WorldWideScience

Sample records for acute tritium releases

  1. ACUTRI a computer code for assessing doses to the general public due to acute tritium releases

    Yokoyama, S; Noguchi, H; Ryufuku, S; Sasaki, T

    2002-01-01

    Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety assessment of a nuclear fusion experimental reactor such as ITER. Thus, a computer code, ACUTRI, which calculates the radiological impact of tritium released accidentally to the atmosphere, has been developed, aiming to be of use in a discussion of licensing of a fusion experimental reactor and an environmental safety evaluation method in Japan. ACUTRI calculates an individual tritium dose based on transfer models specific to tritium in the environment and ICRP dose models. In this calculation it is also possible to analyze statistically on meteorology in the same way as a conventional dose assessment method according to the meteorological guide of the Nuclear Safety Commission of Japan. A Gaussian plume model is used for calculating the atmospheric dispersion of tritium gas (HT) and/or tritiated water (HTO). The environmental pathway model in ACUTRI considers the following internal exposures: i...

  2. Tritium transport calculations for the IFMIF Tritium Release Test Module

    Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

  3. Tritium transport calculations for the IFMIF Tritium Release Test Module

    Freund, Jana, E-mail: jana.freund@kit.edu; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

    2014-10-15

    Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

  4. Evaluation of tritium release properties of advanced tritium breeders

    Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li2TiO3) advanced tritium breeders with excess Li (Li2+xTiO3+y) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li2+xTiO3+y pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li2+xTiO3+y pebbles exhibited good tritium release properties similar to the Li2TiO3 pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)

  5. Biological effects of tritium releases from fusion power plants

    Tritium released as tritium oxide is a much more significant potential hazard to the environment than is elemental tritium. Although most biochemical reactions discriminate against the incorporation of tritium in favor of hydrogen, the possibility of some concentration should not be overlooked. A fraction of tritium accumulated as tritiated water becomes organically bound, that is, exchanges with hydrogen bound in organic molecules. The rate and extent of incorporation are dependent upon metabolic activity of the organism. On this basis, the highest concentration of organically-bound tritium would be expected in tissues and population segments which are in formative or growth stages at the time of exposure. Furthermore, as exposure duration increases from acute to chronic situations, tritium concentrations are shown to approach equilibrium levels with a single tritium-to-hydrogen ratio common to all parts of the hydrogen pool. Organic binding would not be expected to result in significant bioaccumulation of tritium from tritiated water. Tritium loss, both from tissue-free water and the tissue-bound fraction, depends upon metabolic activity. Processes that allow accumulation and incorporation of tritium also assist its elimination. Tritium which is organically bound demonstrates a longer half-time, but it would appear to constitute a small fraction of the total tritium label. The radiation exposure of all living organisms by environmentally dispersed tritium, in whatever form, is essentially a whole body exposure. Uncertainties in the individual parameters, involved in converting measured intake to estimated dose equivalent are probably no larger than a factor of three or four. If fusion reactors hold tritium releases with ICRP standards, no significant adverse impact to the environment from those releases are expected

  6. Thermal release of tritium from SS316

    In an effort to improve current understanding of the mechanisms controlling the long-term release of tritium incorporated thermally into stainless steel SS316 and to develop reliable as well as economically feasible techniques for the conditioning of tritium-containing metallic wastes, a systematic investigation is underway in Toyama under carefully controlled conditions. The release rate of tritium from SS316 at ambient pressure was determined experimentally in a flow system at several constant temperatures within the range 287-573 K for rather extended periods of time. Under these conditions HTO was found to constitute by far the most important tritium-containing species being released, i.e. approx. 99 %. Much data has accumulated in recent years with a variety of specimens, i.e. type of stainless steel and specimen dimension, loaded with tritium under different pressure and temperature conditions. Dynamic behavior of long-term tritium release has been successfully modeled using a onedimensional diffusion equation and assuming that the release rate is governed by the tritium flux at the metal surface boundary. The implications of the results for interim storage and thermal conditioning of stainless steel waste will be discussed. (orig.)

  7. Tritium transport and release from lithium ceramic breeder materials

    In an operating fusion reactor,, the tritium breeding blanket will reach a condition in which the tritium release rate equals the production rate. The tritium release rate must be fast enough that the tritium inventory in the blanket does not become excessive. Slow tritium release will result in a large tritium inventory, which is unacceptable from both economic and safety viewpoints As a consequence, considerable effort has been devoted to understanding the tritium release mechanism from ceramic breeders and beryllium neutron multipliers through theoretical, laboratory, and in-reactor studies. This information is being applied to the development of models for predicting tritium release for various blanket operating conditions

  8. Tritium release from neutron irradiated beryllium pebbles

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to 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 response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  9. Release of gaseous tritium during reprocessing

    About 50% of the tritium put through an LWR reprocessing plant is obtained as tritium-bearing water, HTO. Gaseous tritium, HT has a radiotoxicity which is by 4 orders of magnitude lower than that of HTO. A possibility for the removal of HTO could therefore be its conversion into the gas phase with subsequent emission of the HT into the atmosphere. However, model computations which are, in part, supported by experimental data reveal that the radiation exposure caused by HT release is only by about one order of magnitude below that caused by HTO. This is being attributed to the relatively quick reoxidation of HT by soil bacteria. Two alternatives for producing HT from HTO (electrolysis; voloxidation with subsequent electrolysis) are presented and compared with the reference process of deep-well injection of HTO. The authors come to the conclusion that tritium removal by HT release into the atmosphere cannot be recommended at present under either radiological or economic aspects. (orig.)

  10. Tritium releases, birth defects and infant deaths

    The AECB has published a report 'Tritium releases from the Pickering Nuclear Generating Station and Birth Defects and Infant Mortality in Nearby Communities 1971-1988' (report number INFO-0401). This presents the results of a detailed analysis of deaths and birth defects occurring in infants born to mothers living in the area (25 Km radius) of the Pickering nuclear power plant, over an 18-year period. The analysis looked at the frequency of these defects and deaths in comparison to the general rate for Ontario, and also in relation to airborne and waterborne releases of tritium from the power plant. The overall conclusion was that the rates of infant death and birth defects were generally not higher in the study population than in all of Ontario. There was no prevalent relationship between these deaths and defects and tritium releases measured either at the power plant or by ground monitoring stations t some distance from the facility

  11. Estimated Release of Tritium from 232-F Concrete Rubble

    This report describes an estimate of the release of tritium from contaminated concrete from the demolition of the old 232-F Tritium Facility at the Savannah River Site. The estimate uses data from the scientific literature and information about tritium migration in concrete developed during studies of tritium in concrete at SRS

  12. Development of online measurement system of tritium in out-of-pile tritium release experiment

    It is very important to accurately measure tritium concentration and morphology for mastering tritium release behavior of tritium production breeder and improving the performance of tritium production breeder in out-of-pile tritium release experiment. According to the characteristics of small flow of carrier gas, small quantity of gas and argon as carrier gas, based on the ionization chamber, the digital online tritium measurement system was developed. The sensitive volume of ionization chamber is 50 mL, the digital instrument can automatically obtain the tritium concentration, at the same time, process and display it. The results show that the saturated zone of ionization chamber is about 35 V in argon, and the detection limit is 3.7 × 107 Bq/m3, which is satisfactory for online tritium measurement in out-of-pile tritium release experiment. (authors)

  13. CANDU 6 nuclear power plant tritium control and release

    The issues drawing people's attention, such as ways of CANDU plant tritium generation, measures to control tritium release to environment in the design of nuclear power plants as well as public dose due to tritium released to the environment are presented

  14. Tritium release from irradiated lithium aluminate, can it be improved?

    Lithium aluminate is an attractive material (in terms of its chemical, mechanical and irradiation properties) for breeding tritium in fusion reactors; however, its tritium release characteristics are not as good as those of other candidate materials. To investigate whether tritium release from lithium aluminate can be improved, we have studied the tritium release from irradiated samples of pure lithium aluminate, lithium aluminate doped with Mg, and lithium aluminate with a surface deposit of platinum. The release was studied by the temperature programmed desorption (TPD) method. Both the platinum coating and magnesium doping were found to improve the tritium release characteristics, as determined by TPD. Tritium release shifted to states with lower activation energies for the altered materials

  15. Chronic release of tritium from stainless steel 316

    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. (author)

  16. Doses due to tritium releases by NET - data base and relevant parameters on biological tritium behaviour

    This study gives an overview on the current knowledge about the behaviour of tritium in plants and in food chains in order to evaluate the ingestion pathway modelling of existing computer codes for dose estimations. The tritium uptake and retention by plants standing at the beginning of the food chains is described. The different chemical forms of tritium, which may be released into the atmosphere (HT, HTO and tritiated organics), and incorporation of tritium into organic material of plants are considered. Uptake and metabolism of tritiated compounds in animals and man are reviewed with particular respect to organically bound tritium and its significance for dose estimations. Some basic remarks on tritium toxicity are also included. Furthermore, a choice of computer codes for dose estimations due to chronic or accidental tritium releases has been compared with respect to the ingestion pathway. (orig.)

  17. Tritium releases and impact about EDF nuclear power stations

    After a description of the different ways of formation of tritium in the nuclear power stations (either by fission or by activation), the authors discuss the levels of tritium releases by these power stations, indicate the tritium average activities in liquid and gaseous radioactive releases in 2008. They indicate the choices made by EDF and the actions performed to control these releases. They describe how the presence of tritium in the environment is monitored and how measurements are published. They discuss the interpretation of these measurements (in water streams, water sheets, sediments, along the Channel French coasts), and the impact of the tritium released by the nuclear power stations. They evoke modelling studies and researches supported by EDF on the impact of tritium on mankind

  18. Preliminary analysis of public dose from CFETR gaseous tritium release

    Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230027 (China); Ni, Muyi, E-mail: muyi.ni@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Lian, Chao; Jiang, Jieqiong [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2015-02-15

    Highlights: • Present the amounts and limit dose of tritium release to the environment for CFETR. • Perform a preliminary simulation of radiation dose for gaseous tritium release. • Key parameters about soil types, wind speed, stability class, effective release height and age were sensitivity analyzed. • Tritium release amount is recalculated consistently with dose limit in Chinese regulation for CFETR. - Abstract: To demonstrate tritium self-sufficiency and other engineering issues, the scientific conception of Chinese Fusion Engineering Test Reactor (CFETR) has been proposed in China parallel with ITER and before DEMO reactor. Tritium environmental safety for CFETR is an important issue and must be evaluated because of the huge amounts of tritium cycling in reactor. In this work, different tritium release scenarios of CFETR and dose limit regulations in China are introduced. And the public dose is preliminarily analyzed under normal and accidental events. Furthermore, after finishing the sensitivity analysis of key input parameters, the public dose is reevaluated based on extreme parameters. Finally, tritium release amount is recalculated consistently with the dose limit in Chinese regulation for CFETR, which would provide a reference for tritium system design of CFETR.

  19. Preliminary analysis of public dose from CFETR gaseous tritium release

    Highlights: • Present the amounts and limit dose of tritium release to the environment for CFETR. • Perform a preliminary simulation of radiation dose for gaseous tritium release. • Key parameters about soil types, wind speed, stability class, effective release height and age were sensitivity analyzed. • Tritium release amount is recalculated consistently with dose limit in Chinese regulation for CFETR. - Abstract: To demonstrate tritium self-sufficiency and other engineering issues, the scientific conception of Chinese Fusion Engineering Test Reactor (CFETR) has been proposed in China parallel with ITER and before DEMO reactor. Tritium environmental safety for CFETR is an important issue and must be evaluated because of the huge amounts of tritium cycling in reactor. In this work, different tritium release scenarios of CFETR and dose limit regulations in China are introduced. And the public dose is preliminarily analyzed under normal and accidental events. Furthermore, after finishing the sensitivity analysis of key input parameters, the public dose is reevaluated based on extreme parameters. Finally, tritium release amount is recalculated consistently with the dose limit in Chinese regulation for CFETR, which would provide a reference for tritium system design of CFETR

  20. New techniques for the measurement of tritium activity releases in the air in heavy water reactors

    Tritium constitutes the principal potential source of ionizing radiation in heavy water reactors, playing the role of the dominant radionuclide of concern. In such reactors, the tritium hazard under normal working conditions is as important as all other radionuclide hazards combined. This calls for the employment of tritium monitoring systems capable of giving information about any releases in the shortest possible time so that immediate corrective measures could be adopted and operating personnel informed before they receive doses greater than permissible. Monitoring of tritium becomes complicated because of very low energy of tritium betas and presence of large concentrations of interfering activity and high background gamma fields inside the reactors. At the Rajasthan Atomic Power Station, for example, gamma compensated ionization chamber type of tritium monitors have been used but they have not been found entirely satisfactory. In this way, a period at least 30 minutes elapses before the level of tritium release could be known. This time delay in assessing the containment levels can prove critical in the event of acute releases and can result in excessive exposure to the operating personnel from tritium contamination. The need to undertake the present work arose. 4

  1. Tritium release from SS316 under vacuum condition

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

  2. Tritium release from SS316 under vacuum condition

    Torikai, Y.; Penzhorn, R.D. [Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan)

    2015-03-15

    The plasma facing surface of the ITER vacuum vessel, partly made of low carbon austenitic stainless steel type 316L, will incorporate tritium during machine operation. In this paper the kinetics of tritium release from stainless steel type 316 into vacuum and into a noble gas stream are compared and modelled. Type 316 stainless steel specimens loaded with tritium either by exposure to 1.2 kPa HT at 573 K or submersion into liquid HTO at 298 K showed characteristic thin surface layers trapping tritium in concentrations far higher than those determined in the bulk. The evolution of the tritium depth profile in the bulk during heating under vacuum was non-discernible from that of tritium liberated into a stream of argon. Only the relative amount of the two released tritium-species, i.e. HT or HTO, was different. Temperature-dependent depth profiles could be predicted with a one-dimensional diffusion model. Diffusion coefficients derived from fitting of the tritium release into an evacuated vessel or a stream of argon were found to be (1.4 ± 1.0)*10{sup -7} and (1.3 ± 0.9)*10{sup -9} cm{sup 2}/s at 573 and 423 K, respectively. Polished surfaces on type SS316 stainless steel inhibit considerably the thermal release rate of tritium.

  3. Helium irradiation effects on tritium retention and long-term tritium release properties in polycrystalline tungsten

    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 × 1017 He/cm2, and at 1 × 1018 He/cm2 it became smaller compared to 1 × 1017 He/cm2. 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

  4. Tritium

    This report contains information on chemical and physical properties, occurence, production, use, technology, release, radioecology, radiobiology, dose estimates, radioprotection and legal aspects of tritium. The objective of this report is to provide a reliable data base for the public discussion on tritium, especially with regard to its use in future nuclear fusion plants and its radiological assessment. (orig.)

  5. Transport phenomena and occupational consequences of a tritium release

    Tritium gas, normally in sealed containers, will be present in the US Department of Energy's (DOE's) facilities conducting fusion energy research. A probability of tritium release, however small, exists in these facilities. Once released, tritium can back-diffuse against ventilation flow to contaminate other areas of the facility. Tritium can also be released to the environment by exhaust blowers. The problem of back-diffusion of tritium released in a typical DOE facility was examined as a function of flow rates of the ventilation system. The source term (release to the environment) in the emergency ventilation flow was also calculated. The consequences to personnel in the release room and in an adjacent corridor due to back-diffusion were determined. It was shown that for credible release scenarios, the consequences in the adjacent corridor from tritium back-diffusion were negligible. Higher doses in the release room can be avoided by well-planned emergency evacuation procedures. The source term was calculated, but the on- and off-site consequences were not determined

  6. Low temperature tritium release experiment from lithium titanate breeder material

    Engineering data of neutron irradiation performance are needed to design a fusion blanket. Of the engineering data, tritium release characteristic is one of the most important data. Therefore, tritium release experiments of the tritium breeding materials were carried out to evaluate the effects of various parameters, i.e. sweep-gas flow rate, irradiation temperature, hydrogen content in sweep gas and so on, on tritium release. Lithium titanate (Li2TiO3) is a candidate tritium breeding material for the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to enhance tritium release from the breeder and to reduce the induced thermal stress in the breeder. Li2TiO3 pebbles with a diameter of 1mm and a total weight of ∼134g have been fabricated, and a pebble-pac assembly of the Li2TiO3 pebbles was irradiated in the Japan Materials Testing Reactor (JMTR), for 3 cycles (about 75 days). The tritium generated in breeder, and released from the breeder was swept downstream by the sweep gas for on-line analysis of tritium content. The total concentration and gaseous concentration of tritium released from the Li2TiO3 pebbles were measured, and HT/(HT+HTO) ratio was evaluated. The sweep-gas flow rate was changed from 10 to 1,000cm3/min, and hydrogen concentration in the sweep gas was changed from 100 to 10,000 ppm. The irradiation temperature of the outer region of the pebble-pac assembly was held below 450degC. The results showed that tritium release from the Li2TiO3 pebbles was started between 100 and 140degC and that the amount of released with increasing the irradiation temperature. The sweep-gas flow rate did not have an effect on tritium release from the Li2TiO3 pebble bed in the steady state. On the other hand, the hydrogen content in the sweep gas had an effect on the tritium release from the Li2TiO3 pebble bed. (author)

  7. Modeling unusual tritium release behavior from Li2O

    This paper presents a diffusion-desorption tritium release model in which the unusual tritium-release behavior observed in the CRITIC experiment is accounted for by an activation energy of desorption that is surface coverage dependent. Desorption and adsorption activation energies which are dependent on the amount of surface coverage have been reported. The current model is capable of reproducing both the unusual and the normal tritium release observed in CRITIC and predicts other regions where the surface-coverage-dependent release behavior may be observed. Results from the CRITIC experiment and our calculations imply that the details of the surface phenomena must be known to accurately predict the tritium inventory and changes in inventory that occur with changes in the breeder-material environment. 29 refs., 4 figs

  8. An atmospheric tritium release database for model comparisons. Revision 1

    A database of vegetation, soil, and air tritium concentrations at gridded coordinate locations following nine accidental atmospheric releases is described. While none of the releases caused a significant dose to the public, the data collected are valuable for comparison with the results of tritium transport models used for risk assessment. The largest, potential, individual off-site dose from any of the releases was calculated to be 1.6 mrem. The population dose from this same release was 46 person-rem which represents 0.04% of the natural background radiation dose to the population in the path of the release

  9. An atmospheric tritium release database for model comparisons

    A database of vegetation, soil, and air tritium concentrations at gridded coordinate locations following nine accidental atmospheric releases is described. While none of the releases caused a significant dose to the public, the data collected is valuable for comparison with the results of tritium transport models used for risk assessment. The largest, potential, individual off-site dose from any of the releases was calculated to be 1.6 mrem. The population dose from this same release was 46 person-rem which represents 0.04% of the natural background radiation dose to the population in the path of the release

  10. Description of tritium release from lithium titanate at constant temperature

    Pena, L.; Lagos, S.; Jimenez, J.; Saravia, E. [Comision Chilena de Energia Nuclear, Santiago (Chile)

    1998-03-01

    Lithium Titanate Ceramics have been prepared by the solid-state route, pebbles and pellets were fabricated by extrusion and their microstructure was characterized in our laboratories. The ceramic material was irradiated in the La Reina Reactor, RECH-1. A study of post-irradiation annealing test, was performed measuring Tritium release from the Lithium Titanate at constant temperature. The Bertone`s method modified by R. Verrall is used to determine the parameters of Tritium release from Lithium Titanate. (author)

  11. Tritium release behavior from SS316 under vacuum

    The release behavior of tritium from stainless steel 316 previously loaded with tritium gas was investigated isothermally under an argon gas flow or under vacuum conditions. Depth profiles of the heated specimens were determined by chemical etching. The experimental release rate and tritium depth profiles in the solid could be simulated with a model based on bulk diffusion. The diffusion coefficients used to fit the depth profiles of specimens heated under vacuum conditions were not discernible from those obtained for specimens heated under an argon gas flow at the same temperature. (author)

  12. Tritium release during RB reactor operation

    Tritium content in daily precipitation, ground condensation and atmospheric water vapor samples was monitoring during normal RB reactor operating conditions in November and December 1994. Generated fission energy from the reactor was ranging between 1.7 MWh and 14.62 Wh per day. Tritium concentrations in precipitation and ground condensation were 1.4-132.6 Bq/l and 47.2-1204 Bq/l, respectively. Tritium content in atmospheric water vapor in HTO from varied from 2.8 to 6.2 Bq/m3. (author)

  13. MODELING ATMOSPHERIC RELEASES OF TRITIUM FROM NUCLEAR INSTALLATIONS

    Okula, K

    2007-01-17

    Tritium source term analysis and the subsequent dispersion and consequence analyses supporting the safety documentation of Department of Energy nuclear facilities are especially sensitive to the applied software analysis methodology, input data and user assumptions. Three sequential areas in tritium accident analysis are examined in this study to illustrate where the analyst should exercise caution. Included are: (1) the development of a tritium oxide source term; (2) use of a full tritium dispersion model based on site-specific information to determine an appropriate deposition scaling factor for use in more simplified, broader modeling, and (3) derivation of a special tritium compound (STC) dose conversion factor for consequence analysis, consistent with the nature of the originating source material. It is recommended that unless supporting, defensible evidence is available to the contrary, the tritium release analyses should assume tritium oxide as the species released (or chemically transformed under accident's environment). Important exceptions include STC situations and laboratory-scale releases of hydrogen gas. In the modeling of the environmental transport, a full phenomenology model suggests that a deposition velocity of 0.5 cm/s is an appropriate value for environmental features of the Savannah River Site. This value is bounding for certain situations but non-conservative compared to the full model in others. Care should be exercised in choosing other factors such as the exposure time and the resuspension factor.

  14. Tritium levels in milk in the vicinity of chronic tritium releases.

    Le Goff, P; Guétat, Ph; Vichot, L; Leconte, N; Badot, P M; Gaucheron, F; Fromm, M

    2016-01-01

    Tritium is the radioactive isotope of hydrogen. It can be integrated into most biological molecules. Even though its radiotoxicity is weak, the effects of tritium can be increased following concentration in critical compartments of living organisms. For a better understanding of tritium circulation in the environment and to highlight transfer constants between compartments, we studied the tritiation of different agricultural matrices chronically exposed to tritium. Milk is one of the most frequently monitored foodstuffs in the vicinity of points known for chronic release of radionuclides firstly because dairy products find their way into most homes but also because it integrates deposition over large areas at a local scale. It is a food which contains all the main nutrients, especially proteins, carbohydrates and lipids. We thus studied the tritium levels of milk in chronic exposure conditions by comparing the tritiation of the main hydrogenated components of milk, first, component by component, then, sample by sample. Significant correlations were found between the specific activities of drinking water and free water of milk as well as between the tritium levels of cattle feed dry matter and of the main organic components of milk. Our findings stress the importance of the metabolism on the distribution of tritium in the different compartments. Overall, dilution of hydrogen in the environmental compartments was found to play an important role dimming possible isotopic effects even in a food chain chronically exposed to tritium. PMID:26551587

  15. Tritium and helium retention and release from irradiated beryllium

    Anderl, R.A.; Longhurst, G.R.; Oates, M.A.; Pawelko, R.J. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

    1998-01-01

    This paper reports the results of an experimental effort to anneal irradiated beryllium specimens and characterize them for steam-chemical reactivity experiments. Fully-dense, consolidated powder metallurgy Be cylinders, irradiated in the EBR-II to a fast neutron (>0.1 MeV) fluence of {approx}6 x 10{sup 22} n/cm{sup 2}, were annealed at temperatures from 450degC to 1200degC. The releases of tritium and helium were measured during the heat-up phase and during the high-temperature anneals. These experiments revealed that, at 600degC and below, there was insignificant gas release. Tritium release at 700degC exhibited a delayed increase in the release rate, while the specimen was at 700degC. For anneal temperatures of 800degC and higher, tritium and helium release was concurrent and the release behavior was characterized by gas-burst peaks. Essentially all of the tritium and helium was released at temperatures of 1000degC and higher, whereas about 1/10 of the tritium was released during the anneals at 700degC and 800degC. Measurements were made to determine the bulk density, porosity and specific surface area for each specimen before and after annealing. These measurements indicated that annealing caused the irradiated Be to swell, by as much as 14% at 700degC and 56% at 1200degC. Kr gas adsorption measurements for samples annealed at 1000degC and 1200degC determined specific surface areas between 0.04 m{sup 2}/g and 0.1 m{sup 2}/g for these annealed specimens. The tritium and helium gas release measurements and the specific surface area measurements indicated that annealing of irradiated Be caused a porosity network to evolve and become surface-connected to relieve internal gas pressure. (author)

  16. Release of tritium from fuel and collection for storage

    Burger, L.L.; Trevorrow, L.E.

    1976-04-01

    Recent work is reviewed on the technology that has been suggested as applicable to collection and storage of tritium in anticipation of the necessity of that course of action. Collection technology and procedures must be adapted to the tritium-bearing effluent and to the facility from which it emerges. Therefore, this discussion of tritium collection technology includes some information on the processes from which release is expected to occur, the amounts, the nature of the effluent media, and the form in which tritium appears. Recent work on collection and storage concepts has explored, both by experimentation and by feasibility analyses, the operations generally aimed at producing recycle, collection, or storage of tritium from these streams. Storage concepts aimed specifically at tritium involve plans to store volumes ranging from that of the entire effluent stream to only that of a small volume of a concentrate. Decisions between storage of unconcentrated streams and storage of concentrates are expected to be made largely by weighing the cost of storage space against the cost of concentration. The storage of tritium concentrate requires the selection of a form of tritium possessing physical and chemical properties appropriate for the expected storage conditions. This selection of an appropriate storage form has occupied a major portion of recent work concerned with tritium storage concepts. In summary, within the context of present regulations and expected amounts of waste tritium; this waste can be disposed of by dilution and dispersal to the environment. In the future, however, more restrictive regulations might be introduced that could be satisfied only by some collection and storage operations. Technology for this practice is not now available, and the present discussion reviews recent activities devoted to its development.

  17. Enhancing tritium release from diffusion-limited solid lithium compounds

    Mathematical modeling and numerical calculations have been performed to examine methods for exploiting recoil effects to increase the release of tritium from solid lithium compounds whose release rates are limited by the diffusion process. The basic concept is to employ the kinetic energy of the tritons from the exothermic 6L(n,4He)T reaction in order to move them out of the low-diffusivity region where they are born and into a thin, high-diffusivity region from which they can more easily migrate for eventual removal by a stream of purge gas. In the recoil-enhanced release approach, the lithium-containing blanket particles would consist of coated spheres. The inner region of the spherical particles would have a small diameter (30 to 40 μm) and would contain the lithium compound for tritium production. The outer region of the spherical particles would consist of a thin, highly diffusive coating whose thickness would be approximately one-half the range of a 2.7-MeV triton in the coating material. Tritium concentration profiles are presented parametrically in terms of dimensionless space and time variables and in terms of the ratio of the tritium diffusion coefficients for the inner and outer materials of a spherical particle. Calculations of tritium diffusion were performed for lithium-compound-to-coating diffusion coefficient ratios of 1.0,0.5,0.1, and 0.05. The results indicate that, at steady state, the tritium inventory is directly proportional to the diffusion coefficient in the coating and the time to reach steady state is reduced as the diffusion coefficient ratio is decreased

  18. Retention and release of tritium in aluminum clad, Al-Li alloys

    Tritium retention in and release from aluminum clad, aluminum-lithium alloys is modeled from experimental and operational data developed during the thirty plus years of tritium production at the Savannah River Site. The model assumes that tritium atoms, formed by the 6Li(n,α)3He reaction, are produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly becomes supersaturated in tritium. Newly produced tritium atoms are trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability is the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release is determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. This model is used to calculate tritium release from aluminum clad, aluminum-lithium alloys. 9 refs., 3 figs

  19. Tritium release from neutron irradiated lithium inorganic compounds

    Tritium release from irradiated lithium inorganic compounds Li/sub 2/O, Li/sub 2/SO/sub 4/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/, LiA10/sub 2/, Li/sub 2/TiO/sub 3/, LiNbO/sub 3/ was studied in isochronic and isothermal conditions in the temperature range 200-9000C. The samples were prepared by outgassing at 600-6500 in quartz ampules, sealed and then irradiated under thermal neutron flux about 1.10/sup 13/n cm/sup -2/s/sup -1/ with tritium concentration ≅3.7 10/sup 8/-3.7 10/sup 9/Bq/g. The initial stage of the tritiated water recovery from complex lithium inorganic compounds is characterized by a rapid tritium release accompanied by defect annealing and release energy accumulated by a solid state during irradiation. The two temperature ranges were observed as a result of the OT' groups formation which are in two energy states due to the formation of bonds M-O-T(H) (where M-Si, A1, Ti, Nb...) and Li-O-T(h). Symbol H emphasizes the importance of the residual content of the OH groups in the initial materials which should be considered as an inorganic polymers having the properties of solid acids and bases. The process of tritium release from irradiated lithium inorganic compounds is a many stage process including the following steps: tritium (ion) diffusion inside a crystal lattice; formation of OT groups on the surface of oxygen compounds, recombination of OT and OH groups to form a water molecule that is detected in a gas phase as a product of the annealing process

  20. Computer simulation of tritium releases in inertial fusion reactors

    Perlado, J.M.; Velarde, M. [Universidad Politecnica de Madrid, Instituto de Fusion Nuclear, DENIM (Spain)

    2000-07-01

    Accidental releases of tritium from Inertial Fusion reactors are presented. A well-established computer code, MACCS2, is used with realistic models. Release fractions of 1 - 10 - 50 - 100 % of inventories are considered, with height of emissions 10, 30, 60 m, and duration of 10 min. and 2 hours. Only early emergency phase is considered with mitigative actions and shielding factors. It is concluded that except in 100 % releases for some reactors and heights the effective doses to workers and general population does not exceed the regulatory limits. Differences with very conservative results can attain 2 orders of magnitude. (authors)

  1. Measurement of uptake and release of tritium by tungsten

    Tungsten is currently contemplated as plasma facing material for the divertor of future fusion machines. In this paper the uptake of tritium by tungsten and its release behavior have been investigated. Tungsten samples have been annealed at various temperatures and loaded at also different temperatures with deuterium containing 7.2 % tritium at a pressure of 1.2 kPa. A specific system was designed to assess the release of tritiated water and molecular tritium by the samples. Due to the rather low solubility of hydrogen isotopes in tungsten it is particularly important to be aware of the presence of hydrogen traps or thin oxide films. As shown in this work, traps or oxide films may affect the retention capability of tungsten and lead to significantly modified release properties. It became clear that there were capture sites that had different thermal stability and different capture intensity in tungsten after polishing, or oxide films that were grown on the surface of tungsten and had barrier effects

  2. Environmental health-risk assessment for tritium releases from the National Tritium Labeling Facility (NTLF) at Lawrence Berkeley Laboratory

    This report is a health risk assessment that addresses continuous releases of tritium to the environment from the National Tritium Labeling Facility (NTLF) at the Lawrence Berkeley Laboratory (LBL). The NTLF contributes approximately 95% of all tritium releases from LBL. Transport and transformation models were used to determine the movement of tritium releases from the NRLF to the air, surface water, soils, and plants and to determine the subsequent doses to humans. These models were calibrated against environmental measurements of tritium levels in the vicinity of the NTLF and in the surrounding community. Risk levels were determined for human populations in each of these zones. Risk levels to both individuals and populations were calculated. In this report population risks and individual risks were calculated for three types of diseases--cancer, heritable genetic effects, and developmental and reproductive effects

  3. Environmental health-risk assessment for tritium releases from the National Tritium Labeling Facility (NTLF) at Lawrence Berkeley Laboratory

    McKone, T.E.; Brand, K.P.

