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The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain ...

A PVT-c method, which is widely used to measure the amount of tritium, is not practical for a UHx bed because the process is time-consuming and includes tritium desorption and transfer processes, which can increase the potential risk of tritium release. In this study, a natural circulation (NC) loop, which can compensate for the issues of a PVT-c method, was numerically designed and evaluated for tritium accountability (TA) of a UHx bed. The numerical evaluation shows that the NC loop can be applied to a UHx bed for TA. Moreover, the characteristics of various parameters including the geometry, and initial pressure of the NC loop, were obtained. These characteristics can be commonly applied to any type of NC loops using heat sources and sinks.

We investigated the impact of the prolonged exposure of rat hippocampal synaptosomes to CXCL12 (3 nM) on the NMDA-mediated release of [3H]D-aspartate ([3H]D-Asp) or [3H]noradrenaline ([3H]NA). Synaptosomes were stimulated twice with NMDA/CXCL12 and the amount of the NMDA-evoked tritium release (S1 and S2) quantified to calculate the S2/S1 ratio. The S2/S1 ratio for both transmitters was drastically decreased by 3 nM CXCL12 between the two stimuli (CXCL12-treated synaptosomes) in a AMD3100-sensitive manner. The phosphorylation of the GluN1 subunit in Ser 896 was reduced in CXCL12-treated synaptosomes, while the overall amount of GluN1 and GluN2B proteins as well as the GluN2B insertion in synaptosomal plasmamembranes were unchanged. We conclude that the CXCR4/NMDA cross-talk is dynamically regulated by the time of activation of the CXCR4s. Our results unveil a functional cross-talk that might account for the severe impairments of central transmission that develop in pathological conditions characterized by CXCL12 overproduction.

International audience; ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m. Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

The effects of silicon (Si) on the release of tritium over lithium titanate (Li2TiO3) surface (001) are investigated using periodic density functional theory (DFT) calculations. DFT results show th...

Tritium release behavior in a tungsten sample after exposing to tritium ions with energy about 200 eV created by glow discharge has been studied by both β-ray–induced X-ray spectrometry (BIXS) and ...

The Chinese (CN) Helium Cooled Ceramic Breeding (HCCB) Test Blanket Module (TBM) (CN HCCB TBM) set with its ancillary systems will demonstrate the feasibility of in-pile tritium production/breeding in ITER for fuel self-sufficiency and high-grade fusion energy conversion to heat and extraction for a future magnetic confined fusion reactor. Tritium release in some major components of the recently designed TBM systems through permeation and natural leakage was estimated with simple diffusion/permeation and leak rate calculation models. Results showed that because of the tritium permeation barrier coating for tritium confinement in some tritium containments, total tritium release to the environment by permeation in the CN HCCB TBM and ancillary systems will be kept well below 2 Ci/full-power day. However, tritium release through natural leakage from components can be neglected compared with permeation. Equipped with ITER tritium safety guarantee facilities like the tritium monitoring and detritiation...

Tritium release behavior from the Li 4 SiO 4 pebbles with high densities (∼96%TD) and large grain sizes (100–300 μm) fabricated by a melt-based method (the M-OSi sample) was investigated through out-of-pile experiments. Another batch of Li 4 SiO 4 pebbles with relatively low densities (∼86%TD) and small grain sizes (10–50 μm) fabricated by a wet method (the W-OSi sample) was used for comparative study. Comparing with the W-OSi sample, the temperature of tritium release from the M-OSi sample was found much higher. Moreover, the fraction of tritium gas released from the M-OSi sample was much larger, especially under helium purge gas. The big differences between the characteristics of tritium release from the two batches of samples can be explained reasonably by the effect of grain size, implying that the grain size played an important role in the tritium release behavior. This study can provide a guideline for optimizing the fabrication process of Li 4 SiO 4 pebbles.

