Specific Activity Tritium Research Articles

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

Abstract 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.

Environmental transportation of tritium and estimation of site-specific model parameters for Kaiga site, India

Tritium content in air moisture, soil water, rain water and plant water samples collected around the Kaiga site, India was estimated and the scavenging ratio, wet deposition velocity and ratio of specific activities of tritium between soil water and air moisture were calculated and the results are interpreted. Scavenging ratio was found to vary from 0.06 to 1.04 with a mean of 0.46. The wet deposition velocity of tritium observed in the present study was in the range of 3.3E-03 to 1.1E-02 m s(-1) with a mean of 6.6E-03 m s(-1). The ratio of specific activity of tritium in soil moisture to that in air moisture ranged from 0.17 to 0.95 with a mean of 0.49. The specific activity of tritium in plant water in this study varied from 73 to 310 Bq l(-1). The present study is very useful for understanding the process and modelling of transfer of tritium through air/soil/plant system at the Kaiga site.

Study of ratio of tritium concentration in plants water to tritium concentration in air moisture for chronic atmospheric release of tritium

Abstract Specific activity models (SA) are often used to estimate tritium concentration in the plants for chronic-release of atmospheric HTO in some regulatory models by some countries and commissions. In such models, a major assumption is that the value of specific activity of tritium of tritium oxide in vegetation to the specific activity of tritium of tritium oxide in air moisture is maintained at a constant ratio (R). The value of R is an important factor in determining tritium concentration and dose from chronic atmospheric release. But the value of R recommended is different from different models. Concentrations of tritium in plants will be have huge difference in plants because of the difference of the value of R, and this in turn would result in difference of ingestion dose via food chain. Some studies suggested that a site-specific distribution of R should be developed in suing a specific activity model. In this study, distribution of R is established for the Qinshan NPP Base. The environmental monitoring data of tritium concentration in five type plants (rapeseed, tea, cabbage, radish and rice) and air at three sampling points (Xiajiawan, Qinlian and Ganpu) around Qinshan NPP Base(QNNP) over a 4 years period as the basis for analysis, and the tritium ratio(R) between plant water and air moisture were determined. The results showed the average value of R of five plants were 0.103, 0.687, 1.055, 0.695 and 0.183 respectively. These values of R are mostly consistent with the law presented by foreign literature, only the value of R for cabbage is greater than the value of R for foliage vegetation presented by foreign reports. This is partly attributable to the difference of experimental conditions. The concentration of HTO of vegetations around QNNP could be assessed using the values of R recommended by this report for chronic release of atmospheric HTO.

Detection of tritium sorption on four soil materials

In order to measure groundwater age and design nuclear waste disposal sites, it is important to understand the sorption behavior of tritium on soils. In this study, batch tests were carried out using four soils from China: silty clays from An County and Jiangyou County in Sichuan Province, both of which could be considered candidate sites for Very Low Level Waste disposal; silty sand from Beijing; and loess from Yuci County in Shanxi Province, a typical Chinese loess region. The experimental results indicated that in these soil media, the distribution coefficient of tritium is slightly influenced by adsorption time, water/solid ratio, initial tritium specific activity, pH, and the content of humic and fulvic acids. The average distribution coefficient from all of these influencing factors was about 0.1–0.2 mL/g for the four types of soil samples. This relatively modest sorption of tritium in soils needs to be considered in fate and transport studies of tritium in the environment.

Structural changes and distribution of accumulated tritium in the carbon based JET tiles

In this study the tritium distribution and the effect of structural changes thereon have been analyzed in the bulk of the tile selected from the JET Mark II SRP divertor. Tritium content has been analyzed by the full combustion technique [1]. The structure has been investigated by the method of Scanning Electron Microscopy.Tritium depth profiles have been measured at different poloidal positions. A high specific activity of tritium (up to 156MBqg−1) was found at the plasma-facing surface. At some tile positions up to 98–99% of the T can be in the surface slice of 1mm thickness, whereas in other poloidal positions there can be more T in the bulk than at the surface. The structural changes of the tile material both at the surfaces and in the bulk have been measured.

