Tritium Release Rate Research Articles

Analysis of tritium evolution in Al2O3-coated FeCrAl fuel rods for PWRs

FeCrAl is one of the ATF cladding solutions due to its excellent oxidation resistance. In this work, a 2D numerical simulation method was developed to study tritium release from the nuclear fuel pellets to the fuel-cladding gap, tritium adsorption on the inner surface of the cladding, and permeation through Al2O3-coated FeCrAl cladding into the coolant. For Al2O3-coated fuel rods, based on Sievert’s law, tritium concentration at the pellet-cladding gap is much higher than that at the inner cladding surface for most of the time under normal conditions, which slows down the tritium release rate from the fuel rods. With Al2O3 coating, tritium release from the fuel rods remains rising throughout the lifetime. The release rate is calculated to be low at the end of the lifetime, indicating that the Al2O3 coating prevents tritium release. As the coating thickness decreases, the flux of tritium release is found to be increased. Locally uncoated drawbacks lead tritium in the gap to be preferentially adsorbed by the inner surface of FeCrAl, then diffused in FeCrAl, and finally released outside the fuel rod. Still, more actual irradiation operations and test data are needed.

Tritium and deuterium release behavior of Li2TiO3-0.5Li4SiO4–Pb ceramic

The neutron multiplier lead (Pb) containing Li2TiO3-0.5Li4SiO4–Pb was prepared, which aimed at the improvement of the tritium release performance of the lithium ceramic without an increase in lithium density. Li2TiO3-0.5Li4SiO4–Pb and that without Pb were irradiated at Kyoto University Research Reactor (KUR) with the thermal neutron flux of 5.5 × 1012 n s−1 cm−2 and the fast neutron flux of 1.2 × 1012 n s−1 cm−2. Fast neutrons can trigger the Pb (n, 2n) reaction. The tritium release experiments were conducted using the tritium thermal desorption spectroscopy (tritium-TDS) system. Compared with Li2TiO3-0.5Li4SiO4, the tritium release rate at the lower temperature range was higher than that for Li2TiO3-0.5Li4SiO4–Pb. The tritium-TDS spectrum was also shifted to the lower temperature side. The isothermal heating experimental results confirmed that tritium in the form of gas species (HT) was controlled by the diffusion process, while the trapping and de-trapping process contributed to the total tritium release. Further, 3 keV D2+ implantation with the fluence of 2.0 × 1022 D m−2 and subsequent TDS measurements were performed. It was shown that the D2 release at the lower temperature side was increased for Li2TiO3-0.5Li4SiO4–Pb. Meanwhile, HD and HDO release at higher temperature side was significantly decreased.

Tritium decontamination in vacuum vessel after deuterium plasma experiment due to air exposure

The deuterium plasma experiment using Large Helical Device (LHD) has been conducted since 2017. A small amount of tritium was produced in the vacuum vessel and a part of tritium remained in the plasma-facing materials. After the deuterium plasma experimental campaign, the vacuum vessel was opened for the LHD maintenance activity. Then, the plasma-facing material was exposed to room air and remained tritium was released at ambient temperature. The tritium release rate increased rapidly immediately after the vacuum vessel was opened to the atmosphere, but it turned to decrease soon after. After several weeks of opening the vacuum vessel, the tritium release rate became low and almost constant. The total tritium decontamination rate, defined as the ratio of tritium released by air exposure to tritium produced during the plasma experimental campaign, was achieved at several percent. The air exposure operation is expected to have a decontamination effect due to isotope exchange reactions on the surface of the plasma-facing material.

Experimental Measurement and Multiphysics Simulation of Tritium Transport in Neutron-Irradiated Flibe Salt

Predicting the distribution and release of tritium remains a technical challenge for advanced nuclear reactors with molten Flibe (2LiF-BeF2) salt coolants. Tritium transport models, which are currently used to forecast release behavior, are limited by uncertainty in Flibe-related tritium transport properties and by a lack of relevant benchmark experiments to test input parameters and solution methods. A new test facility has been developed at the Massachusetts Institute of Technology Research Reactor (MITR) to irradiate a molten Flibe target in an ex-core neutron beam port to further investigate tritium transport mechanisms at prototypical reactor conditions. The experiment monitored the time-dependent release of tritium from the salt-free surface and the permeation rate of tritium through the stainless steel Flibe-containing test stand. Measured results were benchmarked with a multiphysics tritium transport simulation to resolve complex effects in the test. Trends in tritium release rates from the irradiation were in agreement with the multiphysics simulation of the test, which combined computational fluid dynamics, radiative heat transfer in participating media, and tritium transport in STAR-CCM+.