Identifications of tritium trapping states in neutron-irradiated Li 1.8 TiO 2.9 (lithium-depleted Li 2 TiO 3 ) were carried out by the out-of-pile tritium release behavior. Tritium release behaviors for neutron-irradiated Li 2 TiO 3 and tritium gas-exposed TiO 2 were also measured for comparison. Among the tritium release spectra for these samples, three tritium release peaks were appeared. By the kinetic analyses of tritium release behaviors, the Arrhenius parameters for three peaks were evaluated. Especially for Li 1.8 TiO 2.9 , there were two tritium release peaks, and the peak in lower temperature region was assigned to the tritium release controlled by the diffusion process in Li 2 TiO 3 structure. The other tritium release peak, which was hardly appeared for Li 2 TiO 3 , was assigned to the release of tritium trapped as hydroxyl groups in Li 1.8 TiO 2.9 , indicating that lithium-depletion would result in the formation of hydroxyl groups in Li 2 TiO 3 . Lithium vacancies existed in Li 2 TiO 3 crystal structure would promote the tritium trapping as hydroxyl groups due to the decreased charge repulsion between lithium ions and tritium ion, resulting in the difficulty of recovering tritium from Li 2 TiO 3 effectively.

Solid tritium breeding materials are expected to be used in high-temperature conditions for a long time in a fusion DEMO reactor. Thus, it is important to understand how bred tritium releases from the long-term heated material under the environment as close to a DEMO condition as possible to establish a tritium fuel cycle and keep it safe. In this work, the tritium release behavior from the irradiated Li2TiO3 pebbles that was preheated for 720 h at most was observed by heating at 1000 °C under 1000 Pa H2/Ar gas flow. The release peaks of HTO and HT were observed around 300 °C due to the desorption of the chemisorbed water. Also, the broad HTO peak was observed in a higher temperature region despite purging H2/Ar gas. This result suggests that this tritium was released without exchanging with H2 but with combining with oxygen in the pebbles. Moreover, the released tritium amount decreased as the pre-heating time. Finally, the amount of tritium that could not be released by the heating experiment was quantified by dissolving samples with an acid solution. Besides, the total tritium release ratio was discussed.

Lithium-based tritium breeding materials will be adopted to produce tritium in the D-T fusion reactor. Li2TiO3 and Li4SiO4 pebbles have been proposed as candidates in water-cooled ceramic breeder blanket (WCCB) and helium-cooled ceramic breeder blanket (HCCB) respectively. Biphasic ceramics of core–shell Li2TiO3-Li4SiO4 have been considered as advanced breeding materials due to combining the superiority of Li2TiO3 and Li4SiO4. The defects will be introduced into tritium breeders in the operation of fusion reactor. They have important effects on tritium release. It is necessary to carry out the experiment on irradiation defects evolution. The annihilation kinetics of defects induced by γ-ray irradiation were investigated. Fluka and Flair were used to calculate the DPA (displacement per atom) of these irradiated pebbles. EPR (Electron Paramagnetic Resonance) characterization and EasySpin simulation were adopted to analyze the evolution and annihilation kinetics processes of irradiation defects. The concentration of defects decreased as the annealing temperature increased. There were still a certain amount of defects in Li4SiO4 and Li2TiO3-Li4SiO4 when the defects in Li2TiO3 disappeared by annealing. Li2TiO3 pebbles have a better irradiation stability than that of Li4SiO4 pebbles. According to the results of EasySpin simulation, the defect concentration of E’-center and O-related was obtained respectively. The kinetics parameters for the defects of E’-center and O-related center in Li2TiO3 and Li4SiO4 were acquired. The evolution behavior of Ti3+ in core–shell Li2TiO3-Li4SiO4 contributes to the recovery of defects. The correlation between annihilation of irradiation defects and tritium release was discussed. The anti-irradiation damage stability of three kinds of ceramic breeders were evaluated. This work carries out the research on the defect kinetics of the new core–shell structure tritium breeder. Meanwhile, a comprehensive comparison of the annihilation properties of three promising breeders has been made. Li2TiO3 pebbles present excellent tritium release performance due to higher annihilation rate constant. Compared the performances of these breeder pebbles, Li2TiO3 plays a positive role in irradiation tolerance and tritium release performance, and Li4SiO4 has higher lithium density which benefit for tritium production.

