novel kind of gamma ray insensitive detector, Planar-type Ionization Chamber (PIC), has been developed and it can be used to measure tritium surface contamination directly with air as working gas. Theoretical calculations on both energy deposition and electric field were used in the design and several kinds of experiments were carried out to check the performance of the PIC detector. Results indicate that the PIC detector can be operated at gamma dose as high as 35.5 μGy/h. With the sensitive area 100 cm2, the lower detection limit of the detector is around 16.1 Bq/cm2, and the detection efficiency is 11.5 %.

Investigation of the distribution of remaining tritium in divertor in LHD

Abstract To reveal the triton transport and the tritium migration in a deuterium plasma experiment in the Large Helical Device (LHD), the distribution of the remaining tritium in divertor tiles made of graphite after the first deuterium plasma experimental campaign in 2017 was investigated. In this study, tritium contents in divertor tiles have been measured by using a full-combustion method. The asymmetric tritium retention in divertor tiles located at symmetric positions, which was found in the previous study by the surface tritium measurement using an imaging plate technique, has been confirmed by the results of the full-combustion method. The asymmetry is considered to be attributed to the asymmetric distribution of lost-points of energetic tritons in divertor. A depth profile of remaining tritium in a divertor tile estimated by using a combination of the imaging plate technique and a sputtering treatment shows that the peak of the profile locates at several micro-meters from the surface. This result suggests that the majority of the remaining tritium impinged upon the divertor tile as energetic tritons. In this study, a distribution of lost-points of energetic tritons has been calculated by using a Lorentz orbit following code (LORBIT) with taking into account divertor components. The obtained distribution has been compared with measured tritium distributions on divertor tiles. The measured and calculated distributions are similar to each other, but they are not the same. The difference between them can be attributed to plasma exposures during and after the deuterium plasma campaign.

Synergistic effects of high energy helium irradiation and damage introduction at high temperature on hydrogen isotope retention in plasma facing materials

In this study, energetic helium (He) ion irradiation was performed to obtain bulk He distribution in tungsten (W) materials, concurrent with damage introduction at high temperature. Then, deuterium (D) implantation and thermal desorption spectrometry were performed to evaluate D retention. At the same time, the surface tritium (T) concentration and depth distribution were evaluated by imaging plate (IP) and β-ray induced X-ray spectroscopy (BIXS) measurements after mixed D-T gas exposure. Numerical simulations were applied to evaluate changes in binding energies, diffusion depths, and trapping sites under different irradiation conditions. The results showed that weak trapping sites with higher concentration, such as vacancies, were produced during only energetic He+ irradiation events, leading to enhancement of D retention. Fe3+-He+ simultaneous irradiation promoted the formation of HexVy complexes, which reduced the concentration of vacancy trapping sites and changed the stress field around defects, leading to the suppression of D trapping behavior. From the reduced effects of D retention caused by HexVy complexes at higher temperatures, the results suggested that defect recovery was the dominant mechanism. With increasing damage level at higher temperatures, more weak trapping sites, such as dislocations and vacancies sites, were produced, leading to a more dominant influence on D retention than HexVy complex effects. It was also found that HexVy complexes prevented D diffusion to the bulk and that simulation results showed that the damage level had little impact on D diffusion depth.

The Area Measurements in the Tritium Laboratory Within NIPNE Magurele

This paper reports the results of radiological measurements taken for the last 3 years in the controlled area and adjacent zone of the Tritium Laboratory of the the Horia Hulubei National Institute for Physics and Nuclear Engineering, Magurele. The radiological characterization has been performed by determination of total and removed tritium contamination for the following surfaces: pavement, walls, windows, radiochemical tables, radionuclide fume cupboards, glove boxes, and sinks. The fixed tritium contamination does not present representative radiological risk because beta particles emitted by tritium are unable to penetrate the skin. The removed tritium contamination represents that component of the total surface contamination that can be taken by mechanical processes. The removed contamination was analyzed as a priority because it represents the main radiological risk factor in tritium laboratories. The determination of surface contamination has been carried out by scanning of the analyzed surfaces using an LB 1230 UMo tritium surface monitor with an LB 1230 detector and by a smear test using extruded polystyrene smears followed by measurement of the removed activity with a liquid scintillation counter. The total surface contamination values, obtained by scanning, were below the detection limit of the equipment, except for radiochemical hood surfaces. The removed tritium contamination determined values are in the domain of 5 … 450 Bq/dm2. At the department level, the obtained values for surface tritium contamination are at the background level.