Pre-study of a detritiation process for plasma facing components in the ITER hot cells

In ITER and future fusion reactors, high heat flux (plasma facing) components will be replaced at regular intervals. Before being stored in hot cells or sent to final disposal, these components or parts of them must be detritiated to recover the large quantities of tritium still contained in the matrix (recycling tritium as fuel) and facilitate the waste acceptance in storage. This paper aims at evaluating the capability of a detritiation process to match the ITER requirements.For that purpose potential detritiation methods have been analysed in terms of detritiation performance leading to the choice of a reference process based on thermal treatment. In parallel, some detailed data concerning the waste have been compiled mainly concerning the amount of waste to treat, and its tritium specific activity with its profile in bulky materials. Since no tritium profile is currently available in ITER documents, an estimation of such profile has been performed based on the modelling of tritium diffusion into the plasma facing components. Thus, the reference detritiation process has been evaluated by a simple but representative modelling. Basically, this first approach enables to confirm the interest of such thermal process. Nevertheless, its design in the ITER hot cells requires a more sophisticated approach.

In vitro dissolution of tritium-loaded particles from the JET fusion machine

With the experimental evolution of fusion power the levels of tritium used will increase as will the potential for human exposure. Tritium-loaded carbon particles produced during the experimental operation of the Joint European Torus fusion tokamak have been characterised in terms of size, elemental composition and specific activity of tritium elsewhere. The aim of this study was to characterise the dissolution of tritium from these particles in order to derive dose coefficients for this material and provide guidance on monitoring procedures should it be inhaled accidentally. The dissolution of tritium was measured for 100 d in lung serum simulant from two batches of materials, SG1 and SG2, which were obtained from carbon tiles originating from different positions in the reactor. Retention over this period followed a three-component exponential. About 1-5% dissolved within a minute, and up to a further 20% dissolved over 100 d for the SG1 materials but <1% for the SG2 materials. Dissolution between the SG1 materials varied greatly, whereas the SG2 materials were similar. As a result of this variability, the assessed dose from urinary excretion could be in error by up to two orders of magnitude depending on the material inhaled. It is recommended that (i) the dissolution is measured for a wider range of materials, preferably dusts collected in working areas, and (ii) in vivo studies are performed to characterise fully the urine excretion of tritium from these materials. This information could be used to provide improved guidance on dose assessment after special or routine monitoring, taking account of the likely variation of particle size and biological retention half times.

Modelling the spatio-temporal evolution of 3H in the waters of the River Tagus

Measurements of tritium specific activity levels and of temperatures in waters of the Torrejón–Tagus reservoir (Spain) showed that their radioactive characteristics were basically influenced by the radioactive liquid effluent from the Almaraz Nuclear Power Plant. This enters the Torrejón–Tagus reservoir via the Arrocampo cooling reservoir, which exchanges water with the first. We studied the temporal and spatial (in two dimensions) evolution of the mentioned parameters for years 1997 and 1998. The tritium levels were found to be significantly correlated with temperature. Two numerical models were constructed for a quantitative study of the tritium levels along Torrejón reservoir: a 1D model was used for the dispersion of tritium along the whole length of the reservoir, and a 2D depth-averaged model was used for a detailed study of the area where tritium is released into the reservoir. Both models solve the hydrodynamic equations, to obtain the currents induced by the exchanges of water between the reservoirs in the River Tagus and Arrocampo, and the advection/diffusion equation to calculate the dispersion of tritium. In general, the model results were in agreement with the experimental observations.

Development of a Low-Level Tritium Air Monitor

There are three components which contain tritium in the air, i.e. water vapor, molecular hydrogen and methane. However, there is no tritium monitor, which is able to measure the concentration separately. In the light of it, we are developing a new monitor, which will be enable us to measure the tritium concentrations in two of these species, i.e. molecular hydrogen and methane. The specific activities of tritium in molecular hydrogen and in methane are very high. Then if we can collect the hydrogen and the methane, we will easily measure the tritium concentrations. We have conceptually designed a new tritium monitor and carried out several basic experiments. Based on the results of these experiments, we have estimated the performance of the monitor. For an hour counting, the monitor will be able to measure the tritium concentrations of molecular hydrogen and methane using about 4m3 air. Since the total performance of the monitor mainly depends on the capability of the proportional counter, we are developing an appropriate counter.

Tritium activity induced in the accelerator building and its correlation to radioactivity of gamma-nuclides

An infrared furnace (ULVAC RHL-410P) was newly applied to the extraction of tritium from concrete samples. After studying the tritium recovery yield regarding temperature and time, the best extraction conditions were set to 800 °C (setting temperature) for 30 minutes under Ar-gas flow of 200 ml/min. Tritium was collected in two cold traps and transferred to a vial for liquid scintillation counting. It took about one hour for the extraction of tritium. Reproducibility and recovery yield of tritium were about 100% compared to the values obtained by the ordinary heating method using an electric furnace. Gamma-ray emitters and tritium of concrete samples collected from several accelerator facilities have been determined. The specific activity of tritium strongly correlated with that of 152Eu and 60Co, so it was found that tritium was produced by thermal neutron reaction by the 6Li(n,α)3H reaction. The results indicate that the tritium specific activity in concrete can be estimated from the 60Co specific activity obtained easily by γ-ray measurement.

