Measurement of tritium concentration in water by imaging plate

Concentration of tritium in water (4–400 kBq cm^?3) was measured by exposing an imaging plate without protection layer (Fujifilm, BAS-IP TR) to vapor for 2–48 h. It was found that tritium gradually penetrated into Eu-doped BaFBr phosphor and induced sufficiently intense photostimulated luminescence (PSL) even at the concentration of 4 kBq cm^?3. The intensity of PSL was proportional to tritium concentration in water. In addition, tritium absorbed in phosphor was reversibly released by keeping IP in air, and IP was able to be used repeatedly if total duration of exposure was ca. 24 h or less. The contamination of IP with tritium was not serious. It was concluded that IP technique has potential to measure tritium concentration in water without direct handling of tritiated water and with a minimum amount of radioactive waste.     

Design of a multipurpose laboratory scale apparatus for the investigation of hydrogen isotopes in PbLi and permeation technologies

The knowledge of the tritium transport parameters in lead lithium is fundamental for the design of the HCLL (helium cooled lead lithium) blanket. In fact, the inventory of tritium in fusion reactors blankets and the permeation of tritium into the blanket coolant, with the consequent leaks toward the environment, are strongly depending on its solubility and diffusivity in the lead alloy PbLi. Several experiments, devoted to investigate the function linking the tritium solubilised in lead lithium with the corresponding tritium partial pressure at equilibrium, were carried out in the past, but significant uncertainties still remain.A detailed analysis of the past experimental works is carried out in this paper with the aim to investigate the main problems occurred in the facilities used to measure the tritium solubility in PbLi that caused such a big spread in the achieved results. On the basis of this analysis, a new a multipurpose laboratory scale apparatus has been designed. The apparatus is able to measure the tritium solubility and diffusivity in PbLi in the range of temperature 300–550 °C and it will be operated with hydrogen partial pressure in the range 10²–10⁴ Pa. The facility can work with desorption and absorption technique.Moreover, the apparatus has been designed to allow the testing of H/D concentration sensors in Pb–15.7Li in operative conditions relevant to the HCLL–TBM and the characterisation of hydrogen permeation barrier.     

Hydrogen isotope exchange in tungsten: Discussion as removal method for tritium

Hydrogen isotope exchange in re-crystallized polycrystalline tungsten was investigated at 320 and 450 K. In a first step the tungsten samples were loaded with deuterium to a fluence of 10²⁴ D/m² from a low-temperature plasma at 200 eV/D particle energy. In a second step, H was implanted at the same particle energy and similar target temperature with a mass-separated ion beam at different ion fluences ranging from 2 × 10²⁰ to 7.5 × 10²³ H/m². The analytic methods used were nuclear reaction analysis with D(³He,p)α reaction and elastic recoil detection analysis with ⁴He. In order to determine the D concentration at depths of up to 7.4 μm the ³He energy was varied from 0.5 to 4.5 MeV. It was found that already at an H fluence of 2 × 10²⁰ H/m², i.e. at 1/5000 of the initial D fluence, about 30% of the retained D was released. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface at 320 K, but extending to all analyzable depths at 450 K especially at high fluences, leading to higher release efficiency. The reverse sequence of hydrogen isotopes allowed the analysis of the replacing isotope and showed that the release of D is balanced by the uptake of H. It also shows that hydrogen does not diffuse through a region of filled traps into a region were unfilled traps can be encounter but transport is rather a dynamic process of trapping and de-trapping even at 320 K. Initial D retention in H loaded W is an order of magnitude higher than in pristine W, indicating that every H-containing trap is a potential trap for D. In consequence, hydrogen isotope exchange is not a viable method to significantly enhance the operation time before the tritium inventory limit is reached but should be considered an option to reduce the tritium inventory in ITER before major interventions at the end of an operation period.     