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.

Tritium release into the primary coolant of the JMTR and the JRR-3 M had been studied, and it is found that tritium recoil release from the chain reaction of beryllium neutron reflectors is dominant. To prevent the tritium recoil release, Al, Ti, V, Ni and Zr are selected as the candidate tritium recoil barrier materials in this feasibility study. It is clear that 20∼40 μm thickness is required depending on the material to reduce by 3 orders, and that an impact on the effective multiplication factor is about 0.2 % at most. Total evaluation including the activities, fabrication and usage experiences, suggests the selection of Al as the first candidate may have the least development risk as the tritium recoil barrier.

The tritium release experiments of Li4SiO4 pebbles, Li2TiO3 pebbles and Li4SiO4-Li2TiO3 biphasic pebbles were conducted to investigated the tritium release behavior of Li4SiO4-Li2TiO3 biphasic pebbles. The results indicate that the biphasic ceramic pebbles with uniform structure and small grain size exhibit good tritium release performance. The tritium release temperature is around 450 °C, and most of tritium is released in the form of tritiated water. The tritium release behavior of Li4SiO4-Li2TiO3 biphasic pebbles is controlled by the release of both Li2TiO3 and Li4SiO4, and the interfaces between Li2TiO3 and Li4SiO4 can promote the tritium release. Besides, the surface reaction is the key process affected the tritium release of the Li4SiO4-Li2TiO3 biphasic pebbles, because the release behavior is significantly affected by the composition of purge gas. Under 0.1% H2+He purge gas, with the combined effect of the two-phase interface and the isotope exchange reaction, the tritium release properties of the biphasic pebbles are promoted.

The correlation between water desorption and tritium release from Li 4 SiO 4 pebbles was studied by temperature programmed desorption. The released water and tritium from irradiated samples were monitored simultaneously. The main peak for tritium release from the irradiated samples that were exposed to air for more than a month, was shifted from 500 to about 250 °C, as compared to that from the unexposed samples. The peak temperatures for water desorption and tritium release overlapped very well, suggesting a strong correlation between the two processes. Accordingly, a two-step mechanism, involving isotope exchange between the tritium trapped on the grain surface and the surface hydroxyls (–OH), and subsequent desorption of tritiated water through recombination of the –OH/–OT groups, was proposed to explain the tritium release behavior for the air-exposed samples. It is believed that the formation and desorption of surface hydroxyl groups at 200–300 °C can affect the behavior of tritium release from Li 4 SiO 4 significantly.

DT + ion irradiation with energy of 0.5 and 1.0 keV was performed on helium pre-irradiated tungsten and the amount of retained tritium and the long-term release of retained tritium in vacuum was investigated using an IP technique and BIXS. Tritium retention and long-term tritium release were significantly influenced by helium pre-irradiation. The amount of retained tritium increased until it reached 1 × 10 17 He/cm 2 , and at 1 × 10 18 He/cm 2 it became smaller compared to 1 × 10 17 He/cm 2 . The amount of retained tritium in tungsten without helium pre-irradiation largely decreased after several weeks preservation in vacuum, and the long-term release rate during vacuum preservation was retarded by helium pre-irradiation. The results indicate that the long-term tritium release and the helium irradiation effect on it should be taken into account for more precise estimation of tritium retention in the long-term use of tungsten in fusion devices.

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...

Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA.