Assessment of the level of annual intake of tritium into the air from the surface of the lake Kyzyltash

Сайт «Озёра России». URL: https://ozera.info/lakes/ about/gvr/list-lakes/kyzyl-tash (Дата обращения 25.03.2022 г.) Екидин А.А., Антонов П.Л., Жуковский М.В. Оценка загрязнения атмосферы тритием при испарении воды с поверхности промышленных водоемов // Вопросы радиационной безопасности. 2012. № 3. С. 3-10. Андреева М.А. Озера среднего и Южного Урала. Челябинск, 1975. 270 с. Толстикова В.С. Кузнецов В.Н. Ядерное наследие на Урале: исторические оценки и документы. Екатеринбург, 2017. 400 с. Баторшин Г.Ш., Мокров Ю.Г., Алексахин А.И. Концепция вывода из эксплуатации водое мов-хранилищ ЖРО ФГУП «ПО «Маяк». VII Международный промышленный форум «АТОМЭКО – 2013». М., 2013. Радиационная обстановка на территории России и сопредельных государств в 2020 г. Ежегодник. Обнинск, 2021. 239 с. Чеботина М.Я., Николин О.А., Смагин А.И., Мурашова Е. Тритий в водоёмах производственного и комплексного назначения в районе ПО «Маяк» на Урале // Водное хозяйство России. 2011. № 4. С. 75-84. Vostrotin V.V., Yanov A.Y., Finashov L.V. Tritium in environmental objects in the area affected by FSUE Mayak Production Association in 2014-15 // Journal of Radiological Protection. 2021. Vol. 41, № 2. P. S56-S66. DOI 10.1088/1361-6498/ abe8c8. Иванов Н.Н. Об определении величины испаряемости // Известия Всесоюзного географического общества. 1954. Т. 86, № 2. вып. 1. С. 189–196. Сайт компании ООО «Расписание Погоды». URL: https:// rp5.ru. (Дата обращения 25.03.2022). Busigin A., Kalyanam K.M., Sood S.K. Estimation of the tritium release rate from a spill of tritiated water // Fusion Technology. 1992. Vol. 21, No 2P2. P. 512-517. DOI: 10.13182/FST92-A29798. Yoshinori S. Lake water mixing and tritium balance in lake Chuzenji // Geographical Review of Japan. 1983. Vol. 56-10. P. 667-678. The paper presents the results of research of the levels of tritium volume activity in lake Kyzyltash, a water storage reservoir for liquid radioactive waste from Mayak Production Association. Insufficient knowledge of the ways of migration of tritium from the reservoir and their quantitative assessment in modern conditions does not allow to fully objectively assess and predict the impact of the object of “nuclear heritage” on the population of Ozersk and the environment. The research was aimed at obtaining scientific results, which will be the starting point for the formation of a tritium monitoring program for industrial reservoirs by Mayak Production Association. The purpose of the study was assessement of the dynamics of tritium volume activity in the lake Kyzyltash, assessment of water balance parameters and the level of tritium intake into the air from its surface during the ice-free period. A total of 33 water samples were collected. The average tritium volume activity in 2021 made 1,57Е+04 Bq/l, SD – 1,1Е+03 Bq/l that more than 2 times exceeds the intervention level. A trend of decrease in the average monthly tritium volume activity in the reservoir during the observation period was revealed. In the course of calculations, it was stated that in the course of evaporation during the ice-free period the 196 TBq of tritium entered the air from lake Kyzyltash surface that is comparable with releases from all the Russia facilities (179 TBq) except Mayak Production Association. The obtained values of evaporated tritium make 13% from atmospheric tritium releases of Mayak Production Association in 2020. A total of ~12,2 million m3 had evaporated from the water reservoir surface in the study period. The model of dependence of tritium volume activity from time in lake Kyzyltash in 2021 and using of exponentially decreasing function allowed defining water balance of lake Kyzyltash in the ice-free period with 6,3 million m3 deficiency. Research of the model parameters allow prediction of sanitary and hygienic characteristics of the water reservoir from the point of view of radiation safety and calculation of water balance of the reservoir in certain periods of time.