Penetration of tritiated water vapor through hydrophobic paints for concrete materials

Behavior of tritium transfer through hydrophobic paints of epoxy and acrylic-silicon resin was investigated experimentally. The amounts of tritium permeating through their paint membranes were measured under the HTO concentration condition of 2–96 Bq/cm³. Most of tritium permeated through the paints as a molecular form of HTO at room temperature. The rate of tritium permeating through the acrylic-silicon paint was correlated in terms of a linear sorption/release model, and that through the epoxy paint was controlled by a diffusion model. Although effective diffusivity estimated by a diffusion model was obtained 1.1 × 10⁻¹³–1.8 × 10⁻¹³ m²/s for epoxy membranes at the temperature of 21–26 °C, its value was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, resistance of tritium diffusion through interface between cement-paste and the epoxy paint was considered to be the most effective in the overall tritium transfer process. Clarification of tritium transfer behavior at the interface is important to understand the mechanism of tritium transfer in concrete walls coated with various paints.     

Overview of the US–Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US–Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 °C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe²⁺, 20 MeV W²⁺, and 700 keV H^?) with that from neutron-irradiated tungsten to identify the similarities and differences among them.     

Application of laser-induced breakdown spectroscopy for characterization of material deposits and tritium retention in fusion devices

Laser-induced breakdown spectroscopy (LIBS) is discussed as a possible method to characterize the composition, tritium retention and amount of material deposits on the first wall of fusion devices. The principle of the technique is the ablation of the co-deposited layer by a laser pulse with P (power density) ≥ 0.5 GW/cm² and the spectroscopic analysis of the light emitted by the laser induced plasma. The typical spatial extension of the laser plasma plume is in the order of 1 cm with typical plasma parameters of n_e ≈ 3 × 10²² m^?3 and T_e ≈ 1–2 eV averaged over the plasma lifetime which is below 1 μs. In this study “ITER-Like” mixed deposits with a thickness of about 2 μm and consisting of a mixture of W/Al/C and D on bulk tungsten substrates have been analyzed by LIBS to measure the composition and hydrogen isotopes content at different laser energies, ranging from about 2 J/cm² (0.3 GW/cm²) to about 17 J/cm² (2.4 GW/cm²) for 7 ns laser pulses. It is found that the laser energies above about 7 J/cm² (1 GW/cm²) are needed to achieve the full removal of the deposit layer and identify a clear interface between the deposit and the bulk tungsten substrate by applying 15–20 laser pulses while hydrogen isotopes decrease strongly after the first laser pulse. Under these conditions, the evolution of the spectral line intensities of W/Al/C/hydrogen can be used to evaluate the layer composition.     

Assessment of tritium levels in rivers and precipitation in north-western Greece before the ITER operation

The project ITER aims to demonstrate that fusion is the energy source of the future. The prototype Tokamak machine is intended to start operation at about 2019 and tritium is one of the major contaminants that can be accidentally released in the environment. Nowadays environmental tritium levels are of natural origin except in the vicinity of nuclear facilities. The evaluation of background tritium levels is essential in the context of a future possibility of accidental tritium releases. For this purpose and also because of the lack of relevant information, an extended programme of river and rain water sampling was implemented in north-western Greece. Water samples from six major rivers in this area and rain water samples were analysed for tritium content. The rivers under investigation were Aliakmonas River, Pinios River, Arachthos River, Kalamas River, Aoos River and Louros River, which originate from the central Greek mountain range Pindos, and flow to Aegean and Ionian Sea.The tritium concentrations were determined by the Liquid Scintillator Analyser Tri-Carb 3170TR/SL. The statistical analysis of data revealed that there is a seasonal variation of tritium concentration in rain samples and a less pronounced seasonal variation in river samples. The weighted mean tritium concentration for rain samples was determined equal to 0.90 ± 0.08 Bq L^?1 (7.6 ± 0.7 TU) and the respective mean value for river samples was 0.94 ± 0.04 Bq L^?1 (7.9 ± 0.3 TU). Further analysis has demonstrated that river waters tend to show lower tritium concentrations than the concurrently measured tritium concentrations in rain samples, during spring and summer (at 47% and 71% of the sampling stations, respectively), while this observation is reversed during autumn and winter (at 44% and 35% of the sampling stations, respectively). This may be attributed to rain water remaining underground for a long period allowing tritium to decay and when it reappears as river water, the tritium concentration values are lower when compared to the rain water concentrations. Rough estimates of the residence time of underground waters in the study area provided values, which ranged from 0.5 to 11.7 years, with a mean value of 5.2 ± 0.9 years.     