Effective removal of deuterium and tritium accumulated in fusion reactor materials is crucial for practical and safe fusion energy. In this paper, investigations on tritium removal in China Low Activation Martensitic (CLAM) steel and China Low-activation Ferritic (CLF-1) steel through thermal desorption with/without hydrogen isotope exchange have been performed. The results indicate that the tritium radioactivity in CLF-1 steel is higher than that of CLAM steel. The tritium removal efficiency is both less than 50% at 473 K, while significantly improved after introduction of hydrogen, 82.24% for CLAM steel and 70.34% for CLF-1 steel respectively. With the increase of temperature, the tritium removal efficiency can be further improved, however, hydrogen isotope exchange can slightly enhance tritium release at 673 K. In consequence, tritium retained in both steels can be removed effectively at relatively moderate temperature (

Post-irradiation tritium release from the Pebble Bed Assembly (PBA) neutron-irradiated beryllium (Be) pebbles (∅ ≈ 1 mm; the total tritium inventory 2–4 GBq/g) was investigated on annealing in He + 0.1% H 2 . Temperature ramps of 2.3 K/min followed by annealing for 1 h at 1310–1320 K resulted in complete detritiation of the PBA Be pebbles. The tritium burst release occurred after a time lag of 1–6.5 h during constant temperature anneals at 1089–1180 K, with the release time measured after reaching the annealing temperature. Tritium burst release was also observed to occur during cool down. Anneals at 1089 K for 8 h and at 1045 K for 23 h caused detritiation of the PBA Be pebbles by ⩾75%. Effective values of the tritium diffusivity were estimated for the PBA Be pebbles as 2.6 × 10 −13 m 2 /s for 1045 K; 4.7–5.1 × 10 −13 m 2 /s for 1089 K.

In the HIDOBE (HIgh DOse irradiation of BEryllium) irradiation program, various grades of constrained and unconstrained beryllium pebbles, beryllium pellets and titanium-beryllide samples are irradiated in the High Flux Reactor (HFR) in Petten at four different temperatures (between 698 K and 1023 K) for 649 days [1] . The first of two HIDOBE irradiation experiments, HIDOBE-01, was completed after achieving a DEMO relevant helium production level of 3000 appm and the samples are retrieved for postirradiation examination (PIE). This work shows preliminary results of the out-of-pile tritium release analysis performed on different grades of irradiated beryllium pebbles (different in size). Relationships between irradiation temperature, tritium inventory and microstructural evolution have been observed by light microscopy and scanning electron microscopy.

Out-of-pile tritium release experiments under different water uptake contents and purge gas chemistry were performed on Li4SiO4. Water measurement was performed on samples under different experimental procedures. It was found that water was adsorbed on the sample during its transferring and storage process. A strong dependence of tritium release behavior on water uptake was determined. By doping H2 in the sweep gas, the formation of water in orthosilicate was observed in addition to the isotope exchange reaction with H2 gas. Thermal desorption peaks of the water formation reaction and H2 isotope exchange reaction appeared at 668 °C and 463 °C, respectively, at ramping rate of 5 °C/min.

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.

Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the feasibility of any PbLi based tritium breeding blanket (BB). Particularly in DEMO, high tritium extraction efficiency will be required in order to keep low the tritium concentration in the Pb-16Li loop. This is essential to minimize tritium release into the environment and tritium permeation from BB into the primary cooling system. In addition, the tritium extraction process needs to be highly reliable in order not to impact negatively on the operation of the whole fusion reactor, ITER or DEMO.In the present paper, a critical review of the main candidate technologies for tritium extraction from Pb-16Li, particularly gas liquid contactors and vacuum permeators, is accomplished. The intrinsic limits and possible advantages of these technologies are presented and discussed, in the light of considerations coming directly from mathematical models describing their behaviour as well as from the experimental results so fa...