Assessment of tritium effluent from Prototype Generation IV Sodium-cooled Fast Reactor

A tritium transport model for a pool-type sodium cooled fast reactor was developed for estimation of tritium effluents from the Prototype Generation IV Sodium-cooled Fast Reactor (PGSFR). For this purpose, the transport model developed for Experimental Breeder Reactor II and Fast Flux Test Facility was updated and validated. The tritium production rate from ternary fission and activations in PGSFR were calculated by the Argonne Reactor Computation (ARC) code suite for fast reactor analysis, and due to the potential uncertainties in source terms and variations of cold trap performance and material permeability during reactor operation, the tritium distributions and tritium effluents were estimated for several bounding cases. The upper bounding gaseous and liquid tritium release rates from PGSFR are ~ 440.7 Ci/year and 14.7 Ci/year, respectively. The gaseous tritium effluent is about few orders of magnitude smaller than the regulatory constraint of the tritium concentration in the air. The peak tritium concentration in the steam-generator feedwater is ~ 1.1x10-3 μCi/g, which is required to be diluted prior to releasing as tritiated water. The accumulated tritium in the cold trap after five cycle operations is ~ 52,000 Ci. Because of significant conservatism in these upper bounding estimations, it is expected that the tritium effluent at the nominal case with plausible source terms would be about 2 orders of magnitude smaller than that of the upper bounding case.

Tritium release behavior from neutron-irradiated FLiNaBe mixed with titanium powder

FLiNaBe is a promising liquid blanket material for a fusion reactor. Since hydrogen solubility for FLiNaBe is low, one concern is the permeation of bred tritium. In order to increase effective solubility of hydrogen, the addition of Ti powder was proposed. In this study, the influence of Ti addition on the behavior of tritium generated inside FLiNaBe by neutron irradiation was investigated. The samples of FLiNaBe and FLiNaBe mixed with Ti of 5 wt% and 0.5 wt% were prepared and the solid-state samples were irradiated by thermal neutrons at Kyoto University Research Reactor. The neutron-irradiated samples were heated to 700 °C in an Ar gas flow. The released tritium was collected by a water bubbler and the temporal variation of TF and HT release was observed. Initially, the most of tritium was released as HT from both FLiNaBe with and without Ti but eventually, the most of released tritium was HT after long time heating from FLiNaBe with Ti. This result suggests that the addition of Ti can decrease the release of TF very much. The amount of tritium released from FLiNaBe with Ti was lower than that from FLiNaBe without Ti. This result suggests that the addition of Ti can suppress the release rate of tritium from FLiNaBe.

Evaluation of public dose from FHR tritium release with consideration of meteorological uncertainties

Tritium management is a potentially significant issue in fluoride-salt-cooled high-temperature reactors (FHRs), as these reactors can produce tritium at high rates. Potential impact of the tritium released into the environment needs to be investigated to help determine the maximum-allowable tritium-release rate from an FHR plant. In this study, a dose assessment on the public resulting from FHR tritium release was performed via computational modeling. Three potential locations for FHR construction, i.e., the Hanford site, Idaho Falls in Idaho, and Oak Ridge in Tennessee, were selected. Atmospheric tritium dispersion was modeled using computer code family GENII and a parametric study of key meteorological variables was carried out. An uncertainty analysis was performed to examine the reliability of the prediction of dose for the year 2020. It is discovered that conditions in favor of lower public dose level from FHR tritium release include low atmosphere temperature, high wind speed, high relative humidity, and high tritium release point. It is also discovered that for different geological locations, the dominance of meteorological parameters differs significantly. Among the three locations modeled, although the Hanford site might be the most suitable location for FHR construction in the past, in the near future, Oak Ridge would possess advantages in the dose assessment aspect over the other two. We assumed that the tritium release rate from an FHR plant is given at 18.5 TBq/day and compared the probability of the maximum individual dose exceeding the regulatory limit (0.4 mSv/y). According to the prediction of dose for the year 2020, this probability is extremely low. While for Idaho Falls, it is 91.62% and for the Hanford site, 44.27%. The results indicate that effective measures should be taken for tritium control in FHR.