    1994-12-01

    This report is a health risk assessment that addresses continuous releases of tritium to the environment from the National Tritium Labeling Facility (NTLF) at the Lawrence Berkeley Laboratory (LBL). The NTLF contributes approximately 95% of all tritium releases from LBL. Transport and transformation models were used to determine the movement of tritium releases from the NRLF to the air, surface water, soils, and plants and to determine the subsequent doses to humans. These models were calibrated against environmental measurements of tritium levels in the vicinity of the NTLF and in the surrounding community. Risk levels were determined for human populations in each of these zones. Risk levels to both individuals and populations were calculated. In this report population risks and individual risks were calculated for three types of diseases--cancer, heritable genetic effects, and developmental and reproductive effects.

  4. Chronic Release of Tritium from SS316 at Ambient Temperature: Correlation Between Depth Profile and Tritium Liberation

    One conceivable option for the disposal of tritium-contaminated stainless steel consists in its storage at ambient temperature in a purged containment. To assess this option several stainless steel 316 specimens, previously loaded at elevated temperatures with 0.8-8.5 MBq of tritium, were flushed continuously with dry argon (water partial pressure 0.073 Pa) for extended periods of time. The released tritium (more than 99 % in the form of tritiated water (HTO)) was collected in bubblers and monitored periodically by liquid scintillation counting. After an initial fast liberation a fairly constant rate of the order of 0.2 % per day established. Tritium depth profile in the SS specimens could be simulated by a diffusion limited desorption model. The rate determining step for tritium release appears to be bulk diffusion

  5. Tritium release from lithium orthosilicate pebbles deposited with palladium

    Full text of publication follows: Slightly over-stoichiometric lithium orthosilicate pebbles have been selected as one optional breeder material for the European Helium Cooled Pebble Bed (HCPB) blanket. This material has been developed in collaboration of Research Center Karlsruhe and the Schott Glass, Mainz. The lithium orthosilicate pebbles are fabricated from lithium hydroxide and silica by a melting and spraying method in a semi-industrial scale facility. Lithium hydroxide was selected as the precursor since enriched lithium hydroxide is commercially available. The lithium orthosilicate pebbles produced by the process contains oxide phases besides orthosilicate, but it was also found that the oxide phases can be decomposed by annealing at high temperatures. The lithium orthosilicate pebbles produced in this way possesses satisfactory pebble characteristics. Therefore, the authors performed out-of-pile annealing tests using the lithium orthosilicate pebbles irradiated in a research reactor. Moreover, the effect of the deposition of palladium in the lithium orthosilicate pebbles on the behavior of tritium release was investigated. Palladium was deposited in the lithium orthosilicate pebbles by the incipient wet impregnation method using a solution of a palladium amino complex. The lithium orthosilicate pebbles were submitted to neutron irradiation at the Kyoto university research reactor. In the out-of-pile annealing experiments, the temperature of the breeder material placed in a tubular reactor made of quartz was raised from ambient temperature to 1173 K at a constant rate of 5 K/min under the stream of sweep gases. The tritium concentration in the outlet stream of the reactor was traced with two ionization chambers. The ionization chambers were used with a water bubbler, which enables to measure the concentrations of molecular form of tritium (HT) and tritiated water vapor (HTO) separately. In the experiments, a 0.1 % hydrogen/nitrogen sweep gas was used. The

  6. Chemical form of tritium released from solid breeder materials and the influences of it on a bred tritium recovery systems

    The ratio of HTO in total tritium was measured at release of the bred tritium to the purge gas with hydrogen using the thermal release after irradiation method, where neutron irradiation was performed at JRR-3 reactor in JAERI or KUR reactor in Kyoto University. It is experimentally confirmed in this study that not a small portion of bred tritium is released to the purge gas in the form of HTO form ceramic breeder materials even when hydrogen is added to the purge gas. The chemical composition is to be decided by the competitive reaction at the grain surface of a ceramic breeder material where desorption reaction, isotope exchange reaction 1, isotope exchange reaction 2 and water formation reaction are considered to take part. Observation in this study implies that it is necessary to have a bred tritium recovery system applicable for both HT and HTO form to recover whole bred tritium. The chemical composition also decides the amount of tritium transferable to the cooling water of the electricity generation system through the structural material in the blanket system. Permeation behavior of tritium through some structural materials at various conditions are also discussed. (author)

  7. Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

    Robinson, Sharon M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chattin, Marc Rhea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giaquinto, Joseph [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). 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. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T2O. In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding

  8. Environmental effects of a tritium release from the Savannah River Plant

    On March 27, 1981, a small amount of tritiated water was inadvertently released from the tritium-processing facility during a routine maintenance operation. This report describes the environmental effects of this release both on the SRP site and offsite. Also, the operation of the WIND (Wind Information and Display) emergency response system during the incident is discussed, and the predicted and diagnosed behavior of the tritium plume is compared with tritium concentrations deduced from air, vegetation, soil, and bioassay samples

  9. Environmental effects of a tritium release from the Savannah River Plant

    Garrett, A.J.; Wilhite, E.L.; Buckner, M.R.

    1981-11-01

    On March 27, 1981, a small amount of tritiated water was inadvertently released from the tritium-processing facility during a routine maintenance operation. This report describes the environmental effects of this release both on the SRP site and offsite. Also, the operation of the WIND (Wind Information and Display) emergency response system during the incident is discussed, and the predicted and diagnosed behavior of the tritium plume is compared with tritium concentrations deduced from air, vegetation, soil, and bioassay samples.

  10. Model parameters and validation for tritium transfer in plants from atmospheric release

    Model parameters and validation for tritium transfer in plants from atmospheric release are examined in different effluent modes. In most cases, tritium uptake by plants can be explained using simple models based on the flux of transpiration and/or vapor diffusion. But, concerning the organically bound tritium in plants, the production rate of it differed with different plant species and plant parts. So, the modeling of the production rate of OBT in target plants and parts still needs experimental results and theoretical consideration. For the release of atmospheric tritiated organic material, the mechanism of tritium incorporation into plant should be known. Tritium was detected in the plant leaves which were exposed to tritiated methane, not only in the water soluble form but also in the organically bound tritium form. The mechanism of this tritium accumulation in plant leaves is still uncertain. (author)

  11. Tritium breeding and release-rate kinetics from neutron-irradiated lithium oxide

    The research encompasses the measurement of the tritium breeding and release-rate kinetics from lithium oxide, a ceramic tritium-breeding material. A thermal extraction apparatus which allows the accurate measurement of the total tritium inventory and release rate from lithium oxide samples under different temperatures, pressures and carrier-gas compositions with an uncertainty not exceeding 3% was developed. The goal of the Lithium Blanket Module program was to determine if advanced computer codes could accurately predict the tritium production in the lithium oxide blanket of a fusion power plant. A fusion blanket module prototype was built and irradiated with a deuterium-tritium fusion-neutron source. The tritium production throughout the module was modeled with the MCNP three dimensional Monte Carlo code and was compared to the assay of the tritium bred in the module. The MCNP code accurately predicted tritium-breeding trends but underestimated the overall tritium breeding by 30%. The release rate of tritium from small grain polycrystalline sintered lithium oxides with a helium carrier gas from 300 to 450 C was found to be controlled by the first order surface desorption of monotritiated water. When small amounts of hydrogen were added to the helium carrier gas, the first order rate constant increased from the isotopic exchange of hydrogen for tritium at the lithium oxide surface occurring in parallel with the first order desorption process. The isotopic-exchange first order rate constant temperature dependence and hydrogen partial pressure dependence were evaluated

  12. Development of LiF tile neutron shield and measurement of tritium release from it

    For neutron capture therapy of cancer, the neutron irradiation field with low gamma-ray is essential for selective treatment. From various lithium compounds, lithium fluoride LiF was selected as the shielding material for the present purpose, because of 1 large lithium density, 2 chemical stability, 3 easy treatment for nonpoison, and 4 small induced activity. In order to utilized LiF in pure chemical form, we have developed LiF tile, although as yet few investigations have been made of sintering a material in fluoride form. From viewpoint of ceramic technology, some new facts have been observed. The behavior of tritium, produced by the reaction of 6Li(n, α)T, in LiF tile was experimentally clarified. Tritium is released from LiF tile in two processes: (1) tritium released by recoil immediately after neutron irradiation, (2) tritium released with temperature condition and elasped time after temporaty containment in LiF tile. The former tritium was trapped in ethanol and measured with a liquid scintillator. While the latter was released in a loop by heating the irradiated tile and measured with a gas-flow-type tritium monitor. The following facts have been clarified: (1) Amount of recoiled tritium from LiF tile is --0.11 μCi/cm2/1014 nvt. (2) Most of tritium produced in LiF tile is contained in the tile itself. (3) Tritium contained in LiF tile is not released at temperature less than 3000C. (4) Contained tritium is released mainly at temperature between 400 -- 6500C. (5) The higher temperature at which LiF tile was sintered, the better containment of tritium. We have finally succeeded in developing LiF tile with low tritium release as neutron shielding material, which is now on sale in commercial base. (author)

  13. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Kurinskiy, P., E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Rolli, R. [Karlsruhe Institute of Technology, Institute for Applied Materials – Materials and Biomechanics (IAM-WBM), P.O. Box 3640, Karlsruhe 76021 (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

    2014-10-15

    Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: 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.

  14. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: 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

  15. International comparison of computer codes for modelling the dispersion and transfer of tritium released to the atmosphere

    Computer codes for modelling the dispersion and transfer of tritium released to the atmosphere were compared. The codes originated from Canada, the United States, Sweden and Japan. The comparisons include acute and chronic emissions of tritiated water vapour or elemental tritium from a hypothetical nuclear facility. Individual and collective doses to the population within 100 km of the site were calculated. The discrepancies among the code predictions were about one order of magnitude for the HTO emissions but were significantly more varied for the HT emissions. Codes that did not account for HT to HTO conversion and cycling of tritium in the environment predicted doses that were several orders of magnitude less than codes that incorporate this feature into the model

  16. Tritium release from beryllium discs and lithium ceramics irradiated in the SIBELIUS experiment

    The SIBELIUS experiment was designed to obtain information on the compatibility between beryllium and ceramics, as well as beryllium and steel, in a neutron environment. This experiment comprised irradiation of eight capsules, seven of which were independently purged with a He/0.1% H2 gas mixture. Four capsules were used to examine beryllium/ceramic (Li2O, LiAlO2, Li4SiO4, and Li2ZrO3) and beryllium/steel (Types 316L and 1.4914) compacts. Isothermal anneal experiments have been run on representative beryllium and ceramic disks from each of the four capsules at 550 degrees C to 850 degrees C in steps of 100 degrees C. The results indicate that tritium release from the beryllium did not exhibit burst release behavior, as previously reported, but rather a progressive release with increasing temperature. Generally, ∼99% of the tritium was released by 850 degrees C. Tritium release from the ceramic discs was quite similar to the behavior shown in other dynamic tritium release experiments on lithium ceramics. The tritium content in beryllium discs adjacent to a steel sample was found to be significantly lower than that found in a beryllium disc adjacent to a ceramic sample. Recoil of tritium from the ceramic into the beryllium appears to be the source of tritium entering the beryllium, probably residing in the beryllium oxide layer

  17. Development of dose assessment code for release of tritium during normal operation of nuclear power plants

    A computer code PTMHTO has been developed to assess tritium doses to the general public. The code enables to simulate the behavior of tritium in the environment released into the atmosphere under normal operation of nuclear power plants. Code can calculate the doses for the three chemical and physical forms: tritium gas (HT), tritiated water vapor and water drops (HTO). The models in this code consist of the tritium transfer model including oxidation of HT to HTO and reemission of HTO from soil to the atmosphere, and the dose calculation model

  18. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Kurinskiy, P.; Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. for Applied Materials - Applied Materials Physics (IAM-AWP); Rolli, R. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. for Applied Materials - Materials Biomechanics (IAM-WBM); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, Barcelona (Spain)

    2013-07-01

    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 main concept of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by company NGK, Japan. It is notable that beryllium pebbles of other types are commercially available at the market. Presented work is dedicated to a study of characteristics of microstructure, packaging density and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Company Materion, USA. (orig.).

  19. Modeled concentrations in rice and ingestion doses from chronic atmospheric releases of tritium

    The expansion of nuclear power programs in Asia has stimulated interest in the improved modeling of concentrations of tritium in rice, a staple crop grown throughout the far east. Normally, the specific activity model is used to calculate concentrations of tritium in the tissue water of edible plants to assess ingestion dose from chronic releases. However, because rice, like other grains, has much lower water content than most crops, the calculation must also account for organically bound tritium. This paper reviews ways to calculate steady-state concentrations of tritium in rice, including the methods of Canadian and US regulatory models, and the assumptions behind them. Concentrations in rice and resulting ingestion doses are compared for the various methods, and equations for calculating concentrations are recommended. The regulatory models underestimate doses received from ingestion of rice contaminated with tritium since they do not account explicitly for organically bound tritium. The importance of including organically bound tritium is illustrated in a comparison of doses from rice, leafy vegetables and milk for an Asian diet. Dose factors from tritium for these foods are estimated to be 135, 47, and 20 nSv y-1/(Bq m-3), respectively. Assuming known air concentrations, tritium concentrations in rice, calculated with the recommended equations, are uncertain by less than a factor 2 when tritium concentrations in the rice paddy water are known, and by less than a factor of 2.3 when concentrations in paddy water are unknown

  20. Behavior of tritium release from thin boron films deposited on SS316

    Release and diffusion behavior of tritium implanted into thin boron films were examined by isochronal and isothermal heating. For comparison, a polycrystal boron plate was also employed for the same examinations. Changes in the residual amount of tritium with heating were measured by β-ray-induced X-ray spectrometry (BIXS). Most of the tritium desorbed at room temperature was in HTO form, and the residual amount decreased to 20-30% of the initial amount loaded at 773 K. The time-course of the tritium reduction was well represented by an exponential function, suggesting that the tritium release obeys first order reaction kinetics and the rate-determining step is a diffusion process. The apparent activation energy of diffusion was determined to be 0.17 eV. Both the depth profiles calculated from a diffusion equation and determined by computer simulation of X-ray spectra agreed quite well for polycrystal boron

  1. Behavior of tritium release from thin boron films deposited on SS316

    Nakagawa, S.; Matsuyama, M.; Kodama, H.; Oya, Y.; Okuno, K.; Sagara, A.; Noda, N.; Watanabe, K.

    2004-08-01

    Release and diffusion behavior of tritium implanted into thin boron films were examined by isochronal and isothermal heating. For comparison, a polycrystal boron plate was also employed for the same examinations. Changes in the residual amount of tritium with heating were measured by β-ray-induced X-ray spectrometry (BIXS). Most of the tritium desorbed at room temperature was in HTO form, and the residual amount decreased to 20-30% of the initial amount loaded at 773 K. The time-course of the tritium reduction was well represented by an exponential function, suggesting that the tritium release obeys first order reaction kinetics and the rate-determining step is a diffusion process. The apparent activation energy of diffusion was determined to be 0.17 eV. Both the depth profiles calculated from a diffusion equation and determined by computer simulation of X-ray spectra agreed quite well for polycrystal boron.

  2. Modelling accidental releases of tritium in the environment: application as an excel spreadsheet

    An application as an Excel spreadsheet of the simplified modelling approach of tritium transfer in the environment developed by Tamponnet (2002) is presented. Based on the use of growth models of biological systems (plants, animals, etc.), the two-pool model (organic tritium and tritiated water) that was developed estimates the concentration of tritium within the different compartments of the food chain and in fine the dose to man by ingestion in the case of a chronic or accidental release of tritium in a river or the atmosphere. Data and knowledge have been implemented on Excel using the object-oriented programming language VisualBasic (Microsoft Visual Basic 6.0). The structure of the conceptual model and the Excel sheet are first briefly exposed. A numerical application of the model under a scenario of an accidental release of tritium in the atmosphere is then presented. Simulation results and perspectives are discussed. (author)

  3. The influence of irradiation defects on tritium release from Li{sub 2}O

    Tanaka, Satoru; Grishmanov, V. [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    During reactor irradiation of Li{sub 2}O defects are introduced by neutrons, triton and helium ions produced by {sup 6}Li(n, {alpha}){sup 3}H reactions and {gamma}-rays. Simultaneous measurements of luminescence emission and tritium release were performed under various conditions (temperature, sweep gas chemical composition) for Li{sub 2}O single crystal and polycrystal in order to elucidate possible influence of defects on tritium release. (author)

  4. Tritium surface loading due to contamination of rainwater from atmospheric release at NAPS

    Annual tritium (HTO) surface loading has been measured and calculated for the year 1998-99 within 0.8 km distance from 145m high stack of Narora Atomic Power Station (NAPS) at eight locations in different directions. The technique for measured values consists of the summation of product of tritium concentration (Bq/l) in daily rainfall samples and daily rainfall (mm) whereas that for calculated values having the use of prevailing meteorological conditions and average tritium release rate during a year. The ratios of measured and calculated values of tritium surface loading during the years 1998-99 are found to be in the range of 0.18 to 6.97. Tritium surface loading studies at NAPS reveal that a fraction 1.7E-03 of total annual tritium released through stack gets deposited on the surface due to washout / rainout of plume within 0.8 km radial distance from stack. The range of deposition velocity, V w (m.s-1) i.e the ratio of annual tritium surface loading W(Bq.m-2.s-1) and annual mean tritium concentration in air, χo(Bq.m-3) at three locations for the years 1998-99 is found to be 5.59E-04 to 5.99E-03 ms-1. The average value for wet deposition velocity V barw for NAPS site is estimated as 2.92E-03 m.s-1. (author)

  5. Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor

    As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. - Highlights: • Dynamic tritium concentration in the air and human body evaluated on the time scale. • Different intake forms and relevant radiation dose assessed to find out the potential relationships. • HT and HTO release amount limits for arbitrary accidents proposed for the fusion reactor according to dose limit

  6. Progress in tritium retention and release modeling for ceramic breeders

    Tritium behavior in ceramic breeder blankets is a key design issue for this class of blanket because of its impact on safety and fuel self-sufficiency. Over the past 10-15 years, substantial theoretical and experimental efforts have been dedicated world-wide to develop a better understanding of tritium transport in ceramic breeders. Models that are available today seem to cover reasonably well all the key physical transport and trapping mechanisms. They have allowed for reasonable interpretation and reproduction of experimental data and have helped in pointing out deficiencies in material property data base, in providing guidance for future experiments, and in analyzing blanket tritium behavior. This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described along with the more recent and sophisticated models developed to help understand them. Recent experimental data are highlighted and model calibration and validation discussed. Finally, example applications to blanket cases are shown as illustration of progress in the prediction of ceramic breeder blanket tritium inventory

  7. Progress in tritium retention and release modeling for ceramic breeders

    Tritium behavior in ceramic breeder blankets is a key design issue for this class of blanket because of its impact on safety and fuel self-sufficiency. Over the past 10-15 years, substantial theoretical and experimental effort has been dedicated worldwide to the development of a better understanding of tritium transport in ceramic breeders. The models available today seem to cover reasonably well all of the key physical transport and trapping mechanisms. They allow for reasonable interpretation and reproduction of experimental data, help to point out deficiencies in the material property database, provide guidance for future experiments and aid in the analysis of blanket tritium behavior.This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described, together with the more recent, sophisticated models which have been developed to help understand them. Recent experimental data are highlighted and model calibration and validation are discussed. Finally, example applications to blanket cases are shown as an illustration of the progress in the prediction of ceramic breeder blanket tritium inventory. (orig.)

  8. Tritium release from a nonevaportable getter-pump cartridge exposed to moist air at ambient temperature

    The amount of tritium released when a commercially available getter-pump cartridge was exposed to moist air at ambient temperatures was measured. The cartridge consisted of Zr-Al powder pressed onto an iron substrate, which is the type of cartridge proposed for use in the Tokomak Fusion Test Reactor. While the initial release of tritium was rapid the total activity released was lss than 0.005% of the cartridge loading. Of this amount, at least 80% was released as tritiated water. 8 figures

  9. Environmental effects of a tritium gas release from the Savannah River Plant on December 31, 1975

    Jacobsen, W.R.

    1976-03-01

    At 10:00 p.m. EST on December 31, 1975, 182,000 Ci of tritium gas was released within about 1.5 min from a tritium processing facility at the Savannah River Plant. The release was caused by the failure of a vacuum gage and was exhausted to the atmosphere by way of a 200-ft-high stack. Winds averaging 20 mph carried the tritium offplant toward the east. Calculations indicate that the puff passed out to sea about 35 miles north of Charleston, South Carolina, about 7 hr after the release occurred. Samples from the facility exhaust system indicated that 99.4 percent of the tritium was in elemental form and 0.6 percent was in the more biologically active oxide (water) form. The maximum potential dose to a person (from inhalation and skin absorption) at the puff centerline on the plant boundary was calculated to be 0.014 mrem, or about 0.01 percent of the annual dose received from natural radioactivity. The integrated dose to the population under the release path was calculated to be 0.2 man-rem before the tritium passed out to sea. Over 300 environmental samples were collected and analyzed following the release. These samples included air moisture, atmospheric hydrogen, vegetation, soil, surface water, milk, and human urine. Positive results were obtained in some onplant and plant perimeter samples; these results aided in confirming the close-in puff trajectory. Tritium concentrations in nearly all samples taken beyond the plant perimeter fell within normal ranges; no urine samples indicated any tritium uptakes as a result of the release. Two milk samples did indicate a measurable tritium uptake; the maximum potential dose to an individual drinking this milk was calculated to be about 0.1 mrem. Because calculated doses from assumed exposure to the tritium are low and analyses of environmental samples indicated no significant accumulation of tritium, it is concluded that no significant environmental effects resulted from the December 31, 1975, tritium release. (auth)

  10. Development, description and validation of a Tritium Environmental Release Model (TERM)

    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. -- Highlights: • A short review of existing tritium models from literature is presented. • A new model for tritium release and transfer in the environment is presented. • The model is freely available and extensively documented. • The model intent is to supplement and bridge the capabilities of existing models. • The model is compared and validated to published data sets and other models

  11. Research of tritium gas releasing from lithium ceramics in the course of reactor irradiation

    The results of experimental researches of tritium gas releasing from ceramic breeder ITER material- lithium ceramics Li2TiO3 - under the conditions of irradiation on the reactor IVG.1M are presented in the work. The lithium ceramics samples were of sphere shape with diameter 2 mm with enrichment on isotope 6Li 7%. Gas releasing was measured using radio-frequency mass-spectrometer MH-6407P at the temperature range from 20 deg.C to 1000 deg.C. The outcomes of the researches showed that tritium gas releasing process depends on temperature and fluence of neutrons and at that helium generated in the samples simultaneously with tritium almost does not release. author)

  12. Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

    Chakin, V., E-mail: vladimir.chakin@kit.edu; Rolli, R.; Vladimirov, P.; Moeslang, A.

    2015-06-15

    Highlights: • For the irradiated constrained beryllium pebbles, the tritium release occurs easier than for the unconstrained ones. • Tritium retention in the irradiated constrained and unconstrained beryllium pebbles decreases with increasing irradiation temperature. • Formation of sub-grains in the constrained beryllium pebbles facilitate the open porosity network formation. - Abstract: Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1 mm at 686–1006 K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020 appm helium and 298 appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release.

  13. Tritium release from highly neutron irradiated constrained and unconstrained beryllium pebbles

    Highlights: • For the irradiated constrained beryllium pebbles, the tritium release occurs easier than for the unconstrained ones. • Tritium retention in the irradiated constrained and unconstrained beryllium pebbles decreases with increasing irradiation temperature. • Formation of sub-grains in the constrained beryllium pebbles facilitate the open porosity network formation. - Abstract: Beryllium is the reference neutron multiplier material in the Helium Cooled Pebble Bed (HCPB) breeding blanket of fusion power plants. Significant tritium inventory accumulated in beryllium as a result of neutron-induced transmutations could become a safety issue for the operation of such blankets as well as for the nuclear waste utilization. To provide a related materials database, a neutron irradiation campaign of beryllium pebbles with diameters of 0.5 and 1 mm at 686–1006 K, the HIDOBE-01 experiment, has been performed in the HFR in Petten, the Netherlands, producing up to 3020 appm helium and 298 appm tritium. Thermal desorption tests of irradiated unconstrained and constrained beryllium pebbles were performed in a purge gas flow using a quadrupole mass-spectrometer (QMS) and an ionization chamber. Compared to unconstrained pebbles, constrained beryllium pebbles have an enhanced tritium release at all temperatures investigated. Small elongated sub-grains formed under irradiation in the constrained pebbles promote formation of numerous channels for facilitated tritium release

  14. Simple fast micromethod for measuring enzyme activities which release tritium as tritium water

    Hughes, W.L.

    1987-03-01

    The entire procedure is carried out in a counting vial by mixing the reagents as a 20- to 30-microliters drop in the cap of a counting vial, incubating, quenching the reaction, and then distilling the tritium water produced into the chilled vial, in which it is assayed after the addition of scintillation solvent and a clean cap. The application of this technique to the analysis of serum transaminases is described.

  15. Tritium release kinetics of Li{sub 2}O with radiation defects

    Grishmanov, V.; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1998-03-01

    The study of an influence of radiation defects on tritium release behavior from polycrystalline Li{sub 2}O was performed by the in-pile and out-of-pile tritium release experiments. The samples were pre-irradiated by accelerated electrons to various absorbed doses up to 140 MGy and then exposed to the fluence of 10{sup 17} thermal neutrons/m{sup 2}. The radiation defects introduced by electron irradiation in Li{sub 2}O cause the retention of tritium. The linear temperature increase of the electron-irradiated samples disclosed two tritium release peaks: first starts at {approx}600 K with the maximum at {approx}800 K and second appears at {approx}950 K with the maximum at {approx}1200 K. It is thought that the tritium release at high temperatures (> 950 K) is due to the thermal decomposition of LiT. In order to further investigated the formation of lithium hydrides, the diffuse-reflectance Fourier transform infrared (FTIR) absorption spectroscopy was applied. The Li{sub 2}O powder was irradiated by electron accelerator under D{sub 2} containing atmosphere (N{sub 2} + 10% D{sub 2}). An absorption band specific to the Li{sub 2}O was observed at 668 cm{sup -1} and attributed to the Li-D stretching vibration. (author)

  16. Computer program for assessing the human dose due to stationary release of tritium

    The computer program TriStat (Tritium dose assessment for stationary release) has been developed to assess the dose to humans assuming a stationary release of tritium as HTO and/or HT from nuclear facilities. A Gaussian dispersion model describes the behavior of HT gas and HTO vapor in the atmosphere. Tritium concentrations in soil, vegetables and forage were estimated on the basis of specific tritium concentrations in the free water component and the organic component. The uptake of contamination via food by humans was modeled by assuming a forage compartment, a vegetable component, and an animal compartment. A standardized vegetable and a standardized animal with the relative content of major nutrients, i.e. proteins, lipids and carbohydrates, representing a standard Japanese diet, were included. A standardized forage was defined in a similar manner by using the forage composition for typical farm animals. These standard feed- and foodstuffs are useful to simplify the tritium dosimetry and the food chain related to the tritium transfer to the human body. (author)

  17. Modeling and validating tritium transfer in a grassland ecosystem in response to 3H releases

    Tritium (3H) is a major radionuclide released in several forms (HTO, HT) by nuclear facilities under normal operating conditions. In terrestrial ecosystems, tritium can be found under two forms: tritium in tissue free water (TFWT) following absorption of tritiated water by leaves or roots and Organically Bound Tritium (OBT) resulting from TFWT incorporation by the plant organic matter during photosynthesis. In order to study transfers of tritium from atmospheric releases to terrestrial ecosystem such as grasslands, an in-situ laboratory has been set up by IRSN on a ryegrass field plot located 2 km downwind the AREVA NC La Hague nuclear reprocessing plant (North-West of France), as was done in the past for the assessment of transfer of radiocarbon in grasslands. The objectives of this experimental field are: (i) to better understand the OBT formation in plant by photosynthesis, (ii) to evaluate transfer processes of tritium in several forms (HT, HTO) from the atmosphere (air and rainwater) to grass and soil, (iii) to develop a modeling allowing to reproduce the dynamic response of the ecosystem to tritium atmospheric releases depending of variable environmental conditions. For this purpose, tritium activity measurements will be carried out in grass (monthly measurements of HTO, OBT), in air, rainwater, soil (daily measurements of HT, HTO) and CO2, H2O fluxes between soil and air compartments will be carried out. Then, the TOCATTA-c model previously developed to simulate 14C transfers to pasture on a hourly time-step basis will be adapted to take account for processes specific to tritium. The model will be tested by a comparison between simulated results and measurements. The objectives of this presentation are (1) to present the organization of the experimental design of the VATO study (Validation of TOCATTA) dedicated to transfers of tritium in a grassland ecosystem, (2) to document the major assumptions, conceptual modelling and mathematical formulations of

  18. Modeling and validating tritium transfer in a grassland ecosystem in response to 3H releases

    In this paper a radioecological model (TOCATTA) for tritium transfer in a grassland ecosystem developed on an hourly time-step basis is proposed and compared with the first data set obtained in the vicinity of the AREVA-NC reprocessing plant of La Hague (France). The TOCATTA model aims at simulating dynamics of tritium transfer in agricultural soil and plant ecosystems exposed to time-varying HTO concentrations in air water vapour and possibly in irrigation and rain water. In the present study, gaseous releases of tritium from the AREVA NC nuclear reprocessing plant in normal operation can be intense and intermittent over a period of less than 24 hours. A first comparison of the model predictions with the field data has shown that TOCATTA should be improved in terms of kinetics of tritium transfer

  19. Modeling and validating tritium transfer in a grassland ecosystem in response to {sup 3}H releases

    Le Dizes, S. [Institute for Radioprotection and Nuclear Safety, IRSN/PRP-ENV/SERIS/LM2E, Centre de Cadarache, Saint-Paul-lez-Durance (France); Maro, D.; Rozet, M.; Hebert, D. [IRSN/PRP-ENV/SERIS/LRC, Cherbourg-Octeville (France)

    2015-03-15

    In this paper a radioecological model (TOCATTA) for tritium transfer in a grassland ecosystem developed on an hourly time-step basis is proposed and compared with the first data set obtained in the vicinity of the AREVA-NC reprocessing plant of La Hague (France). The TOCATTA model aims at simulating dynamics of tritium transfer in agricultural soil and plant ecosystems exposed to time-varying HTO concentrations in air water vapour and possibly in irrigation and rain water. In the present study, gaseous releases of tritium from the AREVA NC nuclear reprocessing plant in normal operation can be intense and intermittent over a period of less than 24 hours. A first comparison of the model predictions with the field data has shown that TOCATTA should be improved in terms of kinetics of tritium transfer.