The present authors have developed the tritium release model to represent the tritium release behavior from solid breeder materials (Li2ZrO3, Li2TiO3, Li4SiO4, LiAlO2 and Li2O). It has been found that water is released from solid breeder materials into the purge gas due to desorption of physically and chemically adsorbed water and water generation reaction and that this water affects the tritium release behavior. In this study, the amount of adsorbed water and its desorption rate for Li2ZrO3 were quantified. It was found in this experiment that Li2ZrO3 has the largest adsorption amount among the solid breeder candidates. It was also observed that Li2ZrO3 has the largest water generation capacity among the solid breeder candidates. A unique reaction at around 550°C which made up approximately 80% of the capacity of water generation was also observed. It is considered that the phase change of ZrO2 at around 550°C supplies oxygen to promote water generation reaction. Tritium release behavior from Li2...

The liquid Li17Pb83 alloy is a prominent breeder material for use in a fusion reactor. In the design of an effective tritium extraction system for the liquid lithium lead bubbler of the test blanket module of such a reactor, finding ways to strictly limit the losses of tritium and to minimize radioactive risks is very important. For this purpose, the isotope exchange process has been investigated as a means of trace tritium recovery from a model of the residue from Li17Pb83 alloy. The results indicate that the isotope exchange process is an effective means of tritium recovery from the residue of Li17Pb83 alloy, and the optimum composition of the exchange carrier gas is He + 0.1% D2. The exchange temperature and number of exchange steps are the main factors influencing the efficiency of tritium recovery from the residue. Trace tritium recovery efficiency increases with increasing exchange temperature and number of times of exchange. Tritium recovery efficiency can approach 80% when the residue is treated six times at 823 K. A gas-liquid two-phase contact model to describe the proceeding of tritium release from the liquid Li17Pb83 alloy has been derived on the basis of this experiment.

Results obtained on radiation and magnetic field (MF) effects on tritium release at annealing of the beryllium pebbles from the EXOTIC-8-3/13 irradiation are presented in this study and compared with those for other irradiated beryllium materials. Abundance ratios of chemical forms of tritium in the EXOTIC-8-3/13 beryllium pebbles were determined: T 2 – 65%, T 0 – 23%, T + – 12%. A complete detritiation of these pebbles was achieved at 1123 K for 240 min; MF of 2.35 T had no appreciable effect on the tritium release. At 991 K for 240 min, the degree of detritiation was 96.6% without MF; MF of 2.35 T decreased it to 86.7%. At 940 K for 47 min, the degree of detritiation was 60%, 5 MeV fast-electron radiation of 14 MGy/h increased it to 76%, but the simultaneous action of the fast-electron radiation and MF of 1.7 T increased it to 88%.

The need of less conservative but reliable risk assessment for accidental tritium release was recognized in the frame of present debate on nuclear energy future. While the fusion reactor is only in research phase and risk assessment are not extremely stringent, Romania faces a debate on near future of nuclear energy with heavy water reactors, known to have a large load and release of tritium. Both tritium and 14 C are key radionuclide for CANDU and are also directly involved in the life cycle of plants and animals. An interdisciplinary approach is proposed in order to better handle the environmental transfer of tritium under accidental conditions and preparatory steps are presented.

A tritium release model has been developed by the present authors. The tritium release curves estimated by this tritium model give good agreement with experimental curves for L1 4 SiO 4 , Li 2 TiO 3 , Li 2 ZrO 3 or LiAlO 2 under various purge gas conditions in our out-of-pile bred tritium release. The characteristics of tritium release behavior from various solid breeder materials carried out by us and in EXOTIC experiments at Petten are discussed in this study.

To construct a fusion reactor with high safety and acceptability, it is necessary to establish and to ensure tritium safe handling technology. Tritium should be well-controlled not to be released to the environment excessively and to prevent workers from excess exposure. It is especially important to grasp tritium behavior in the final confinement area, such as the room and/or building. In order to obtain data for actual tritium behavior in a room and/or building, a series of intentional Tritium Release Experiments (TREs) were planned and carried out within a radiologically controlled area (main cell) at Tritium System Test Assembly (TSTA) in Los Alamos National Laboratory (LANL) under US-JAPAN collaboration program. These experiments were carried out three times. In these experiments, influence of a difference in the tritium release point and the amount of hydrogen isotope for the initial tritium behavior in the room were suggested. Tritium was released into the main cell at TSTA/LANL. The released tritium reached a uniform concentration about 30 ∼ 40 minutes in all the experiments. The influence of the release point and the amount of hydrogen isotope were not found to be important in these experiments. The experimental results for the initial tritium behavior in the room were also simulated well by the modified three-dimensional eddy flow analysis code FLOW-3D.