Low-temperature preparation of nanostructured Li2TiO3 tritium breeder ceramic pebbles using CTAB-modified ultrafine powders by a mixed solvent-thermal method

Effects of cetyl trimethyl ammonium bromide (CTAB) on the formation of Li2TiO3 nanocrystals during ethanol-water solvothermal reactions were investigated. Addition of CTAB changed Li2TiO3 crystalline morphology from large to small aggregate particles due to the steric repulsion and electrostatic interaction of the CTAB ligands. Aggregate particle size of the Li2TiO3 precursor was reduced from 455.2 to 138.7 nm by preventing their agglomeration after CTAB addition. This was followed by sintering the green body prepared with the CTAB-modified precursor powders using an indirect wet method, then the nanostructured Li2TiO3 ceramic pebbles were obtained at relatively low temperatures from 750 to 800 °C for 4 h. Nanostructured Li2TiO3 ceramic pebbles had small grains and a large number of grain boundaries, which might potentially improve the radiation tolerance, tritium release rate, and mechanical properties of Li2TiO3 tritium breeder pebbles. In addition, grain growth mechanisms as function of different sintering temperatures, crush load as well as relative density of fabricated nanostructured Li2TiO3 ceramic pebbles were also investigated.

Sol-gel synthesis of lithium metatitanate as tritium breeding material under different sintering conditions

Lithium metatitanate (Li2TiO3) is a promising tritium breeding material candidate for solid blanket of D-T fusion reactors, due to its high mechanical strength, chemical stability, and tritium release rate. In this paper, Li2TiO3 powder with homogeneous crystal structure is synthesized by sol-gel method. The chemical reactions in gel thermal cracking and sintering process are studied by thermo gravimetric/differential scanning calorimetry (TG-DSC). The relationship between the sintering condition and the particle/grain size is characterized by X-ray diffraction (XRD) and scanning electron microscopy (SEM). Results show that below 673 K the gel precursor is completely decomposed and Li2TiO3 phase initially forms. The LiTiO2 by-product formed under the reductive atmosphere in muffle furnace, could be oxidized continually to Li2TiO3 at higher sintering temperature (≥1273 K) for longer sintering time (≥10 h). Both grain and particle sizes rely on a linear growth with the increase of sintering time at 1273 K. Over 1473 K, significant agglomerations exist among particles. The optimal sintering condition is selected as 1273 K for 10 h, for the purer Li2TiO3 phase (>99%), smaller grain and particle size.

Calculation of tritium release from driver fuels into primary coolant of research reactors

Increasing of tritium concentration in the primary coolant of the research and test reactors during operationhad been reported. To check the source for tritium release into the primary coolant during operation of theJMTR and the JRR-3M, the tritium release from the driver fuels was calculated by MCNP6 and PHITS. Itis clear that the calculated values of tritium release from fuels are as about 107 and 106 Bq for the JMTR andJRR-3M, respectively, and that calculated values are about 4 order of magnitude smaller than that of themeasured values. The JMTR is a tank type, and the JRR-3M is a pool type research reactor. The primarycoolants are light water, in which the coolant temperature is lower than 50ºC. These results show that thetritium release from fuels is negligible for both the reactors. In the calculations of tritium release rates intothe water by PHITS, a gaussian triton energy by ternary fissions of 235U.

Evaluation of tritium release curve in primary coolant of research reactors

Increase of tritium concentration in the primary coolant for the research and testing reactors during reactoroperation had been reported. To clarify the tritium sources, a curve of the tritium release rate into the primarycoolant for the JMTR and the JRR-3M are evaluated. It is also observed that the amount of released tritium islower in the case of new beryllium components installation, and increases with the reactor operating cycle.These results show the beryllium components in core strongly affect to the tritium release into the primarycoolant. As a result, the tritium release rate is related with produced 6Li by (n,α) reaction from 9Be, andevaluation results of tritium release curve are shown as the dominant source of tritium release into the primarycoolant for the JMTR and the JRR-3M are beryllium components. Scattering of the tritium release rate withirradiation time were observed, and this phenomena in the JMTR occurred in earlier time than that of the JRR-3M.

Analysis of tritium production in a 2 MW liquid-fueled molten salt experimental reactor and its environmental impact

Tritium release is one of the most concerning topics in nuclear power plants. Here, the tritium production in a 2 MW liquid-fueled molten salt experimental reactor (TMSR-LF1) was calculated by ORIGEN-S with an updated cross-section library generated by TRITON in SCALE 6.1.3 code system. The results show that the tritium production rate and normalized tritium production rate of TMSR-LF1 are 8.90 × 1011 Bq/day and 4.45 × 1011 Bq/(MW day), respectively. The environmental impact of tritium was analyzed via PC-CREAM 08 with an assumed 36 % release rate of tritium referring to the molten salt reactor experiment. During normal operations, the maximum tritium concentration is 1.4 Bq/m3 under normal condition, and the corresponding individual dose to the public is about 1 μSv/a; under extreme conditions, the maximum concentration and corresponding individual doses are 11.8 Bq/m3 and 9 μSv/a, respectively. Ingestion is the main exposure pathway and accounts for 62 % of the total dose. Of this, 35 % is from organically bound tritium.

