Development of an integrated sampler based on direct 222Rn/220Rn progeny sensors in flow-mode for estimating unattached/attached progeny concentration

A flow-mode integrated sampler consisting of a wire-mesh and filter-paper array along with passive solid state nuclear track detectors has been developed for estimating unattached and attached fraction of ²²²Rn/²²⁰Rn progeny concentration. The essential element of this sampler is the direct ²²²Rn/²²⁰Rn progeny sensor (DRPS/DTPS), which is an absorber-mounted-LR115 type nuclear track detector that selectively registers the alpha particles emitted from the progeny deposited on its surface. During sampling at a specified flow-rate, the unattached progeny is captured on the wire-mesh; while the attached progeny gets transmitted and is captured on the filter-paper. The alpha particles emitted by the deposited progeny atoms are registered on the sensors placed at a specified distance facing the wire-mesh and the filter-paper, respectively. The various steps involved in the development of this flow-mode direct progeny sampler such as the optimization of the sampling rate and the distance between the sensor and the deposition substrate are discussed. The sensitivity factor of the DTPS-loaded sampler for ²²⁰Rn progeny deposited on the wire-mesh and filter-paper is found to be 23.77 ± 0.64 (track cm^?2 h^?1) (Bq m^?3)^?1 and 22.30 ± 0.18 (track cm^?2 h^?1) (Bq m^?3)^?1, respectively; while that of DRPS-loaded sampler for ²²²Rn progeny deposition, is 3.03 ± 0.14 (track cm^?2 h^?1) (Bq m^?3)^?1 and 2.08 ± 0.07 (track cm^?2 h^?1) (Bq m^?3)^?1, respectively. The highlight of this flow-mode sampler is its high sensitivity and that it utilizes the passive technique for estimating the unattached and attached progeny concentration, thus doing away with the alpha counting procedures.     

Inactive commissioning of a micro channel catalytic reactor for highly tritiated water production in the CAPER facility of TLK

In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose.Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min^?1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 10¹⁶ Bq kg^?1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water.The combination of the μCCR with the O₂ sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R&D activity on HTW processing with PERMCAT.     

Processing highly tritiated water desorbed from molecular sieve bed using PERMCAT

Tritium handling facilities use molecular sieve beds (MSB) to collect and recover tritiated water. After reaching the capacity limit of the MSB, the water is desorbed and decontaminated in a water detritiation system (WDS). In the case of highly tritiated water (HTW) absorbed into a MSB, an inherent safe option for processing is necessary due to the HTW specific properties. Ideally, HTW should be processed immediately in a continuous mode. With this in consideration, the water desorption process from a zeolite bed was developed and optimized in a dedicated non active facility. The results of this experiments were applied into the regeneration of a MSB previously loaded with HTW containing an activity of 1.9 × 10¹⁴ Bq kg^?1. The water was desorbed, by step increasing the temperature bed and fed by helium carrier gas into the PERMCAT for detritiation and tritium recovery. The processed water was collected in a dry MSB downstream of the PERMCAT. These initial studies successfully demonstrate the viability of the process. The obtained results of the preliminary study and the subsequent tests with tritium, will provide useful information for the design of tritium processes relying on MSB, such as the water processing foreseen for the test blanket modules in ITER.     

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.     

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.     

Tritium release experiments on Li4SiO4 pebbles from TRINPC-I experiments: Effects of water adsorption and hydrogen gas

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.     

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.     

Conceptual design of a water cooled breeder blanket for CFETR

China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by ⁶Li(n,α)T reaction. Li₂TiO₃ pebbles and Be₁₂Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li₂TiO₃ and Be₁₂Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be₁₂Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported.     