Study of tritium and helium release from beryllium pebbles with diameters of 0.5 and 1 mm after high-dose neutron irradiation at temperatures of 686–968 K was performed. The release rate always has a single peak, and the peak temperatures at heating rates of 0.017 K/s and 0.117 K/s lie in the range of 1100–1350 K for both tritium and helium release. The total tritium release from 1 mm pebbles decreases considerably by increasing the irradiation temperature. The total tritium release from 0.5 mm pebbles is less than that from 1 mm pebbles and remains constant regardless of the irradiation temperature. At high irradiation temperatures, open channels are formed which contribute to the enhanced tritium release.

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.

CAP88-PC, Version 2.0, and NORMTRI, which are based on the Gaussian plume model, were used to estimate the chronic dose due to a continuous, ground-level, atmospheric release of tritium as tritiated water. For the same conditions the predictions of CAP88-PC were found to be higher by a factor of 3 or less than those of NORMTRI. The major differences are due to the use of higher dose coefficients in CAP88-PC andNORMTRI's method of calculating the tritium content of food.

The computer program TriStat (Tritium dose assessment for stationary release) was used to estimate the human dose under stationary release and to obtain a conservative estimate of the dose after an accidental release as well. The atmospheric behavior of tritium is described by a Gaussian dispersion model. The tritium concentration in the atmosphere, soil, vegetables and cereals were estimated on the basis of tritium inventory of the facility and the release rate of tritium. In the model description, the specific tritium concentrations for the free water component and the organic component are essential. The food chain for humans was modeled by assuming a forage compartment, a plant compartment and an animal compartment. In the model, a virtual plant and a virtual animal were defined. The calculation revealed that the exchange of HTO between atmosphere and plant leaves has a critical role for increasing the human dose both for stationary and accidental release of tritium.

The effects of irradiation damages induced by neutron irradiation on tritium release processes were studied in Li4SiO4. It was shown that the thermal annealing process of the damages consisted of two processes of a fast and a slow process. The activation energies were determined to be 0.21 eV and 1.1 eV, respectively. From the experimental results, it was suggested that the migration of tritium and displaced oxygen into the damages plays an important role in both thermal annealing processes.

To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500oC, which is within the temperature range (480 - 600oC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700oC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

Tritium release behavior from Flibe (Li2BeF4) a potential liquid tritium breeding material was investigated. The main chemical form of tritium-containing species in Flibe was controlled to HT or TF...

At the Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI), Caisson Assembly for Tritium Safety study (CATS) with 12 m3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate tritium behavior in the case where a tritium leak event should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak event should happen in a ventilated room. The RNG model was found to be valid for eddy flow calculation in the 50m3/h ventilated Caisson with acceptable engineering precision. The calculated initial and removal tritium concentration histories after intended tritium release were consistent with the experimental observations in the 50 m3/h ventilated Caisson. It is found that the flow near a wall plays an important role for the tritium transport in the ventilated room. On the other hand, tritium behavior intentionally ...

Among the various tritium transport processes in lithium ceramics, the importance and the detailed mechanism of surface reactions remain to be elucidated. The dynamic adsorption and desorption model for tritium desorption from lithium ceramics, especially Li 2 O was constructed. From the experimental results, it was considered that both H 2 and H 2 O are dissociatively adsorbed on Li 2 O and generate OH − on the surface. In the first model developed in 1994, it was assumed that either the dissociative adsorption of H 2 or H 2 O on Li 2 O generates two OH − on the surface. However, recent calculation results show that the generation of one OH − and one H − is more stable than that of two OH − s by the dissociative adsorption of H 2 . Therefore, assumption of H 2 adsorption and desorption in the first model is improved and the tritium release behavior from Li 2 O surface is evaluated again by using the improved model. The tritium residence time on the Li 2 O surface is calculated using the improved model, and the results are compared with the experimental results. The calculation results using the improved model agree well with the experimental results than those using the first model.