HTO Emission from Contaminated Surfaces and Distribution in Environmental Media Around Nuclear Facilities

ABSTRACTField studies were undertaken on both a wetland, which is downgrabient from a radioactive waste storage area at Chalk River Laboratory, and a grassland near Pickering nuclear power station. The purpose for this study was to quantify HTO (tritiated water/vapor) transfer in the land surface environment. The amounts of evaporation and transpiration were separately estimated, because the specific activity of tritium (HTO) is commonly different between surface and subsurface soil. Most of the water and tritium fluxes are attributed to transpiration in the season examined. Measured tritium concentrations in leaves at daytime agreed with predicted values, which are based on a simple equation. The ratios of organically bound tritium (OBT) to free water tritium ranged from 0.2 to 0.8 in this study.

Correlation between Tritium and 152Eu Induced in Various Types of Concrete by Thermal Neutron Irradiation

Tritium induced in BWR reactor shielding wall and PWR biological shielding wall irradiated by thermal neutrons is considered to be the critical radionuclide for trench disposal management in Japan. For the purpose of tritium quantification, preparing a specimen of activated concrete without letting tritiated water scatter away is, however, practically impossible. This calls for an alternative means of tritium quantification through indirect estimation from measurement of a substitute radionuclide, such as 152Eu, that is easy to quantify. Many types of concrete and their raw materials were sampled and irradiated by thermal neutrons of the JRR-4 reactor to find out the correlation existing in irradiated concrete between the specific activities of tritium and of 152Eu. The induced activity of for 152Eu was measured by Ge detector, and that of tritium by liquid scintillation counter. From the results measured on the various types of concrete, the formula obtained empirically between 3H and 152Eu was 3H(Bq/g)=152Eu(Bq/g)x(5–27). Another noteworthy finding is that the ratio of the quantity of tritium induced by thermal neutrons is 1:46 between low-activation concrete and ordinary concrete.

Correlation between Tritium and 152Eu Induced in Various Types of Concrete by Thermal Neutron Irradiation.

Tritium induced in BWR reactor shielding wall and PWR biological shielding wall irradiated by thermal neutrons is considered to be the critical radionuclide for trench disposal management in Japan. For the purpose of tritium quantification, preparing a specimen of activated concrete without letting tritiated water scatter away is, however, practically impossible. This calls for an alternative means of tritium quantification through indirect estimation from measurement of a substitute radionuclide, such as 152Eu, that is easy to quantify. Many types of concrete and their raw materials were sampled and irradiated by thermal neutrons of the JRR-4 reactor to find out the correlation existing in irradiated concrete between the specific activities of tritium and of 152Eu. The induced activity of for 152Eu was measured by Ge detector, and that of tritium by liquid scintillation counter. From the results measured on the various types of concrete, the formula obtained empirically between 3H and 152Eu was 3H(Bq/g)=152Eu(Bq/g)x(5–27). Another noteworthy finding is that the ratio of the quantity of tritium induced by thermal neutrons is 1:46 between low-activation concrete and ordinary concrete.

Dose contribution from metabolized organically bound tritium after chronic tritiated water intakes in humans.

Our earlier study of acute tritiated water intakes in humans has demonstrated that the dose contribution from metabolized organically bound tritium is less than 10% of the body water dose. To further demonstrate that the dose contribution from the organically bound tritium per unit intake of tritiated water is the same, regardless of whether the intake is acute (all at once) or chronic (spread over time), urine samples from six male radiation workers with chronic tritiated water intakes were collected and analyzed for tritium. These workers have a well-documented dose history and a well-controlled tritium bioassay database, providing assurance that their tritium intakes were in the form of tritiated water. Each month for a full calendar year, urine samples were collected from each exposed worker. The monthly concentration of tritiumin-urine for each exposed worker was no lower than 10(4) Bq L(-1) but no higher than 10(5) Bq L(-1). These urine samples were analyzed for tritiated water and organically bound tritium to determine the ratio of these tritiated species in urine. The average ratio of tritiated water to organically bound tritium in urine for each exposed worker was 330 +/- 129 (range, 297-589). In calculating the dose to these workers, we assumed that, under steady-state conditions, the ratio of the specific activity of tritium (3H activity per gH) in the organic matter and water fractions of urine is representative of the ratio of the specific activity of tritium in the organic matter and water fractions of soft tissue. A mathematical model was developed and used to estimate the dose increase from the metabolized organically bound tritium based on the ratio of tritiated water to organically bound tritium in urine. The resulting average dose from the organically bound tritium was 6.9 +/- 3.1% (range, 4.7-9.9%) of the body water dose for the six male workers, and agrees well with the value obtained from our acute tritiated water intakes study in humans. The observed dose contribution from organically bound tritium, relative to body water dose, is in agreement with current recommendations of assigning 10% of total body water dose for organically bound tritium in soft tissues after tritiated water intakes.