Development of a technical scale PERMCAT reactor for processing of highly tritiated water

In view of future fusion rectors fueled by deuterium–tritium mixtures, highly tritiated water (HTW) of up to 5.2·10¹⁶ Bq kg^?1 will be produced, during routine operation and scenarios as an accidental release of tritium into a glove box. Also in the solid breeder blanket concept, a non-negligible fraction of the tritium produced will be in the tritiated water fraction. To decontaminate HTW the PERMCAT using isotope swamping in a Pd/Ag membrane reactor has been identified as a robust and reliable solution. In order to investigate the decontamination of HTW at flow rates relevant for future fusion power plants, a technical scale, fully tritium compatible PERMCAT consisting of a bundle of finger-type membranes inserted in a single catalyst bed was developed. Nevertheless, it represents only one part of a PERMCAT cascade necessary to achieve the required performance to process HTW on technical scale. By improving the existing PERMCAT geometry using experimental data obtained from isotopic exchange between D₂O and H₂, the performance of the existing PERMCAT reactor was optimised. Based on the optimised geometry a new fully tritium compatible technical scale PERMCAT cascade comprising of two PERMCAT reactors in series was designed, manufactured and commissioned as presented in this paper.     

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.     

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.     

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.     

AMS measurements of 10Be and 14C in loess profile at Donglingshan,Beijing

A loess profile in Donglingshan site (40°02′N, 115°27′E) near Beijing was chosen to study the loess formation process and paleo-climate variation. Thirty eight samples were collected and analyzed for ¹⁴C, ¹⁰Be as well as MS, TOC and δ¹³C. Based on ¹⁴C measurements, we established a time scale for this loess profile during Holocene. The averaged ¹⁰Be deposition flux was found to be 4.87 × 10⁶ atoms/cm² year. This is similar to the flux of 4.2 × 10⁶ atoms/cm² year estimated for Chinese Loess Plateau in central China. High ¹⁰Be concentrations of 3.85–5.66 × 10⁸ atoms/g for the samples in layer 23–39 cm from 2965 to 528 years BP suggest a warm and humid weather during this period. MS values have similar variation with ¹⁰Be and reflect the similar paleo-climate information. TOC and δ¹³C suggest that the vegetation around Donlingshan area was C₃ type plants during entire Holocene.     

Small-mass graphite preparation by sealed tube zinc reduction method for AMS 14C measurements

Previous work has demonstrated that the sealed tube Zn reduction method for converting CO₂ to graphite for AMS ¹⁴C measurements produces targets that can be measured with high precision and low background for samples of about 1 mg C down to approximately 0.1 mg C at the Keck Carbon Cycle AMS facility at the University of California, Irvine (KCCAMS). Now a modified method has been developed to prepare small-mass samples ranging from 0.015 to 0.1 mg C. In this modified method, the volume of the sealed reactor tube is reduced to ～1.9 cm³, and the amounts of Zn and TiH₂ reagents are reduced proportionally. The amount of Fe catalyst used remains the same to ensure a long lasting current in the AMS. Small-mass samples prepared by this method generally yield ¹²C⁺¹ currents of about 0.5 μA per 1 μg C. An in situ simultaneous AMS δ¹³C measurement allows for correction of both graphitization and machine-induced isotopic fractionation, and is a prerequisite for high precision and accurate measurements using the Zn reducing method. Corrections for modern-carbon and dead-carbon background components are applied to samples based on small-mass samples of a ¹⁴C free material and of a modern standard covering the sample size range. It was discovered during additional investigation into lowering the modern-carbon background component that baking assembled reactor tubes at 300 °C for 1 h prior to use resulted in significantly lower modern-carbon background values. The accuracy and precision of small-mass samples prepared by this method are size dependent, but is usually ±10–15‰ for the smallest samples (0.015–0.02 mg C), based on duplicate measurements of primary and secondary standards.     

