Tritium Desorption Research Articles

Tritium permeation and desorption in reduced activation martensitic steels studied in EXOTIC-9/1 irradiation experiment

The EXOTIC-9/1 has been irradiated in the High Flux Reactor in Petten within the European framework for the development of the helium cooled pebble bed breeder concept. The EXOTIC-9/1 irradiation assembly consists of a pebble bed with Li2TiO3 pebbles enclosed in a EUROFER97 containment. The irradiation of EXOTIC-9/1 proceeded for 301 full power days reaching 1.3dpa in steel. Irradiation temperatures varied between 613 and 853K.Tritium permeation in a EUROFER97 containment wall was studied, in-pile, during temperature and purge gas chemistry transients. Under reference purge gas conditions (He+0.1%H2) permeation proceeded in the diffusion-limited regime. Tritium permeability in EUROFER97 was obtained from analysis of the in-pile data.During the post irradiation examination program a number of EUROFER97 containment wall segments were ramp annealed in a temperature programmed desorption setup. This study determines the amount of tritium retained in the samples and presents analysis of the desorption spectra.

Analysis of tritium retention in beryllium pebbles in EXOTIC, PBA and HIDOBE-01 experiments

Tritium desorption studies were performed on beryllium pebbles irradiated in three different irradiation campaigns (i.e., EXOTIC-8, PBA and HIDOBE-01), which were carried out in the High Flux Reactor in Petten. Irradiation doses corresponded to the following helium production levels: 100 appm in EXOTIC-8/3, 284 appm in EXOTIC-8/7, 260 appm in EXOTIC-8/8, 400–500 appm in PBA, and 3090 appm in HIDOBE-01. The beryllium pebbles used in this study were manufactured by melt spray and rotating electrode methods. Measured tritium inventories were compared with the MCNP/FISPACT calculations, and it was concluded that for the irradiation temperatures below 900K the tritium retention is nearly 100% of tritium production. At the irradiation temperatures above 1000K, tritium escaped from beryllium during irradiation and the retention fraction, in case of the HIDOBE-01 experiment, was about 5% or less.Analysis of the release spectra revealed a number of characteristic peaks measured as gas (HT) and water (HTO). Below 500K, tritium release takes place in the HTO form and can be ascribed to the exchange processes taking place at the beryllium surface and tritium dissociation from small clusters. At higher temperatures, tritium release takes place in the HT form and in most spectra represented as one peak centered between 1100 and 1200K.Optical and scanning electron microscopy studies demonstrated that between 1073 and 1173K, helium bubbles form an extended network of interconnected channels, which are apparently responsible for the observed tritium release.

Pilot determinations of 3H and 14C in gaseous releases from Kozloduy nuclear power plant

Methods for routine assessment of 3H and 14C content in gaseous releases from ventilation stacks of Kozloduy Nuclear Power Plant (Bulgaria) were developed. Technique for correction of incomplete desorption of tritium from exposed silica gel was proposed. The distribution and the concentrations of both nuclides in various chemical forms were constantly monitored for a period of 1 year. The results for annual normalized gaseous discharges were assessed for the fifth unit at 173 GBq/(GW.a) for 3H and 369 GBq/(GW.a) for 14C, while for the sixth unit—3H—98 GBq/(GW.a) and 14C—289 GBq/(GW.a).

Matter Sampling, Tritium Removal, Inventory and Management with Versatile Embedded System Called LASK

The integration of laser techniques in an embedded system is studied in this paper to propose system able to perform tritium inventory (with in-line and on-line methods) and tritium removal (by ablation of layers or by tritium desorption) in a tokamak. Two complementary embedded systems are currently studied: a first innovative system which can be used between two plasma sessions (without tokamak pressure and temperature breaks) and a second system which can be used during maintenance phases (at atmospheric pressure). The two systems will be presented in this paper with their advantages and disadvantages.