Influx and efflux of moisture in a desert soil during a 1 year period

Spatial and temporal variation of moisture flux in soil over a 1 year period was studied in the Thar desert of India. Tritium was injected at a depth of 0.8 m at 18 sites distributed over an area of 50 km2 and at 20 adjacent points at a single site over a sand dune, before the 1994 monsoon. Vertical soil profiles were collected from the 18 sites after the end of monsoon and at the end of the year and from the multiple‐injections site at different time intervals throughout the year. Moisture flux values, computed from tritium and moisture distribution in soil profiles, show that the moisture influx at a site depends upon the grain size of the top 20 cm layer and that about 68% of the total influx of water below the injection depth is lost during the summer months. The data for the multiple‐injections site show that the tritium spike was displaced to a depth of 4 m by the end of the monsoon, and it remained stationary during the postmonsoön period of 8 months, although about 64% of the soil moisture was lost. The spiked layer showed enrichment in the specific activity of tritium during summer months. These observations suggest that vapor diffusion is the dominant process for moisture loss from the Thar desert soil. The net input (natural recharge) to groundwater due to 1994 precipitation was 62 mm.

Tritium fractionation in anomalous water bound to environmental samples

The quantification of the hydrogen isotope effect involved in the physi- and chemisorption of water on to environmental samples provides the possibility for the characterization of their bound water. This is realized by distilling suspensions of samples in tritiated water ( HTO H 2O ) under near-equilibrium conditions at 5 °C and 5 mbar, and measuring the specific tritium activity in fractions of the condensed gas phase in the course of water removal. The tritium fractionation is caused by isotope effects during the evaporation of water as well as during the sorption of water on to the sample matrix. The two isotope effects are separated by comparing the fractionation in the sample—water mixtures with that in pure water. Three typical soils and maize plant seeds are chosen for investigation. The experimental procedure is tested for model substances with well known amounts of bound water, e.g. Na-tartratedihydrate and molecular sieve. The investigated examples show a different water sorption capacity from 4 to 60% related to the dry substance. Dehydration kinetics further demonstrate the existence of at least four types of sorbed water with distinct fractionation factors. They are characterized by the specific-activity ratio of sorbed water to free water, which ranges from 0.97 to 1.46.

トリチウムおよび｀３５´Ｓトレーサー法による石炭液化反応機構の解明 鉄‐硫黄系触媒および硫黄の添加効果

Effects of addition of the catalyst (pyrrhotite) and sulfur on hydrogen transfer in liquefaction of Taiheiyo coal were investigated using tritium and 35S. The coal liquefaction was performed at the initial pressure of 5.9 MPa and at 400°C for 30 min with tetralin solvent and tritium-labelled hydrogen, with or without the synthesized pyrrhotite catalyst and sulfur (or 35S-labelled sulfur). The specific activities of tritium and 35S in the reaction products were measured with a liquid scintillation counter. Amounts of exchanged and transferred hydrogens between the gas phase and coal/solvent, were calculated from the distributions of tritium and changes in the composition of products.In the reaction with tritiated hydrogen and solvent, the dehydrogenation of tetralin to produce naphthalene and the hydrogen exchange reaction between gas phase and solvent were promoted by added catalyst and sulfur. Added sulfur produced hydrogen sulfide mainly with hydrogen of solvent. A part of added sulfur participated in the sulfur exchange reaction with the pyrrhotite catalyst. In the reaction with tritiated hydrogen, solvent and coal, the hvdrogen addition and exchange to coal and liquefaction products increased with the addition of catalyst and sulfur. It was suggested the sulfur promoted the formation of tetralyl radical in the hydrogen transfer from solvent to coal.