  20. Effects of helium and ambient water vapor on tritium release from Li2TiO3

    The effects of hydrogen isotopes in tritium recovery gas and the presence of helium on the tritium release behaviors were investigated for lithium meta-titanate (Li2TiO3) to develop a tritium migration model. The tritium trapping sites as oxygen vacancies and oxygen atoms with dangling bonds were formed by energetic tritium ion implantation. Isotope exchange processes with water vapor in purge gas enhanced the recovery of tritium adsorbed on the surface of Li2TiO3 and trapped in oxygen vacancies. Replacement of tritium as hydroxyl groups with hydrogen isotopes from purge gas was hardly occurred due to the higher trapping energy of tritium as hydroxyl groups. Helium implantation would induce the formation of helium bubbles, which would occupy irradiation damages, contributing to the decrease in the densities of trapping site for tritium. Consequently, the addition of hydrogen isotopes in the tritium recovery gas and the presence of helium in tritium breeding materials can contribute to the efficient tritium recovery for tritium breeding materials

  1. Estimation of tritium release during refueling operation in research reactor Dhruva

    Dhruva is 100 MW (th) research reactor. Heavy water is used as coolant, moderator as well as reflector. Refueling of the reactor is carried out in reactor shutdown condition using fueling machine-A (FM/A) which is designed for handling heavy water cooled assemblies. FM /A has two barrels for handling irradiated and fresh fuel assemblies during refueling operation. Spent fuel in fueling machine is cooled using heavy water which has tritium activity concentration of ∼ 185 MBqml-1. FM/A exhaust fan B29/B30 provides a negative pressure of 180 mm of water gauge to avoid any activity buildup in working area. As exhaust tritium release during the refueling operation is always on higher side, an attempt is made to quantify the tritium release from FM/A exhaust system

  2. Investigation of fire at Council, Alaska: A release of approximately 3000 curies of tritium

    On September 6, 1987, about 6:00 a.m., a fire was discovered in the community building at Council, Alaska, where 12 radioluminescent (RL) light panels containing approximately 3000 Ci were stored. All of the tritium in the panels was released as a result of the fire. This report summarizes the recovery of the remains of the panels destroyed in the fire and investigations completed to evaluate the fire site for possible exposure of community residents or contamination by tritium release in the environment. Based on the analysis of urine samples obtained from individuals in the community and from Pacific Northwest Laboratory personnel participating in the recovery operation, no evidence of exposure to individuals was found. No tritium (above normal background) was found in water and vegetation samples obtained at various locations near the site. 12 figs., 3 tabs

  3. Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor.

    Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

    2015-10-01

    As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. PMID:26164282

  4. ITER task title - source term data, modelling, and analysis. ITER subtask no. S81TT05/5 (SEP 1-1). Global tritium source term analysis basis document. Subtask 2: tritium releases due to accident conditions. Final task report

    This report presents the methodology and the key assumptions that are adopted in preparing the preliminary estimates of accidental tritium release terms. A room-by room map/table, at the subsystem level, identifying the major equipment and their failure modes, secondary containments, maximum releasable tritium inventory, duration of tritium releases and the tritium release pathways etc. is also included. The tritium release calculations and the room-by-room map/table have been prepared in EXCEL spreadsheets, so that the estimates can be refined easily, and the approach is relatively adaptable to changes of the ITER design information. (author). 22 refs., 7 tabs

  5. Tritium surface loading due to contamination of rainwater from atmospheric release at NAPS (2011)

    Annual tritium (HTO) surface loading has been measured and calculated for the year 2011 within 0.8 km distance from 145 m high stack of Narora Atomic Power Station (NAPS) at eight locations in different directions. The technique for measured values consists of the summation of product of tritium concentration (Bq/l) in daily rainfall samples and daily rainfall (mm). Tritium surface loading studies at NAPS reveal that a fraction 1.01E-03 of total annual tritium released through stack gets deposited on the surface due to washout/rainout of plume within 0.8 km radial distance from stack. The range of deposition velocity, Vw (m.s-1) i.e., the ratio of annual tritium surface loading W (Bq. m-2.s-1) and annual mean tritium concentration in air, c0(Bq.m-3) at three locations for the years 2011 is found to be 6.12E-04 to 2.89E-03. The average value for wet deposition velocity Vw for NAPS site is estimated as 3.17E-03 m.s-1. (author)

  6. Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding

    Robinson, Sharon M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chattin, Marc Rhea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giaquinto, Joseph M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-09-01

    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.

  7. Thermal ramp tritium release in COBRA-1A2 C03 beryllium pebbles

    Baldwin, D.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Tritium release kinetics, using the method of thermal ramp heating at three linear ramp rates, were measured on the COBRA-1A2 C03 1-mm beryllium pebbles. This report includes a brief discussion of the test, and the test data in graph format.

  8. Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li2TiO3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li2TiO3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328degC in

  9. Relation between the tritium in continuous atmospheric release and the tritium contents of fruits and tubers

    Concentrations of organically bound tritium (OBT) and tissue-free water tritium (TFWT, also referred to as HTO) in fruits and tubers were measured at a garden plot in the vicinity of the source of chronic airborne tritium emissions during the 2008, 2010, and 2011 growing seasons. A continuous record of HTO concentration in the air moisture was reconstructed from the continuous record of Ar-41 ambient gamma radiation, as well as from frequent measurements of air HTO by active samplers at the garden plot and Ar-41 and air HTO monitoring data from the same sector. Performed measurements were used for testing the modified Specific Activity (SA) model based on the assumption that the average air HTO during the pod-filling period provides an appropriate basis for estimating the levels of OBT present in pods, fruits and tubers. It is established that the relationship between the OBT of fruits and tubers and the average air HTO from a 15–20 day wide window centred at the peak of the pod-filling period is consistent throughout the three analysed years, and could be expressed by the fruit or tuber's OBT to air-HTO ratio of 0.93 ± 0.21. For all three years, the concentration of HTO in fruits and tubers was found to be related to levels of HTO in the air, as averaged within a 3-day pre-harvest window. The variability in the ratio of plant HTO to air HTO appears to be three times greater than that for the OBT of the fruits and tubers. It is concluded that the OBT of fruits and tubers adequately follows an empirical relationship based on the average level of air HTO from the pod-filling window, and therefore is clearly in line with the modified SA approach. -- Highlights: ► This paper provides the first account of the relationship between the concentration of HTO in the air and OBT in fruits and tubers, as measured during three growing seasons. ► It was found that the relation between fruit/tuber OBT and the average air HTO for 15-day wide window centered at the peak of

  10. Assessment of the dose to a representative Japanese due to stationary release of tritium to the environment

    The computer program TriStat was applied to estimate the dose to a representative Japanese due to a stationary release of tritium as HTO and/or HT to the atmosphere from nuclear facilities. In TriStat, the air tritium concentration is estimated by a Gaussian dispersion model. The tritium deposition to the soil was assumed to occur both by dry and wet deposition processes of atmospheric tritium. The primary process of tritium transfer to human body is assumed to take place through a local food-chain in the contaminated area. Tritium concentrations in soil, vegetables and forage were estimated as the tritium concentration per water equivalent. The food chain was modeled by assuming a vegetable compartment and an animal-food compartment. By using TriStat the annual dose to the representative Japanese was evaluated for stationary release of tritium as a function of the distance from a release point. The dose contribution from drinking water was neglected, since the drinking water is generally supplied as tap water or as commercial bottled water. In the case of HT release, the committed dose due to tritium intake through breathing and skin absorption was found to be of minor importance. (author)

  11. Assessment of airborne release methodology for tritium from US DOE facilities

    There are numerous nonreactor facilities in the U.S. Department of Energy (DOE) complex that contain and/or process large inventories of tritium. The primary safety analysis context for these facilities above key threshold quantities of tritium is the safety analysis report (SAR) format. As a means of identifying safe operating margins and ensuring low levels of risk to on-site workers and the general public, a licensing process analogous to a commercial reactor format is followed. The ultimate goal of this process is a DOE-approved SAR, developed in the framework contained in DOE Order 5480.23. The SAR for the particular DOE facility in question addresses dose consequences and subsequent health effects caused by the release of tritium under normal, abnormal, and accident conditions. Dose assessment codes are thus critical to the completion of the overall safety evaluation

  12. Influence of start up and pulsed operation on tritium release and inventory of NET ceramic blanket

    A first estimate for the tritium release behaviour of a ceramic breeder blanket in pulsed operation is obtained by assuming a linear steady state temperature distribution and taking into account the time constant of the thermal behaviour. The release behaviour of the breeder exposed to consecutive periods of tritium generation is described with an analytical solution of the diffusion equation. The results are compared with a simple exponential approach valid for surfacte desorption controlled release. The exponential model is used to simulate a blanket with aluminate as breeder material, which takes longest to reach steady state. The simulation demonstrates that a significant fraction (>67%) of steady state can be achieved after a testing time of about one day. (author). 7 refs.; 8 figs.; 3 tabs

  13. The VOM/JRR-2 experiments; performance of in-situ tritium release from the lithium ceramics

    In-situ tritium release experiments on lithium ceramics used as tritium breeding materials have been carried out in Japan Research Reactor 2 (JRR-2) to support fusion reactor design activity. The in-situ tritium measurement system was specifically designed for the VOM experiment and several techniques in ceramic electrolysis cell, ionization chamber, capsule and associated components were utilized. The knowledge and experience gained from these experiments have been very useful for the design and fabrication of the IEA collaborative irradiation experiment, BEATRIX-II. This report compares the tritium release behavior between single crystal, ring monolithic and sintered pebble of Li2O in VOM-34 and 44 experiments. The tritium release behavior of Li2ZrO3, Li4SiO4 and Li2Be2O3 have been investigated in VOM-32 and 48 experiments. ((orig.))

  14. Tritium release from neutron irradiated beryllium: Kinetics, long-time annealing and effect or crack formation

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe, (Germany)

    1995-09-01

    Since beryllium is considered as one of the best neutron multiplier materials in the blanket of the next generation fusion reactors, several studies have been started to evaluate its behaviour under irradiation during both operating and accidental conditions. Based on safety considerations, tritium produced in beryllium during neutron irradiation represents one important issue, therefore it is necessary to investigate tritium transport processes by using a comprehensive mathematical model and comparing its predictions with well characterized experimental tests. Because of the difficulties in extrapolating the short-time tritium release tests to a longer time scale, also long-time annealing experiments with beryllium samples from the SIBELIUS irradiation. have been carried out at the Forschungszentrum Karlsruhe. Samples were annealed up to 12 months at temperatures up to 650{degrees}C. The inventory after annealing was determined by heating the samples up to 1050{degrees}C with a He+0.1 vo1% H{sub 2} purge gas. Furthermore, in order to investigate the likely effects of cracks formation eventually causing a faster tritium release from beryllium, the behaviour of samples irradiated at low temperature (40-50{degrees}C) but up to very high fast neutron fluences (0.8-3.9{center_dot}10{sup 22} cm{sup -2}, E{sub n}{ge}1 MeV) in the BR2 reactor has been investigated. Tritium was released by heating the beryllium samples up to 1050{degrees}C and purging them with He+0.1 vo1% H{sub 2}. Tritium release from high-irradiated beryllium samples showed a much faster kinetics than from the low-irradiated ones, probably because of crack formation caused by thermal stresses in the brittle material and/or by helium bubbles migration. The obtained experimental data have been compared with predictions of the code ANFIBE with the goal to better understand the physical mechanisms governing tritium behaviour in beryllium and to assess the prediction capabilities of the code.

  15. The effect of oxygen on the release of tritium during baking of TFTR D-T tiles

    A series of tests involving 10 h baking under the current ITER design conditions (240 deg. C with 933 Pa O2) was performed using a cube of a carbon fiber composite tile that had been used in Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium burning operation. The removal rate of the codeposits was about 3 μm/h near the surface and 0.9 μm/h in the deeper region. Total amount of tritium released from the cube during 10 h baking was 202 MBq, while remaining tritium in the cube after baking was 403 MBq. Thus 10 h baking at 240 deg. C with 933 Pa O2 removed 1/3 of tritium from the cube. After 10 h baking, the tritium concentration on the cube surface also dropped by about 1/3. In addition, some tritium was released from another cube of the tile during baking at 240 deg. C in pure Ar, and a rapid increase of tritium release was observed when the purging gas was shifted from pure Ar to Ar-1%O2. When a whole TFTR tile was baked in air at 350 deg. C for 1 h and then at 500 deg. C for 1 h, the ratios of tritium released were 53 and 47%, respectively. Oxygen reacted with carbon to produce carbon monoxide during baking in air

  16. Dose from organically bound tritium after an acute tritiated water intake in humans

    We have analyzed the urinary excretion data from eight male workers following an acute intake of tritiated water (HTO) and assessed the dose contribution from organically bound tritium (OBT) in the body. The individuals affected increased their fluid intakes during the first month or more post-exposure, to accelerate the turnover of tritium in the body water for dose mitigation purposes. The volumes of cumulative 24 h urine samples were similar to Reference Man in the latter part of the study (100-300 d post-exposure). The workers' urine samples were analyzed for total tritium up to 300 d post-exposure. The results suggest that a measurement of the tritium activity per unit mass of organic matter in urine can provide an assessment of the specific activity of tritium in the organic fraction of the soft tissue, providing an equilibrium condition exists. A mathematical model is proposed to estimate the dose increase from the retained OBT by examining the kinetics of total tritium excretion in urine. The model accounts for the variable rates of fluid intake. The influence of measurement errors and the limited duration of the study (0-300 d post-exposure) on the OBT dose contribution was assessed through statistical analysis, while the role of direct OBT excretion in urine was estimated by using metabolic models. Based on the time series of tritium concentration in urine, the average dose increase to the workers from the metabolised OBT was calculated as 6.2 ± 1.3% of the HTO dose. 78 refs., 36 tabs., 11 figs

  17. Developing a tritium release model for Li2TiO3 with irradiation-induced defects

    The annihilation kinetics of irradiation-induced defects in Li2TiO3 was evaluated by electron spin resonance. The radiation defects were stabilized with increasing defect density by interaction with neighboring defects. Subsequently, the tritium migration model in Li2TiO3 was established by integrating the kinetics of tritium diffusivity in Li2TiO3 crystalline grains, tritium trapping/detrapping at oxygen vacancies and hydroxyl groups, and annihilation of irradiation-induced defects. The contribution of hydrogen isotopes contained in tritium recovery gas was also considered in this model. The model can demonstrate an overall profile of out-of-pile tritium release for Li2TiO3 with various neutron fluences. Tritium release behavior under neutron irradiation was also estimated by the present established model. The tritium inventory increased under high neutron flux because of the continuous generation of tritium trapping sites, and the addition of hydrogen isotopes to the purge gas could reduce the tritium inventory for Li2TiO3

  18. User guide for UFOTRI: A program for assessing the off-site consequences from accidental tritium releases

    The computer program UFOTRI for assessing the consequences of accidental tritium releases from fusion facilities has been developed; its first version is now being released. Processes such as the conversion of tritium gas (HT) into tritiated water (HTO) in the soil, reemission after deposition and the conversion of HTO into organically bound tritium are considered. During a time period of some days, all the relevant transfer processes between the compartments of the biosphere (atmosphere, soil, plants, animals) are described dynamically. A first order compartment model calculates the longer term pathway of tritium in the foodchains. Additionally, UFOTRI allows probabilistic assessments of the tritium impact in the environment. This report contains a description of the main processes which are important for the understanding of the input parameter list, as well as a detailed listing of the input parameters which can be changed by the user. Additionally, an input and output description with examples completes the report. (orig.)

  19. Lithium orthosilicate ceramics: sol-gel preparation, lithium dynamics and tritium release

    Ceramics based on the lithium orthosilicate (Li4SiO4) are candidates as blanket materials for forthcoming fusion reactors. Lithium orthosilicate powders, with controlled stoichiometry, were prepared from sol-gel route. This method of processing powders makes possible the preparation of monophase ceramics with fine-grained uniform microstructure by sintering at 650-8000C, without prior calcination. Lithium transport properties were investigated from complex impedance spectroscopy and 7Li NMR spin-lattice relaxation measurements. The enhancement of the lithium conductivity in the orthosilicate type structure was realized by introducing mobile ion vacancies in the lithium sites, as noted in the Li4SiO4-Li3PO4 system. Concerning tritium release properties, deduced from out-of-pile experiments, no relation was found between the tritium behavior and the lithium bulk-diffusion within the grains. However, a large effect of the microstructure was displayed. The release rate appeared much faster for microporous fine-grained ceramics than for dense coarse-grained ones. In fact, the tritium release is controlled, at least at low temperature, by water chemistry and can be very well described by OH-/OT- recombination and desorption

  20. Study of lithium materials for tritium release, evolution in the CCHEN

    The Chilean Nuclear Energy Commission decided in 1993 to study lithium ceramic materials with tritium properties, ceramics breeder, using neutron irradiation in order to allow the transition from the conceptual to the engineering design of the ceramic breeder blanket fusion reactor. The project 'Development of Lithium Ceramic Materials for Fusion Reactors' was defined leading to the construction of the Tritium Handling Laboratory with the technical assistance of the Argonne National Laboratory and with Canadian equipment. Later, it was presented to the International Atomic Energy agency, IAEA, as a technical cooperation project for obtaining a tritium device in real time, irradiation loop. This was approved with a significant budget for equipment, training and experts. The project was presented to the international Ceramic Breeder Blanket Interactions community at the International Workshop on Ceramic Breeder Blanket Interactions. To date several investigations in this subject have been carried out, producing the release of tritium in the laboratory, in batch, and recently in continuous with the construction, installation and operation of an irradiation loop in the RECH-1, loading with ceramic material prepared and fabricated in the our laboratories.This work presents the genesis, evolution and current state, of this study, from 1993 to date

  1. Effects of helium production and radiation damage on tritium release behavior of neutron-irradiated beryllium pebbles

    The tritium release from neutron-irradiated beryllium pebbles, irradiated under different helium production (0.5-1.0 x 103 appm He) and dpa (4.2-8.6) conditions, was studied. From these results, it was clear that the apparent diffusion coefficient at 600 deg. C was significantly affected by irradiation conditions, but returned to normal values at 900 deg. C, apparently due to thermal annealing. Multiple peaks in the tritium release curve at 900 deg. C were observed

  2. A prototype wearable tritium monitor

    Sudden unexpected changes in tritium-in-air concentrations in workplace air can result in significant unplanned exposures. Although fixed area monitors are used to monitor areas where there is a potential for elevated tritium in air concentrations, they do not monitor personnel air space and may require some time for acute tritium releases to be detected. There is a need for a small instrument that will quickly alert staff of changing tritium hazards. A moderately sensitive tritium instrument that workers could wear would bring attention to any rise in tritium levels that were above predetermined limits and help in assessing the potential hazard therefore minimizing absorbed dose. Hand-held instruments currently available can be used but require the assistance of a fellow worker or restrict the user to using only one hand to perform some duties. (authors)

  3. Post-irradiation exam of tritium release from long-term irradiated Li2TiO3 ceramics

    Full text of publication follows: Lithium ceramics is planned to be used in tritium breeding systems of future fusion reactors. To provide effective tritium generation while obeying the ecological and safety restrictions on tritium processing it is necessary to investigate tritium interaction with elements of proposed breeding systems. Therefore tritium-ceramics interaction is of most interest in such systems. Presented work describes experimental studies of tritium yield from lithium ceramics (Li2TiO3+5 mol.% TiO2 ) after long-term neutron irradiation. Initially ceramics was 96% enriched with Li6 and irradiated with neutrons (about 220 days) in research water-water reactor of Kazakh National Nuclear Center (WWRK) till the 20% burn-up of Li6. Examinations of residual tritium yield from irradiated lithium ceramics were conducted using thermodesorption method with linear heating rates from 2 to 10 K/min up to ceramics melting point temperature. The experiments were carried-out under continuous pump-out and mass-analysis of desorbed gases in experimental chamber. As the result the data on tritium (and other gases) release rates from irradiated ceramics are obtained. Preliminary results on estimations of residual tritium content in irradiated lithium ceramics and its thermodesorption data are presented in given report. (authors)

  4. ITER task title - source term data, modelling, and analysis. ITER subtask no. S81TT05/5 (SEP 1-1). Global tritium source term analysis basis document. Subtask 1: operational tritium effluents and releases. Final report (1995 TASK)

    This document represents the final report for the global tritium source term analysis task initiated in 1995. The report presents a room-by-room map/table at the subsystem level for the ITER tritium systems, identifying the major equipment, secondary containments, tritium release sources, duration/frequency of tritium releases and the release pathways. The chronic tritium releases during normal operation, as well as tritium releases due to routine maintenance of the Water Distillation Unit, Isotope Separation System and Primary and Secondary Heat Transport Systems, have been estimated for most of the subsystems, based on the IDR design, the Design Description Documents (April - Jun 1995 issues) and the design updates up to December 1995. The report also outlines the methodology and the key assumptions that are adopted in preparing the tritium release estimates. The design parameters for the ITER Basic Performance Phase (BPP) have been used in estimating the tritium releases shown in the room-by-room map/table. The tritium release calculations and the room-by-room map/table have been prepared in EXCEL, so that the estimates can be refined easily as the design evolves and more detailed information becomes available. (author). 23 refs., tabs

  5. Development of the IFMIF Tritium Release Test Module in the EVEDA phase

    Abou-Sena, Ali, E-mail: ali.abou-sena@kit.edu [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Arbeiter, Frederik [Institute for Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany)

    2013-10-15

    This paper presents the engineering design of the IFMIF (International Fusion Materials Irradiation Facility) Tritium Release Test Module (TRTM). The objectives of the TRTM are: (i) in-situ measurements of the tritium released from lithium ceramics and beryllium pebble beds during irradiation, (ii) studying the chemical compatibility between lithium ceramics and structural materials under irradiation, and (iii) performing post irradiation examinations for the irradiated materials. The TRTM has eight rigs which are arranged in two rows (2 × 4) perpendicular to the beam axis and enclosed by a structural container. Each rig includes one capsule that contains lithium ceramic or beryllium pebbles for irradiation. Neutrons reflectors are implemented at different locations to reflect the scattered neutrons back to the active region aiming to improve the tritium production. The TRTM is required to provide irradiation temperature range of 400–900 °C for the capsules filled with lithium ceramics and 300–700 °C for the ones packed with beryllium. The engineering design of the TRTM components such as container, rigs, capsules, pebble beds, neutrons reflectors, and purge gas and coolant tubes are presented. In addition a test matrix for the irradiation campaign is proposed.

  6. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

  7. Thermal Release of 3He from Tritium Aged LaNi4.25Al0.75 Hydride

    Recently, the demand for He-3 has increased dramatically due to widespread use in nuclear nonproliferation, cryogenic, and medical applications. Essentially all of the world's supply of He-3 is created by the radiolytic decay of tritium. The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 in the tritium process to store hydrogen isotopes. The vast majority of He-3 ''born'' from tritium stored in LANA.75 is trapped in the hydride metal matrix. The SRS-TF has multiple LANA.75 tritium storage beds that have been retired from service with significant quantities of He-3 trapped in the metal. To support He-3 recovery, the Savannah River National Laboratory (SRNL) conducted thermogravimetric analysis coupled with mass spectrometry (TGA-MS) on a tritium aged LANA.75 sample. TGA-MS testing was performed in an argon environment. Prior to testing, the sample was isotopically exchanged with deuterium to reduce residual tritium and passivated with air to alleviate pyrophoric concerns associated with handling the material outside of an inert glovebox. Analyses indicated that gas release from this sample was bimodal, with peaks near 220 and 490°C. The first peak consisted of both He-3 and residual hydrogen isotopes, the second was primarily He-3. The bulk of the gas was released by 600 °C

  8. Influence of the rate of conversion of HT and HTO on projected radiation doses from release of molecular tritium

    Releases of tritium in the past have been largely in the form of tritiated water, and the projected radiation doses could be estimated by assuming tritium behaviour to parallel that of water. There is increasing interest in potential releases of tritium in the form of HT because of significant recent advances in fusion reactor research. Several recent studies have shown that bacteria containing the enzyme hydrogenase can catalyse the conversion of HT to HTO at rates several orders of magnitude faster than the rates measured in atmospheric systems. Rates of conversion in the soil have been combined with estimates of rates of permeation of HT into the soil and with global and local models depicting tritium transport and cycling. The results suggest that for the expected conversion rates, the impact on projected radiation doses should be relatively minor. (author)

  9. Tritium release experiment in France results concerning HT/HTO conversion in the air and soil

    A controlled release of 256 TBq of pure dry tritium through a 40-m high stack was performed in France, on October 15, 1986. Air measurements were assigned to the Health Physics Division (SPR) from the Centre of Bruyeres-le-Chatel and soil measurements in particular for deposition velocity and reemission rate, to the Division for Studies and Research on the Environment (SERE) of the Institute of Protection and Nuclear Safety (IPSN). The main conclusions were: concordance between code predictions and air measurements; HTO increase in the air originating from soil effects; HT deposition velocity: 1 → 5 E-4 ms/sup -1/, reemission rate: 5 % per hour

  10. Modeling of the dispersion of tritium from postulated accidental releases from nuclear power plants

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models of hydrodynamics and transport for the simulation of tritium dispersion caused by an accident in a CANDU reactor located in the ongoing Angra 3 site. It was postulated, then, the LOCA - Loss of Coolant Accident -, accident in the emergency cooling system of the nucleus ( without fusion), where was lost 66m3 of soda almost instantaneously. This inventory contained 35 PBq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios ( plant stopped or operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1MBq/m3), during the first 14 days after the accident. The main difference between the scenario without and with seawater recirculation (pumping and discharge) is based on the enhancement of dilution of the highest concentrations in the last one. This dilution enhancement resulting in decreasing concentrations was observed only during the first two weeks, when they ranged from 1x109 to 5x105 Bq/m3 close to the Itaorna beach spreading just to Sandri Island. After 180 days, the plume could not be detected anymore in the bay, because their activities would be lower than the minimum detectable value (3). (author)

  11. Modelling of tritium dispersion from postulated accidental release of nuclear power plants

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models of hydrodynamics and transport for the simulation of tritium dispersion caused by an accident in a CANDU reactor located in the ongoing Angra 3 site. This exercise was accomplished with the aid of a code system (SisBAHIA) developed in the Rio de Janeiro Federal University (COPPE/UFRJ). The CANDU reactor is one that uses heavy water (D2O) as moderator and coolant of the core. It was postulated, then, the LOCA (Loss of Coolant Accident) accident in the emergency cooling system of the nucleus (without fusion), where was lost 66 m3 of soda almost instantaneously. This inventory contained 35 PBq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios (plant stopped and operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1 MBq/m3 ) during the first 14 days after the accident. The main difference between the scenario without and with seawater recirculation (pumping and discharge) is based on the enhancement of dilution of the highest concentrations in the last one. This dilution enhancement resulting in decreasing concentrations was observed only during the first two weeks, when they ranged from 1x109 to 5x105 Bq/m3 close to the Itaorna beach spreading just to Sandri Island. After 180 days, the plume could not be detected anymore in the bay, because their activities would be lower than the minimum detectable value (3). (author)

  12. Mesoscale atmospheric modeling of the July 12, 1992 tritium release from the Savannah River Site

    In August of 1991, the Environmental Transport Group (ETG) began the development of an advanced Emergency Response (ER) system based upon the Colorado State University Regional Atmospheric Modeling System (RAMS). This model simulates the three-dimensional, time-dependent, flow field and thermodynamic structure of the planetary boundary layer (PBL). A companion Lagrangian Particle Dispersion Model (LPDM) simulates contaminant transport based on the flow and turbulence fields generated by RAMS. This paper describes the performance of the advanced ER system in predicting transport and diffusion near the SRS when compared to meteorological and sampling data taken during the July 12, 1992 tritium release. Since PUFF/PLUME and 2DPUF are two Weather INformation and Display (WIND) System atmospheric models that were used to predict the transport and diffusion of the plume at the time of the release, the results from the advanced ER system are also compared to those produced by PUFF/PLUME and 2DPUF

  13. Environmental aspects of tritium release in the Argentinas nuclear power station

    The Argentine nuclear power facilities of Atucha (CNA) and Embalse (CNE) use natural uranium as fuel and heavy water as coolant. During the normal operation a fraction of generated tritium by neutron activation of the deuterium of heavy water, is released into the atmosphere by means of liquid and gaseous discharged. From the radiological protection point of view, it is one of the main effluents. Therefore, from the start in operation of the CNA (1974) and the CNE (1984), continuous determinations of the gaseous and liquid discharged are made, as well as the environmental monitoring in the surroundings of the facilities is realized. In this work, the values of release and the concentration integrated along the time, obtained from the environmental monitoring in both power facilities are correlated. The results obtained with the model of transference used in the preoperational studies, are compared considering the parameters of atmospheric dispersion corresponding to the meteorological conditions of the own site. (author)

  14. Thermal release of 3He from tritium aged LaNi4.25Al0.75 hydride

    The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 (LaNi4.25Al0.75)in the tritium process to store hydrogen isotopes. The vast majority of 3He born from the radioactive decay of tritium stored in LANA.75 is trapped in the hydride metal matrix. The SRS-TF has multiple LANA.75 tritium storage beds that have been retired from service with significant quantities of He-3 trapped in the metal. To support He-3 recovery, the Savannah River National Laboratory (SRNL) conducted thermogravimetric analysis coupled with mass spectrometry (TGA-MS) on a tritium aged LANA.75 sample. TGA-MS testing was performed in an argon environment. Prior to testing, the sample was isotopically exchanged with deuterium to reduce residual tritium and passivated with air to alleviate pyrophoric concerns associated with handling the material outside of an inert glovebox. Analyses indicated that gas release from this sample was bimodal, with peaks near 220 and 490 C. degrees. The first peak consisted of both 3He and residual hydrogen isotopes, the second was primarily 3He. The bulk of the gas was released by 600 Celsius degrees. (author)

  15. Tritium accumulation and release from Li{sub 2}TiO{sub 3} during long-term irradiation in the WWR-K reactor

    Tazhibayeva, I., E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy of National Nuclear Center RK, Krasnoarmeyskaya str.-10, 071100 Kurchatov (Kazakhstan); Beckman, I., E-mail: info@rector.msu.ru [Moscow State University, Leninskie Gory, 119991 Moscow (Russian Federation); Shestakov, V. [Kazakh State University, Tole bi str., 96, Almaty (Kazakhstan); Kulsartov, T. [Institute of Atomic Energy of National Nuclear Center RK, Krasnoarmeyskaya str.-10, 071100 Kurchatov (Kazakhstan); Chikhray, E. [Kazakh State University, Tole bi str., 96, Almaty (Kazakhstan); Kenzhin, E. [Institute of Atomic Energy of National Nuclear Center RK, Krasnoarmeyskaya str.-10, 071100 Kurchatov (Kazakhstan); Kiykabaeva, A. [Kazakh State University, Tole bi str., 96, Almaty (Kazakhstan); Kawamura, H.; Tsuchiya, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan)

    2011-10-01

    Proposed mathematical and software analysis of reactor experiments allowed interpretation of the experimental results of a tritium release study. Tritium was continuously generated by the reaction of lithium-6 with thermal neutrons for various thermal conditions of lithium metatitanate (Li{sub 2}TiO{sub 3}). The main gas release parameters were calculated in order to assess the potential use of lithium metatitanate in tritium breeders. These parameters were: gas release rate, tritium retention, retention time, activation energy for thermal desorption as HT, activation energy for volume diffusion as T{sup +}, and the corresponding pre-exponential (frequency) indexes.