Evaluation of tritium diffusion through the Neutral Beam Injector calorimeter panel

The Neutral Beam Test Facility (NBTF) to be realized in Padoa will test the Neutral Beam Injection (NBI), one of the Heating and Current Drive Systems foreseen for ITER. The NBI is based on the acceleration of hydrogen or deuterium negative ions up to 1 MeV. This work has been aimed at assessing the tritium release from the NBTF in order to provide data for the safety analysis. In particular, the diffusion of the tritium through the neutral beam target material (the CuCrZr alloy calorimeter panels) has been assessed by using literature data of the diffusion coefficient. The tritium generated inside the calorimeter panels moves into both the vacuum and water side: the tritium diffusion flux has been evaluated during the beam-on (200 °C) and the beam-off (20 °C) phases of the NBTF experiments consisting of an interim campaign and a final test. The penetration depth of the tritium through the 2 mm thick CuCrZr alloy material has been also evaluated by using a Monte-Carlo code. As main result, the assessed diffusion flux of tritium during both the beam-on and the beam-off phases are modest. In fact, at the end of the interim campaign (100 days), about the 96% of the all generated tritium (626.5 MBq) exits the calorimeter while the residual tritium inventory (25 MBq) leaves the copper alloy with a diffusion time of about 1 month. At the end of the final test (14 days) about the 99% of the total generated tritium (1.023 × 10⁴ MBq) leaves the copper alloy and the remaining tritium inventory (152.2 MBq) is released by about 32 days. In both the interim campaign and the final test, more than the 99% of the total tritium is transferred into the vacuum side of the calorimeter panel while negligible tritium amounts enter the cooling water system thus showing a very low impact on the environment.     

Study on the retention behavior of hydrogen isotopes and the change of chemical states of boron film exposed to hydrogen plasma in LHD

The behavior of hydrogen retention and the change of chemical states of boron film exposed to hydrogen plasma in LHD were investigated. The sample was prepared in LHD, and atomic concentrations for the boron film after hydrogen plasma exposure were changed from 75% for boron, 15% for carbon and 8% for oxygen to 53%, 18% and 22%, respectively. BC bond was a major chemical state of the boron film after hydrogen plasma exposure, although abundance of BB bond was the highest before the plasma exposure. Total hydrogen retention measured by TDS was evaluated to be 1.7 × 10²⁰ H m^?2, and the retentions of hydrogen as BHB, BH and BCH bonds were, respectively, 4.8 × 10¹⁹, 7.2 × 10¹⁹ and 5.2 × 10¹⁹ H m^?2. It was concluded that the hydrogen retention could be estimated by taking account not only of chemical states of impurities, but also of hydrogen depth profile.     

Tritium permeation experiments using reduced activation ferritic/martensitic steel tube and erbium oxide coating

Low concentration tritium permeation experiments have been performed on uncoated F82H and Er₂O₃-coated tubular samples in the framework of the Japan-US TITAN collaborative program. Tritium permeability of the uncoated sample with 1.2 ppm tritium showed one order of magnitude lower than that with 100% deuterium. The permeability of the sample with 40 ppm tritium was more than twice higher than that of 1.2 ppm, indicating a surface contribution at the lower tritium concentration. The Er₂O₃-coated sample showed two orders of magnitude lower permeability than the uncoated sample, and lower permeability than that of the coated plate sample with 100% deuterium. It was also indicated that the memory effect of ion chambers in the primary and secondary circuits was caused by absorption of tritiated water vapor that was generated by isotope exchange reactions between tritium and surface water on the coating.     

Erosion behavior of soft,amorphous deuterated carbon films by heat treatment in air and under vacuum

The erosion of soft a-C:D films by heat treatment in air and under vacuum is studied by ion-beam analysis. When the films are heated in air above 500 K, the film thickness and the areal densities of C and especially D decrease, and oxygen is incorporated in the films. The initial atomic loss rates of carbon and deuterium from the films are 2.6 × 10¹⁷ C atoms cm⁻² h⁻¹ and 4.8 × 10¹⁷ D atoms cm⁻² h⁻¹ at 550 K. However, after D depletion the films show a resistivity against further erosion due to annealing in air. When the films are heated under vacuum erosion starts at about 600 K and all components including D decrease proportionally to the film thickness. Thermal desorption spectroscopy of the films reveals the evolution of C_xD_y type hydrocarbons. Infrared analysis shows that the incorporated oxygen is chemically bonded to carbon. The thermally-activated decomposition of the soft a-C:D films is compared to that of hard a-C:D films and a reaction scheme is suggested.     