Inventory and distribution of tritium in the oceans in 2016

Tritium concentrations in oceans were compiled from the literature, online databases and original measurements in order to determine the global distribution of tritium concentrations according to latitude and depth in all oceans.The total inventory of tritium decay corrected in 2016 has been estimated using evaluation of the natural and artificial contributions in 23 spatial subdivisions of the total ocean. It is determined equal to 26.8 ± 14 kg including 3.8 kg of cosmogenic tritium. That is in agreement with the total atmospheric input of tritium from nuclear bomb tests and the natural inventory at steady-state estimated from natural production rates in the literature (27.8–29.3 kg in the Earth). We confirm the global increase in tritium according to latitude observed in the Northern hemisphere since 1967 with a maximum in the Arctic Ocean. The minimum tritium concentrations observed in the Southern Ocean were close to steady-state with known natural tritium deposition.We focused on the temporal evolution of surface (0 to 500 m) tritium concentrations in a selected area of the North Atlantic Ocean (30°N–60°N) where we found the 2016 concentration to be 0.60 ± 0.10 TU (1σ). Results showed that in that area, between 1988 and 2013, tritium concentrations: i) decreased faster than the sole radioactive decay, due to a mixing with lower and lateral less concentrated waters, and ii) decreased towards an apparent steady state concentration. The half-time mixing rate of surface waters and the steady state concentration were respectively calculated to be 23 ± 5 years (1σ) and 0.38 ± 0.07 TU (1σ). This apparent steady-state concentration in the North Atlantic Ocean implies a mean tritium deposition of 1870 ± 345 Bq·m−2 (1σ), five folds higher than the known inputs (natural, nuclear tests fallout and industrial releases, ~367 Bq·m−2) in this area.

Tritium permeation from the primary helium cooling loop for the ESS target

The tritium permeation from the target cooling primary helium loop to the intermediate water loop through heat exchangers is studied, for the 5 MW operation of the target station at European Spallation Source (ESS). Both bi-metallic and mono-metallic heat exchanger concepts are studied, for the co-current and counter flow cases. The quasi two-dimensional temperature profiles in the helium, in the water and in the metallic barrier separating the two flow regions are calculated. The surface temperature of the metallic barrier facing to the helium side determines the tritium concentration dissolved in this region. The diffusion flux of the surface tritium through the wall thickness is calculated. The permeation rates via these different heat exchanger configurations are compared and a recommendation is made. It is shown that the tritium contamination level in the intermediate cooling loop allows controlled discharge of water to the city sewage system. The methodology presented in this paper can be applied to general engineering analyses of the tritium permeation problems in spallation facilities.

Geant4 Monte Carlo simulations for sensitivity investigations of an experimental facility for the measurement of tritium surface contaminations by BIXS

A new experimental facility was built at the Tritium Laboratory Karlsruhe (TLK) to investigate tritium surface contamination on arbitrary solid samples. With the TRIADE facility samples can be exposed to tritium in a pressure range of 10−7–102Pa and a temperature range of 170–470K. Surface contamination which accumulate on the sample by adsorption or absorption processes of T2 are determined by β-induced X-ray spectrometry (BIXS). To quantify the amount of surface contamination and to study systematic effects of the setup Monte Carlo simulations have been performed. Detection efficiencies for tritium adsorption on all inner surfaces of the setup were determined. From that, a minimum detectable activity on the sample surface of 600Bq/cm2 has been calculated which can be translated into a minimum measurable surface contamination of 5.5×10−4 monolayer for ideal (100) gold surfaces. First tritium measurements with a gold-coated beryllium sample at exposure pressures of 5×10−3Pa showed an increasing background signal. After a dosage of p×t=(1.6±0.1)×10−2Pah the background signal was 1.10±0.04cps. Assuming an ideal plane (100) gold surface the results found can be translated into a surface contamination of ≈5.4×104Bq/cm2. Further measurements must be performed as a linear increase of the count rate with the applied dosage was observed.