Behavior of Tritiated Water on Concrete Materials

A fusion reactor requires high levels of safety and public acceptability, so safeconfinement of tritium is one of the key issues. Tritium must be well controlled with no excessive release to environment and no excessive workers exposure. Especially, the hot cell and tritium facility of ITER will use various construction materials such as the concrete and the organic materials. Since the concrete materials will be contaminated by tritium compared with the metal materials such as SS, it is very important to study the tritium behavior on the materials from the viewpoint of the excessive exposure protection to workers. Therefore, in order to understand the tritium behavior on the concrete materials, the sorption and desorption experiment was carried out as a function of the exposure time and temperature of water in the desorption experiment. The used samples were cement paste, mortar and concrete. These samples were exposed into 740 ˜ 1110 Bq/cm3 of tritiated water vapor at room temperature. The exposed time was from a day to several weeks. The exposed samples for a certain period were soaked into water at 277 K, 298 K and 343 K, and then the water was periodically measured by Liquid Scintillation Counter (LSC) and the amount of tritium sorbed on the concrete materials were evaluated. The amount of the tritium sorbed in the concrete materials reached equilibrium less than 2 months. In the desorption behavior from concrete materials to water at 277K, 298K and 343K, the tritium sorbed in the concrete materials was desorbed about 99 % for 2 days at 343 K of water. However, the tritium desorption from concrete materials was sufficiently detected though 3 months passed. In addition, the tritium profile on the surface concrete materials was measured by a tritium imaging plate.

A Prototype Four-Inch Short Hydride (FISH) Bed as a Replacement Tritium Storage Bed

The Savannah River Site (SRS) tritium facilities have used 1st generation (Gen1) metal hydride storage bed assemblies with process vessels (PVs) fabricated from 3 inch nominal pipe size (NPS) pipe to hold up to 12.6 kg of LaNi4.25Al0.75 metal hydride for tritium gas absorption, storage, and desorption for over 15 years. The 2nd generation (Gen2) of the bed design used the same NPS for the PV, but the added internal components produced a bed nominally 1.2 m long, and presented a significant challenge for heater cartridge replacement in a footprint limited glove-box.A prototype 3rd generation (Gen3) metal hydride storage bed has been designed and fabricated as a replacement candidate for the Gen2 storage bed. The prototype Gen3 bed uses a PV pipe diameter of 4 inch NPS so the bed length can be reduced below 0.7 m to facilitate heater cartridge replacement. For the Gen3 prototype bed, modeling results show increased absorption rates when using hydrides with lower absorption pressures. To improve absorption performance compared to the Gen2 beds, a LaNi4.15Al0.85 material was procured and processed to obtain the desired pressure-composition-temperature (PCT) properties. Other bed design improvements are also presented.

Retention and Desorption Behaviors of Hydrogen Isotopes in Gamma-Ray Irradiated Li2TiO3

Annihilation behaviors of irradiation defect and, correlation of these behaviors with deuterium trapping and desorption in gamma-ray irradiated Li2TiO3, which is one of the candidates for tritium breeding material, were studied by means of an ESR(Electron Spin Resonance) method and TDS (Thermal Desorption Spectroscopy). From the ESR spectra, gamma-ray irradiation induced irradiation defects such as E’-centers, oxygen-hole centers which were expected to be tritium trapping sites. These irradiation defects were annihilated in the temperature range of 500-650 K. From the TDS spectra for Li2TiO3 exposed to D2 gas, the deuterium desorption behavior was found to consist of four stages, corresponding retention as the surface, in E’-center and as hydroxides bound with Ti or Li. In addition, most of deuterium was released as water form around 400, 550 and 650K. By comparison of the amounts of the deuterium retentions with or without the gamma-ray irradiation, the retention of deuterium trapped with the irradiation defects was increased by gamma-ray irradiation, indicating that the irradiation defects like E’-centers induced by gamma-ray irradiation would be one of the tritium trapping sites in tritium breeding materials. The activation energy of hydrogen isotope desorption from the E’-center was estimated to be 0.63 eV for gamma-ray irradiated Li2TiO3, showing good agreement with that of the recombination reaction between the E’-center and the oxygen-hole center. These results indicated that the tritium desorption was governed by the annihilation of the E’-centers.