Feasibility of a multi-purpose demonstration neutron source based on a compact superconducting spherical tokamak

Tokamak neutron sources would allow near term applications of fusion such as fusion–fission hybrid reactors, elimination of nuclear wastes, production of radio-isotopes for nuclear medicine, material testing and tritium production. The generation of neutrons with fusion plasmas does not require energetic efficiency; thus, nowadays tokamak technologies would be sufficient for such purposes. This paper presents some key technical details of a compact (～1.8 m³ of plasma) superconducting spherical tokamak neutron source (STNS), which aims to demonstrate the capabilities of such a device for the different possible applications already mentioned. The T-11 transport model was implemented in ASTRA for 1.5 D simulations of heat and particle transport in the STNS core plasma. According to the model predictions, total neutron production rates of the order of ～10¹⁵ s^?1 and ～10¹³ s^?1 can be achieved with deuterium/tritium and deuterium/deuterium respectively, with 9 MW of heating power, 1.4 T of toroidal magnetic field and 1.5 MA of plasma current. Engineering estimates indicate that such scenario could be maintained during ～20 s and repeated every ～5 min. The viability of most of tokamak neutron source applications could be demonstrated with a few of these cycles and around ～100 cycles would be required in the worst cases.     

Ion beam mixing in uranium nitride thin films studied by Rutherford Backscattering Spectroscopy

Thickness, composition, concentration depth profile and ion irradiation effects on uranium nitride thin films deposited on fused silica have been investigated by Rutherford Backscattering Spectroscopy (RBS) using 2 MeV He⁺ ions. The films were prepared by reactive DC sputtering at the temperatures of ?200 °C, +25 °C and +300 °C. A perfect 1U:1N stoichiometry with a layer thickness of 660 nm was found for the film deposited at ?200 °C. An increase of the deposition temperature led to an enhancement of surface oxidation and an increase of the thickness of the mixed U–N–Si–O layers at the interface. The sample irradiation by 1 MeV Ar⁺ ion beam with ion fluence of about 1.2–1.7 × 10¹⁶ ions/cm² caused a large change in the layer composition and a large increase of the total film thickness for the films deposited at ?200 °C and at +25 °C, but almost no change in the film thickness was detected for the film deposited at +300 °C. An enhanced mixing effect for this film was obtained after further irradiation with ion fluence of 2.3 × 10¹⁶ ions/cm².     

Carbon film growth and hydrogenic retention of tungsten exposed to carbon-seeded high density deuterium plasmas

Tungsten (W) targets have been exposed to high density (n_e ? 4 × 10¹⁹ m^?3), low temperature (T_e ? 3 eV) CH₄-seeded deuterium (D) plasma in Pilot-PSI. The surface temperature of the target was ～1220 K at the center and decreased radially to ～650 K at the edges. Carbon film growth was found to only occur in regions where there was a clear CII emission line, corresponding to regions in the plasma with T_e ? 2 eV. The maximum film thickness was ～2.1 μm after a plasma exposure time of 120 s. ³He nuclear reaction (NRA) analysis and thermal desorption spectroscopy (TDS) determine that the presence of a thin carbon film dominates the hydrogenic retention properties of the W substrate. Thermal desorption spectroscopy analysis shows retention increasing roughly linearly with incident plasma fluence. NRA measures a C/D ratio of ～0.002 in these films deposited at high surface temperatures.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

Development of a plasma driven permeation experiment for TPE

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1], [2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.     