In the EXOTIC-8 irradiation experiment in the High Flux Reactor (HFR) in Petten, lithium orthosilicate (OSi) pebbles with 50% 6Li-content were irradiated with the aim at investigating the behaviour of breeder ceramics at DEMO relevant burn-up. Specifically, changes in the mechanical stability as well as in the tritium release properties (in-pile and out-of-pile) had to be studied. After the completion of the irradiation, lithium orthosilicate pebble samples were shipped to the Research Centre of Karlsruhe (FZK) for post-irradiation examinations. The pebbles irradiated up to about 11% 6Li burn-up showed an increase in smaller cracks in the bulk and the presence of larger through-cracks, but the amount of fragments is quite small and the pebbles maintained a very good crush load. On the basis of these results, it can be concluded that the OSi pebbles can withstand high burn-up irradiation without an unduly high fragmentation.

A new rapid, sensitive and convenient procedure is presented allowing determination of dUTPase activity. With [5-(3)H]dUTP used as the substrate, dUTPase, converts it to the corresponding monophosphate and is coupled with thymidylate synthase-catalyzed reaction, resulting in tritium release from [5-(3)H]dUMP. Following charcoal absorption of the labeled nuleotides, radioactivity of tritiated water is determined. The new assay was tested to show comparable results with a previously described assay, based on measuring dUTPase-catalyzed [5-(3)H]dUMP production.

Lithium metatitanate (Li2TiO3) is considered as one of the candidate materials for the ceramic breeder in both the ITER Breeding Blanket and the European Helium Cooled Pebble-Bed Blanket for DEMO. A wet process based on powder-gelation for the manufacture of Li2TiO3-pebbles is described. Pebble characteristics and its basic properties are given along with results of out-of-pile tritium release experiments.

To understand the release mechanism of tritium from solid materials, release rate of tritium was measured when a tritium loaded 316 stainless steel specimen was put in dry argon gas flow of atmospheric pressure at room temperature. During blowing of argon gas released products from the specimen were collected in water bubblers which were set in the downstream of the blowing circuit, and the tritium content in the bubbler water was periodically measured as a function of time by scintillation counter. More than 99% of released species from the specimen was tritiated water, HTO. The measured result of tritium release rate showed that tritium is released chronically for a long time. The chronic release rate of tritium was evaluated using the diffusion model reported by Calder and Lewin, and it was found that when a reasonable value for the bulk diffusion coefficient of tritium is assumed, the tritium release rate can be described with the diffusion flux at the surface boundary of the specimen.

A model including an isotopic exchange reaction of T+, HT, H2 and H+ dissolved in Flibe is proposed to explain the tritium release behavior from Flibe in an in-pile experiment. The temporal change ...

The two still open issues related to beryllium for fusion applications are its mechanical response under neutron irradiation and the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the “Beryllium” experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium pebble response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation on essential material properties and trying to elucidate the processes controlling the property changes.

Activities regarding tritium safety technology in the Tritium Process Laboratory (TPL) at Tokai Establishment of Japan Atomic Energy Research Institute are reviewed. Research and development of a new tritium removal system is being carried out by using a gas separation membrane which enable to make the ITER atmosphere detritiation system more compact and cost-effective. Techniques of gas flowing calorimetry and laser Raman spectroscopy are applied to develop new tritium accountancy methods. Studies of tritium-material interaction, such as plasma material interactions, radiochemical reaction of tritium in gas phase, radiolysis of tritiated water, and waste processing are being carried out under ITER/EDA and U.S.-Japan collaboration. Tritium release experiments for research of tritium behavior in confinements and environment and demonstration of safety related components are planned.