Operating Experience with Tritium Accounting and Analysis Systems in Our Tritium-Handling Facilities

An integral part of the operation of a tritium-handling facility is the need to maintain a tritium accounting system which meets both operational and regulatory requirements. The Darlington Tritium Removal Facility processes heavy water to remove tritium for storage or sale. The Chalk River Tritium Laboratory is a general-purpose experimental facility, handling various quantities and specific activities of tritium in separate systems, including re-packaging of tritium from large metal-getter storage beds to smaller aliquots. Tritium accounting procedures, analytical techniques and operational experience gained to date in the operation of these tritium-handling facilities are described.

Dynamics of Tritiated Water and Organically Bound Tritium after an Acute Tritiated Water Intake in Humans

Tritiated water° (HTO) represents the most important occupational and environmental exposures to tritium, as far as radiation protection is concerned. We have studied the dynamics of HTO and organically bound tritium* (OBT) in urine, feces and blood from male workers at 100-300 d following an acute intake of HTO. Blood and cumulative 24 h urine and fecal samples were collected and analyzed for HTO and OBT. The activity concentrations of HTO in urine and HTO in feces were in equilibrium and were representative of HTO in the body water (HTO in blood). The ratio of bound tritium per gram hydrogen (Bq.gH -1 ) in organic matter in feces (OBT feces ) to urine (OBT urine ) was 1.0 ± 0.1. Similarly, the ratios of OBT blood to OBT urine and OBT blood to OBT feceS were 0.9 ± 0.2 and 1.1 ± 0.2, respectively. These results suggest that, providing an equilibrium condition exists, a measurement of the tritium activity per unit mass of organic matter in urine or feces can provide an assessment of the specific activity of tritium in the organic fraction of the soft tissue. °HTO in this paper means the sum of tritium as water and exchangeable bound tritium. The exchangeable tritium is bound to inorganic moieties (e.g. N, O, P and S) of the biomolecules and is readily exchangeable with the body fluids. *OBT means the tritium bound covalently to carbon in biomolecules and the stable carbon-tritium bond is dissociated by metabolic reactions. The activity concentrations of HTO and OBT in urine samples, from a few days up to 300 d post-exposure, were examined for the clearance kinetics of HTO in urine and OBT in urine. The early presence of OBT in urine indicated that a portion of tritium from the ingested HTO is rapid fixed into organic constituents of the body. The half-life for the longer-term component of OBT in urine is comparable to the half-life for the longer-term component of HTO in urine. This close relationship between the longer-term excretion of HTO in urine and OBT in urine suggests that most of the HTO produced in its longer-term component is a by-product of metabolized OBT. This work has demonstrated that OBT is excreted in urine in all examinations, up to 300 d post-exposure to HTO, and that a fraction of the metabolized OBT is also excreted in feces.

The radiochemical characterization of regular- and high-density concrete from a decommissioned reactor

Samples of high- and regular-density concrete exposed to neutron and gamma irradiation as well as radioactive contamination were obtained from the Nuclear Power Demonstration (NPD) reactor. The radionuclide composition and distribution in these specimens was determined to provide important information on the management of decommissioning waste, concrete decontamination, and the behaviour of radionuclides in cement-based engineered barriers. With the exception of tritium, all radionuclides in regular-density concrete have specific activities less than 5 Bq/g. The tritium-specific activity varied along the length of the core section, but its maximum activity was 82 kBq/g. This was five times the maximum tritium activity found in high-density concrete (15 kBq/g). The effective diffusivity, D e , of tritiated water in high- and regular-density concrete is 5.2 × 10 −9cm 2 · 6.6 × 10 −8cm 2 · s −1 respectively. Radiation fields on contact with the ends of the high-density cores ranged from a low of 150 cpm above background at the boiler-room end of the core to a high of 125 mR/h at the vault-end closest to the neutron flux. The radiation field is due to Co-60, which has a specific activity of 1800 Bq/g when averaged over a distance of 20 cms from the reactor-vault end. Beyond 30 cm into the core the neutron flux attenuation was such that the Co-60 specific activity is <5 Bq/g. Other gamma-emitting radionuclides present in high-density concrete are Mn-54, Cs-134/137, Eu-152/154, and Ce-144, but the specific activity of these is <35 Bq/g. Of the pure beta emitters C-14, Ca-41, and Ni-63, C-14 was the most abundant, with a specific activity of 300 Bq/g. It originates from heavy water containing dissolved 14CO 2.