Lithium concentration dependence of implanted helium retention in lithium silicates

Helium ions of 500 keV were implanted with a fluence of 1.4 × 10¹⁷ ion/cm² into various lithium silicates to investigate whether a threshold level of helium retention exists in Li-containing silicate ceramics similar to that found in SiO_x in previous work. The composition and phases of the as prepared lithium silicates were determined by proton backscattering spectrometry (p-BS) and X-ray diffraction (XRD) methods with an average error of ±10%. Electrostatic charging of the samples was successfully eliminated by wrapping the samples in Al foil. The amounts of the retained helium within the samples were determined by subtracting the non-implanted spectra from the implanted ones. The experimental results show a threshold in helium retention depending on the Li concentration. Under 20 at.% all He is able to escape from the material; at around 30 at.% nearly half of the He, while over 65 at.% all implanted He is retained. With compositions expressed in SiO₂ volume percentages, a trend similar to those reported of SiO_x previously is found.     

Residual defects in Cz-silicon after low dose self-implantation and annealing from 400°C to 800°C

Si(1 0 0) samples of n-type have been implanted with 5.6 MeV ²⁸Si³⁺ at room temperature using a dose of 1 × 10⁸ cm⁻². Thereafter the samples were annealed in nitrogen ambient for 30 min at temperatures from 400°C to 800°C using steps of 50°C. Deep level transient spectroscopy (DLTS) was used for sample analysis. Deep levels, not previously reported in the literature, are observed to arise, evolve, and to vanish again in the described temperature interval, while above 650°C only two defects levels remain. Depth profiles of these levels show, that the maximum concentrations of the defects are located deeper than the projected range, as is characteristic for secondary implantation defects. Their possible identities are discussed.     

The chemical durability of glass and graphite–glass composite doped with cesium oxide

The role of temperature in determining the chemical stability of a waste form, as well as its leach rate, is very complex. This is because the dissolution kinetics is dependent both on temperature and possibility of different rate-controlling mechanisms that appear at different temperature regions. The chemical durability of Alumina-Borosilicate Glass (ABG) and Glass–Graphite Composite (GGC), bearing Tristructural Isotropic (TRISO) fuel particles impregnated with cesium oxide, were compared using a static leach test. The purpose of this study is to examine the chemical durability of glass–graphite composite to encapsulate coated fuel particles, and as a possible alternative for recycling of irradiated graphite. The test was based on the ASTM C1220-98 methodology, where the leaching condition was set at a temperature varying from 298 K to 363 K for 28 days. The release of cesium from ABG was in the permissible limit and followed the Arrhenius’s law of a surface controlled reaction; its activation energy (E_a) was 65.6 ± 0.5 kJ/mol. Similar values of Ea were obtained for Boron (64.3 ± 0.5) and Silicon (69.6 ± 0.5 kJ/mol) as the main glass network formers. In contrast, the dissolution mechanism of cesium from GGC was a rapid release, with increasing temperature, and the activation energy of Cs (91.0 ± 5 kJ/mol) did not follow any model related to carbon kinetic dissolution in water. Microstructure analysis confirmed the formation of Crystobalite SiO₂ as a gel layer and Cs⁺¹ valence state on the ABG surface.     

Comparison of neutron and gamma irradiation effects on KU1 fused silica monitored by electron paramagnetic resonance

Electron paramagnetic resonance (EPR) studies have been carried out on KU1 fused silica irradiated with neutrons at fluences 10²¹ and 10²² n/m², and gamma-ray doses up to 12 MGy. The effects of post-irradiation thermal annealing treatments, up to 850 °C, have also been investigated. Paramagnetic oxygen-related defects (POR and NBOHC) and E′-type defects have been identified and their concentration has been measured as a function of neutron fluence, gamma dose and post-irradiation annealing temperature. It is found that neutrons at the highest fluence generate a much higher concentration of defects (mainly E′ and POR, both at concentrations about 5 × 10¹⁸ spins/cm³) than gamma irradiations at the highest dose (mainly E′ at a concentration about 4 × 10¹⁷ spins/cm³). Moreover, for gamma-irradiated samples a lower treatment temperature (about 400 °C) is required to annihilate most of the observed defects than for neutron-irradiated ones (about 600 °C).