Advanced β-ray-induced X-ray spectrometry for non-destructive measurement of tritium retained in fusion related materials

A new β-ray-induced X-ray measurement system equipped with a silicon drift detector, which was named “Advanced-BIXS”, was constructed to study in detail retention behavior of surface tritium by measurements of low energy X-rays below 1keV such as C(Kα) and O(Kα) as well as high energy X-rays induced by β-rays from tritium. In this study, basic performance of the present system has been examined using various tritium-containing samples. It was seen that energy linearity, energy resolution and sensitivity were quite enough for measurements of low energy X-rays induced by β-rays. Intensity of characteristic X-rays emitted from the surface and/or bulk of a tritium-containing sample was lowered by argon used as a working gas of the Advanced-BIXS. Pressure dependence of transmittance of C(Kα) and Fe(Kα) was examined as examples of low and high energy X-rays, and it was able to represent by using the mass absorption coefficient in argon. It was concluded, therefore, that the present system has high potentiality for nondestructive measurements of tritium retained in surface layers and/or bulk of fusion related materials.

Solubility of Tritium in Cu-Be Alloy

Solubility of tritium in Cu-Be(2 mass%) alloy was determined by means of measurement of a tritium depth profile in the alloy. Tritium exposure to the samples was conducted under the following conditions: pressure, 0.4 to 2.6 kPa; temperature, 350 to 450°C; exposure time, 4 to 11 hours. Tritium depth profiles were obtained by chemical etching after the exposure. Remarkably high tritium concentration appeared in surface layers within 0.5 μm, whereas almost constant concentration was observed from 10 μm to the bulk. It was found, therefore, that surface tritium should be omitted in evaluation of the solubility of tritium. In addition, it was seen that dissolution of tritium into Cu-Be alloy obeys the Sieverts’ law from the pressure dependence, and the solubility of tritium in Cu-Be alloy was lower than that in pure copper. From the temperature dependence of solubility, the heat of solution of tritium was determined as 17 kJ/mol.

Influence of Chromium Oxide Layer on Surface Tritium Contamination of Type 316 Stainless Steel

Sheet type specimens of type 316 stainless steel covered by Cr oxide layers were exposed to tritium gas at 300 °C together with the specimens covered by Fe-rich oxide layers and pure Fe specimens to compare tritium behaviors in oxide films with different Cr contents. Tritium concentration in the oxide layers decreased with increasing Cr content due to reduction in concentrations of hydroxyl species and adsorbed water. In addition, release rate of tritium from Cr oxide layers was larger than that from Fe-rich oxide layers. It was concluded that preparation of thin Cr oxide layer is effective to reduce surface tritium contamination of austenitic stainless steels.

Trapping and depth profile of tritium in surface layers of metallic materials

Tritium amount retained in surface layers and release behavior from surface layers were examined using SS316L samples exposed to plasmas in the Large Helical Device and a commercial Cu–Be alloy plate. BIXS analyses and observation by SEM indicate that carbon and titanium deposited on the plasma-facing surface of the SS316L samples. Larger amount of tritium was trapped in the plasma-facing surface in comparison with the polished surface. Higher enrichment of tritium in surface layers was similarly found in the polished surface of both samples. The amount of surface tritium in both samples was almost same, while the bulk concentration of tritium in Cu–Be was much lower than that in SS316L. Tritium release from the SS316L and Cu–Be samples into water was examined by immersion experiments. Tritium elution was observed for both samples, but changes in the residual tritium amount in surface layers were different from each other.

Application of tritium imaging plate technique to examine tritium behaviors on the surface and in the bulk of plasma-exposed materials

We have applied a tritium imaging plate technique to measure the tritium distribution profile on surface and in bulk of various metal materials after exposure to a deuterium–tritium plasma in a linear plasma experimental apparatus. The experimental tritium concentration profiles in mm range are interpreted according to a simple hydrogen diffusion model in each metal. We found that a significant amount of tritium is localized in near surface regions and is clearly distinguishable from tritium diffused in the bulk. The amount of surface tritium is not likely correlated to bulk properties (diffusivity and solubility), but is related to trapping in surface defects or metal impurities such as oxide and carbide.