Environmental impacts from tritium and active waste are examined and compared for four heavy-ion-driven inertial confinement fusion (IFC) reactor concept studies. The radiation exposure from tritium emissions during normal routine operation and for accidential release of 10%, of the tritium inventory in the target fabrication is determined and compared with regulatory limits. In normal operation, two of the ICF reactor concepts are, independently of the tritium model applied, always below the regulatory limit. For the 10% tritium release from the target fabrication, only one ICF reactor concept is around the limit for the early off-site dose (5 rem). Active waste from structural material of the reactor chamber and the coolant-breeder qualifies either for the waste classes A to C defined by IOCFR61 or geological disposal by deep geological burial. Chambers with carbon-based structural material (SiC and C-C composites) qualify mainly for waste class A. Stainless steel as a structural material leads to active waste that is at least partly subject to geological storage and the disposal conditions depend strongly on the type of stainless steel. Coolant-breeder media are class C and mixed waste.

In recent years tritium has been used in plastic case watches as permanent light sources on watch dials. To measure the release of tritium through the plastic cases, 82 different waterproofed watches were immersed in a water bath for 24 h, and the tritium concentration of the water was measured. The mean tritium release rate was 24,400 Bq d -1 ranging from 110-162,000 Bq d -1 . Parallel measurements were made to determine the tritium concentration in the urine of 108 wearers of plastic case watches. The mean tritium concentration in urine was 197 Bq L -1 up to 1,133 Bq L -1 . The whole body dose resulting from exposure to plastic case watches is negligibly small but given the pathway of skin absorption, the annual skin dose is 3-4 times higher than the dose limit for the public. Plastic case watches are collector's items and are often kept with other watches in glass cabinets or other containers. Storage of a large number of such watches in one container causes tritium to diffuse through the plastic cases and to contaminate watches that did not contain tritium at first. If the container is more or less airtight, the tritium concentration in the container and the tritium release rate from the watches can reach levels up to 4 MBq d -1 .

Tritium is a radioisotope of hydrogen that exists naturally in the environment and may also be released through anthropogenic activities. It bonds readily with hydrogen and oxygen atoms to form tritiated water, which then cycles through the hydrosphere. This paper seeks to model the migration of tritiated species throughout the environment – including atmospheric, river and coastal systems – more comprehensively and more consistently across release scenarios than is currently in the literature. A review of the features and underlying conceptual models of some existing tritium release models was conducted, and an underlying aggregated conceptual process model defined, which is presented. The new model, dubbed ‘Tritium Environmental Release Model’ (TERM), was then tested against multiple validation sets from literature, including experimental data and reference tests for tritium models. TERM has been shown to be capable of providing reasonable results which are broadly comparable with atmospheric HTO release models from the literature, spanning both continuous and discrete release conditions. TERM also performed well when compared with atmospheric data. TERM is believed to be a useful tool for examining discrete and continuous atmospheric releases or combinations thereof. TERM also includes further capabilities (e.g. river and coastal release scenarios) that may be applicable to certain scenarios that atmospheric models alone may not handle well.

An analysis of the heat transport and water detritiation systems of ITER has been performed in order to determine major pathways for tritium loss and estimate releases during normal operation (operational tritium release). Heavy water escape and tritium release estimates compiled on the basis of operating experiences of typical CANDU PWR and the Darlington Tritium Removal Facility (DTRF) have been appropriately scaled on the basis of water and tritium inventories and tritium concentrations to fit ITER design and operating conditions. The paper estimates the chronic and acute tritium releases to the environment in elemental and oxide forms, via waterborne and airborne pathways of the ITER water systems. The results of the analysis will be used to demonstrate that the ITER design will meet the dose limits for occupational and accidental tritium releases. 5 refs., 5 tabs.