  16. Chemical form of released tritium from molten Li{sub 2}BeF{sub 4} salt under neutron irradiation at elevated temperatures

    Suzuki, Akihiro; Terai, Takayuki; Yoneoka, Toshiaki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    Chemical forms of released tritium from FLIBE (the 2:1 mixture of LiF and BeF{sub 2}) by in-pile tritium release experiment were HT and TF and their proportion depended on the chemical composition of purge gas and the dehumidification time of specimen at high temperatures. The chemical form of tritium was determined by the thermodynamic equilibrium of the isotopic exchange reaction (T{sup +} + H{sub 2} {yields} H{sup +} + HT). (author)

  17. Tritium release behavior from neutron-irradiated Li{sub 2}TiO{sub 3} single crystal

    Tanifuji, Takaaki; Yamaki, Daiju; Noda, Kenji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nasu, Shoichi

    1998-03-01

    Li{sub 2}TiO{sub 3} single-crystals with various size (1-2mm) were used as specimens. After the irradiation up to 4 x 10{sup 18} n/cm{sup 2} with thermal neutrons in JRR-2, tritium release from the Li{sub 2}TiO{sub 3} specimens in isothermal heating tests was continuously measured with a proportional counter. The tritium release in the range from 625K to 1373K seems to be controlled by bulk diffusion. The tritium diffusion coefficient (D{sub T}) in Li{sub 2}TiO{sub 3} was evaluated to be D{sub T}(cm{sup 2}/sec) = 0.100exp(-104(kJ/mol)/RT), 625Ktritium diffusion coefficients in Li{sub 2}TiO{sub 3} is almost equal to those of Li{sub 2}O irradiated with thermal neutrons up to 2 x 10{sup 19} n/cm{sup 2}. It indicates that the tritium release performance of Li{sub 2}TiO{sub 3} is essentially good as Li{sub 2}O. (author)

  18. Levels of tritium in soils and vegetation near Canadian nuclear facilities releasing tritium to the atmosphere: implications for environmental models

    Concentrations of organically bound tritium (OBT) and tritiated water (HTO) were measured over two growing seasons in vegetation and soil samples obtained in the vicinity of four nuclear facilities and two background locations in Canada. At the background locations, with few exceptions, OBT concentrations were higher than HTO concentrations: OBT/HTO ratios in vegetation varied between 0.3 and 20 and values in soil varied between 2.7 and 15. In the vicinity of the four nuclear facilities OBT/HTO ratios in vegetation and soils deviated from the expected mean value of 0.7, which is used as a default value in environmental transfer models. Ratios of the OBT activity concentration in plants ([OBT]plant) to the OBT activity concentration in soils ([OBT]soil) appear to be a good indicator of the long-term behaviour of tritium in soil and vegetation. In general, OBT activity concentrations in soils were nearly equal to OBT activity concentrations in plants in the vicinity of the two nuclear power plants. [OBT]plant/[OBT]soil ratios considerably below unity observed at one nuclear processing facility represents historically higher levels of tritium in the environment. The results of our study reflect the dynamic nature of HTO retention and OBT formation in vegetation and soil during the growing season. Our data support the mounting evidence suggesting that some parameters used in environmental transfer models approved for regulatory assessments should be revisited to better account for the behavior of HTO and OBT in the environment and to ensure that modelled estimates (e.g., plant OBT) are appropriately conservative. - Highlights: • We measured tritium in soils and plants near four nuclear facilities in Canada. • OBT/HTO ratios in plants are higher than default value in environmental models. • OBT/HTO ratios in background soils reflect historically higher atmospheric tritium. • Implications for environmental transfer models are discussed

  19. Historical Doses To The Public from Routine and Accidental Releases of Tritium - Lawrence Livermore National Laboratory, 1953 - 2005

    Peterson, S; Raskob, W

    2007-08-15

    Throughout fifty-three years of operations, an estimated 29,300 TBq of tritium have been released to the atmosphere at the Livermore site of Lawrence Livermore National Laboratory; about 75% of this was released accidentally as gaseous tritium in 1965 and 1970. Routine emissions contributed slightly more than 3,700 TBq gaseous tritium and about 2,800 TBq tritiated water vapor to the total. Mean annual doses (with 95% confidence intervals) to the most exposed member of the public were calculated for all years using the same model and the same assumptions. Because time-dependent tritium models require detailed meteorological data that were unavailable for the large releases, ingestion/inhalation dose ratios were derived from experience with UFOTRI. Even with assumptions to assure that doses would not be underestimated, all doses from routine and accidental releases were below the level (3.6 mSv) at which adverse health effects have been documented, and most were below the current regulatory limit of 100 {micro}Sv per year from releases to the atmosphere.

  20. Pantex Plant Cell 12-44-1 tritium release: Re-assessment of environmental doses for 1990 to 1992

    A release of tritium gas occurred within Cell 12-44-1 at the Pantex Plant on May 17, 1989. The release was the result of a nuclear component containment failure. This document summarizes past assessments and characterization of the release. From 1990 to 1992, the average annual dose to the offsite maximally exposed individual (MEI), re-assessed using updated methods and data, ranged from 9E-6 to 2E-4 mrem/y. Doses at this level are well below the regulatory dose limit and support the discontinuation of the distinct calculation of the MEI doses from the cell's tritium releases in future Pantex Annual Site Environmental Reports. Additional information provides guidance for the evaluation of similar releases in the future. Improved Environmental Protection Department sampling plans and assessment goals will increase the value of the data collected during future incidents

  1. Maximum permissible amounts of accidentally released tritium derived from an environmental experiment to meet dose limits for public exposure

    This paper reports that it is important in the design of future fusion reactors and associated facilities that incorporate passive safety to take account of the possible environmental impact of accidental tritium release. Reliable information on dose consequences can be obtained by evaluating urine samples from persons exposed to tritium. Translating the results of the environmental HT experiment performed in France in 1986 into worst-case exposure conditions, the effective dose equivalent to an individual with highest exposure at a distance of 800 m (typical for site boundaries) is ∼1 x 10-4 Sv per gram of tritium emitted as HT when inhalation and skin absorption are considered. From this value, maximum permissible amounts of accidentally released HT can be derived on the basis of regulatory or anticipated dose limits

  2. Tritium release kinetics from Li2TiO3 pebbles as prepared by soft-wet-chemistry

    Lithium meta titanate pebbles has been prepared from agglomeration-sintering powders which were obtained by Li-Ti-peroxo-complex solution precursor (Li2TiO3 dissolved at room temperature in H2O + 40% H2O2 and stabilized with citric acid). Through this wet route Li2TiO3 pebbles with high density(∼92% of T.D.) has been obtained and the tritium release behavior has been tested 'in-pile' by the EXOTIC-8.9 experiment (∼440 days of irradiation at full power in the high neutron flux of HFR-Petten). Tritium residence times (τ) in the pebbles has been measured during irradiation between 550 and 400 deg. C and He + 0.1%H2 purge gas composition. By a thermally activated process (activation energy=111 kJ/mol) with 410 deg. C as minimum temperature the tritium residence time is found to be about 1 day, which places this specimen in a good ranking position among those tested by the EXOTIC-series. A clear increase of the tritium release rate has been observed by increased H2 concentration (up to 1%) in the He purge. Out-of-pile ramp-annealing tritium desorption (TPD) tests on short-time irradiated pebbles has been also performed by various devices and conditions. The kinetic parameters from the TPD investigation gave consistent results with those characterizing the equilibrium times of tritium release rate after the gas composition and temperature transients imposed on the specimen during the in-pile experiment

  3. Tritium release kinetics from Li2TiO3 pebbles as prepared by soft-wet-chemistry

    Lithium meta titanate pebbles has been prepared from agglomeration-sintering powders which were obtained by Li-Ti-peroxo-complex solution precursor (Li2TiO3 dissolved at room temperature in H2O +40% H2O2 and stabilized with citric acid). Through this wet route Li2TiO3 pebbles with high density (90/93% of T.D.) has been obtained and the tritium release behaviour has been tested 'in-pile' by the EXOTIC-8.9 experiment (∼440 days of irradiation at full power in the high neutron flux of HFR-Petten). Tritium residence times (τ) in the pebbles has been measured during irradiation between 550 and 400degC and He + 0.1% H2 purge gas composition. By a thermally activated process (activation heat = 111 kJ/mol) with 410degC as minimum temperature the tritium residence time is found to be about 1 day, which places this specimen in a good ranking position among those tested by the EXOTIC-series. A clear increase of the tritium release rate has been observed by increased H2 concentration (up to 1%) in the He purge. Out-of-pile ramp-annealing tritium desorption (TPD) tests on shortly irradiated pebbles has been also performed by various devices and conditions. The kinetic parameters from the TPD spectra were consistent with those characterizing the equilibrium times of tritium release rate after the gas composition and temperature transients imposed to the specimen during the in-pile experiment. (author)

  4. Migration and release behavior of tritium in SS316 at ambient temperature

    Torikai, Y.; Murata, D.; Penzhorn, R.-D.; Akaishi, K.; Watanabe, K.; Matsuyama, M.

    2007-06-01

    BIXS measurements indicate that immersion into water or chemical etching of SS316 contaminated with tritium at moderate temperatures causes an immediate reduction of the outermost surface concentration of tritium. The fraction of surface tritium removed by water, i.e. 30-50%, is small in comparison to the total tritium present in the specimen. Allowing a specimen to age whose surface and subsurface had been removed by etching up to a depth where the concentration of tritium is mostly constant revealed that within a few months a re-growth of tritium up to a saturation value higher than half of that originally present on the specimen takes place. Concurrently, a small but steady liberation of tritium at rates increasing from 0.1 to 0.3 kBq/h was noticed.

  5. In-pile tritium release behaviour of lithiummetatitanate produced by extrusion-spheroidisation-sintering process in EXOTIC-9/1 in the high flux reactor, Petten

    Peeters, M.M.W. [N.R.G., P.O. Box 25, 1755 ZG Petten (Netherlands)], E-mail: peeters@nrg-nl.com; Magielsen, A.J.; Stijkel, M.P.; Laan, J.G. van der [N.R.G., P.O. Box 25, 1755 ZG Petten (Netherlands)

    2007-10-15

    The irradiation programme EXOTIC (extraction of tritium in ceramics) is carried out within the European framework for the development of the helium cooled pebble bed concept. The EXOTIC-9/1 is the latest experiment in the series of EXOTICs that are irradiated in the high flux reactor in Petten. Tritium release and inventory in lithium containing ceramic pebbles are key properties to be tested in a TBM. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The objective of the EXOTIC-9/1 is to study in-pile tritium release behaviour of the latest developed lithiummetatitanate pebbles (Li{sub 2}TiO{sub 3}). The pebbles are produced by a extrusion-spheroidisation-sintering process at CEA. The new pebbles differ with respect to porosity from the lithiummetatitanate ceramics tested in the previous EXOTIC 8 programme. The pebbles have diameter in the range from 0.6 to 0.8 mm. Irradiation of EXOTC-9/1 started at 24 March 2005, and will continue until the end of 2006, in total about 400 irradiation days. The temperature is varied between 340 and 580 deg. C. Begin of Life (BOL) tritium production rate is 0.56 mCi/min. Based upon the in-pile tritium release measurements and the analysis of the tritium residence time it can be concluded that tritium release in the new batch of the high density Li{sub 2}TiO{sub 3} pebbles irradiated in EXOTIC 9/1 is rather slow compared to the ceramics irradiated in the EXOTIC 8 irradiation campaign. In this paper, the in-pile tritium behaviour will be reported during normal operation and during transients in temperature, purge gas chemistry and gasflow. The collected data is compared to tritium release data from ceramics irradiated in previous EXOTIC experiments with respect to tritium inventory, residence time and porosity.

  6. Thermal release of tritium implanted in graphite studied by T(d,α)n nuclear reaction depth profiling analysis

    Specimens of graphite from limiter tiles in JET were implanted at room temperature with HT+ and DT+ ions at energies in the range 10-50 keV and at fluences between 5x1014 and 3x1016 ions/cm2, using the isotope separator. Depth profiles of tritium were measured by the T(d,α)n reaction using glancing incidence of the incoming 500 keV D2+ ions and detecting the outgoing α particles in a forward direction. Considerable broadening of the experimentally obtained depth profiles was observed as compared to calculated ones. For a specimen implanted with 5x1015 HT+ ions/cm2 at 40 keV, the depth profiles of tritium were studied as a function of isochronal annealing in vacuum up to ≅ 700 K. It was found that the release of tritium proceeds essentially without any change in the depth profile. Implanted tritium stars to be detrapped around 600 K, reaches a maximum in its release rate at 1100-1400 K and is 95% released at 1600 K. The obtained release curve is consistent with that obtained by other investigators for deuterium implanted at fluences of ≅ 1016-1017 D+ ions/cm2, and it is shifted to ≅ 200-300 K higher temperatures compared to the case of ≅ 1018-1019/cm2 deuterium fluences. The observed behaviour is believed to be indicative of a single-step detrapping and recombination mechanism of implanted tritium ions, and its subsequent fast transgranular diffusion as molecules. (orig.)

  7. Environmental health-risk assessment for tritium releases at the National Tritium Labeling Facility at Lawrence Berkeley National Laboratory

    McKone, T.E.; Brand, K.P. [Lawrence Livermore National Lab., CA (United States). Health and Ecological Assessment Div.; Shan, C. [Lawrence Berkeley National Lab., CA (United States). Earth Sciences Div.

    1997-04-01

    This risk assessment calculates the probability of experiencing health effects, including cancer incidence due to tritium exposure for three groups of people: (1) LBNL workers near the LBNL facility--Building 75--that uses tritium; (2) other workers at LBNL and nearby neighbors; and (3) people who use the UC Berkeley campus area, and some Berkeley residents. All of these groups share the same probability of health effects from the background radiation from natural sources in the Berkeley area environment, including an increased risk of developing a cancer of 11,000 chances per million. In calculating risk the authors assumed continuous operation in Building 75 for at least a human lifetime. Under this assumption, LBNL workers located near Building 75 have an additional risk of 60 chances out of one million to suffer a cancer; other workers at LBNL and people who live near LBNL have an additional risk of six chances out of one million over a lifetime of exposure; and users of the UC Berkeley campus area and other residents of Berkeley have an additional risk of less than once chance out of one million over a lifetime.

  8. Environmental health-risk assessment for tritium releases at the National Tritium Labeling Facility at Lawrence Berkeley National Laboratory

    This risk assessment calculates the probability of experiencing health effects, including cancer incidence due to tritium exposure for three groups of people: (1) LBNL workers near the LBNL facility--Building 75--that uses tritium; (2) other workers at LBNL and nearby neighbors; and (3) people who use the UC Berkeley campus area, and some Berkeley residents. All of these groups share the same probability of health effects from the background radiation from natural sources in the Berkeley area environment, including an increased risk of developing a cancer of 11,000 chances per million. In calculating risk the authors assumed continuous operation in Building 75 for at least a human lifetime. Under this assumption, LBNL workers located near Building 75 have an additional risk of 60 chances out of one million to suffer a cancer; other workers at LBNL and people who live near LBNL have an additional risk of six chances out of one million over a lifetime of exposure; and users of the UC Berkeley campus area and other residents of Berkeley have an additional risk of less than once chance out of one million over a lifetime

  9. Sources of tritium

    A review of tritium sources is presented. The tritium production and release rates are discussed for light water reactors (LWRs), heavy water reactors (HWRs), high temperature gas cooled reactors (HTGRs), liquid metal fast breeder reactors (LMFBRs), and molten salt breeder reactors (MSBRs). In addition, release rates are discussed for tritium production facilities, fuel reprocessing plants, weapons detonations, and fusion reactors. A discussion of the chemical form of the release is included. The energy producing facilities are ranked in order of increasing tritium production and release. The ranking is: HTGRs, LWRs, LMFBRs, MSBRs, and HWRs. The majority of tritium has been released in the form of tritiated water

  10. Accidental tritium release from nuclear technologies and a radiobiological survey of the impact of low dose tritium on the developing mouse brain

    Full text: The Atomic Energy Act, 1962 provides for the development of the peaceful uses of atomic energy for the welfare of the people in India. The licensing policy adopted for nuclear power stations in India requires that the plants meet stringent requirements based on the system of dose limitation, recommended by the International Commission of Radiological Protection (ICRP). Currently, nuclear energy is contributing just 3% of the country's power generation. The share of nuclear power is proposed to be increased to 10% in the near future. With the introduction of nuclear energy, the need to assess the radioecological and radiobiological impact of radionuclides of long half- life existing in the environment for longer duration has appeared. Tritium, a radioactive by-product of power reactors is one of such major radionuclides of concern. In the world, routine releases and accidental spills of tritium from nuclear power plants pose a growing health and safety concern. Tritium has been observed in ground water in the vicinity of several nuclear stations. Exposure to tritium has been clinically proven to cause deleterious and detectable effects such as teratogenesis, cancer and life shortening in laboratory animals. There is, now, a growing emphasis on tritium in radiation protection as the challenge of nuclear fusion comes nearer. Present investigation is an attempt to elucidate the effects of low dose tritiated water exposure on developing mouse cerebellum. Pregnant Swiss albino mice (12-15 in number were given a priming injection 7.4 and 74 kBq/ml of body water) of tritiated water (HTO) on 16th day of gestation. From the same day onward, through parturition, till the last interval studied, the pregnant females were continuously maintained respectively on 11.1 and 111 kBq/ml of tritiated drinking water provided ad libidum. After cervical dislocation the litters were autopsied on 1, 3, 5 and 6 weeks post- partum. Brains were fixed and then cerebellum from each of

  11. The evaluation of the nuclides migration from Maishiagala radioactive waste repository taking into consideration the actual tritium release

    The evaluation of the nuclides migration from Maishiagala repository was performed taking into consideration the actual tritium release. The waste activity includes the activity of nuclides from the sealed containers. It was shown that the most dangerous nuclides are 3H, 36Cl and 239Pu which concentrations in the groundwater 35 m from the repository (at fence) and annual effective dose to population can exceed acceptable limits. (author)

  12. Validation of phenomena relative to tritium transport in irradiated closed capsules filled with Pb-17Li through modelling tritium release in LIBRETTO-2 experiment

    The control of tritium permeation through the cooling tubes into the cooling water is a key feasibility issue for the European water-cooled Pb-17Li blanket for the DEMO reactor. Modelling of the LIBRETTO-2 experiment is outlined. A selection of possible release mechanisms followed by their evaluation impose to consider specific interface and irradiation phenomena. From the confidence on the final approaches, values for empirical parameters as alloy/cladding wetting factors, weld diffusivities or sticking factors are provided. In-pile correlations under HFR irradiation conditions are validated. An average value of 90 for the barrier permeation reduction factor is obtained for the PC-Al2O3 coating through the LIBRETTO-2 irradiation history. (author) 19 refs.; 4 figs

  13. Tritium releases from the Pickering Nuclear Generating Station and birth defects and infant mortality in nearby communities 1971-1988

    This study was commissioned to examine whether there were elevated rates of stillbirth, birth defects, or death in the first year of life between 1971 and 1988 among offspring of residents of communities within a 25-kilometre radius of the Pickering Nuclear Generating Station. The study was also to investigate whether there were any statistical associations between the monthly airborne or waterborne tritium emissions from the Pickering Nuclear Generating Station and the rates of these reproductive outcomes. Overall analysis did not support a hypothesis of increased rates of stillbirths, neonatal mortality or infant mortality near the Pickering Nuclear Generating Station, or a hypothesis of increased birth prevalence of birth defects for 21 of 22 diagnostic categories. The prevalence of Down Syndrome was elevated in both Pickering and Ajax; however, there was no consistent pattern between tritium release levels and Down Syndrome prevalence, chance could not be ruled out for the associations between Down Syndrome and tritium releases or ground-monitored concentrations, the association was detected in an analysis where multiple testing was done which may turn up significant associations by change, and maternal residence at birth and early in pregnancy needs to be verified. The association between Down Syndrome and low-level radiation remains indeterminate when existing evidence from epidemiological studies is summed. The estimated radiation exposure from the nuclear plant for residents of Pickering and Ajax is lower by a factor of 100 than the normal natural background radiation. Further study is recommended. (21 tabs., 29 figs., 5 maps, 37 refs.)

  14. Welsh tritium

    Of all radioactive isotopes, tritium and carbon-14 have a special status because of the possibility of their intimate involvement in the biosphere. Both are formed naturally in the upper atmosphere but both are also anthropogenic and discharged into the environment. Tritium has engendered considerably greater notoriety as it has been released into the environment in quite large amounts during nuclear weapons testing and subsequently from nuclear plants. The natural tritium inventory of about 1.3 EBq was dwarfed by contributions from weapons testing. In the 1960s this added about 186 EBq to the global inventory which even today remains at about 50 EBq. In contrast the nuclear industry has contributed about 0.43 EBq but the rate of discharge from some plants is far from insignificant - for instance, the Savannah River site in South Carolina (which is responsible for about 90% of the US tritium releases) discharged about 0.02 EBq in 1987. Currently the major sources of anthropogenic tritium in the UK are [4] the BNF plants at Sellafield (2756 TBq/year, 91% as liquid) and Chapelcross (1421 TBq/year, 0.05% as liquid). As described in the paper there have been unexpected levels of tritium in fish caught in the Bristol Channel in the vicinity of the outfall of the discharge from the Cardiff factory. This tritium is 'unexpected' because the levels in sea water in the area have been measured at around 10 Bq/l [4] and a greater part (90%) of the uptake into fish has been shown to be organically bound tritium (OBT) rather than as part of the body water

  15. Behaviour of three chemical forms of tritium in the environment after release from inertial fusion reactors

    Velarde, M.; Perlado, J.M. [Instituto de Fusion Nuclear (DENIM)/ETSII/Univ. Politecnica Madrid (Spain); Sedano, L. [CIEMAT, Madrid (Spain)

    2006-06-15

    In order to fully simulate the behaviour of elementary tritium (HT), tritiated water vapour (HTO) in the environment, it is necessary to take into account diffusion and deposition processes in the soil and vegetables. In addition this work also incorporates the penetration in the underground, re-emission and later conversion to organic tritium (OBT). The whole study has led to the conclusion that the behaviour of the tritium should be simulated using two well-differentiated studies: deterministic and probabilistic. Deterministic calculations are based on a fixed meteorological data given 'a priori'. The probabilistic study is based on measured real meteorological analysis every hour, and the probability that individuals can present dose for internal irradiation. Both options have been considered for a specific mediterranean environment of the system. Once the elementary tritium has been deposited in the soil, it can be oxidized by microbial action of the enzymes of the soil, and the resulting tritium form (in its oxidize form) goes back to the atmosphere. This process of re-emission is shown to be very important since it has been typically considered that the inhaled tritium is only, HTO, when, in fact part of that account is due to the HT converted to HTO and re-emitted to the atmosphere. Our calculations demonstrate that the HT contributes very significantly to the dose for inhalation through the re-emission processes. A final aspect of this work is the dosimetric analysis of the contamination through all ways: inhalation, re-emission and ingestion. Early and chronic doses have been assessed.

  16. Comparison of the Regulatory Models Assessing Off-Site Radiological Dose due to the Routine Releases of Tritium

    Methodologies of NEWTRIT model, NRC model and AIRDOS-EPA model, which are off-site dose assessment models for regulatory compliance from routine releases of tritium into the environment, were investigated. Using the domestic data, if available, the predictive results of the models were compared. Among them, recently developed NEWTRIT model considers only doses from organically bounded tritium (OBT) due to environmental releases of tritiated water (HTO) . A total dose from all exposure pathways predicted from AIRDOS-EPA model was 1.03 and 2.46 times higher than that from NEWTRIT model and NRC model, respectively. From above result, readers should not have an understanding that a predictive dose from NRC model may be underestimated compared with a realistic dose. It is because of that both mathematical models and corresponding parameter values for regulatory compliance are based on the conservative assumptions. For a dose by food consumption predicted from NEWTRIT model, the contribution of OBT was nearly equivalent to that of HTO due to relatively high consumption of grains in Korean. Although a total dose predicted from NEWTRIT model is similar to that from AIRDOS-EPA model, NEIIfTRIT model may be have a meaning in the understanding of phenomena for the behavior of HTO released into the environment.

  17. Kinetics of tritium release from irradiated Li2TiO3 pebbles in out-of-pile TPD tests

    The rate of tritium release from Li2TiO3 pebbles was examined by post irradiation thermal desorption spectroscopy (the Temperature Programmed Desorption (TPD) method). Pre-treatments before and even after irradiation were found useful to gain insight on the behavior of these pebbles at different temperatures, as good spectrum de-convolution is achieved and kinetic parameters for the rate determining pseudo-first-order steps can be estimated. We show the results concerning Li2TiO3 pebbles bed specimens developed in the frame of the European fusion technology program

  18. Studies on the biological half-lives of tritium released at Wolsong Nuclear Power Plants

    The one of important parameter involved in the calculation of internal radiation dose to the human body is the biological half-life of the radionuclide. The biological half-life is population specific and may differ from one population group to another. So the effective half-life of tritium exposure based on urinal bioassay measurement of Wolsong Nuclear Power Plants was investigated and studied

  19. Studies on the biological half-lives of tritium released at Wolsong Nuclear Power Plants

    Kim, H. G.; Eum, H. M. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Cha, S. C.; Kim, M. C. [Wolsong Nuclear Power Plants, Gyungju (Korea, Republic of)

    2001-09-15

    The one of important parameter involved in the calculation of internal radiation dose to the human body is the biological half-life of the radionuclide. The biological half-life is population specific and may differ from one population group to another. So the effective half-life of tritium exposure based on urinal bioassay measurement of Wolsong Nuclear Power Plants was investigated and studied.

  20. Description of NORMTRI: a computer program for assessing the off-site consequences from air-borne releases of tritium during normal operation of nuclear facilities

    The computer program NORMTRI has been developed to calculate the behaviour of tritium in the environment released into the atmosphere under normal operation of nuclear facilities. It is possible to investigate the two chemical forms tritium gas and tritiated water vapour. The conversion of tritium gas into tritiated water followed by its reemission back to the atmosphere as well as the conversion into organically bound tritium is considered. NORMTRI is based on the statistical Gaussian dispersion model ISOLA, which calculates the activity concentration in air near the ground contamination due to dry and wet deposition at specified locations in a polar grid system. ISOLA requires a four-parametric meteorological statistics derived from one or more years synoptic recordings of 1-hour-averages of wind speed, wind direction, stability class and precipitation intensity. Additional features of NORMTRI are the possibility to choose several dose calculation procedures, ranging from the equations of the German regulatory guidelines to a pure specific equilibrium approach. (orig.)

  1. Preliminary study of the impact of tritium and carbon 14 releases from the Saint-Alban nuclear power plant. CRIIRAD N.04-20 V1 Report

    After having recalled the results of previous studies on the radioactivity in surface water and land environments, and outlined the need of an investigation of the tritium and carbon 14 contamination, this report defines the objectives of this investigation, the adopted methodology (choice of plants, tritium and carbon 14 dose measurements, and sampling to study time variations). It recalls some aspects of tritium and carbon 14 releases (production of radionuclides, origins of emissions in the environment, assessments by EDF). It reports the investigation and the assessment of tritium activity in a land environment and in rain waters about the investigated site, and the investigation and the assessment of carbon 14 activity within the same environment. It reports preliminary results concerning the aquatic environment

  2. Dose assessment for releases of tritium and activation products into the atmosphere performed in the frame of two fusion related studies: ITER-EDA and SEAFP

    Within the SEAFP and ITER studies dose calculations have been performed for tritium and activation products. Unit release rates as well as preliminary activation product source terms have been investigated. The individual dose values at the fence of the site together with the collective dose to the public have been obtained. Worst case and typical release conditions have been investigated. Additionally, various release durations under accidental conditions, ranging from 1 hour up to 7 days, have been considered. (orig.)

  3. Dose assessment for releases of tritium and activation products into the atmosphere performed in the frame of two fusion related studies: ITER-EDA and SEAFP

    Raskob, W. [Kernforschungszentrum Karlsruhe, Abt. INR (Germany)]|[D.T.I. Dr. Trippe Ingenieurgesellschaft, Karlsruhe (Germany)

    1995-12-31

    Within the SEAFP and ITER studies dose calculations have been performed for tritium and activation products. Unit release rates as well as preliminary activation product source terms have been investigated. The individual dose values at the fence of the site together with the collective dose to the public have been obtained. Worst case and typical release conditions have been investigated. Additionally, various release durations under accidental conditions, ranging from 1 hour up to 7 days, have been considered. (orig.).

  4. Effect of purge gas oxidizing potential on tritium release from Li-ceramics and on its permeation through 316L SS clads under irradiation (TRINE experiment)

    The effect of red-ox potential of helium purge gas (variously doped with H2, H2O and O2) was examined on tritium release from Li-ceramics (LiAlO2 and Li2ZrO3 pellets) and on its permeation rate through the 316L stainless steel clads (bare and coated) held at 500 C. Decreasing the H2 content from 1000 vpm (reference 'R' gas mixture) to 100 vpm, and substituting H2O for H2, the tritium permeation rate (ca. 1.41010 atoms cm-2 s-1 in R-gas) increases. Tritium inventories in the Li ceramics were increased too. When a strong oxidizing purge (1000 vpm O2 added to He containing 100 vpm H2O) was used, a retention time (τ) of two days at 400 C was measured for Li2ZrO3. In this oxidizing environment the tritium permeation loss dropped by a factor five for the uncoated capsules while an aluminide coating became a very effective tritium barrier: tritium permeation flux at 550 C fell below the measurable limit. (orig.)