F4E R&D programme and results on in-vessel dust and tritium

In a Tokamak vacuum vessel, plasma–wall interactions can result in the production of radioactive dust and H isotopes (including tritium) can be trapped both in in-vessel material and in dust. The vacuum vessel represents the most important confinement barrier to this radioactive material. In the event of an accident involving ingress of steam to the vacuum vessel, hydrogen could be produced by chemical reactions with hot metal and dust. Hydrogen isotopes could also be desorbed from in-vessel components, e.g. cryopumps. In events where an ingress of air to the vacuum vessel occurs, reaction of the air with hydrogen and/or dust therefore cannot be completely excluded. Due to the radiological risks highlighted by the safety evaluation studies for ITER in normal conditions (e.g. in-vessel maintenance chronic release) and accidental ones (e.g. challenge of vacuum vessel tightness in the event of a hydrogen/dust explosion with air), limitations on the accumulation of dust and tritium in the vacuum vessel are imposed as well as controls over the maximum extent of the quantity of accidental air ingress. ITER IO has defined a strategy for the control of in-vessel dust and tritium inventories below the safety limits based primarily on the measurement and removal of dust and tritium. In this context, this paper will report on the efforts under F4E responsibility to develop a number of the new ITER baseline systems. In particular this paper, after a review of safety constraints and ITER strategy, provides the status of: (1) tasks being launched on diagnostics for in-vessel dust inventory measurement, (2) experiments to enrich the data about the effectiveness of desorption of tritium from Be at 350 °C (divertor baking aiming to release significant amount of tritium trapped in Be co-deposit), (3) on-going R&D programme (experimental and numerical simulation) at FZK, CEA and ENEA on in-vacuum vessel H2 dust explosion.

Studies on the Characteristic of Zirconium–Tritium Reaction

The p–c–T curves of tritium absorption and desorption of zirconium were measured using the method of step equilibrium by stepping up the tritium quantity on an experimental apparatus of metal hydride. The p–c–T curves for tritium have one plateau at temperature range from 450 to 500°C and two plateaus at temperature above 600°C. The thermodynamic parameters of the different phases were determined according to the van’t Hoff equation. The hysteresis effect was observed in reversible process of tritium absorption and desorption of zirconium on our experimental condition. The tritium absorption behavior by zirconium in the temperature range from 450 to 620°C and desorption behavior of zirconium in the temperature range from 775 to 875°C have been investigated. A method of the reaction rate analysis was proposed and examined for determining the rate constant. The apparent activation energy obtained by this analysis for the absorption and the desorption were (−16.8 ± 0.8) kJ·mol−1 and (57.7 ± 1.6) kJ·mol−1, respectively.

Effective desorption of tritium from diverse solid matrices and its application to routine analysis of decommissioning materials

Tritium extraction from materials is most commonly carried out using oxidative thermal desorption in purpose-built furnace systems and typically involves trapping the product in a water bubbler which is sampled for measurement using liquid scintillation counting (LSC). The performance of perhaps the most widely used commercial system, the Raddec Pyrolyser, has been evaluated for a broad range of sample types. Several parameters that were expected to affect tritium desorption and recovery were systematically studied. These included sample heating rates and end-point temperatures, carrier/oxidant gas flows, catalyst temperature, bubbler trapping and carry-over/memory effects. A catalyst such as platinised-alumina is used to ensure the quantitative oxidation of volatile combustion products to HTO and CO 2. This also ensures that the trapped decomposition products do not colourise the bubbler solutions that are subsequently sampled for LSC. Tritium evolution profiles were determined for a range of sample types and were obtained by systematically changing bubblers at a set of progressively increasing temperatures. These experiments showed the maximum heating temperature and total combustion time required for the complete recovery of tritium from samples was dependent on the sample matrix types and the 3H form. These evolution profiles need only be determined once and are readily transferable to other Pyrolyser systems. For example tritiated water is rapidly liberated from samples at temperature around 100 °C whereas 3H substituted for structural H in organic species can require a temperature in excess of 300 °C to be released. Higher temperatures (up to 900 °C) are needed to liberate 3H originating from neutron capture reactions on trace Li or B within a material (e.g. reactor graphite or concrete). The furnace system investigated is highly effective at extracting tritium and 14C from all sample types studied (soil, sediment, biota, wood, metal, plastic, concrete, graphite, etc.) and overall it demonstrates high and reproducible recoveries.