Tritium Release from SS316 under Vacuum Condition

Type 316 stainless steel specimens loaded with tritium either by exposure to 1.2 kPa HT at 573 K or submersion into liquid HTO at 298 K showed characteristic thin surface layers trapping tritium in concentrations far higher than those determined in the bulk. The evolution of the tritium depth profile in the bulk during heating under vacuum was non-discernable from that of tritium liberated into a stream of argon. Only the relative amount of the two released tritium-species, i.e. HT or HTO, was different. Temperature-dependent depth profiles could be predicted with a one-dimensional diffusion model. Diffusion coefficients derived from fitting of the tritium release into an evacuated vessel or a stream of argon were found to be (1.4 ± 1.0) × 10-7 and (1.3 ± 0.9) × 10-9 cm2/s at 573 and 423 K, respectively. Polished surfaces on type SS316 stainless steel inhibit considerably the thermal release rate of tritium.

Preparation and Characterization of the Lithium Metatitanate Ceramics by Solution-Combustion Method for Indian LLCB TBM

In fusion DEMO reactors, the blanket requires lithium-containing ceramics as the tritium breeder material. Lithium metatitanate (Li2TiO3) is being considered as a promising tritium breeding material for thermonuclear fusion reactors because of its reasonable lithium atom density, prominent tritium release rate at low temperatures, low activation characteristics, low thermal expansion coefficient, high thermal conductivity, etc. Li2TiO3 will be used in the Indian Lead-Lithium–Cooled Ceramic Breeder concept to be tested in ITER. Li2TiO3 powder has been synthesized by the solution-combustion technique using a less expensive precursor of titanium, i.e., titanium dioxide (TiO2), at Institute for Plasma Research. Titanium oxynitrate [TiO(NO3)2] and lithium carbonate (Li2CO3) with citric acid fuel are used as the raw materials. The combustion reaction was carried out at citrate-to-metal ratios of 0.8 to 1.5, as well as for various pH values ranging from 1 to 5. Citric acid was used as a fuel material for the reaction. Calcination of the powder was carried out at 600°C. The powders were characterized for phase purity, grain size, and surface area using X-ray diffraction, scanning electron microscopy, and a Brunauer-Emmett-Teller surface area analyzer. Finally, Li2TiO3 pebbles were prepared by extrusion followed by spheronization with a diameter range from 1 to 1.5 mm. The details of the powder systemization, pebble formation, and their various characterizations are discussed in this paper.

Elaidate, an 18‐Carbon trans‐Monoenoic Fatty Acid, Inhibits β‐Oxidation in Human Peripheral Blood Macrophages

Consumption of trans-unsaturated fatty acids promotes atherosclerosis, but whether degradation of fats in macrophages is altered by trans-unsaturated fatty acids is unknown. We compared the metabolism of oleate (C18:1Δ9-10 cis; (Z)-octadec-9-enoate), elaidate (C18:Δ9-10 trans; (E)-octadec-9-enoate), and stearate (C18:0, octadecanoate) in adherent peripheral human macrophages. Metabolism was followed by measurement of acylcarnitines in cell supernatants by MS/MS, determination of cellular fatty acid content by GC/MS, and assessment of β-oxidation rates using radiolabeled fatty acids. Cells incubated for 44 h in 100 µM elaidate accumulated more unsaturated fatty acids, including both longer- and shorter-chain, and had reduced C18:0 relative to those incubated with oleate or stearate. Both C12:1 and C18:1 acylcarnitines accumulated in supernatants of macrophages exposed to trans fats. These results suggested β-oxidation inhibition one reaction proximal to the trans bond. Comparison of [1-(14)C]oleate to [1-(14)C]elaidate catabolism showed that elaidate completed the first round of fatty acid β-oxidation at rates comparable to oleate. Yet, in competitive β-oxidation assays with [9,10-(3)H]oleate, tritium release rate decreased when unlabeled oleate was replaced by the same quantity of elaidate. These data show specific inhibition of monoenoic fat catabolism by elaidate that is not shared by other atherogenic fats.