  5. Leukaemia in the vicinity of two tritium-releasing nuclear facilities: a comparison of the Kruemmel Site, Germany, and the Savannah River Site, South Carolina, USA

    In 1991, an increased rate of childhood leukaemia was reported from the small northern German community of Elbmarsch, which is located on the banks of the River Elbe opposite the Kruemmel nuclear power plant. Owing to the fact that the increase occurred six years after the start-up of the plant, radioactive discharges were suspected as being implicated in the development of the cases. Previous investigations have failed to identify any exposure which might be associated with the cluster. Nonetheless, concern regarding the increased tritium burden in the environment remains. To further assess the impact of tritium releases to the environment upon population cancer rates, the releases and leukaemia rates at the Savannah River site, USA, were compared with the Kruemmel site. Based on the data from 1991 to 1995, the incidence of childhood leukaemia in the vicinity of the Savannah River site was non-significantly less than expected compared with the significantly higher than expected rates close to the German plant. In contrast, tritium releases from the Savannah River site exceed those from the Kruemmel site by several orders of magnitude. The results of this observational study suggest that factors other than environmental tritium releases are associated with the increased number of leukaemia cases near the Kruemmel site. (author)

  6. TFTR tritium handling concepts

    The Tokamak Fusion Test Reactor, to be located on the Princeton Forrestal Campus, is expected to operate with 1 to 2.5 MA tritium--deuterium plasmas, with the pulses involving injection of 50 to 150 Ci (5 to 16 mg) of tritium. Attainment of fusion conditions is based on generation of an approximately 1 keV tritium plasma by ohmic heating and conversion to a moderately hot tritium--deuterium ion plasma by injection of a ''preheating'' deuterium neutral beam (40 to 80 keV), followed by injection of a ''reacting'' beam of high energy neutral deuterium (120 to 150 keV). Additionally, compressions accompany the beam injections. Environmental, safety and cost considerations led to the decision to limit the amount of tritium gas on-site to that required for an experiment, maintaining all other tritium in ''solidified'' form. The form of the tritium supply is as uranium tritide, while the spent tritium and other hydrogen isotopes are getter-trapped by zirconium--aluminum alloy. The issues treated include: (1) design concepts for the tritium generator and its purification, dispensing, replenishment, containment, and containment--cleanup systems; (2) features of the spent plasma trapping system, particularly the regenerable absorption cartridges, their integration into the vacuum system, and the handling of non-getterables; (3) tritium permeation through the equipment and the anticipated releases to the environment; (4) overview of the tritium related ventilation systems; and (5) design bases for the facility's tritium clean-up systems

  7. Uncertainties in modeling of consequences of tritium release from fusion reactors. Plasma Fusion Center No. PFC/TR-79-5

    The bases for various models concerned with all phases of estimating doses from routine tritium releases from fusion reactors have been examined. The implications of uncertainties in parameters and assumptions for the uncertainty of the calculated doses and resulting maximum permissible releases are presented. Global dispersion models are most affected by the assumptions made concerning movement, such as the role of the ocean as a sink. Dose models were generally found to agree within a factor of two, with the largest variation due to agricultural data. Plant tritium flow studies are the least developed and require substantial improvement in the data base. Based on two possible arbitrary global standards, the maximum allowable releases were found to range from 1.6 to 20,000 Ci/day. The local criteria imply releases between 5 and 20 Ci/day

  8. Uncertainties in modeling of consequences of tritium release from fusion reactors. Plasma Fusion Center No. PFC/TR-79-5

    Piet, S.J.; Kazimi, M.S.

    1979-07-01

    The bases for various models concerned with all phases of estimating doses from routine tritium releases from fusion reactors have been examined. The implications of uncertainties in parameters and assumptions for the uncertainty of the calculated doses and resulting maximum permissible releases are presented. Global dispersion models are most affected by the assumptions made concerning movement, such as the role of the ocean as a sink. Dose models were generally found to agree within a factor of two, with the largest variation due to agricultural data. Plant tritium flow studies are the least developed and require substantial improvement in the data base. Based on two possible arbitrary global standards, the maximum allowable releases were found to range from 1.6 to 20,000 Ci/day. The local criteria imply releases between 5 and 20 Ci/day.

  9. In-situ tritium release (CORELLI-2 experiment) and ex-reactor ionic conductivity of substoichiometric LiAlO2 breeder ceramics

    LiAlO2 pellets with about 5% Li deficiency, prepared by a ''wet'' and a ''dry'' route were tested in situ for tritium release properties in nearly the same environmental conditions (CORELLI-2 experiment). Both the ''wet'' and ''dry'' route specimens were characterized by 80% of theoretical density (TD), almost fully open porosity and grain size ≤0.5 μm. The tritium removal rate evolution, following temperature or sweep gas changes during the irradiation, were observed to be nearly the same for both materials, in spite of their different preparation routes and impurities concentration. The ionic conductivities, as determined by impedance spectroscopy, were also similar. The presence of LiAl5O8 spinel phase in both samples apparently influenced the defect structure related transport properties of both lithium and tritium in these materials. (orig.)

  10. Nuclear graphite waste's behaviour under disposal conditions: Study of the release and repartition of organic and inorganic forms of carbon 14 and tritium in alkaline media

    23000 tons of graphite wastes will be generated during dismantling of the first generation of French reactors (9 gas cooled reactors). These wastes are classified as Long Lived Low Level wastes (LLW-LL). As requested by the law, the French National Radioactive Waste Management Agency (Andra) is studying concepts of low-depth disposals.In this work we focus on carbon 14, the main long-lived radionuclide in graphite waste (5730 y), but also on tritium, which is the main contributor to the radioactivity in the short term. Carbon 14 and tritium may be released from graphite waste in many forms in gaseous phase (14CO2, HT...) or in solution (14CO32-, HTO...). Their speciation will strongly affect their migration from the disposal site to the environment. Leaching experiments, in alkaline solution (0.1 M NaOH simulating repository conditions) have been performed on irradiated graphite, from Saint-Laurent A2 and G2 reactors, in order to quantify their release and characterize their speciation. The studies show that carbon 14 exists in both gaseous and aqueous phases. In the gaseous phase, release is weak (≤0.1%) and corresponds to oxidizable species. Carbon 14 is mainly released into liquid phase, as both inorganic and organic species. 65% of released fraction is inorganic and 35% organic carbon. Two tritiated species have been identified in gaseous phase: HTO and HT/Organically Bond Tritium. More than 90% of tritium in that phase corresponds to HT/OBT. But release is weak (≤0.1%). HTO is mainly in the liquid phase. (author)

  11. Intercomparison of model predictions of tritium concentrations in soil and foods following acute airborne HTO exposure

    This paper describes the results of a model intercomparision exercise for predicting tritium transport through foodchains. Modellers were asked to assume that farmland was exposed for one hour to an average concentration in air of 104 MBq tritium m-3. They were given the initial soil moisture content and 30 days of hourly averaged historical weather and asked to predict HTO and OBT concentrations in foods at selected times up to 30 days later when crops were assumed to be harvested. Two fumigations were postulated, one at 10.00 h (i.e., in day-light), and the other at 24.00 h (i.e., in darkness).Predicted environmental media concentrations after the daytime exposure agreed within an order of magnitude in most cases. Important sources of differences were variations in choices of numerical values for transport parameters. The different depths of soil layers used in the models appeared to make important contributions to differences in predictions for the given scenario. Following the night-time exposure, however, greater differences in predicted concentrations appeared. These arose largely because of different ways key processes were assumed to be affected by darkness. Uptake of HTO by vegetation and the rate it is converted to OBT were prominent amongst these processes. Further research, experimental data and modelling intercomparisons are required to resolve some of these issues. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)

  12. Environmental Impact of a Tritium Extraction System Small Pipe Break by the Atmospheric Modelling of Elemental Tritium Gas transport with Flexpart

    Castro, Paloma; Ardao, Jose; Velarde, Marta; Xiberta, Jorge; Sedano, Luis

    2014-05-01

    In the case of a little Tritium-Extraction-System (TES) pipe break (with critical failure of a fuelling line), the tritium source term has not yet been determined in the frame of European Test Blanket Systems, as Design Basis Accident (DBA) but it is expected to be in the order of a few grams. In this critical scenario acute modeling of environmental tritium transport forms (HT and HTO) for the assessment of fusion facilities dosimetric impact appears as of major interest. This paper considers different term releases of tritium-forms to the atmosphere from ITER which has experienced a frequent failure of a fueling line, due the little TES pipe break affecting a Helium-Cooled-Lithium-Lead Test-Blanket-Module. In case of 24.3 g of tritium were released from the broken fuelling-line directly into the gallery found only 0.5 g was released to the environment, assuming a little rupture in the TES piping located in the Port Cell. In this paper we assume a hypothetical daily release of one gram of tritium in HT and HTO forms. The daily failure is taken just in order to evaluate different meteorological scenarios or weather conditions. The FLEXPART working model simulates the tritium forms dispersion and environmental impact out of the complex ITER-tokamak (and its safeguards) of selected environmental patterns both inland and in-sea using ECMWF/FLEXPART model. We explore specific values of this ratio at different levels. We examine the influence of meteorological conditions of the tritium behavior during 48 hours after the release. For this purpose we have FLEXPART version 9.2 numerical weather model which is useful to follow real-time releases of tritium at low levels of the boundary layer to provide an approximation of tritium cloud behavior ranging from 3 to 48 hours.

  13. Public Health Consequences on Vulnerable Populations from Acute Chemical Releases

    Perri Zeitz Ruckart

    2008-01-01

    Full Text Available Data from a large, multi-state surveillance system on acute chemical releases were analyzed to describe the type of events that are potentially affecting vulnerable populations (children, elderly and hospitalized patients in order to better prevent and plan for these types of incidents in the future. During 2003–2005, there were 231 events where vulnerable populations were within ¼ mile of the event and the area of impact was greater than 200 feet from the facility/point of release. Most events occurred on a weekday during times when day care centers or schools were likely to be in session. Equipment failure and human error caused a majority of the releases. Agencies involved in preparing for and responding to chemical emergencies should work with hospitals, nursing homes, day care centers, and schools to develop policies and procedures for initiating appropriate protective measures and managing the medical needs of patients. Chemical emergency response drills should involve the entire community to protect those that may be more susceptible to harm.

  14. Li4SiO4 pebbles reduction in He + 0.1% H2 purge gas and effects on tritium release properties

    Lithium orthosilicate reduction was examined by Temperature Programmed Reaction (TPR) and Temperature Programmed Desorption (TPD) methods performed in He (or Ar) + H2 purge gas flowing through pebble bed specimens. The parameters governing the kinetics and the steady-state of the reduction process to Li4SiO4-x were determined at 800 deg. C. The level x of the O-vacancy concentration at steady-state (of the order of 1.5x10-3 mole fraction) was found to be compatible with the impurities content in the specimens. Pebble pre-annealing treatments were found to affect the microstructure and the reduction mechanism. Post-irradiation tritium release by TPD tests were performed on both stoichiometric and reduced pebbles with similar results. Tritium release properties of this breeder system seem to be independent from the material reduction state (x)

  15. Historical Doses from Tritiated Water and Tritiated Hydrogen Gas Released to the Atmosphere from Lawrence Livermore National Laboratory (LLNL) Part 1. Description of Tritium Dose Model (DCART) for Routine Releases from LLNL

    Peterson, S R

    2006-09-27

    DCART (Doses from Chronic Atmospheric Releases of Tritium) is a spreadsheet model developed at Lawrence Livermore National Laboratory (LLNL) that calculates doses from inhalation of tritiated hydrogen gas (HT), inhalation and skin absorption of tritiated water (HTO), and ingestion of HTO and organically bound tritium (OBT) to adult, child (age 10), and infant (age 6 months to 1 year) from routine atmospheric releases of HT and HTO. DCART is a deterministic model that, when coupled to the risk assessment software Crystal Ball{reg_sign}, predicts doses with a 95% confidence interval. The equations used by DCART are described and all distributions on parameter values are presented. DCART has been tested against the results of other models and several sets of observations in the Tritium Working Groups of the International Atomic Energy Agency's programs, Biosphere Modeling and Assessment and Environmental Modeling for Radiation Safety. The version of DCART described here has been modified to include parameter values and distributions specific to conditions at LLNL. In future work, DCART will be used to reconstruct dose to the hypothetical maximally exposed individual from annual routine releases of HTO and HT from all LLNL facilities and from the Sandia National Laboratory's Tritium Research Laboratory over the last fifty years.

  16. Development of a code to simulate dispersion of atmospheric released tritium gas in the environmental media and to evaluate doses. TRIDOSE

    A computer code (TRIDOSE) was developed to assess the environmental impact of atmospheric released tritium gas (T2) from nuclear fusion related facilities. The TRIDOSE simulates dispersion of T2 and resultant HTO in the atmosphere, land, plant, water and foods in the environment, and evaluates contamination concentrations in the media and exposure doses. A part of the mathematical models in TRIDOSE were verified by comparison of the calculation with the results of the short range (400 m) dispersion experiment of HT gas performed in Canada postulating a short-time (30 minutes) accidental release. (author)

  17. Development of a code to simulate dispersion of atmospheric released tritium gas in the environmental media and to evaluate doses. TRIDOSE

    Murata, Mikio [Nuclear Engineering Co., Ltd., Hitachi, Ibaraki (Japan); Noguchi, Hiroshi; Yokoyama, Sumi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-11-01

    A computer code (TRIDOSE) was developed to assess the environmental impact of atmospheric released tritium gas (T{sub 2}) from nuclear fusion related facilities. The TRIDOSE simulates dispersion of T{sub 2} and resultant HTO in the atmosphere, land, plant, water and foods in the environment, and evaluates contamination concentrations in the media and exposure doses. A part of the mathematical models in TRIDOSE were verified by comparison of the calculation with the results of the short range (400 m) dispersion experiment of HT gas performed in Canada postulating a short-time (30 minutes) accidental release. (author)

  18. Modelling of the tritium dispersion from postulated accidental release of nuclear power plants; Modelagem da dispersao de tritio a partir de liberacoes acidentais postuladas de centrais nucleares

    Soares, Abner Duarte; Simoes Filho, Francisco Fernando Lamego; Cunha, Tatiana Santos da [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Aguiar, Andre Silva de; Lapa, Celso Marcelo Franklin, E-mail: asoares@cnen.gov.b, E-mail: flamego@ien.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models to simulation of tritium dispersion caused by an accident in a Candu reactor located in the ongoing Angra 3 site. The Candu reactor is one that uses heavy water (D{sub 2}O) as moderator and coolant of the core. It was postulated, then, the LOCA accident (without fusion), where was lost 66 m3 of soda almost instantaneously. This inventory contained 35 P Bq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios (plant stopped or operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1 MBq/m{sup 3}) during the first 14 days after the accident. (author)

  19. Acute Immobilization Stress Modulate GABA Release from Rat Olfactory Bulb: Involvement of Endocannabinoids—Cannabinoids and Acute Stress Modulate GABA Release

    Alejandra Delgado; Erica H. Jaffé

    2011-01-01

    We studied the effects of cannabinoids and acute immobilization stress on the regulation of GABA release in the olfactory bulb. Glutamate-stimulated 3H-GABA release was measured in superfused slices. We report that cannabinoids as WIN55, 212-2, methanandamide, and 2-arachidonoylglycerol were able to inhibit glutamate- and KCl-stimulated 3H-GABA release. This effect was blocked by the CB1 antagonist AM281. On the other hand, acute stress was able per se to increase endocannabinoid activity. Th...

  20. Tritium release of Li4SiO4, Li2O and beryllium and chemical compatibility of beryllium with Li4SiO4, Li2O and steel (SIBELIUS irradiation)

    The objective of the SIBELIUS irradiation, a joint EC-US project performed at CEN Grenoble, was to investigate the oxidation kinetics of beryllium in contact with ceramic and the nature and extent of beryllium in contact with ceramic and the nature and extent of beryllium interaction with (316 L and 1.4914) steel in a neutron environment. In this work post irradiation examinations of SIBELIUS specimens performed at KfK are described. Tritium release of Li4SiO4, Li2O and beryllium was studied by out-of-pile annealing and chemical compatibility of beryllium with Li4SiO4, Li2O and steel by microscopic examinations. Tritium release of the ceramics was found to be consistent with SIBELIUS inpile observations and previous tests. Release of tritium generated in beryllium was found to be very slow, in accordance with previous work. For beryllium which was in contact with ceramic during irradiation, a second type of tritium, caused by injection of 2.7 MeV tritons generated in the ceramic, is observed. Release of injected tritium is faster than that of generated. Evidence for injected tritium in beryllium was also found in the microscopic studies. The observed minor chemical reactions of beryllium with steel and probably also those with breeder materials under neutron irradiation are consistent with the results of laboratory annealing tests. (orig.)

  1. A dynamic model for assessing radiological consequences of tritium routinely released in rivers. Application to the Loire River

    A dynamic model for assessing the transfer of tritium in a food chain was applied to the Loire River, where 14 nuclear power plants situated on five different sites operate. The model considers several potential exposure pathways in the aquatic and terrestrial ecosystems: transfer of tritium through the aquatic food chain (especially fish); use of river water for agricultural purposes (irrigation) and transfer of radionuclides through the terrestrial food chain (vegetables, meat, milk); subsequent internal exposure of humans due to ingestion of contaminated foodstuffs. For biological environmental compartments, the transfer of tritium to organic matter (i.e. OBT) was simulated. For each of the parameters introduced in this model, a probability density function, allowing further uncertainty and sensitivity analyses, was proposed. Uncertainty/sensitivity analyses were performed to determine a confidence interval for the mean annual dose to critical groups and to identify the parameters responsible for the uncertainty and subsequent research priorities

  2. Tritium behaviors in plants

    The tritium intake of plants was briefly reviewed in this report. The major chemical forms of tritium released from nuclear facilities are HTO and HT and in the natural environment, tritium is also found in various OBT such as CH3T. The exposure dose to HTO by inhalation exposure in humans was evaluated by ICRP to be 104 fold higher than HT and 102 fold than CH3T. Whereas for the organic compound binding form, it was evaluated to be 2.3 times higher than that of HTO. To study the tritium transition into plants, especially edible parts such as vegetables and fruits and the transition process were thought important and many studies including theoretical analysis have been done mainly regarding HTO, HT and CH3T. The transition of HT tritium into plants was negligible. However, it was reported that the released HT was converted to HTO by microorganisms in surface soil and incorporated into plants. But, the HTO concentration of the leaves in potted plants always lower than that of water in the soil of the pot, suggesting that tritium was not concentrated by the plant. However, there are few studies on tritium transition via photosynthesis into plant tissues. (M.N.)

  3. Tritium in the aquatic environment

    Tritium is of environmental importance because it is released from nuclear facilities in relatively large quantities and because it has a half life of 12.26 y. Most of the tritium released into the atmosphere eventually reaches the aqueous environment, where it is rapidly taken up by aquatic organisms. This paper reviews the current literature on tritium in the aquatic environment. Conclusions from the review, which covered studies of algae, aquatic macrophytes, invertebrates, fish, and the food chain, were that aquatic organisms incorporate tritium into their tissue-free water very rapidly and reach concentrations near those of the external medium. The rate at which tritium from tritiated water is incorporated into the organic matter of cells is slower than the rate of its incorporation into the tissue-free water. If organisms consume tritiated food, incorporation of tritium into the organic matter is faster, and a higher tritium concentration is reached than when the organisms are exposed to only tritiated water alone. Incorporation of tritium bound to molecules into the organic matter depends on the chemical form of the ''carrier'' molecule. No evidence was found that biomagnification of tritium occurs at higher trophic levels. Radiation doses from tritium releases to large populations of humans will most likely come from the consumption of contaminated water rather than contaminated aquatic food products

  4. Tritium in the aquatic environment

    Blaylock, B.G.; Hoffman, F.O.; Frank, M.L.

    1986-02-01

    Tritium is of environmental importance because it is released from nuclear facilities in relatively large quantities and because it has a half life of 12.26 y. Most of the tritium released into the atmosphere eventually reaches the aqueous environment, where it is rapidly taken up by aquatic organisms. This paper reviews the current literature on tritium in the aquatic environment. Conclusions from the review, which covered studies of algae, aquatic macrophytes, invertebrates, fish, and the food chain, were that aquatic organisms incorporate tritium into their tissue-free water very rapidly and reach concentrations near those of the external medium. The rate at which tritium from tritiated water is incorporated into the organic matter of cells is slower than the rate of its incorporation into the tissue-free water. If organisms consume tritiated food, incorporation of tritium into the organic matter is faster, and a higher tritium concentration is reached than when the organisms are exposed to only tritiated water alone. Incorporation of tritium bound to molecules into the organic matter depends on the chemical form of the ''carrier'' molecule. No evidence was found that biomagnification of tritium occurs at higher trophic levels. Radiation doses from tritium releases to large populations of humans will most likely come from the consumption of contaminated water rather than contaminated aquatic food products.

  5. Tritium accounting during the first tritium experiment at JET

    This paper summarises the measuring procedures and the results of the tritium accounting during the first tritium experiment at JET, carried out in November 1991. The measurement of the amount of tritium injected into the Torus and of the quantity recovered from the Torus and the Neutral Injector Boxes is described and the accuracy of the data assessed. The new Gas Collection System used during the experiment is briefly described. The tritium recovery data taken in the months following the experiment are reviewed, with special attention to the first three weeks after the experiment. The total amount of tritium collected in the Gas Collection System is compared with the data of tritium release from the Torus and the Neutral Injection Boxes. The analysis of the data allows us to estimate the residual tritium inventory in the Injection System and in the Torus. (orig.)

  6. Modelling of tritium dispersion from postulated accidental release of nuclear power plants; Modelagem da dispersao de tritio a partir de liberacoes acidentais postuladas de centrais nucleares

    Soares, Abner Duarte

    2010-07-01

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models of hydrodynamics and transport for the simulation of tritium dispersion caused by an accident in a CANDU reactor located in the ongoing Angra 3 site. This exercise was accomplished with the aid of a code system (SisBAHIA) developed in the Rio de Janeiro Federal University (COPPE/UFRJ). The CANDU reactor is one that uses heavy water (D{sub 2}O) as moderator and coolant of the core. It was postulated, then, the LOCA (Loss of Coolant Accident) accident in the emergency cooling system of the nucleus (without fusion), where was lost 66 m{sup 3} of soda almost instantaneously. This inventory contained 35 PBq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios (plant stopped and operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1 MBq/m{sup 3} ) during the first 14 days after the accident. The main difference between the scenario without and with seawater recirculation (pumping and discharge) is based on the enhancement of dilution of the highest concentrations in the last one. This dilution enhancement resulting in decreasing concentrations was observed only during the first two weeks, when they ranged from 1x10{sup 9} to 5x10{sup 5} Bq/m{sup 3} close to the Itaorna beach spreading just to Sandri Island. After 180 days, the plume could not be detected anymore in the bay, because their activities would be lower than the minimum detectable value (< 11 kBq/m{sup 3}). (author)

  7. Validation of phenomena relative to tritium transport in irradiated closed capsules filled with Pb-17Li through modelling tritium release in LIBRETTO-2 experiment

    The present paper is centred in the modelling of LIBRETTO-2 experiment. An open-minded selection of possible release mechanisms followed by its evaluation impose to consider specific interface and irradiation phenomena. From the confidence on the final approaches, values for empirical parameters as alloy/cladding wetting factors, weld diffusivities or sticking factors are provided. In-pile correlations under HFR irradiation conditions are validated. An average value of 90 for the barrier permeation reduction factor is obtained for the PC-Al2O3 coating through the LIBRETTO-2 irradiation history. (orig.)

  8. Tritium oxidation and exchange: preliminary studies

    The radiological hazard resulting from an exposure to either tritium oxide or tritium gas is discussed and the factors contributing to the hazard are presented. From the discussion it appears that an exposure to tritium oxide vapor is 104 to 105 times more hazardous than exposure to tritium gas. Present and future sources of tritium are briefly considered and indicate that most of the tritium has been and is being released as tritium oxide. The likelihood of gaseous releases, however, is expected to increase in the future, calling to task the present general release assumption that 100% of all tritium released is as oxide. Accurate evaluation of the hazards from a gaseous release will require a knowledge of the conversion rate of tritium gas to tritium oxide. An experiment for determining the conversion rate of tritium gas to tritium oxide is presented along with some preliminary data. The conversion rates obtained for low initial concentrations (10-4 to 10-1 mCi/ml) indicate the conversion may proceed more rapidly than would be expected from an extrapolation of previous data taken at higher concentrations

  9. Tritium oxidation and exchange: preliminary studies

    Phillips, J. E.; Easterly, C. E.

    1978-05-01

    The radiological hazard resulting from an exposure to either tritium oxide or tritium gas is discussed and the factors contributing to the hazard are presented. From the discussion it appears that an exposure to tritium oxide vapor is 10/sup 4/ to 10/sup 5/ times more hazardous than exposure to tritium gas. Present and future sources of tritium are briefly considered and indicate that most of the tritium has been and is being released as tritium oxide. The likelihood of gaseous releases, however, is expected to increase in the future, calling to task the present general release assumption that 100% of all tritium released is as oxide. Accurate evaluation of the hazards from a gaseous release will require a knowledge of the conversion rate of tritium gas to tritium oxide. An experiment for determining the conversion rate of tritium gas to tritium oxide is presented along with some preliminary data. The conversion rates obtained for low initial concentrations (10/sup -4/ to 10/sup -1/ mCi/ml) indicate the conversion may proceed more rapidly than would be expected from an extrapolation of previous data taken at higher concentrations (10/sup -1/ to 10/sup 2/ mCi/ml).

  10. Tritium discharge during nuclear power plant operation in China

    This paper summarized tritium discharges from gaseous and liquid effluents of nuclear power plants in China prior to 2005. The differences of tritium discharges for various nuclear power plants were also evaluated. For all pressurized water reactors, tritium releases are lower than national authorized limits. For heavy water reactors, the average normalized tritium release is less than 10% of global average. The tritium releases are mainly from liquid pathways for pressurized water reactors, but gaseous and liquid tritium releases have are comparable for heavy water reactor. The gaseous release from the heavy water reactor is obviously higher than that from the pressurized water reactor. (authors)

  11. Oxides as barriers to tritium permeation in steam generators and tritium content in CTR coolants

    The primary release of tritium from a fusion reactor complex into the environment is via the steam generator system. Tritium in the coolant can permeate through the heat exchanger into the steam cycle, and is trapped in the steam as HTO. Subsequent recovery of tritium from the steam is impractical. The amount of tritium that permeates into the steam cycle will depend on the concentration of tritium in the coolant, or more significantly the amount of tritium that can be allowed in the coolant will depend on the rate of tritium permeation that can be tolerated

  12. Overview of tritium: characteristics, sources, and problems.

    Okada, S; Momoshima, N

    1993-12-01

    Tritium has certain characteristics that present unique challenges for dosimetry and health-risk assessment. For example, in the gas form, tritium can diffuse through almost any container, including those made of steel, aluminum, and plastics. In the oxide form, tritium can generally not be detected by commonly used survey instruments. In the environment, tritium can be taken up by all hydrogen-containing molecules, distributing widely on a global scale. Tritium can be incorporated into humans through respiration, ingestion, and diffusion through skin. Its harmful effects are observed only when it is incorporated into the body. Several sources contribute to the inventory of tritium in our environment. These are 1) cosmic ray interaction with atmospheric molecules; 2) nuclear reactions in the earth's crust; 3) nuclear testing in the atmosphere during the 1950s and 1960s; 4) continuous release of tritium from nuclear power plants and tritium production facilities under normal operation; 5) incidental releases from these facilities; and 6) consumer products. An important future source will be nuclear fusion facilities expected to be developed for the purpose of electricity generation. The principal health physics problems associated with tritium are 1) the determination of the parameters for risk estimation with further reduction of their uncertainties (e.g., relative biological effectiveness and dose-rate dependency); 2) risk estimation from complex exposures to tritium in gas form, tritium in oxide form, tritium surface contamination, and other tritium-contaminated forms, with or without other ionizing radiations and/or nonionizing radiations; 3) the dose contributions of elemental tritium in the lung and from its oxidized tritium in the gastrointestinal tract; 4) prevention of tritium (in oxide form) intake and enhancement of tritium (oxide form) excretion from the human body; 5) precise health effects information for low-level tritium exposure; and 6) public

  13. Tritium accounting for PHWR plants

    Tritium, the radioactive isotope of hydrogen, is produced as a byproduct of the nuclear reactions in the nuclear power plants. In a Pressurized Heavy Water Reactor (PHWR) tritium activity is produced in the Heat Transport and Moderator systems due to neutron activation of deuterium in heavy water used in these systems. Tritium activity build up occurs in some of the water systems in the PHWR plants through pick up from the plant atmosphere, inadvertent D2O ingress from other systems or transfer during processes. The tritium, produced by the neutron induced reactions in different systems in the reactor undergoes multiple processes such as escape through leaks, storage, transfer to external locations, decay, evaporation and diffusion and discharge though waste streams. Change of location of tritium inventory takes place during intentional transfer of heavy water, both reactor grade and downgraded, from one system to another. Tritium accounting is the application of accounting techniques to maintain knowledge of the tritium inventory present in different systems of a facility and to construct activity balances to detect any discrepancy in the physical inventories. It involves identification of all the tritium hold ups, transfers and storages as well as measurement of tritium inventories in various compartments, decay corrections, environmental release estimations and evaluation of activity generation during the accounting period. This paper describes a methodology for creating tritium inventory balance based on periodic physical inventory taking, tritium build up, decay and release estimations. Tritium accounting in the PHWR plants can prove to be an effective regulatory tool to monitor its loss as well as unaccounted release to the environment. (author)

  14. Li2TiO3 pebbles tritium release mechanism and kinetics by post-irradiation 'temperature programmed desorption' (TPD) spectroscopy

    The interaction of gaseous environments such as air (with moisture and CO2 impurities) and the purge gas mixture He+H2 (1000 vpm) (R-gas) with Li2TiO3 pebbles was examined by TPD/TPR methods. Surface ''cleaning'' processes were stated and the titanate reduction kinetics to Li2TiO3-x concerning free and grain boundary surfaces were determined by TPR spectra de-convolution-fitting analysis performed by assuming the reaction rate signal as the overlap of independent first-order steps. Formally near all the Ti4+ on both the grain and grain boundary surfaces around 800 and 950 K respectively was found reduced to Ti3+ (i.e x ∼1). Several specimens were shortly irradiated and examined for tritium release by previously annealing them for 1-2 hours at 473 K in flowing R-gas and then by heating at the rate β=5 K/min (TPD) method. Specimens with density lower than about 82% TD and nearly full open porosity showed a single broad peak (Tp=760±20 K) whose first-order deconvolution gave a main contribution of a desorption site-peak characterized by the kinetic rate constant k=9.5105(exp(-1.52.104/T)) [s-1]. Diffusion control was however found to be more compatible with experimental data for these porous materials. Tritium removal from pebbles with density higher than 90% TD was found to be limited by diffusion within the grain boundary interface. This transport stage was found to be affected by the above mentioned ''reduction'' steps, tritium trapped on grain boundaries resulting removable above 900 K where the pebbles reduction occurs. (orig.)