Temperature-programmed desorption of tritium loaded into beryllium

The influence of grain size and amount of beryllium oxide BeO on the tritium release characteristics of the S-65H and I-220H beryllium grades was investigated. The beryllium samples were loaded with hydrogen at a temperature of 1123 K for 6 h at a pressure of 4 bar in a gas mixture of molecular protium with 45 wppm tritium. The tritium release measurements were performed using a stepped heating ramp with steps of 50 K in the temperature range of 473–1373 K with a duration of each step of 1 h. Five major release peaks were observed. Interpretations of the nature of the peaks are suggested. In particular, we ascribe the low-temperature peaks (300–723 K) to desorption from the surface and connect the shift of the peak at elevated temperature (900–1000 K) with the enhanced diffusion of hydrogen along grain boundaries. The high temperature peak (1123–1373 K) is attributed to dissociation of beryllium hydroxide Be(OH)2, which is formed on the surface and grain boundaries of a sample in a reaction of hydrogen isotopes with beryllium oxide.

Tritium sorption and desorption from JET beryllium tiles under temperature, electron radiation and magnetic field

Tritium release at annealing of samples cut from beryllium tiles exposed to D + D, D + T plasma in the Joint European Torus (JET) was investigated under 5 MeV fast-electron radiation at the dose rate 14 MGy h −1 and in a magnetic field of 1.7 T separately and simultaneously in order to evaluate possible effects of these factors. Abundances of chemical forms of tritium—molecular T 2 (44%), atomic T 0 (42%), and ionic T + (14%) and their distribution were determined in the plasma-exposed beryllium samples with lyomethods. Fast-electron radiation considerably increased the fractional tritium release at annealing by a factor of approximately 5. The magnetic field increased the fractional tritium release at annealing both without and under electron radiation by approximately 25%. Sorption of tritium in samples cut from an unused beryllium tile was investigated at about 1 Pa of T 2 at 773 K for 3 h both with and without simultaneous electron radiation and/or magnetic field. The amount of the tritium sorbed was found to be 3.3–6.2 kBq cm −2. Most of the tritium sorbed was localized in subsurface layers up to a depth of 80 μm of the beryllium samples.

Effects of H2O and H2O2 on Thermal Desorption of Tritium from Stainless Steel

Tritiated stainless steel was subjected to thermal desorption at various temperatures, different temperature profiles, and in the presence of different helium carrier gas additives. In all cases the identities of the desorbing tritiated species were characterized as either watersoluble or insoluble. The samples were found to contain 1.1 mCi±0.4 mCi. Approximately ninety-five percent of this activity was released in molecular water-soluble form. Additives of H2O or H2O2 to dry helium carrier gas increase the desorption rate and lower the maximum temperature to which the sample must be heated, in order to remove the bulk of the tritium. The measurements validate a method of decontamination of tritiated steel and suggest a technique that can be used to further explore the mechanisms of desorption from tritiated metals.

On the mechanism of tritium desorption from stainless steel

Stainless-steel specimens (15×15×0.5mm3) containing homogeneously distributed tritium were used for the experiments. To investigate the fate of tritium in the surface region, about 70μm were chemically etched and the regrowth of tritium on the top surface layer determined as function of time by β-ray-induced x-ray spectrometry. Simultaneously, the specimen was flushed with an argon gas stream at ambient temperature and the released tritium collected in water bubblers. The chronically liberated tritium retained in the bubblers was quantified in regular intervals using a liquid-scintillation counting technique. It was found to consist predominantly of tritiated water (99%); the rest probably being tritium-containing hydrogen. In this work, a model for the transport of hydrogen is proposed that allows quantitative evaluation of the experimental results. It is demonstrated that the numerical simulation accurately predicts the amount of tritium progressively trapped on the surface and that slowly released at ambient temperature from the bulk of the stainless-steel material.