Behavior of tritium accumulated on materials surface

Tritium release behavior and surface tritium behavior were separately examined for typical fcc and bcc metals by using tritium tracer techniques. Pure copper (Cu), iron (Fe), nickel (Ni) and molybdenum (Mo) were loaded with hydrogen including a trace amount of tritium and then immersed into water at around room temperatures. Then, the tritium release rate into the water was examined by a liquid scintillation counting technique and the surface tritium concentration by a tritium imaging plate technique. The tritium release from the metals is attributed to the release of dissolved tritium by diffusion from the normal interstitial sites, and the first order desorption of trapped one with detrapping energies of 64, 72 and 25 kJ mol −1 for Cu, Fe and Mo, respectively. Overall release behavior is varied depending on the ratio of dissolved and trapped amounts of tritium.

Input function and simulated distributions of tritium in the North Pacific

An ocean general circulation model (OGCM) of the North Pacific with a southern open boundary condition is used to simulate the distribution and transport of tritium in the North Pacific. First, the observed tritium in precipitation dataset of GNIP/ISOHIS is chosen, and used to develop an input function of tritium for the model domain from 1951 to 1997 in terms of the triangulation for interpolation. Then, the input function is used as an upper flux for the tritium model. The OGCM with tritium is integrated for 47 years with the initial condition of zero tritium concentrations. Compared with the observations from GEOSECS and WOCE, the model generates a meridional gradient of surface tritium concentrations and the high tritium concentrations in the subsurface layer, which is consistent with that indicated in the observations. The simulated time series of total tritium fluxes from the atmosphere to the ocean shows a spike in 1963. The total inventory of tritium in the ocean changes slowly with time after 1975. Tritium enters the higher-latitude region, and then is mainly transported by the subsurface process towards the equatorial region, which is well simulated in the model. However, the difference mainly occurs at 30°N–40°N, which can be seen also in the simulation of CFCs. This difference may partly originate from the insufficient description of physical processes in the OGCM. Nevertheless, some uncertainties exist in the development of the tritium input function, which can also influence the simulated results.

Ex-situ tritium removal from JET tiles using RF inductive heating

Radio-frequency (RF) inductive heating has been successfully used as an ex-situ technique for the detritiation of entire tiles retrieved from JET. The assessment of the detritiation process used three different techniques, namely autoradiography, calorimetry and full combustion assisted by the liquid scintillation analysis. Autoradiography showed that using the RF technique more than 99% of the total surface tritium inventory can be efficiently removed from a tile after several heating cycles at the average temperature of only 490 °C. On the other hand, a comparison of the combustion measurements obtained before and after RF heating has shown that the bulk tritium activity decreased significantly (95% of the bulk tritium was released), while at the same time more than 99% of the surface tritium was liberated. The three methods used to assess the RF detritiation process have their advantages and disadvantages. Calorimetry allows a rough estimation of the tritium content of the complete tile, whilst IP imaging allows an approximate estimation of the content before and after detritiation but only for the surface tritium. On the other hand, full combustion associated with scintillation analysis allows very accurate measurements but only for specific samples with reduced dimensions, therefore extrapolation is needed to estimate the tritium content of an entire tile.

Tritium Monitor Based on Gas-flow Proportional Counter

The tritium surface contamination monitor based on gas-flow proportional counter is introduced in this paper. The detector is very advantageous for detecting low energy beta particles. Its detection efficiency is 84%, higher than the similar type monitors. Besides, Plateau Length is longer than 300V, Plateau slope is less than 1%/100V, Lowest Detection Limit is 0.221 Bq/cm2, the relative expansion uncertainty is 0.56%/h. The capabilities are superiority and steadiness. As a result, it can be well used to monitor tritium contamination in working sites or other places, with great value for usage and wide application fields.

Tritium distribution measurement of JET Mk II SRP divertor tiles

Tritium surface distribution on the JET Mark II Septum Replacement Plate (Mk II SRP) divertor tiles was measured by imaging plate technique. It was observed that areal tritium concentration was higher at the entrance of inner/outer pumping slots (so called ‘shadowed area’). The tritium distribution profiles were similar to those obtained in the Mk IIA divertor which was exposed to a series of D–T plasma operation (DTE1). Tritium concentration of the plasma facing surface was lower compared to that of the shadowed area. Particularly, it was very low at the outer divertor surface. The inner divertor surface also showed low level of tritium retention, though it was covered by the thick carbon deposition on that. This could be caused by tritium release due to the temperature rise when the inner strike point was on the tiles. On the plasma shadowed area like tile gaps, high tritium retention owing to the codeposition was observed.