  15. Tritium in the aquatic environment

    Most of the tritium released from nuclear facilities into the atmosphere eventually reaches the aqueous environment where it is rapidly taken up by aquatic organisms. This paper reviews the current literature on tritium in the aquatic environment. Conclusions from the review, which covered algae, aquatic plants, invertebrates, fish, and food chain studies, were that aquatic organisms incorporate tritium into their tissue free water very rapidly and reach concentrations near that of the external medium. Incorporation of tritium from triated water into the organic matter of cells is at a slower rate than incorporation into the tissue free water. If organisms consume tritiated food, incorporation of tritium into the organic matter is faster and a higher tritium concentration is reached than when the organisms are exposed to only tritiated water. Incorporation of tritium bound to molecules into the organic matter depends on the chemical form of the 'carrier' molecule. No evidence was found that biomagnification of tritium occurs at higher tropic levels. Radiation doses to large populations of humans from tritium releases will most likely be from the consumption of contaminated water rather than contaminated aquatic food products. (author)

  16. Tritium. Today's and tomorrow's developments

    Radioactive hydrogen isotope, tritium is one of the radionuclides which is the most released in the environment during the normal operation of nuclear facilities. The increase of nuclear activities and the development of future generations of reactors, like the EPR and ITER, would lead to a significant increase of tritium effluents in the atmosphere and in the natural waters, thus raising many worries and questions. Aware about the importance of this question, the national association of local information commissions (ANCLI) wished to make a status of the existing knowledge concerning tritium and organized in 2008 a colloquium at Orsay (France) with an inquiring approach. The scientific committee of the ANCLI, renowned for its expertise skills, mobilized several nuclear specialists to carry out this thought. This book represents a comprehensive synthesis of today's knowledge about tritium, about its management and about its impact on the environment and on human health. Based on recent scientific data and on precise examples, it treats of the overall questions raised by this radionuclide: 1 - tritium properties and different sources (natural and anthropic), 2 - the problem of tritiated wastes management; 3 - the bio-availability and bio-kinetics of the different tritium species; 4 - the tritium labelling of environments; 5 - tritium measurement and modeling of its environmental circulation; 6 - tritium radio-toxicity and its biological and health impacts; 7 - the different French and/or international regulations concerning tritium. (J.S.)

  17. Transfer of Tritium in the Environment after Accidental Releases from Nuclear Facilities. Report of Working Group 7 Tritium Accidents of EMRAS II Topical Heading Approaches for Assessing Emergency Situations. Environmental Modelling for Radiation Safety (Emras II) Programme

    Environmental assessment models are used for evaluating the radiological impact of actual and potential releases of radionuclides to the environment. They are essential tools for use in the regulatory control of routine discharges to the environment and also in planning measures to be taken in the event of accidental releases. They are also used for predicting the impact of releases which may occur far into the future, for example, from underground radioactive waste repositories. It is important to verify, to the extent possible, the reliability of the predictions of such models by a comparison with measured values in the environment or with predictions of other models. The IAEA has been organizing programmes of international model testing since the 1980s. These programmes have contributed to a general improvement in models, in the transfer of data and in the capabilities of modellers in Member States. IAEA publications on this subject over the past three decades demonstrate the comprehensive nature of the programmes and record the associated advances which have been made. From 2009 to 2011, the IAEA organized a programme entitled Environmental Modelling for RAdiation Safety (EMRAS II), which concentrated on the improvement of environmental transfer models and the development of reference approaches to estimate the radiological impacts on humans, as well as on flora and fauna, arising from radionuclides in the environment. Different aspects were addressed by nine working groups covering three themes: reference approaches for human dose assessment, reference approaches for biota dose assessment and approaches for assessing emergency situations. This publication describes the work of the Tritium Accidents Working Group

  18. Tritium permeation, contamination and decontamination

    As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated three research programs on the tritium permeation, contamination and decontamination have been conducted by the CO2 team. The results are summarized as follows: (1) Study for the development of the tritium permeation barrier was carried out. A ZrO2 film with a magnesium phosphate layer sintered on a SUS 430 steel plate showed excellent reduction in the hydrogen permeation. (2) The non-destructive method using an imaging plate was proposed to monitor tritium release from contaminated materials. The method was applied to SUS 316 steel and revealed that the tritium release from SUS 316 steel was diffusion-limited. (3) As for contamination-protection and decontamination techniques, improvement in the decontamination rate from SUS 316 steel was obtained by providing CrO2 coating. (J.P.N.)

  19. Tritium concentrations in tree ring cellulose

    Measurements of tritium (tissue bound tritium; TBT) concentration in tree rings are presented and discussed. Such measurement is expected to provide a useful means of estimating the tritium level in the environment in the past. The concentration of tritium bound in the tissue (TBT) in a tree ring considered to reflect the environmental tritium level in the area at the time of the formation of the ring, while the concentration of tritium in the free water in the tissue represents the current environmental tritium level. First, tritium concentration in tree ring cellulose sampled from a cedar tree grown in a typical environment in Fukuoka Prefecture is compared with the tritium concentration in precipitation in Tokyo. Results show that the year-to-year variations in the tritium concentration in the tree rings agree well with those in precipitation. The maximum concentration, which occurred in 1963, is attibuted to atmospheric nuclear testing which was performed frequently during the 1961 - 1963 period. Measurement is also made of the tritium concentration in tree ring cellulose sampled from a pine tree grown near the Isotope Center of Kyushu University (Fukuoka). Results indicate that the background level is higher probably due to the release of tritium from the facilities around the pine tree. Thus, measurement of tritium in tree ring cellulose clearly shows the year-to-year variation in the tritium concentration in the atmosphere. (N.K.)

  20. Histamine is not released in acute thermal injury in human skin in vivo: a microdialysis study

    Petersen, Lars J; Pedersen, Juri L; Skov, Per S;

    2009-01-01

    BACKGROUND: Animal models have shown histamine to be released from the skin during the acute phase of a burn injury. The role of histamine during the early phase of thermal injuries in humans remains unclear. PURPOSE: The objectives of this trial were to study histamine release in human skin during...... the acute phase of a standardized thermal injury in healthy volunteers. METHODS: Histamine concentrations in human skin were measured by skin microdialysis technique. Microdialysis fibers were inserted into the dermis in the lower leg in male healthy volunteers. A standardized superficial thermal...... (baseline 11.6 +/- 1.8 nM vs. post-burn values of 14.8 +/- 1.8 nM, n = 8). CONCLUSIONS: Histamine is not released in human skin during the acute phase of a thermal injury....

  1. Tritium monitor and collection system

    Baker, J.D.; Wickham, K.L.; Ely, W.E.; Tuggle, D.G.; Meikrantz, D.H.; Grafwaller, E.G.; Maltrud, H.R.; Bourne, G.L.

    1991-03-26

    This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber. Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next online getter is valved on. Simultaneously, the first getter is heated and a pure helium purge is employed to carry the tritium from the getter. The tritium loaded gas stream is then routed through an ion chamber which measures the tritium activity. The ion chamber effluent passes through a collection getter that readsorbs the tritium and is removable from the system once it is loaded and is then replaced with a clean getter. Prior to removal of the collection getter, the system switches to a parallel collection getter. The effluent from the collection getter passes through a guard getter to remove traces of tritium prior to exiting the system. The tritium loaded collection getter, once removed, is analyzed by liquid scintillation techniques. The entire sequence is under computer control except for the removal and analysis of the collection getter.

  2. Tritium monitor and collection system

    Bourne, Gary L.; Meikrantz, David H.; Ely, Walter E.; Tuggle, Dale G.; Grafwallner, Ervin G.; Wickham, Keith L.; Maltrud, Herman R.; Baker, John D.

    1992-01-01

    This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber. Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next on-line getter is valved on. Simultaneously, the first getter is heated and a pure helium purge is employed to carry the tritium from the getter. The tritium loaded gas stream is then routed through an ion chamber which measures the tritium activity. The ion chamber effluent passes through a collection getter that readsorbs the tritium and is removable from the system once it is loaded and is then replaced with a clean getter. Prior to removal of the collection getter, the system switches to a parallel collection getter. The effluent from the collection getter passes through a guard getter to remove traces of tritium prior to exiting the system. The tritium loaded collection getter, once removed, is analyzed by liquid scintillation techniques. The entire sequence is under computer control except for the removal and analysis of the collection getter.

  3. Historical Doses from Tritiated Water and Tritiated Hydrogen Gas Relesed to the Atmosphere from Lawrence Livermore National Laboratory (LLNL) Part 1. Description of Tritium Dose Model (DCART) for Chronic Releases from LLNL

    Peterson, S

    2004-06-30

    DCART (Doses from Chronic Atmospheric Releases of Tritium) is a spreadsheet model developed at Lawrence Livermore National Laboratory (LLNL) that calculates doses from inhalation of tritiated hydrogen gas (HT), inhalation and skin absorption of tritiated water (HTO), and ingestion of HTO and organically bound tritium (OBT) to adult, child (age 10), and infant (age 6 months to 1 year) from routine atmospheric releases of HT and HTO. DCART is a deterministic model that, when coupled to the risk assessment software Crystal Ball{reg_sign}, predicts doses with a 95th percentile confidence interval. The equations used by DCART are described and all distributions on parameter values are presented. DCART has been tested against the results of other models and several sets of observations in the Tritium Working Group of the International Atomic Energy Agency's Biosphere Modeling and Assessment Programme. The version of DCART described here has been modified to include parameter values and distributions specific to conditions at LLNL. In future work, DCART will be used to reconstruct dose to the hypothetical maximally exposed individual from annual routine releases of HTO and HT from all LLNL facilities and from the Sandia National Laboratory's Tritium Research Laboratory over the last fifty years.

  4. Portable tritium-in-air monitor (TIAM) for monitoring tritium activity during shutdown in PHWR

    This paper discusses an overview of Portable Tritium in Air Monitor and usefulness of the system for health safety in Pressured Heavy Water Reactor based Nuclear Power Plants during Shutdown activity. Tritium in Air Monitor is meant for detection of Tritium activity in air in different accessible area and shutdown area inside the Reactor Building and also in atmospheric release. (author)

  5. Tritium in the environment. Knowledge synthesis

    This report first presents the nuclear and physical-chemical properties of tritium and addresses the notions of bioaccumulation, bio-magnification and remanence. It describes and comments the natural and anthropic origins of tritium (natural production, quantities released in the environment in France by nuclear tests, nuclear plants, nuclear fuel processing plants, research centres). It describes how tritium is measured as a free element (sampling, liquid scintillation, proportional counting, enrichment method) or linked to organic matter (combustion, oxidation, helium-3-based measurement). It discusses tritium concentrations noticed in different parts of the environment (soils, continental waters, sea). It describes how tritium is transferred to ecosystems (transfer of atmospheric tritium to ground ecosystems, and to soft water ecosystems). It discusses existing models which describe the behaviour of tritium in ecosystems. It finally describes and comments toxic effects of tritium on living ground and aquatic organisms

  6. A compact, low cost, tritium removal plant for CANDU-6 reactors

    Tritium concentrations in CANDU-6 reactors are currently around 40 Ci/kg in moderator systems and around 1.5 Ci/kg in primary heat transport (PHT) systems. It is expected that tritium concentrations in moderator systems will continue to rise and will reach about 80 Ci/kg at maturity. A more detailed description of the increase in tritium concentrations in the moderator and PHT systems of CANDU-6 reactors is given in the next section of this paper. While moderator systems currently contribute more than 50% to tritium emissions, the impact of acute releases of moderator water is more severe at higher tritium concentrations. This impact can be substantially reduced by the addition of an isotope separation system for lowering the tritium level in the moderator system. In addition, lower tritium levels in CANDU systems will inevitably result in reduced occupational exposures, or will provide economic benefits due to ease of maintenance because less protective measures are required and maintenance activities can be more efficient

  7. Comparison of the temporal release pattern of copeptin with conventional biomarkers in acute myocardial infarction

    Y.L. Gu (Youlan); A.A. Voors (Adriaan); F. Zijlstra (Felix); H.L. Hillege (Hans); J. Struck (Joachim); S. Masson (Serge); T. Vago (Tarcisio); S.D. Anker (Stefan); A.F.M. van den Heuvel (Ad); D.J. van Veldhuisen (Dirk); B.J.G.L. de Smet (Bart)

    2011-01-01

    textabstractBackground Early detection of acute myocardial infarction (AMI) using cardiac biomarkers of myocardial necrosis remains limited since these biomarkers do not rise within the first hours from onset of AMI. We aimed to compare the temporal release pattern of the C-terminal portion of prova

  8. Comparison of the temporal release pattern of copeptin with conventional biomarkers in acute myocardial infarction

    Gu, Youlan L.; Voors, Adriaan A.; Zijlstra, Felix; Hillege, Hans L.; Struck, Joachim; Masson, Serge; Vago, Tarcisio; Anker, Stefan D.; van den Heuvel, Ad F. M.; van Veldhuisen, Dirk J.; de Smet, Bart J. G. L.

    2011-01-01

    Background Early detection of acute myocardial infarction (AMI) using cardiac biomarkers of myocardial necrosis remains limited since these biomarkers do not rise within the first hours from onset of AMI. We aimed to compare the temporal release pattern of the C-terminal portion of provasopressin (c

  9. TSOAK-M1: a computer code to determine tritium reaction/adsorption/release parameters from experimental results of air-detritiation tests

    Land, R.H.; Maroni, V.A.; Minkoff, M.

    1979-01-01

    A computer code has been developed which permits the determination of tritium reaction (T/sub 2/ to HTO)/adsorption/release and instrument correction parameters from enclosure (building) - detritiation test data. The code is based on a simplified model which treats each parameter as a normalized time-independent constant throughout the data-unfolding steps. Because of the complicated four-dimensional mathematical surface generated by the resulting differential equation system, occasional local-minima effects are observed, but these effects can be overcome in most instances by selecting a series of trial guesses for the initial parameter values and observing the reproducibility of final parameter values for cases where the best overall fit to experimental data is achieved. The code was then used to analyze existing small-cubicle test data with good success, and the resulting normalized parameters were employed to evaluate hypothetical reactor-building detritiation scenarios. It was concluded from the latter evaluation that the complications associated with moisture formation, adsorption, and release, particularly in terms of extended cleanup times, may not be as great as was previously thought. It is recommended that the validity of the TSOAK-M1 model be tested using data from detritiation tests conducted on large experimental enclosures (5 to 10 cm/sup 3/) and, if possible, actual facility buildings.

  10. TSOAK-M1: a computer code to determine tritium reaction/adsorption/release parameters from experimental results of air-detritiation tests

    A computer code has been developed which permits the determination of tritium reaction (T2 to HTO)/adsorption/release and instrument correction parameters from enclosure (building) - detritiation test data. The code is based on a simplified model which treats each parameter as a normalized time-independent constant throughout the data-unfolding steps. Because of the complicated four-dimensional mathematical surface generated by the resulting differential equation system, occasional local-minima effects are observed, but these effects can be overcome in most instances by selecting a series of trial guesses for the initial parameter values and observing the reproducibility of final parameter values for cases where the best overall fit to experimental data is achieved. The code was then used to analyze existing small-cubicle test data with good success, and the resulting normalized parameters were employed to evaluate hypothetical reactor-building detritiation scenarios. It was concluded from the latter evaluation that the complications associated with moisture formation, adsorption, and release, particularly in terms of extended cleanup times, may not be as great as was previously thought. It is recommended that the validity of the TSOAK-M1 model be tested using data from detritiation tests conducted on large experimental enclosures (5 to 10 cm3) and, if possible, actual facility buildings

  11. Estimated radiological doses to the maximumly exposed individual and downstream populations from releases of tritium, strontium-90, ruthenium-106, and cesium-137 from White Oak Dam

    Concentrations of tritium, 90Sr, 106Ru, and 137Cs in the Clinch River for 1978 were estimated by using the known 1978 releases of these nuclides from the White Oak Dam and diluting them by the integrated annual flow rate of the Clinch River. Estimates of 50-year dose commitment to a maximumly exposed individual were calculated for both aquatic and terestrial pathways of exposure. The maximumly exposed individual was assumed to reside at the mouth of White Oak Creek where it enters the Clinch River and obtain all foodstuffs and drinking water at that location. The estimated total-body dose from all pathways to the maximumly exposed individual as a result of 1978 releases was less than 1% of the dose expected from natural background. Using appropriate concentrations of to subject radionuclides diluted downstream, the doses to populations residing at Harriman, Kingston, Rockwood, Spring City, Soddy-Daisy, and Chattanooga were calculated for aquatic exposure pathways. The total-body dose estimated for aquatic pathways for the six cities was about 0.0002 times the expected dose from natural background. For the pathways considered in this report, the nuclide which contributed the largest fraction of dose was 90Sr. The largest dose delivered by 90Sr was to the bone of the subject individual or community

  12. Surface desorption and bulk diffusion models of tritium release from Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} pebbles

    Avila, R.E., E-mail: ravila@cchen.c [Departamento de Materiales Nucleares, Comision Chilena de Energia Nuclear, Cas. 188-D, Santiago (Chile); Pena, L.A.; Jimenez, J.C. [Departamento de Produccion y Servicios, Comision Chilena de Energia Nuclear, Cas. 188-D, Santiago (Chile)

    2010-10-30

    The release of tritium from Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} pebbles, in batch experiments, is studied by means of temperature programmed desorption. Data reduction focuses on the analysis of the non-oxidized and oxidized tritium components in terms of release limited by diffusion from the bulk of ceramic grains, or by first or second order surface desorption. By analytical and numerical methods the in-furnace tritium release is deconvoluted from the ionization chamber transfer functions, for which a semi-empirical form is established. The release from Li{sub 2}TiO{sub 3} follows second order desorption kinetics, requiring a temperature for a residence time of 1 day (T{sub 1dRes}) of 620 K, and 603 K, of the non-oxidized, and the oxidized components, respectively. The release from Li{sub 2}ZrO{sub 3} appears as limited by either diffusion from the bulk of the ceramic grains, or by first order surface desorption, the first possibility being the more probable. The respective values of T{sub 1dRes} for the non-oxidized component are 661 K, according to the first order surface desorption model, and 735 K within the bulk diffusion limited model.

  13. Tritium practices past and present

    History of the production and use of tritium, as well as handling techniques, are reviewed. Handling techniques first used at Lawrence Livermore National Laboratory made use of glass vacuum systems and relatively crude ion chambers for monitoring airborne activity. The first use of inert atmosphere glove boxes demonstrated that uptake through the skin could be a serious personnel exposure problem. Growing environmental concerns in the early 1970's resulted in the implementation by the Atomic Energy Commission of a new criteria to limit atmospheric tritium releases to levels as low as practicable. An important result of the new criteria was the development of containment and recovery systems to capture tritium rather than vent it to the atmosphere. The Sandia National Laboratories, Livermore, Tritium Research Laboratory containment and decontamination systems are presented as a typical example of this technology. The application of computers to control systems is expected to provide the greatest potential for change in future tritium handling practices

  14. Transport of tritium in SS316 at moderate temperatures

    From tritium release experiments with stainless steel 316 carried out at several temperatures and tritium depth profiles of tritium-depleted specimen information on the transport of tritium by two diverse techniques was obtained. The results could be interpreted by a one dimensional diffusion model. The activation energy for the diffusion of tritium through stainless steel was found to be 61.3 kJ/mol. (authors)

  15. Tritium Elimination System Using Tritium Gas Oxidizing Bacteria

    In order to eliminate atmospheric tritium gas (HT) released from tritium handling apparatus, we proposed to use the HT oxidizing ability (hydrogenase enzyme) of bacterial strains isolated from surface soils instead of a high temperature precious metal catalyst. Among the isolated strains with high HT oxidation activity, several strains were selected to develop a tritium elimination (detritiation) system. Bioreactors were made of bacterial cells grown on agar medium on a cartridge filter and stored in a refrigerator until use. The detritiation ability of these bioreactors at room temperature was investigated during the intentional HT release experiments carried out in the Cassion Assembly for Tritium Safety Study (CATS) in TPL/JAERI. When HT contaminated air from the CATS was introduced into the biological detritiation system, in which three bioreactors were connected in series, 86% of HT in air was removed as tritiated water in these bioreactors at a flow rate of 100 cm3/min for 2 hours

  16. Recovery of tritium from a liquid lithium blanket

    Talbot, J.B.

    1981-01-01

    The sorption of tritium on yttrium from liquid lithium and the subsequent release of tritium from yttrium by thermal regeneration of the metal sorbent were investigated to study such a tritium-recovery process for a fusion reactor blanket of liquid lithium. Recent static sorption experiments have shown the effects of lithium temperature and possible impurities on the sorption of tritium. Diffusivity data, obtained from previous tritium recovery experiments, were evaluated to show the importance of the yttrium surface condition in controlling the release of tritium.

  17. HADOC: a computer code for calculation of external and inhalation doses from acute radionuclide releases

    The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested

  18. Tritium calorimetry

    Complete text of publication follows. Future deuterium-tritium fusion experiments (like ITER) will use large amount of tritium. Therefore, it is very important to develop better tritium accountancy methods. Tritium calorimetry is used to measure the heat produced by the beta-decay of tritium. If we consider that all the decay energy is converted into thermal heat, we can calculate the tritium activity and mass from calorimetric measurements. The advantages of calorimetry are that it measures absolute activity, and the physical or chemical composition of the sample is not relevant. For example, tritiated structural components can only be measured in a non-destructive way with calorimeters. Disadvantages are: long measurement time for large sample volumes, and offline sampling. The accepted conversion factor is 0.324W/g ± 0.3%. I have started participation from ATOMKI in an EFDA-GOT program, called TRI-TOFFY (TRITium fOr Fusion Fuel cYcle), in 2010. I have spent 8 months at Tritium Laboratory Karlsruhe (TLK), Germany in 2011, and 9 months in 2012. TLK is a semi-industrial scale facility for processing tritium, the radioactive hydrogen isotope. The main tasks of TLK are fusion research (ITER) and neutrino physics (KATRIN), but also EU projects. The present site inventory is ∼ 25 g T2 (8914 TBq). There are four calorimeters are used for tritium analytics at TLK. My main work was to carry out upgrade on these devices, to deploy new modern control and data acquisition (DAQ) software, and to partly change their hardware. I worked on three calorimeters at the laboratory. The ANTECH-351 is a commercial 20 years old calorimeter. It is a power compensation type isothermal calorimeter. Useful sample volume is 1.2 dm3. This is not a sensitive device (power range is 1 mW - 5 W), mainly used for tritium shipment (from Canadian CANDU reactors) validation, but can measure tritium samples very fast: less than 8 hours. The IGC-V0.5 is a custom made heat flow calorimeter, using a

  19. Tritium contamination and decontamination

    Establishment of tritium safe handling technology is required with the development of fusion reactor research. Tritium is contained by multiple-barriers containment due to the difficulty in perfect containment of hydrogen isotopes. Tritium contamination of materials and subsequent desorption are one of the critical issues in tritium containment. And the development of tritium decontamination technology is also a critical issue in tritium safe handling. The status of tritium contamination study and tritium decontamination technology are reviewed. (author)

  20. NDT and inspection of tritium removal facility

    CANDU heavy water reactors produce tritium in the moderator and coolant circuits through neutron absorption by the deuterium atoms in heavy water. The concentration of tritium, in the form of DTO molecules builds up slowly with time of reactor operation. A typical yearly production rate of tritium is 2400 curie for each megawatt of electricity produced and as a consequence, a 600 megawatt Candu reactor produces 1.4 million curie of tritium per year. Tritium decays to 3He, a non radioactive species, and has a half life of approximately 12 years. Both Ontario Hydro and AECL are constructing plants to remove tritium from heavy water to maintain the tritium concentration below the equilibrium value. This will result in lower radiation doses to operating personnel and reduce the level of radiation in any releases of heavy water to the environment

  1. Acute stress increases interstitial fluid amyloid-β via corticotropin-releasing factor and neuronal activity

    Kang, Jae-Eun; Cirrito, John R.; Dong, Hongxin; John G. Csernansky; Holtzman, David M.

    2007-01-01

    Aggregation of the amyloid-β (Aβ) peptide in the extracellular space of the brain is critical in the pathogenesis of Alzheimer's disease. Aβ is produced by neurons and released into the brain interstitial fluid (ISF), a process regulated by synaptic activity. To determine whether behavioral stressors can regulate ISF Aβ levels, we assessed the effects of chronic and acute stress paradigms in amyloid precursor protein transgenic mice. Isolation stress over 3 months increased Aβ levels by 84%. ...

  2. Safety and Efficacy of Paliperidone Extended-Release in Acute and Maintenance Treatment of Schizophrenia

    Edoardo Spina; Rosalia Crupi

    2011-01-01

    Paliperidone, the major active metabolite of risperidone, is a second-generation antipsychotic that has been developed as an extended-release (ER) tablet formulation that minimizes peak-trough fluctuations in plasma concentrations, allowing once-daily administration and constant drug delivery. Paliperidone ER has demonstrated efficacy in the reduction of acute schizophrenia symptoms in 6-week, placebo-controlled, double-blind trials and clinical benefits were maintained in the longer-term acc...

  3. Tritium processing in JT-60U

    Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7 x 1019 D-D fusion neutrons were produced by D (d, p) T reactions in plasma, which are expected to produce ∼31 GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium Distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ∼ 7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm3 for radiation work permit requirements. (author)

  4. The toxicity of tritium

    Among radionuclides of importance in atomic energy, 3H has relatively low toxicity. There is concern, however, because very large amounts are involved in nuclear fission and fusion, impressive quantities are released to the environment and tritium in its preferred state, water, has free access to living cells and organisms. The main health and environmental worry is the possibility that significant biological effects may follow from protracted exposure to low concentrations in water. To examine this possible hazard and measure toxicity at low tritium concentrations, chronic exposure studies were done on mice and monkeys. During vulnerable developmental periods animals were exposed to 3HOH and mice were exposed also to 60Co gamma irradiation and energy-related chemical agents. The biological endpoint measured was the irreversible loss of female germ cells. Effects from tritium were observed at surprisingly low concentrations where 3H was found more damaging than previously thought. Comparisons between tritium and gamma radiation showed the relative biological effectiveness (RBE) to be greater than 1 and to reach approximately 3 at very low exposures. For perspective, other comparisons were made: between radiation and chemical agents, which revealed parallels in action on germ cells; and between pre- and postnatal exposure, which warn of possible special hazard to the foetus from both classes of energy-related byproducts. (author)

  5. Effluent Treatment Facility tritium emissions monitoring

    An Environmental Protection Agency (EPA) approved sampling and analysis protocol was developed and executed to verify atmospheric emissions compliance for the new Savannah River Site (SRS) F/H area Effluent Treatment Facility. Sampling equipment was fabricated, installed, and tested at stack monitoring points for filtrable particulate radionuclides, radioactive iodine, and tritium. The only detectable anthropogenic radionuclides released from Effluent Treatment Facility stacks during monitoring were iodine-129 and tritium oxide. This paper only examines the collection and analysis of tritium oxide

  6. Impact of blanket tritium against the tritium plant of fusion reactor

    The breeder blanket and the blanket tritium recovery system are tested using test blanket modules during ITER campaign. And then, these are integrated with the tritium plant for the first time at a prototype reactor after ITER. In this work, impact to the tritium plant by integration of the solid breeder blanket was discussed. The method of tritium extraction from the blanket and the choice of the process for breeder blanket interface should be discussed not only from the viewpoint of tritium release but also from the viewpoint of the load of processing. (author)

  7. Historical Doses from Tritiated Water and Tritiated Hydrogen Gas Released to the Atmosphere from Lawrence Livermore National Laboratory (LLNL). Part 5. Accidental Releases

    Peterson, S

    2007-08-15

    Over the course of fifty-three years, LLNL had six acute releases of tritiated hydrogen gas (HT) and one acute release of tritiated water vapor (HTO) that were too large relative to the annual releases to be included as part of the annual releases from normal operations detailed in Parts 3 and 4 of the Tritium Dose Reconstruction (TDR). Sandia National Laboratories/California (SNL/CA) had one such release of HT and one of HTO. Doses to the maximally exposed individual (MEI) for these accidents have been modeled using an equation derived from the time-dependent tritium model, UFOTRI, and parameter values based on expert judgment. All of these acute releases are described in this report. Doses that could not have been exceeded from the large HT releases of 1965 and 1970 were calculated to be 43 {micro}Sv (4.3 mrem) and 120 {micro}Sv (12 mrem) to an adult, respectively. Two published sets of dose predictions for the accidental HT release in 1970 are compared with the dose predictions of this TDR. The highest predicted dose was for an acute release of HTO in 1954. For this release, the dose that could not have been exceeded was estimated to have been 2 mSv (200 mrem), although, because of the high uncertainty about the predictions, the likely dose may have been as low as 360 {micro}Sv (36 mrem) or less. The estimated maximum exposures from the accidental releases were such that no adverse health effects would be expected. Appendix A lists all accidents and large routine puff releases that have occurred at LLNL and SNL/CA between 1953 and 2005. Appendix B describes the processes unique to tritium that must be modeled after an acute release, some of the time-dependent tritium models being used today, and the results of tests of these models.