Tritium Interaction with Surface Layer and Bulk of Type 316 Stainless Steel and Consequences of Aging

Stainless steel exposed to gaseous tritium characteristically shows a firmly trapped fraction of tritium in the surface layer, which is not fully removable by water at ambient temperature. Prolonged thermal treatment of tritium-loaded specimens at <443 K causes substantial depletion of the bulk but almost no depletion of the surface layer. For complete removal of hydrogen isotopes from the bulk and the surface, temperatures exceeding 573 K are necessary. Upon chemical etching virtually all tritium trapped in the surface layer appears in the etching solution as tritiated water. Following removal of the layer by chemical etching, the tritium-rich layer reappears after months of aging at ambient temperature with nearly the original tritium activity. Comparison of chronic tritium release rates into liquid water before and after etching reveals that the surface layer only marginally influences the rate. X-ray photoelectron spectroscopy provides evidence that during prolonged aging the surface layer continues to grow while at the same time trapping a fraction of bulk tritium released at ambient temperature. Experimental results suggest different mechanisms of hydrogen uptake and release by the bulk and surface layers. Inference of tritium activity in the bulk of aged or heat-exposed stainless steel material from surface activity measurements may depart significantly from reality.

Improvement of hydrogen isotope exchange reactions on Li4SiO4 ceramic pebble by catalytic metals

Abstract Li4SiO4 ceramic pebble is considered as a candidate tritium breeding material of Chinese Helium Cooled Solid Breeder Test Blanket Module (CH HCSB TBM) for the International Thermonuclear Experimental Reactor (ITER). In this paper, Li4SiO4 ceramic pebbles deposited with catalytic metals, including Pt, Pd, Ru and Ir, were prepared by wet impregnation method. The metal particles on Li4SiO4 pebble exhibit a good promotion of hydrogen isotope exchange reactions in H2–D2O gas system, with conversion equilibrium temperature reduction of 200–300 °C. The out-of-pile tritium release experiments were performed using 1.0 wt% Pt/Li4SiO4 and Li4SiO4 pebbles irradiated in a thermal neutron reactor. The thermal desorption spectroscopy shows that Pt was effective to increase the tritium release rate at lower temperatures, and the ratio of tritium molecule (HT) to tritiated water (HTO) of 1.0 wt% Pt/Li4SiO4 was much more than that of Li4SiO4, which released mainly as HTO. Thus, catalytic metals deposited on Li4SiO4 pebble may help to accelerate the recovery of bred tritium particularly in low temperature region, and increase the tritium molecule form released from the tritium breeding materials.

On the fate of tritium in thermally treated stainless steel type 316L

Several type 316L stainless steel specimens of 6mm thickness were charged with tritium at 473K at Joint European Torus (JET) using five sets of conditions. Isothermal tritium release rates were investigated at Hydrogen Isotope Research Centre (HRC) over extended periods of time at 473, 573, or 673K constant temperature. The HTO/HT ratio of the liberated tritium was generally high, but decreased with decreasing release temperature. Nearly complete release of tritium required additional prolonged heating at 1073K. Chemical etching and beta-ray-induced X-ray spectrometry measurements carried out at HRC provided complementary information on the tritium distribution in surface and bulk of thermally treated specimens. Whereas the thickness of the material and initial distribution of tritium in its bulk were found to play an important role for expedient thermal decontamination, the influence of the type of purge gas was only minor. Experimental evidence for tritium grain boundary diffusion is provided. Implications of the results for waste conditioning are discussed.

Lithium orthosilicate surfaces: Characterization and effect on tritium release

Within the European Union, slightly hyperstoichiometric lithium orthosilicate has evolved as one of the candidate solid breeder materials for the helium cooled pebble bed blanket, which will be tested in ITER. In the past, several long-term irradiation experiments proved that lithium orthosilicate shows excellent tritium release behavior when purged with helium with 0.1% hydrogen.In this work, short-term irradiation experiments at the High Flux Reactor in Petten with two standard pebble qualities, as-received and heat-treated lithium orthosilicate were carried out to investigate possible differences in tritium release. Since the surface of the pebbles may play a significant role, especially for short-term irradiation, and the mechanism of desorption from the surface determines the chemical form and very likely the release rate of tritium, the aim of this work is to link the chemical composition of the surface to the Thermal Programmed Desorption (TPD) tritium release experiments. Therefore X-ray photoelectron spectroscopy and time-of-flight secondary ion mass spectrometry were applied to characterize the surface of the unirradiated pebbles and to obtain sputter depth profiles of up to 800nm. In addition, X-ray diffractometry was used to determine the composition of phases of the samples before and after irradiation.