Water diffusivity in human nail plate

The diffusivity of water in human nail at 32 degrees C was determined for cadaveric, human finger nails having water contents ranging from 0.536 g H(2)O/g dry nail to 0.035 g H(2)O/g dry nail by measuring the desorption of tritiated water from nails suspended in water or in the vapor phase above salt solutions yielding a range of relative humidities (RH). Diffusivity increased with increasing RH from (7.7 +/- 1.3) x 10(-10) cm(2) s(-1) at 15% RH to (3.2 +/- 1.1) x 10(-7) cm(2) s(-1) in the liquid phase study at 100% RH, a more than 400-fold increase. The diffusivity values, which may be understood in terms of the equilibrium water content of the nail and a free volume theory for diffusion in hydrophilic polymers, were consistent with water diffusivities measured in other keratinized tissues including wool, horn and the corneocyte phase of stratum corneum. Analysis of the tritium desorption data was complicated by a tritium exchange process between (3)H(2)O and nail keratin, the kinetics of which are presented in part. The combination of the concentration-dependent water diffusivity with the natural water activity gradient in nail in vivo leads to the prediction of a nonlinear steady-state water concentration profile in human nail in vivo which, in turn, has implications for ungual drug delivery.

Determination of the order of surface reactions in Li 2O

An analysis of the different surface reactions taking place in Li 2O was performed in order to determine whether adsorption and desorption of tritium are first or second order reactions. Data from BEATRIX-II Phase I and CRITIC-I were used as basis for calculations. It was found that only second order adsorption/desorption on the surface of Li 2O can predict the tritium behavior observed experimentally.

Studies on the Characteristic of Titanium–Tritium Reaction

The p–c–T curves of tritium absorption and desorption of titanium were measured using the method of step equilibrium by stepping up the tritium quantity on an experimental apparatus of metal hydride. The p–c–T curves for tritium have one plateau at temperature below 300°C and two plateaus at temperature above 300°C. The thermodynamic parameters of the different phases were determined according to the van’t Hoff equation. The hysteresis effect was not observed in reversible process of tritium absorption and desorption of titanium on our experimental condition. The tritium absorption behavior of titanium in the temperature ranging from 550°C to 750°C and desorption behavior of titanium in the temperature ranging from 350°C to 550°C have been investigated in a constant volume system. A method of the reaction rate analysis was proposed and examined for determining the rate constant. The apparent activation energy obtained by this analysis for the absorption and the desorption were 155.5 ± 3.2 kJ mol−1 and 62.1 ± 1.6 kJ mol−1 respectively.

Desorption of tritium and helium from high dose neutron irradiated beryllium

The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5–16) × 10 22 cm −2 at 70–100 °C and 380–420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250–1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312–445 °C and 500–740 °C, respectively. It is shown that most tritium (∼82%) from sample irradiated at 70–100 °C releases in temperature range of 312–700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380–420 °C, tritium release starts at a higher temperature ( T s > T ann = 445 °C) and most of the tritium (∼99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at T ann > 600 °C and in helium release – at T ann > 750 °C.

Deuterium, tritium, and helium desorption from aged titanium tritides

Six samples of tritium-aged bulk titanium have been examined by thermal desorption and isotope exchange chemistry. The previously reported discovery of a lower temperature hydrogen desorption state in these materials has been confirmed in one new sample. The samples’ helium release shows the more severe effects obtained from longer aging periods, i.e. higher initial He/M ratios. Several of the more aged samples were spontaneously releasing helium. The hydrogen isotope and helium desorption of the samples will be discussed.

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100–600°C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300–600°C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150°C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.