  8. Development of a tritium recovery system from CANDU tritium removal facility

    The main purpose of the Tritium Recovery System (TRS) is to reduce to a maximum possible extent the release of tritium from the facility following a tritium release in confinement boundaries and also to have provisions to recover both elemental and vapors tritium from the purging gases during maintenance and components replacement from various systems processing tritium. This work/paper proposes a configuration of Tritium Recovery System wherein elemental tritium and water vapors are recovered in a separated, parallel manner. The proposed TRS configuration is a combination of permeators, a platinum microreactor (MR) and a trickle bed reactor (TBR) and consists of two branches: one branch for elemental tritium recovery from tritiated deuterium gas and the second one for tritium recovery from streams containing a significant amount of water vapours but a low amount, below 5%, of tritiated gas. The two branches shall work in a complementary manner in such a way that the bleed stream from the permeators shall be further processed in the MR and TBR in view of achieving the required decontamination level. A preliminary evaluation of the proposed TRS in comparison with state of the art tritium recovery system from tritium processing facilities is also discussed. (authors)

  9. Development of a tritium recovery system from CANDU tritium removal facility

    Draghia, M.; Pasca, G.; Porcariu, F. [SC.IS.TECH SRL, Timisoara (Romania)

    2015-03-15

    The main purpose of the Tritium Recovery System (TRS) is to reduce to a maximum possible extent the release of tritium from the facility following a tritium release in confinement boundaries and also to have provisions to recover both elemental and vapors tritium from the purging gases during maintenance and components replacement from various systems processing tritium. This work/paper proposes a configuration of Tritium Recovery System wherein elemental tritium and water vapors are recovered in a separated, parallel manner. The proposed TRS configuration is a combination of permeators, a platinum microreactor (MR) and a trickle bed reactor (TBR) and consists of two branches: one branch for elemental tritium recovery from tritiated deuterium gas and the second one for tritium recovery from streams containing a significant amount of water vapours but a low amount, below 5%, of tritiated gas. The two branches shall work in a complementary manner in such a way that the bleed stream from the permeators shall be further processed in the MR and TBR in view of achieving the required decontamination level. A preliminary evaluation of the proposed TRS in comparison with state of the art tritium recovery system from tritium processing facilities is also discussed. (authors)

  10. Tritium emission reduction at Darlington tritium removal facility using a Bubbler System

    Ontario Power Generation Nuclear (OPGN) has a 4 x 880 MWe CANDU nuclear station at its Darlington Nuclear Div. located in Bowmanville. The station operates a Tritium Removal Facility (TRF) to reduce and maintain low tritium levels in the Moderator and Heat Transport heavy water systems of Ontario's CANDU fleet by extracting, concentrating, immobilizing and storing as a metal tritide. Minimizing tritium releases to the environment is of paramount importance to ensure that dose to the public is as low as reasonably achievable (ALARA) and to maintain credibility with the Public. Tritium is removed from the Cryogenic Distillation System to the Tritium Immobilization System (TIS) glove box via a transfer line that is protected by a rupture disc and relief valve. An overpressure event in 2003 had caused the rupture disc to blow, resulting in the release of a significant quantity of elemental tritium into the relief valve discharge line, which ties into the contaminated exhaust system. As a result of a few similar events occurring over a number of years of TRF operation, the released elemental tritium would have been converted to tritium oxide in the presence of a stagnant moist air environment in the stainless steel discharge line. A significant amount of tritium oxide hold-up in the discharge line was anticipated. To minimize any further releases to the environment, a Bubbler System was designed to remove and recover the tritium from the discharge line. This paper summarizes the results of several Bubbler recovery runs that were made over a period of a month. Approximately 3500 Ci of tritium oxide and 230 Ci of elemental tritium were removed and collected. The tritium contained in the water produced from the Bubbler system was later safely recovered in the station's downgraded D2O clean-up and recovery system. (authors)

  11. Behaviour of tritium in the environment

    Full text: There is considerable interest in the behaviour of radionuclides of global character that may be released to the environment through the development of nuclear power. Tritium is of particular interest due to its direct incorporation into water and organic tissue. Although there has been a large decrease (more than ten times) in tritium concentration since the stopping of nuclear weapons tests in the atmosphere, the construction in the near future of many water reactors and in the far future of fusion reactors could increase the present levels. Progress has been made during recent years in the assessment of tritium distribution, in detection methods and in biological studies While several meetings have given scientists an opportunity to present papers on tritium, no specific symposium on this topic has been organized by the IAEA since 1961. Thus the purpose of the meeting was to review recent advances and to report on the practical aspects of tritium utilization and monitoring. The symposium was jointly organized with OECD/NEA, in co-operation with the US Department of Energy and the Lawrence Livermore Laboratory. Papers were presented on distribution of tritium, evaluation of future discharges, measurement of tritium, tritium in the aquatic environment, tritium in the terrestrial environment, tritium in man and monitoring of tritium Very interesting papers were given on distribution of tritium and participants got a good idea of the circulation of this radionuclide Some new data were provided on tritium pollution from luminous compounds and we learnt that the tritium release of the Swiss luminous compounds industry is of the same order of magnitude as the tritium release of Windscale. Projections indicate that, in the USA, the total quantity of tritium contained in discarded digital watches will be equal, approximately ten years in the future, to the release of nuclear power reactors Whereas nuclear reactor discharges are controlled there is no control

  12. Tritium safety of fusion power plants

    Tritium systems of a nuclear fusion plant, using the deuterium-tritium fuel cycle, has to ensure tritium safety during plant operation, and activated/tritiated materials management, during plant decommissioning. Accidents resulting in tritium releases to the environment may occur. It is important, therefore, to minimize the mobile tritium inventory by means of an adequate design and optimization of the plant tritium-bearing systems. Furthermore, the behaviour of tritium in the environment must be accurately studied, in order to take into account its oxidation and absorption by different materials. An experimental tritium-breeding module of DEMO fusion reactor is now under development in Russia. We plan to test it in the International Thermonuclear Experimental Reactor (ITER). The ceramic lithium orthosilicate will be used in it as tritium breeder. The Tritium Cycle System (TCS) will ensure tritium extracting and processing of gaseous mixtures containing tritium. The report contains the flow chart of the TCS using alloys producing hydrides. TCS assures highest possible autonomy and independence on ITER tritium plant at technological operations. The classification of the TCS modes of operation, adopted at the present stage of the module development, is described. The main initial events that may result in accidents are analysed. The maximum design accident and its consequences are considered. In particular, the maximum effective dose equivalent to the most exposed individual is calculated by means of the GEN II/FRAMES code. The flow sheet of technological operations at the maintenance and repair works and the System of Radiological Safety ensuring safety during these works is analyzed. Finally, some aspects of tritium decontamination from standpoint of waste handling are developed. In particular, material detritiation should be sufficient to allow clearance and recycling of less activated fusion materials. (orig.)

  13. A study on the safety of tritium storage and treatment

    For reduction of tritium release to the environment and utilization of tritium at industrial application and fusion technology, it is necessary to separate and store tritium. As a tritium separation and storage system, Tritium Removal Facility (TRF) and tritium storage vessel is under development in Korea. For the construction and operation of the system, it is necessary to estimate the safety of tritium storage system. As an isotope of hydrogen, tritium has similar hazards to hydrogen. In addition to the hydrogen hazards, due to radioactive decay of tritium, it is necessary to consider the risk of hydrogen and radioactive decay for the safe storage. In this study, hazards of hydrogen and the risk due to storage of tritium are reviewed. The safety related factors are suggested in terms of classification of hydrogen hazards and problems related to the tritium storage. The major design parameters of the vessel of foreign countries for the storage and transport of tritium are reviewed. By review of major safety parameters related to the tritium storage, the results of this study can be applied and helpful to the development and design of tritium storage vessel in Korea. Also, the results can be useful at design of the tritium treatment facility

  14. Tapentadol immediate release: a new treatment option for acute pain management

    Marc Afilalo

    2010-02-01

    Full Text Available Marc Afilalo1, Jens-Ulrich Stegmann2, David Upmalis31Sir Mortimer B. Davis Jewish General Hospital, Montréal, Canada; 2Global Drug Safety, Grünenthal GmbH, Aachen, Germany; 3Johnson & Johnson Pharmaceutical Research and Development, L.L.C., Raritan, New Jersey, USAAbstract: The undertreatment of acute pain is common in many health care settings. Insufficient management of acute pain may lead to poor patient outcomes and potentially life-threatening complications. Opioids provide relief of moderate to severe acute pain; however, therapy with pure µ-opioid agonists is often limited by the prevalence of side effects, particularly opioid-induced nausea and vomiting. Tapentadol is a novel, centrally acting analgesic with 2 mechanisms of action, µ-opioid receptor agonism and norepinephrine reuptake inhibition. The analgesic effects of tapentadol are independent of metabolic activation and tapentadol has no active metabolites; therefore, in theory, tapentadol may be associated with a low potential for interindividual efficacy variations and drug–drug interactions. Previous phase 3 trials in patients with various types of moderate to severe acute pain have shown that tapentadol immediate release (IR; 50 to 100 mg every 4 to 6 hours provides analgesia comparable to that provided by the pure µ-opioid agonist comparator, oxycodone HCl IR (10 or 15 mg every 4 to 6 hours, with a lower incidence of nausea, vomiting, and constipation. Findings suggest tapentadol may represent an improved treatment option for acute pain.Keywords: tapentadol IR, acute pain, opioid, gastrointestinal tolerability

  15. Tritium migration studies at the Nevada Test Site

    Emanation of tritium from waste containers is a commonly known phenomenon. Release of tritium from buried waste packages was anticipated; therefore, a research program was developed to study both the rate of tritium release from buried containers and subsequent migration of tritium through soil. Migration of tritium away from low-level radioactive wastes buried in Area 5 of the Nevada Test Site was studied. Four distinct disposal events were investigated. The oldest burial event studied was a 1976 emplacement of 3.5 million curies of tritium in a shallow land burial trench. In another event, 248 thousand curies of tritium was disposed of in an overpack emplaced 6 m below the floor of a low-level waste disposal pit. Measurement of the emanation rate of tritium out of 55 gallon drums to the overpack was studied, and an annual doubling of the emanation rate over a seven year period, ending in 1990, was found. In a third study, upward tritium migration in the soil, resulting in releases in the atmosphere were observed in a greater confinement disposal test. Releases of tritium to the atmosphere were found to be insignificant. The fourth event consisted of burial of 2.2 million curies of tritium in a greater confinement disposal operation. Emanation of tritium from the buried containers has been increasing since disposal, but no significant migration was found four years following backfilling of the disposal hole

  16. Titanium for long-term tritium storage

    Heung, L.K.

    1994-12-01

    Due to the reduction of nuclear weapon stockpile, there will be an excess of tritium returned from the field. The excess tritium needs to be stored for future use, which might be several years away. A safe and cost effective means for long term storage of tritium is needed. Storing tritium in a solid metal tritide is preferred to storing tritium as a gas, because a metal tritide can store tritium in a compact form and the stored tritium will not be released until heat is applied to increase its temperature to several hundred degrees centigrade. Storing tritium as a tritide is safer and more cost effective than as a gas. Several candidate metal hydride materials have been evaluated for long term tritium storage. They include uranium, La-Ni-Al alloys, zirconium and titanium. The criteria used include material cost, radioactivity, stability to air, storage capacity, storage pressure, loading and unloading conditions, and helium retention. Titanium has the best combination of properties and is recommended for long term tritium storage.

  17. Involvement of histamine released from mast cells in acute radiation dermatitis in mice

    A possible involvement of histamine in acute radiation dermatitis in mice was investigated. The dose of 40 Gy of gamma irradiation induced erythema and edema in C57BL/6 mice treated with vehicle. However, in C57BL/6 mice treated with chlorpheniramine and WBB6F1-W/WV mice, erythema and edema were not observed. In all of these mice, epilation and dry desquamation were induced, but bepotastine significantly reduced the extent of these areas. These results suggest that gamma irradiation-induced erythema and edema were caused by histamine released from mast cells via histamine H1 receptor, and epilation was induced by other inflammatory mediators. (author)

  18. Management of Tritium in European Spallation Source

    Ene, Daniela; Andersson, Kasper Grann; Jensen, Mikael; Nielsen, Sven Poul; Severin, Gregory

    2015-01-01

    The European Spallation Source (ESS) will produce tritium via spallation and activation processes during operational activities. Within the location of ESS facility in Lund, Sweden site it is mandatory to demonstrate that the management strategy of the produced tritium ensures the compliance with...... the country regulation criteria. The aim of this paper is to give an overview of the different aspects of the tritium management in ESS facility. Besides the design parameter study of the helium coolant purification system of the target the consequences of the tritium releasing into the environment...... were also analyzed. Calculations shown that the annual release of tritium during the normal operations represents a small fraction from the estimated total dose. However, more refined calculations of migration of activated-groundwater should be performed for higher hydraulic conductivities, with the...

  19. Studies on steps affecting tritium residence time in solid blanket

    For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

  20. Universal tritium transmitter

    At the Savannah River Site and throughout the National Nuclear Security Agency (NNSA) tritium is measured using Ion or Kanne Chambers. Tritium flowing through an Ion Chamber emits beta particles generating current flow proportional to tritium radioactivity. Currents in the 1 x 10-15 A to 1 x 10-6 A are measured. The distance between the Ion Chamber and the electrometer in NNSA facilities can be over 100 feet. Currents greater than a few micro-amperes can be measured with a simple modification. Typical operating voltages of 500 to 1000 Volts and piping designs require that the Ion Chamber be connected to earth ground. This grounding combined with long cable lengths and low currents requires a very specialized preamplifier circuit. In addition, the electrometer must be able to supply 'fail safe' alarm signals which are used to alert personnel of a tritium leak, trigger divert systems preventing tritium releases to the environment and monitor stack emissions as required by the United States federal Government and state governments. Ideally the electrometer would be 'self monitoring'. Self monitoring would reduce the need for constant checks by maintenance personnel. For example at some DOE facilities monthly calibration and alarm checks must be performed to ensure operation. NNSA presently uses commercially available electrometers designed specifically for this critical application. The problems with these commercial units include: ground loops, high background currents, inflexibility and susceptibility to Electromagnetic Interference (EMI) which includes RF and Magnetic fields. Existing commercial electrometers lack the flexibility to accommodate different Ion Chamber designs required by the gas pressure, type of gas and range. Ideally the electrometer could be programmed for any expected gas, range and high voltage output. Commercially available units do not have 'fail safe' self monitoring capability. Electronics used to measure extremely low current must have

  1. Interaction of energetic tritium with silicon carbide

    In order to investigate the physical and chemical interactions of energetic hydrogen isotope species with silicon carbide, recoil tritium from the 3He(n,p)T reaction has been allowed to react with K-T silicon carbide and silicon carbide powder. The results show that if the silicon carbide has been degassed and annealed at 14000C prior to tritium bombardment, a considerable fraction of the tritium (ca. 40%) is released as HTO from the SiC upon heating to 13500C under vacuum conditions. Most of the remaining tritium is retained in SiC, e.g., the retention of the tritium in the K-T SiC was found to be 62 and 22% upon heating to 600 and 13500C, respectively. This is in direct contrast to graphite samples in which the tritium is not released to any significant extent even when heated to 13500C. Samples which were exposed to H2O and H2 prior to tritium bombardment were heated to 6000C after the irradiation. The results obtained indicate that a total of 38.7 and 2.49% of the tritium is released in the form of HT and CH3T in the case of H2 or H2O exposure, respectively. Treatment of degassed samples after tritium bombardment with H2O and H2 at temperatures up to 10000C leads to the release of up to 44.9% of the tritium as HT and CH3T. 42 references, 2 figures, 2 tables

  2. Tritium handling in vacuum systems

    Gill, J.T. [Monsanto Research Corp., Miamisburg, OH (United States). Mound Facility; Coffin, D.O. [Los Alamos National Lab., NM (United States)

    1986-10-01

    This report provides a course in Tritium handling in vacuum systems. Topics presented are: Properties of Tritium; Tritium compatibility of materials; Tritium-compatible vacuum equipment; and Tritium waste treatment.

  3. Magmatic tritium

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory. Detailed geochemical sampling of high-temperature fumaroles, background water, and fresh magmatic products from 14 active volcanoes reveal that they do not produce measurable amounts of tritium (3H) of deep origin (2O). On the other hand, all volcanoes produce mixtures of meteoric and magmatic fluids that contain measurable 3H from the meteoric end-member. The results show that cold fusion is probably not a significant deep earth process but the samples and data have wide application to a host of other volcanological topics

  4. Acute stress increases depolarization-evoked glutamate release in the rat prefrontal/frontal cortex: the dampening action of antidepressants.

    Laura Musazzi

    Full Text Available BACKGROUND: Behavioral stress is recognized as a main risk factor for neuropsychiatric diseases. Converging evidence suggested that acute stress is associated with increase of excitatory transmission in certain forebrain areas. Aim of this work was to investigate the mechanism whereby acute stress increases glutamate release, and if therapeutic drugs prevent the effect of stress on glutamate release. METHODOLOGY/FINDINGS: Rats were chronically treated with vehicle or drugs employed for therapy of mood/anxiety disorders (fluoxetine, desipramine, venlafaxine, agomelatine and then subjected to unpredictable footshock stress. Acute stress induced marked increase in depolarization-evoked release of glutamate from synaptosomes of prefrontal/frontal cortex in superfusion, and the chronic drug treatments prevented the increase of glutamate release. Stress induced rapid increase in the circulating levels of corticosterone in all rats (both vehicle- and drug-treated, and glutamate release increase was blocked by previous administration of selective antagonist of glucocorticoid receptor (RU 486. On the molecular level, stress induced accumulation of presynaptic SNARE complexes in synaptic membranes (both in vehicle- and drug-treated rats. Patch-clamp recordings of pyramidal neurons in the prefrontal cortex revealed that stress increased glutamatergic transmission through both pre- and postsynaptic mechanisms, and that antidepressants may normalize it by reducing release probability. CONCLUSIONS/SIGNIFICANCE: Acute footshock stress up-regulated depolarization-evoked release of glutamate from synaptosomes of prefrontal/frontal cortex. Stress-induced increase of glutamate release was dependent on stimulation of glucocorticoid receptor by corticosterone. Because all drugs employed did not block either elevation of corticosterone or accumulation of SNARE complexes, the dampening action of the drugs on glutamate release must be downstream of these processes

  5. ICF tritium production reactor

    The conceptual design of an ICF tritium production reactor is described. The chamber design uses a beryllium multiplier and a liquid lithium breeder to achieve a tritium breeding ratio of 2.08. The annual net tritium production of this 532 MW/sub t/ plant is 16.9 kg, and the estimated cost of tritium is $8100/g

  6. Atrial distension, haemodilution, and acute control of renin release during water immersion in humans

    Gabrielsen, A; Pump, B; Bie, P;

    2002-01-01

    We tested the hypothesis that atrial distension (stimulation of cardiopulmonary baroreceptors) is not the single pivotal stimulus for the acute suppression of renin release during water immersion in humans and that immersion-induced haemodilution constitutes an important additional stimulus. In...... nine healthy male subjects, identical increases in atrial distension were induced by two immersion procedures (of 30 min each); one without (WI) and one with attenuation (WI + cuff) of the concomitant haemodilution (estimated from changes in plasma protein concentration) by inflating thigh cuffs during...... immersion. During WI, central venous pressure (CVP) and left atrial diameter (LAD) increased (P <0.05) by 5.5 +/- 0.4 mmHg and 4.6 +/- 0.5 mm, respectively, and plasma protein concentration and plasma renin activity (PRA) progressively decreased (P <0.05) by 4.8 +/- 0.5 g L(-1) and 1.6 +/- 0.2 ng mL(-1) h...

  7. Effects of pre-treatments of Li2TiO3 pebbles on the release of tritium generated during short irradiations

    The reduction of Li2TiO3 to Li2TiO3-x at 800 deg. C by He+H2 (0.1%) purge gas was studied with Li2TiO3 pebbles. The evolution of O-vacancy concentration (x(t)) allowed extrapolation to the steady-state value x(t→∞)=0.0096 which was found to be independent from the material microstructure, while the process rate and its mechanism were found to depend on the grain size, porosity and sintering temperature of the pebbles. The temperature of tritium removal from the so-treated (reduced) pebbles was found to be higher than those from the oxidized (stoichiometric) ones, the increase of x being accompanied by an increase in the population of more energetic tritium trap sites at the grain boundary interface of the material. However, this phenomenon should not be of concern under 'normal' breeding blanket operating conditions

  8. Handling of tritium-bearing wastes

    The generation of nuclear power and reprocessing of nuclear fuel results in the production of tritium and the possible need to control the release of tritium-contaminated effluents. In assessing the need for controls, it is necessary to know the production rates of tritium at different nuclear facilities, the technologies available for separating tritium from different gaseous and liquid streams, and the methods that are satisfactory for storage and disposal of tritiated wastes. The intention in applying such control technologies and methods is to avoid undesirable effects on the environment, and to reduce the radiation burden on operational personnel and the general population. This technical report is a result of the IAEA Technical Committee Meeting on Handling of Tritium-bearing Effluents and Wastes, which was held in Vienna, 4 - 8 December 1978. It summarizes the main topics discussed at the meeting and appends the more detailed reports on particular aspects that were prepared for the meeting by individual participants

  9. DOE handbook: Tritium handling and safe storage

    NONE

    1999-03-01

    The DOE Handbook was developed as an educational supplement and reference for operations and maintenance personnel. Most of the tritium publications are written from a radiological protection perspective. This handbook provides more extensive guidance and advice on the null range of tritium operations. This handbook can be used by personnel involved in the full range of tritium handling from receipt to ultimate disposal. Compliance issues are addressed at each stage of handling. This handbook can also be used as a reference for those individuals involved in real time determination of bounding doses resulting from inadvertent tritium releases. This handbook provides useful information for establishing processes and procedures for the receipt, storage, assay, handling, packaging, and shipping of tritium and tritiated wastes. It includes discussions and advice on compliance-based issues and adds insight to those areas that currently possess unclear DOE guidance.

  10. DOE handbook: Tritium handling and safe storage

    The DOE Handbook was developed as an educational supplement and reference for operations and maintenance personnel. Most of the tritium publications are written from a radiological protection perspective. This handbook provides more extensive guidance and advice on the null range of tritium operations. This handbook can be used by personnel involved in the full range of tritium handling from receipt to ultimate disposal. Compliance issues are addressed at each stage of handling. This handbook can also be used as a reference for those individuals involved in real time determination of bounding doses resulting from inadvertent tritium releases. This handbook provides useful information for establishing processes and procedures for the receipt, storage, assay, handling, packaging, and shipping of tritium and tritiated wastes. It includes discussions and advice on compliance-based issues and adds insight to those areas that currently possess unclear DOE guidance

  11. Magmatic tritium

    Goff, F.; Aams, A.I. [Los Alamos National Lab., NM (United States); McMurtry, G.M. [Univ. of Hawaii, Honolulu, HI (United States); Shevenell, L. [Univ. of Nevada, Reno, NV (United States); Pettit, D.R. [National Aeronautics and Space Administration (United States); Stimac, J.A. [Union Geothermal Company (United States); Werner, C. [Pennsylvania State Univ., University Park, PA (United States)

    1997-07-01

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory. Detailed geochemical sampling of high-temperature fumaroles, background water, and fresh magmatic products from 14 active volcanoes reveal that they do not produce measurable amounts of tritium ({sup 3}H) of deep origin (<0.1 T.U. or <0.32 pCi/kg H{sub 2}O). On the other hand, all volcanoes produce mixtures of meteoric and magmatic fluids that contain measurable {sup 3}H from the meteoric end-member. The results show that cold fusion is probably not a significant deep earth process but the samples and data have wide application to a host of other volcanological topics.

  12. An Acute Bout of Self-Myofascial Release in the Form of Foam Rolling Improves Performance Testing

    Peacock, Corey A; KREIN, DARREN D.; Silver, Tobin A.; Sanders, Gabriel J; VON CARLOWITZ, KYLE-PATRICK A.

    2014-01-01

    Recent developments in the strength and conditioning field have shown the incorporation of foam rolling self-myofascial release in adjunct with a dynamic warm-up. This is thought to improve overall training performance; however, minimal research exists supporting this theory. Therefore, determining if an acute bout of foam rolling self-myofascial release in addition to a dynamic warm-up could influence performance is of importance. In order to do so, eleven athletically trained male subjects ...

  13. Tritium recovery from ceramic breeder blanket

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  14. Tritium in the Savannah River Site environment

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found

  15. Tritium in the Savannah River Site environment

    Murphy, C.E. Jr.; Bauer, L.R.; Hayes, D.W.; Marter, W.L.; Zeigler, C.C.; Stephenson, D.E.; Hoel, D.D.; Hamby, D.M.

    1991-05-01

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found.

  16. Behavior of tritium in heavy water reactors

    In the ATR Fugen power station, the radiation control regarding the tritium in heavy water has been carried out since the heavy water was filled in the system of the reactor in November, 1977. At first, the concentration of tritium in heavy water was about 60 μCi/cc, but in November, 1981, it increased to about 1.3 mCi/cc, and the saturation concentration after 30 years is estimated to become about 17 mCi/cc. In this report, on the transfer of tritium to the work environment and general environment, its barrier, recovery, measurement and the protection against it, the experience in the Fugen power station is described. The heavy water system was constructed as the perfectly closed circuit by welding stainless steel, and a canned heavy water circulating pump has been used. The leak of heavy water in the steady operation is negligible, but attention must be paid to the transfer of tritium to the environment when the system is disassembled for the regular inspection. The measurement of tritium for individual exposure control, environment and released radioactivity, the tritium-removing equipment and protective suits, and the release of tritium to general environment are reported. (Kako, I.)

  17. Analyzing the Release of Copeptin from the Heart in Acute Myocardial Infarction Using a Transcoronary Gradient Model.

    Boeckel, Jes-Niels; Oppermann, Jana; Anadol, Remzi; Fichtlscherer, Stephan; Zeiher, Andreas M; Keller, Till

    2016-01-01

    Copeptin is the C-terminal end of pre-provasopressin released equimolar to vasopressin into circulation and recently discussed as promising cardiovascular biomarker amendatory to established markers such as troponins. Vasopressin is a cytokine synthesized in the hypothalamus. A direct release of copeptin from the heart into the circulation is implied by data from a rat model showing a cardiac origin in hearts put under cardiovascular wall stress. Therefore, evaluation of a potential release of copeptin from the human heart in acute myocardial infarction (AMI) has been done. PMID:26864512

  18. Effect of Acute Exercise on ANP-Induced Inhibition of Aldosterone Release in Rat Adrenals

    SUDA, Kazuhiro; Hagiwara, Hiromi; Komabayashi, Takao; Izawa, Tetsuya; Imai, Hajime; Hayashi, Tomoya; Era, Seiichi

    2004-01-01

    SUDA, K., HAGIWARA, H., KOMABAYASHI, T., IZAWA, T., IMAI, H., HAYASHI, T. and ERA, s., Effect of Acute Exercise on ANP-Induced Inhibition of Aldosterone Release in Rat Adrenals. Abv. Exerc. Sports Physiol., Vol.10, No.2 pp.43-47, 2004. We intide (ANP)-induced inhibition of aldosterone release in rat adrenals. The rats ran on treadmill for two hours. Immediately after the exercise, the adrenals were excised and used for an aldosterone release experiment, an ANP binding assay, and a guanylate c...

  19. Tritium analysis of fusion-based hydrogen production reactor FDS-III

    A dynamic subsystem model of tritium fuel cycle for the FDS-III was developed, and the required minimum tritium supply for reactor startup and the doubling time for tritium breeding were calculated by using the Tritium Analysis Software (TAS). Some factors which would affect the tritium supply and doubling time were considered, such as the tritium fractional burnup in the plasma, tritium breeding ratio (TBR), the residence time of tritium in all subsystems, and tritium decay, etc. The results showed that the minimum tritium supply for startup was sensitive with the tritium fractional burnup in the plasma, but the effect of the TBR could be neglected. The double time for tritium breeding strongly depended on the TBR and the tritium fractional burnup. Based on the model, the analysis results predicted that the required initial minimum tritium supply was ∼9.9 kg for startup. After one year's operation, the total tritium inventory in fuel cycle system was ∼33 kg. And the total tritium release into environment was ∼4 mg, which was much lower than the allow level, i.e. 1 g-T/year. The tritium in fuel storage system would be doubled and could be extracted to supply for the other fusion power reactor's startup after ∼886 days operation.

  20. Tritium analysis of fusion-based hydrogen production reactor FDS-III

    Song Yong, E-mail: ysong@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Huang Qunying [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Ni Muyi [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2010-12-15

    A dynamic subsystem model of tritium fuel cycle for the FDS-III was developed, and the required minimum tritium supply for reactor startup and the doubling time for tritium breeding were calculated by using the Tritium Analysis Software (TAS). Some factors which would affect the tritium supply and doubling time were considered, such as the tritium fractional burnup in the plasma, tritium breeding ratio (TBR), the residence time of tritium in all subsystems, and tritium decay, etc. The results showed that the minimum tritium supply for startup was sensitive with the tritium fractional burnup in the plasma, but the effect of the TBR could be neglected. The double time for tritium breeding strongly depended on the TBR and the tritium fractional burnup. Based on the model, the analysis results predicted that the required initial minimum tritium supply was {approx}9.9 kg for startup. After one year's operation, the total tritium inventory in fuel cycle system was {approx}33 kg. And the total tritium release into environment was {approx}4 mg, which was much lower than the allow level, i.e. 1 g-T/year. The tritium in fuel storage system would be doubled and could be extracted to supply for the other fusion power reactor's startup after {approx}886 days operation.