Feasibility study of an intense D–T fusion source: “The New Sorgentina”

The international community agrees on the importance to build a large facility devoted to test and validate materials to be used in harsh neutron environments. Such a facility, proposed by ENEA, reconsiders a previous study known as “Sorgentina” but takes into account new technological development so far attained. The “New Sorgentina” Fusion Source (NSFS) project is based upon an intense D–T 14 MeV neutron source achievable with T and D ion beams impinging on 2 m radius rotating targets. NSFS produces about 1 × 10¹³ n cm⁻² s⁻¹ over about 50 cm³. Larger volumes of lower neutron flux will be available (e.g. for TBM experiments) as well as collimated channels to study some features of the ITER neutron camera. The NSFS facility will overcome problems related to the ion source and accelerating system, by means of an upgraded version of the JET–PINI ion beams. NSFS has to be intended as an European facility that may be realized in a few years, once provided a preliminary technological program devoted to the operation of the ion source in continuous mode, target heat loading/removal, target and tritium handling, inventory as well as site licensing. In this contribution, the main characteristics of NSFS project will be presented.     

First-principles study of hydrogen adsorption and permeation in reconstructed cubic erbium oxide surfaces

Understanding surface properties of Er₂O₃, especially in relation to adsorption and permeation of atomic hydrogen, is of considerable importance to the study of tritium permeation barriers. In this work, hydrogen diffusion pathways through the low-index (1 0 0), (1 1 0), and (1 1 1) surfaces of cubic Er₂O₃ have been calculated using density functional theory within the GGA (PBE) + U approach. The dependence of the effective U parameter on lattice constants, bulk moduli, and formation energies of Er₂O₃ has been investigated in detail. The energetics of hydrogen penetration from the surfaces to the solution site in bulk Er₂O₃ were defined using the optimum effective U value of 5.5 eV. For a low surface coverage of hydrogen (0.89 × 10¹⁴ H/cm²), a penetration energy of at least 1.7 eV was found for all the low-index erbium oxide surfaces considered. The results of the present study will provide useful guidance for future studies on modeling defects, such as grain boundaries and vacancies, in tritium permeation barriers.     

Tritium analysis in hydraulic oil waste by oxidation technique

The aim of the present study is to investigate a method to evaluate the tritium activity in hydraulic oil waste generated during the operation of Romanian Cernavoda Nuclear Power Plant.The method is based on a combustion technique using the 307 PerkinElmer^® Sample Oxidizer model.The hydraulic oil samples must be processed prior to counting to avoid color quenching (the largest source of inaccuracy) because these samples absorb in the region of 200–500 nm, where scintillation phosphors emit.Prior to combustion of the hydraulic oil waste, tritium recovery degree and tritium retention degree in the circuits of combustion system were evaluated as higher than 98% and less than 0.08%, respectively.After combustion, tritium activity was measured by a 2100 Tri-Carb^® Packard model liquid scintillation analyzer.The blank counts were 16.25 ± 0.50 counts/min, measured for 60 min. The significant activity level value was 6.53 counts/min, at a preselected confidence level of 95%. The Minimum Detectable Activity of a 0.2 mL hydraulic oil sample was calculated to 1.09 Bq/mL. Therefore, the developed method is sensitive enough for the tritium evaluation in the ordinary hydraulic oil waste samples.     

Soft X-ray fluorescence measurements of irradiated polymer films

Fluorescent soft X-ray carbon Kα emission spectra (XES) have been used to characterize the bonding of carbon atoms in polyimide (PI) and polycarbosilane (PCS) films. The PI films have been irradiated with 40 keV nitrogen or argon ions, at fluences ranging from 1 × 10¹⁴ to 1 × 10¹⁶ cm⁻². The PCS films have been irradiated with 5 × 10¹⁵ carbon ions cm⁻² of 500 keV and/or annealed at 1000°C. We find that the fine structure of the carbon XES of the PI films changes with implanted ion fluence above 1 × 10¹⁴ cm⁻² which we believe is due to the degradation of the PI into amorphous C:N:O. The width of the forbidden band as determined from the high-energy cut-off of the C Kα X-ray excitation decreases with the ion fluence. The bonding configuration of free carbon precipitates embedded in amorphous SiC which are formed in PCS after irradiation with C ions or combined treatments (irradiation and subsequent annealing) is close to either to that in diamond-like films or in silicidated graphite, respectively.