  1. Management of tritium European Spallation Source

    The European Spallation Source (ESS) will produce tritium via spallation and activation processes during operational activities. Within the location of ESS facility in Lund, Sweden site it is mandatory to demonstrate that the management strategy of the produced tritium ensures the compliance with the country regulation criteria. The aim of this paper is to give an overview of the different aspects of the tritium management in ESS facility. Besides the design parameter study of the helium coolant purification system of the target the consequences of the tritium releasing into the environment were also analyzed. Calculations show that the annual release of tritium during the normal operations represents a small fraction from the estimated total dose. However, more refined calculations of migration of activated-groundwater should be performed for higher hydraulic conductivities, with the availability of the results on soil examinations. With the assumption of 100% release of tritium to the atmosphere during the occurring of the extreme accidents, it was found as well that the total dose complies with the constraint. (authors)

  2. Organically bound tritium in fishes at Rana Pratap Sagar lake

    Tritium is mainly released into environment in the form of tritiated water (HTO). It becomes a part of general hydrosphere and gets equilibrated with body water of aquatic organisms within a short period of time, as they are totally immersed in the water. Tritium present in body water is called tissue free tritium (TFT). Tritium gets incorporated in body tissue, either during the formation of the tissues or by the exchange of tritium with the labile hydrogen sites present in the tissues. It can significantly contribute radiation dose to the aquatic organism and public consuming the aquatic food, as organically bound tritium (OBT) shows longer retention period in the body compared to those of TFT. The estimations of TFT and OBT contents in the variety of fish at Rana Pratap Sagar (RPS) are reported in this paper. An attempt is also made to evaluate the doses to the local residents, due to intake of tritium through the consumption of RPS fish. (author)

  3. Acute effects of self-myofascial release using a foam roller on arterial function.

    Okamoto, Takanobu; Masuhara, Mitsuhiko; Ikuta, Komei

    2014-01-01

    Flexibility is associated with arterial distensibility. Many individuals involved in sport, exercise, and/or fitness perform self-myofascial release (SMR) using a foam roller, which restores muscles, tendons, ligaments, fascia, and/or soft-tissue extensibility. However, the effect of SMR on arterial stiffness and vascular endothelial function using a foam roller is unknown. This study investigates the acute effect of SMR using a foam roller on arterial stiffness and vascular endothelial function. Ten healthy young adults performed SMR and control (CON) trials on separate days in a randomized controlled crossover fashion. Brachial-ankle pulse wave velocity (baPWV), blood pressure, heart rate, and plasma nitric oxide (NO) concentration were measured before and 30 minutes after both SMR and CON trials. The participants performed SMR of the adductor, hamstrings, quadriceps, iliotibial band, and trapezius. Pressure was self-adjusted during myofascial release by applying body weight to the roller and using the hands and feet to offset weight as required. The roller was placed under the target tissue area, and the body was moved back and forth across the roller. In the CON trial, SMR was not performed. The baPWV significantly decreased (from 1,202 ± 105 to 1,074 ± 110 cm·s-1) and the plasma NO concentration significantly increased (from 20.4 ± 6.9 to 34.4 ± 17.2 μmol·L-1) after SMR using a foam roller (both p < 0.05), but neither significantly differed after CON trials. These results indicate that SMR using a foam roller reduces arterial stiffness and improves vascular endothelial function. PMID:23575360

  4. The Tritium White Paper

    This publication proposes a synthesis of the activities of two work-groups between May 2008 and April 2010. It reports the ASN's (the French Agency for Nuclear Safety) point of view, describes its activities and actions, and gives some recommendations. It gives a large and detailed overview of the knowledge status on tritium: tritium source inventory, tritium origin, management processes, capture techniques, reduction, tritium metrology, impact on the environment, impacts on human beings

  5. Safety and Efficacy of Paliperidone Extended-Release in Acute and Maintenance Treatment of Schizophrenia

    Spina, Edoardo; Crupi, Rosalia

    2011-01-01

    Paliperidone, the major active metabolite of risperidone, is a second-generation antipsychotic that has been developed as an extended-release (ER) tablet formulation that minimizes peak-trough fluctuations in plasma concentrations, allowing once-daily administration and constant drug delivery. Paliperidone ER has demonstrated efficacy in the reduction of acute schizophrenia symptoms in 6-week, placebo-controlled, double-blind trials and clinical benefits were maintained in the longer-term according to extension studies of up to 52 weeks in duration. Compared with quetiapine, paliperidone ER was associated with a more rapid symptom improvement. In addition, it was more effective than placebo in the prevention of symptom recurrence. Paliperidone ER is generally well tolerated with a predictable adverse event profile. Like risperidone, it is associated with a dose-dependent risk of extrapyramidal symptoms and prolactin elevation. Short- and longer-term studies have indicated a low liability for paliperidone ER to cause metabolic (ie, weight gain, hyperglycaemia and lipid dysregulation) or cardiovascular adverse effects. Available safety data from elderly patients appear to be promising. Due to negligible hepatic biotransformation, paliperidone ER is unlikely to be involved in clinically significant metabolic drug-drug interactions. Additional active comparator trials evaluating efficacy, tolerability and cost-effectiveness are required to better define the role of paliperidone ER in the treatment of schizophrenia in relation to other currently available second-generation antipsychotics, particularly risperidone. PMID:23861636

  6. Safe handling of tritium

    The main objective of this publication is to provide practical guidance and recommendations on operational radiation protection aspects related to the safe handling of tritium in laboratories, industrial-scale nuclear facilities such as heavy-water reactors, tritium removal plants and fission fuel reprocessing plants, and facilities for manufacturing commercial tritium-containing devices and radiochemicals. The requirements of nuclear fusion reactors are not addressed specifically, since there is as yet no tritium handling experience with them. However, much of the material covered is expected to be relevant to them as well. Annex III briefly addresses problems in the comparatively small-scale use of tritium at universities, medical research centres and similar establishments. However, the main subject of this publication is the handling of larger quantities of tritium. Operational aspects include designing for tritium safety, safe handling practice, the selection of tritium-compatible materials and equipment, exposure assessment, monitoring, contamination control and the design and use of personal protective equipment. This publication does not address the technologies involved in tritium control and cleanup of effluents, tritium removal, or immobilization and disposal of tritium wastes, nor does it address the environmental behaviour of tritium. Refs, figs and tabs

  7. Modeling tritium behavior in Li2ZrO3

    Lithium metazirconate (Li2ZrO3) is a promising tritium breeder material for fusion reactors because of its excellent tritium release characteristics. In particular, for water-cooled breeding blankets (e.g., ITER), Li2ZrO3 is appealing from a design perspective because of its good tritium release at low operating temperatures. The steady-state and transient tritium release/retention database for Li2ZrO3 is reviewed, along with conventional diffusion and first-order surface resorption models which have been used to match the database. A first-order surface resorption model is recommended in the current work both for best-estimate and conservative (i.e., inventory upper-bound) predictions. Model parameters we determined and validated for both types of predictions, although emphasis is placed on conservative design predictions. The effects on tritium retention of ceramic microstructure, protium partial pressure in the purge gas and purge gas flow rate are discussed, along with other mechanisms for tritium retention which may not be dominant in the experiments, but may be important in blanket design analyses. The proposed tritium retention/release model can be incorporated into a transient thermal performance code to enable whole-blanket predictions of tritium retention/release during cyclic reactor operation. Parameters for the ITER driver breeding blanket are used to generate a numerical set of model predictions for steady-state operation

  8. Modeling tritium behavior in Li2ZrO3

    Lithium metazirconate (Li2ZrO3) is a promising tritium breeder material for fusion reactors because of its excellent tritium release characteristics. In particular, for water-cooled breeding blankets (e.g., ITER), Li2ZrO3 is appealing from a design perspective because of its good tritium release at low operating temperatures. The steady-state and transient tritium release/retention database for Li2ZrO3 is reviewed, along with conventional diffusion and first-order surface desorption models which have been used to match the database. A first-order surface desorption model is recommended in the current work both for best-estimate and conservative (i.e., inventory upper-bound) predictions. Model parameters are determined and validated for both types of predictions, although emphasis is placed on conservative design predictions. The effects on tritium retention of ceramic microstructure, protium partial pressure in the purge gas and purge gas flow rate are discussed, along with other mechanisms for tritium retention which may not be dominant in the experiments, but may be important in blanket design analyses. The proposed tritium retention/release model can be incorporated into a transient thermal performance code to enable whole-blanket predictions of tritium retention/release during cyclic reactor operation. Parameters for the ITER driver breeding blanket are used to generate a numerical set of model predictions for steady-state operation. (author)

  9. Tritium and radon risks for humans

    Full text: The gaseous and liquid releases into environment from the two CANDU type units of Cernavoda NPP now in operation has more tritium contents than other kind of western power reactors. CANDU type reactor uses heavy water as moderator and primary circuit heat transfer agent. In normal operation deuterium go to tritium by neutron capture, the molecule of tritiated heavy water can escape from nuclear systems in very small amounts and so it is released into environment. After release the tritium follows the way of water into environment. One year ago the antinuclear NGO led a hard attack against Units 3 and 4 during the procedure of public acceptance request. This attack tried to demonstrate the great risk for humans of the tritium released by Cernavoda NPP. Obviously this risk is very low as demonstrated by many years reactor operation. SNN as owner of Cernavoda NPP ensures by all kind of information channels about the radioactive potential risk for humans. By the other hand, ironically, the antinuclear NGO makes nothing to inform the people about radon risk magnitude in some areas. This is a well-known fact but the radon concentration in dwellings can be decreased by some improved building procedures. The radon is the first natural cause of lung cancer. The environmental NGO and Romanian authorities do not have an information service about radon hazard data neither in dwellings or in uranium mining areas. The paper compares the properties and risks for tritium and radon. (authors)

  10. Tritium conference days

    This document gathers the slides of the available presentations given during this conference day. Twenty presentations out of 21 are assembled in the document and deal with: 1 - tritium in the environment (J. Garnier-Laplace); 2 - status of knowledge about tritium impact on health (L. Lebaron-Jacobs); 3 - tritium, discrete but present everywhere (M. Sene); 4 - management of tritium effluents from Areva NC La Hague site - related impact and monitoring (P. Devin); 5 - tritium effluents and impact in the vicinity of EDF's power plants (V. Chretien and B. Le Guen); 6 - contribution of CEA-Valduc centre monitoring to the knowledge of atmospheric tritiated water transfers to the different compartments of the environment (P. Guetat); 7 - tritium analysis in environment samples: constraints and means (N. Baglan); 8 - organically-linked tritium: the analyst view (E. Ansoborlo); 9 - study of tritium transfers to plants via OBT/HTOair and OBT/HTOfree (C. Boyer); 10 - tritium in the British Channel (M. Masson and P. Bailly-Du-Bois); 11 - tritium in British coastal waters (S. Jenkinson); 12 - recent results from epidemiology (R. Wakeford); 13 - effects of tritiated thymidine on hematopoietic stem cells (P.H. Romeo); 14 - tritium management issue in Canada: the point of view from authorities (P. Thompson); 15 - experience feedback of the detritiation process of Valduc centre (D. Leterq); 16 - difficulties linked with tritiated wastes confinement (F. Chastagner); 17 - optimisation of tritium management in the ITER project (P. Cortes); 18 - elements of thought about the management of tritium generated by nuclear facilities (M. Philippe); 19 - CIPR's position about the calculation of doses and risks linked with tritium exposure (F. Paquet); 20 - tritium think tanks (M. Fournier). (J.S.)

  11. On-line tritium monitoring in research reactor Dhruva

    Dhruva is 100 MW (Th) research reactor. Heavy water is used as a coolant, moderator and reflector. Tritium is one of the gaseous effluents released from 100 meter stack. During reactor operation, tritium release is approximately 135 GBqd-1 with a concentration of ∼ 90-100 KBqm-3 (0.33 DAC) and in shutdown condition it is slightly higher and depends upon the heavy water related jobs. Continuous sampling of the exhaust air is carried out using bubbler method and samples are analyzed once in a day to estimate the tritium release during the period. Due to energy interference, it is not possible to monitor tritium in presence of 41Ar using continuous tritium monitor in normal sampling method. Detection of tritium in the presence of other radionuclides is a particular challenge since radiation from these species masks detector response for tritium. In Dhruva main interfering radionuclides is 41Ar which is produced due to activation of 40Ar present in air (∼ 1 %) while cooling shutoff rod assemblies and depends upon reactor power. Concentration of 41Ar in exhaust air is 34.1 MBqm3 at 100 MW. This paper describes the setup proposed for online tritium monitoring using multi-line continuous tritium monitor developed by RSSD

  12. Tritium monitoring in environment at ICIT Tritium Separation Facility

    Full text: The Cryogenic Pilot is an experimental project developed within the national nuclear energy research program, which is designed to develop the required technologies for tritium and deuterium separation by cryogenic distillation of heavy water. The process used in this installation is based on a combination between liquid-phase catalytic exchange (LPCE) and cryogenic distillation. Basically, there are two ways that the Cryogenic Pilot could interact with the environment: by direct atmospheric release and through the sewage system. This experimental installation is located 15 km near the region biggest city and in the vicinity - about 1 km, of Olt River. It must be specified that in the investigated area there is an increased chemical activity; almost the entire Experimental Cryogenic Pilot's neighborhood is full of active chemical installations. This aspect is really essential for our study because the sewerage system is connected with the other three chemical plants from the neighborhood. For that reason we progressively established elements of an environmental monitoring program well in advance of tritium operation in order to determine baseline levels. The first step was the tritium level monitoring in environmental water and wastewater of industrial activity from neighborhood. In order to establish the base level of tritium concentration in the environment around the nuclear facilities, we investigated the sample preparation treatment for different types of samples: onion, green beams, grass, apple, garden lettuce, tomato, cabbage, strawberry and grapes. We used azeotropic distillation of all types of samples, the carrier solvent being toluene from different Romanian providers. All measurements for the determination of environmental tritium concentration were performed using liquid scintillation counting (LSC), with the Quantulus 1220 spectrometer. (authors)

  13. Acute hazardous substance releases resulting in adverse health consequences in children: Hazardous Substances Emergency Events Surveillance system, 1996-2003.

    Wattigney, Wendy A; Kaye, Wendy E; Orr, Maureen F

    2007-11-01

    Because of their small size and ongoing organ development, children may be more susceptible than adults to the harmful effects of toxic chemicals. The objective of the study reported here was to identify frequent locations, released substances, and factors contributing to short-term chemical exposures associated with adverse health consequences experienced by children. The study examined the Hazardous Substances Emergency Events Surveillance (HSEES) system data from 1996-2003. Eligible events involved the acute release of a hazardous substance associated with at least one child being injured. The study found that injured children were predominantly at school, home, or a recreational center when events took place. School-related events were associated with the accidental release of acids and the release of pepper spray by pranksters. Carbon monoxide poisonings occurring in the home, retail stores, entertainment facilities, and hotels were responsible for about 10 percent of events involving child victims. Chlorine was one of the top chemicals harmful to children, particularly at public swimming pools. Although human error contributed to the majority of releases involving child victims, equipment failure was responsible for most chlorine and ammonia releases. The authors conclude that chemical releases resulting in injury to children occur mostly in schools, homes, and recreational areas. Surveillance of acute hazardous chemical releases helped identify contributing causes and can guide the development of prevention outreach activities. Chemical accidents cannot be entirely prevented, but efforts can be taken to provide safer environments in which children can live, learn, and play. Wide dissemination of safety recommendations and education programs is required to protect children from needless environmental dangers. PMID:18044249

  14. Tritium contamination control

    Over the last years, there has been increased importance of tritium (3H or T), the radioactive isotope of hydrogen, in the nuclear power program and environmental studies. Cosmic ray interaction in the atmosphere, nuclear weapons testing, commercial products and nuclear facilities are the sources for environmental tritium. Several routes are available by which tritium as a gas or as tritiated water can reach the body tissues of man. It becomes necessary to constantly control the tritium concentration in the environment. Analytical methods to determine tritium in matrixes such as urine, water, air, fishes by scintillation counting and proportional counting are described. (Author)

  15. Tritium glovebox stripper system seismic design evaluation

    Grinnell, J. J. [Savannah River Site (SRS), Aiken, SC (United States); Klein, J. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    The use of glovebox confinement at US Department of Energy (DOE) tritium facilities has been discussed in numerous publications. Glovebox confinement protects the workers from radioactive material (especially tritium oxide), provides an inert atmosphere for prevention of flammable gas mixtures and deflagrations, and allows recovery of tritium released from the process into the glovebox when a glovebox stripper system (GBSS) is part of the design. Tritium recovery from the glovebox atmosphere reduces emissions from the facility and the radiological dose to the public. Location of US DOE defense programs facilities away from public boundaries also aids in reducing radiological doses to the public. This is a study based upon design concepts to identify issues and considerations for design of a Seismic GBSS. Safety requirements and analysis should be considered preliminary. Safety requirements for design of GBSS should be developed and finalized as a part of the final design process.

  16. Tritium Depth Profiles in 316 Stainless Steel

    Torikai, Yuji; Murata, Daiju; Penzhorn, Ralf-Dieter; Akaishi, Kenya; Watanabe, Kuniaki; Matsuyama, Masao

    To investigate the behavior of hydrogen uptake and release by 316 stainless steel (SS316), as-received and finely polished stainless steel specimens were exposed at 573 K to tritium gas diluted with hydrogen. Then tritium concentration in the exposed specimens was measured as a function of depth using a chemical etching method. All the tritium concentration profiles showed a sharp drop in the range of 10 μm from the top surface up to the bulk. The amount of tritium absorbed into the polished specimens was three times larger than that into the as-received specimen. However, the polishing effects disappeared by exposing to the air for a long time.

  17. Tritium pellet injector results

    Injection of solid tritium pellets is considered to be the most promising way of fueling fusion reactors. The Tritium Proof-of- Principle (TPOP) experiment has demonstrated the feasibility of forming and accelerating tritium pellets. This injector is based on the pneumatic pipe-gun concept, in which pellets are formed in situ in the barrel and accelerated with high-pressure gas. This injector is ideal for tritium service because there are no moving parts inside the gun and because no excess tritium is required in the pellet production process. Removal of 3He from tritium to prevent blocking of the cryopumping action by the noncondensible gas has been demonstrated with a cryogenic separator. Pellet velocities of 1280 m/s have been achieved for 4-mm-diam by 4-mm-long cylindrical tritium pellets with hydrogen propellant at 6.96 MPa (1000 psi). 10 refs., 10 figs

  18. Modeling of tritium behavior in ceramic breeder materials

    The model described in this paper considers diffusion and desorption as the rate-controlling mechanisms for tritium release from a ceramic breeder material. This model was used to predict the tritium release from samples of Li2SiO3 and LiAlO2, given the temperature history of the samples. The diffusion-desorption model did a better job of predicting the tritium release for these samples under pure helium purge gas than did a pure diffusion model using the best values for the diffusivity of these materials available. The activation energies of desorption found from the best fit of the predicted tritium release to the observed release were 105-108 kJ/mol for Li2SiO3 and 85.7 kJ/mol for LiAlO2. These values are in fair agreement with activation energies reported in the literature. 13 refs., 6 figs

  19. Tritium in the environment. The IRSN's opinion on key issues and on research and development perspectives

    This report states the opinion of the IRSN on issues related to the behaviour of tritium in the environment, and to the associated risks. This report is based on a set of studies and researches performed on this radionuclide. Thus, the authors address the status of knowledge on the evolution of tritium released by nuclear activities (measurement techniques), the risk of bioaccumulation of tritium by living organisms within ecosystems (behaviour of tritium in the atmosphere, in soils, in ground plants, in continental and sea aquatic media), and the knowledge of risks due to tritium absorbed by living organisms (dose assessment, knowledge of tritium harmful effects and relative biological effectiveness)

  20. Elemental tritium deposition and conversion in the terrestrial environment

    Studies were undertaken to determine the deposition and conversion of atmospheric elemental tritium in soils and vegetation. In the field tritium deposition velocities ranged between 0.007 and 0.07 cm s-1 during the summer and autumn and were less than 0.0005 cm s-1 during the winter. Deposition velocity was found to depend significantly on soil water content, total pore space and organic content in controlled laboratory experiments. In contrast to soils, exposure of vegetation to atmospheric elemental tritium resulted in negligible uptake and conversion in foliage. These studies are of significance to the assessment of behaviour and impact of elemental tritium releases

  1. Tritium removing device

    Tritium-containing gases in a reactor container are discharged to a gas pressurizer and the gases pressurized there are sent to the primary side of a tritium separation device under a high or low pressure. Polyimide polymer separation membranes having selective permeability to elemental tritium and tritium vapor are coated in the tritium separation device. The separation device is divided into primary and secondary sides by the separation membranes and the pressure in the secondary side is lowered by a vacuum pump, etc. Tritium contained in the tritium-containing gases passes through the separation membranes selectively to be moved into the secondary side. Accordingly, tritium is treated in the elemental form and equipments for regeneration such as an adsorption column, etc. are no more necessary and the space can be saved due to minimization of the removing device. Further, since tritium can be removed continuously without storing a great amount of tritium, it is preferable in view of safety. (T.M.)

  2. Monitoring for Tritium at Low Levels

    Tritium presents little external radiation hazard because the beta rays it emits have very low energy. However, the tritium produced by neutron capture in heavy water reactors is in the form of tritiated water which, if it escapes into the air, may be absorbed into the body through the lungs and skin. Direct measurement of tritiated water vapour (HTO) in air at levels below one (mpc)a is difficult with existing installed or portable monitors. However, an indirect measurement can be obtained by making frequent routine determinations of the amount of tritium absorbed by the workers who are chronically exposed. In addition, individual doses may be estimated directly. For this approach to area monitoring to be useful to a health physicist, rapid, direct measurement of HTO intake is needed. Analyses for tritium carried out in the laboratory usually involve delays, so an automatic urinalyser has been developed. It will detect body burdens of tritium resulting from chronic exposures to 0.05 (mpc)a. Urine is voided into a urinal attached to the analyser, a sample is automatically metred and mixed with liquid scintillator and the tritium activity is then measured. The assay takes less than two minutes. Performance of the instrument is automatically checked by processing standard and background samples. This method may also be used for monitoring planned acute exposures. The sensitivity is 5 (mpc)ah, although accurate measurements can oily be made when the HTO has dispersed throughout the body, one to two hours after the exposure. Tritium hazards at specific places may be monitored by collecting airborne HTO in water bubblers and then measuring the collected tritium with the urinalyser. Concentrations of HTO in air down to 0.01 (mpc)a may be detected. Since no manual processing of urine or other aqueous samples is needed the instrument is well suited to those nuclear installations, such as power stations, where minimal laboratory back-up facilities are available. (author)

  3. Stack and area tritium monitoring systems for the tokamak fusion test reactor (TFTR)

    Pearson, G.G.; Meixler, L.D.; Sissingh, R.A.P.

    1991-01-01

    TFTR Tritium Stack and Area Monitoring Systems have been developed to provide the required level of reliability in a cost effective manner consistent with the mission of the Tritium Handling System on TFTR. Personnel protection, environmental responsibility, and tritium containing system integrity have been the considerations in system design. During the Deuterium-Tritium (D-T) experiments on TFTR, tritium will be used for the first time as one of the fuels. All of the tritium bearing systems will have potentially releasable inventories. Although the tritium inventories (total on-site inventory is limited to 50,000 Ci) are low, the consequences of a release may still be significant. For that reason, a thorough TFTR tritium monitoring program has been initiated. 4 refs., 2 figs.

  4. Investigation of tritium in groundwater at Pickering NGS

    Ontario Power Generation Inc. (OPG) investigated tritium in groundwater at the Pickering Nuclear Generating Station (PNGS). The objectives of the study were to evaluate and define the extent of radio-nuclides, primarily tritium, in groundwater, investigate the causes or sources of contamination, determine impacts on the natural environment, and provide recommendations to prevent future discharges. This paper provides an overview of the investigations conducted in 1999 and 2000 to identify the extent of the tritium beneath the site and the potential sources of tritium released to the groundwater. The investigation and findings are summarized with a focus on unique aspects of the investigation, on lessons learned and benefits. Some of the investigative techniques discussed include process assessments, video inspections, hydrostatic and tracer tests, Helium 3 analysis for tritium age dating, deuterium and tritium in soil analysis. The investigative techniques have widespread applications to other nuclear generating stations. (author)

  5. Using the Tritium Plasma Experiment to evaluate ITER PFC safety

    The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 1019 ions/cm2 · s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment

  6. Environmental control of tritium use at the Tokamak Fusion Test Reactor (TFTR)

    Howe, H.J. Jr.; Lind, K.E.

    1978-12-01

    A primary objective of the Tokamak Fusion Test Reactor Project (TFTR) is to demonstrate the production of fusion energy using the deuterium--tritium fusion reaction in a magnetically confined plasma system. This paper will discuss the various tritium control methods employed to minimize the release of tritium to the environment. The methods to be described include the containment and ALAP philosophy, engineered safety features, redundant tritium cleanup systems, redundant instrumentation and control systems, interlocks, monitoring systems, management controls, and waste handling systems. Estimates will be included concerning the impact of routine and accidental tritium releases with these control systems in place.

  7. Environmental control of tritium use at the Tokamak Fusion Test Reactor (TFTR)

    A primary objective of the Tokamak Fusion Test Reactor Project (TFTR) is to demonstrate the production of fusion energy using the deuterium--tritium fusion reaction in a magnetically confined plasma system. This paper will discuss the various tritium control methods employed to minimize the release of tritium to the environment. The methods to be described include the containment and ALAP philosophy, engineered safety features, redundant tritium cleanup systems, redundant instrumentation and control systems, interlocks, monitoring systems, management controls, and waste handling systems. Estimates will be included concerning the impact of routine and accidental tritium releases with these control systems in place

  8. Tritium: an underestimated health risk- 'ACROnic du nucleaire' nr 85, June 2009

    After having indicated how tritium released in the environment (under the form of tritiated water or gas) is absorbed by living species, the author describes the different biological effects of ionizing radiations and the risk associated with tritium. He evokes how the radiation protection system is designed with respect to standards, and outlines how the risk related to tritium is underestimated by different existing models and standards. The author discusses the consequences of tritium transmutation and of the isotopic effect

  9. On electrochemical tritium production

    This paper reports tritium formed in LiOD-D2O solutions in which Pd cathodes are used to evolve D2. Electrolysis was carried out for up to 4 1/2 months. Excess heat has been observed from 5 electrodes out of 28, tritium in 15 out of 53 but 9 out of 13 if the electrodes are limited to 1 mm diameter. Steady state tritium concentrations were 104-107 disintegrations min-1 ml-1. A weak correlation may exist between heat observed and tritium produced. The rate of production of tritium was ca 1010 atoms cm-2 s-1. The branching ratio of tritium to neutrons was ∼108. A theoretical dendrite enhanced fusion model is suggested. Growing gas layer breakdown occurs at sufficiently high surface potential dendrite tips and correspondingly fusion reactions occur. The model gives quantitative consistence with experiment, especially the sporadic nature and the observed branching ratio. (author)

  10. Tritium monitoring techniques

    As part of their operations, the U.S. Navy is required to store or maintain operational nuclear weapons on ships and at shore facilities. Since these weapons contain tritium, there are safety implications relevant to the exposure of personnel to tritium. This is particularly important for shipboard operations since these types of environments can make low-level tritium detection difficult. Some of these ships have closed systems, which can result in exposure to tritium at levels that are below normally acceptable levels but could still cause radiation doses that are higher than necessary or could hamper ship operations. This report describes the state of the art in commercial tritium detection and monitoring and recommends approaches for low-level tritium monitoring in these environments

  11. Confinement and Tritium Stripping Systems for APT Tritium Processing

    This report identifies functions and requirements for the tritium process confinement and clean-up system (PCCS) and provides supporting technical information for the selection and design of tritium confinement, clean-up (stripping) and recovery technologies for new tritium processing facilities in the Accelerator for the Production of Tritium (APT). The results of a survey of tritium confinement and clean-up systems for large-scale tritium handling facilities and recommendations for the APT are also presented

  12. FDMH - The tritium model in RODOS

    intensively use of interdisciplinary research. It is developed in a modular structure with a variable time grid according with the physical processes. During the release phase, the transfer processes are modelled with a half hour time step using real time meteorological data, whereas in the next few days weather forecast data are used at a 2-3 hour interval. In the long term prognosis, a site specific synoptic data file is used and the transfer rates are weekly or monthly averaged. Different from other models, using generic transfer parameters or parameters fitted on individual experiments, FDMH derives tritium transfer rates based on physical and physiological process analysis, using scientifically accepted results from interdisciplinary research on among others, land-atmosphere interaction, water cycle in the soil-plant-atmosphere system, plant physiology, photosynthesis and growth and hydrogen metabolism in mammals. A unique feature of FDMH is the coherent modelling of tritium uptake by plant canopies and its conversion to organic matter, using a physiological plant parameters data base which can reproduce plant growth under various pedo-climatic conditions. Furthermore, in order to predict tritium transfer in animal products, in the absence of a complete experimental data base, results from basic research on hydrogen metabolism in mammals is applied. Due to this novel approach, FDMH can be easily customised for any European site and can predict the time evolution of tritiated water or organically bound tritium in up to 22 plants, 12 animal products, 35 foodstuff and the public dose for 7 population groups. The code is developed not only under the HP-UNIX platform for RODOS but also as a stand alone PC version which can be easily upgraded for PSA studies in CANDU reactors. Preliminary validation tests of FDMH show remarkable agreement with recent experimental data on tritium transfer in cereals and potato as well as in cow's milk. Future effort is related to customise the

  13. Characterisation of redundant tritium light devices

    Gaseous tritium light devices (GTLDs) are currently used widely as long lasting totally independent sources of illumination. Although tritium is of low radiological significance particularly when in gaseous form, because of their widespread use they could give rise to hazardous situations if action is not taken to provide a sensible recycling and disposal route for redundant devices. As a first step to developing this treatment process a number of GTLDs have been destructively examined to determine the amount and speciation of the remaining tritium. This report covers a further investigation sponsored by HMIP which reviewed the production process for GTLDs to identify a typical GTLD type from which a set of specimens with known ages could be selected. These were then subjected to destructive analysis to measure the total tritium, its speciation and the conditions necessary to effect the release of absorbed tritium. The data provided by the analysis programme has been used in a review of treatment process options for handling redundant GTLDs which ranged from long term storage, tritium recovery and recycle to disposal. In addition the results have been used to assess the possible hazards which could arise from the accidental disposal of typical GTLD packages to an open refuse site. (author)

  14. A New Solid State Tritium Surface Monitor

    Traditionally the amount of tritium on a surface is determined by swiping the surface with a material such as filter paper and counting the removed tritium by scintillation. While effective, this method can be time consuming, can alter the surface, only measures removable tritium and produces radioactive waste. For a given application each of these considerations may or may not be a disadvantage. A solid state monitor, on the other hand, has the potential to provide rapid analysis, not alter the surface, measure all tritium on a surface and produce little or not radioactive waste. This allure has promoted open wall ion chamber and PIN diode-based tritium surface monitor development, and these techniques have enjoyed certain success. Recently the first tests were performed with an avalanche photodiode (APD) for surface tritium measurement. While quite similar in concept to PIN diode based measurements, side-by-side testing showed that the APD provided substantially better counting efficiency. Considerations included count rate, background, sensitivity, stability and effect of ambient light. Of particular importance in the US, the APD was able to measure concentrations down to the 'free release' limit, i.e., the concentration below which items can be removed from radiological control areas

  15. Tritium in metals

    In this Chapter a review is given of some of the important features of metal tritides as opposed to hydrides and deuterides. After an introduction to the topics of tritium and tritium in metals information will be presented on a variety of metal-tritium systems. Of main interest here are the differences from the classic hydrogen behavior; the so called isotope effect. A second important topic is that of aging effects produced by the accumulation of 3He in the samples. (orig.)

  16. Tritium Research Laboratory safety analysis report

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment

  17. Tritium Research Laboratory safety analysis report

    Wright, D.A.

    1979-03-01

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment. (ERB)

  18. A metabolic derivation of tritium transfer factors in animal products

    Tritium is a potentially important environmental contaminant arising from the nuclear industry. Because tritium is an isotope of hydrogen, its behaviour in the environment is controlled by the behaviour of hydrogen. Chronic releases of tritium to the atmosphere, in particular, will result in tritium-to-hydrogen (T/H) ratios in plants and animals that are more or less in equilibrium with T/H ratios in the air moisture. Tritium is thus a potentially important contaminant of plant and animal food products. The transfer of tritium from air moisture to plants is quite well understood. In contrast, although a number of regulatory agencies have published transfer coefficient values for diet tritium transfer for a limited number of animal products, a fresh evaluation of these transfers needs to be made In this paper we present an approach for the derivation of tritium transfer coefficients which is based on the metabolism of hydrogen in animals in conjunction with experimental data on tritium transfer. The derived transfer coefficients separately account for transfer to and from free (i.e. water) and organically bound tritium. The predicted transfer coefficients are compared to available data independent of model development. Agreement is good, with the exception of the transfer coefficient for transfer from tritiated water to organically bound tritium in ruminants, which may be attributable to the particular characteristics of ruminant digestion. We show that transfer coefficients will vary in response to the metabolic status of an animal (e.g. stage of lactation, digestibility of diet, etc.) and that the use of a single transfer coefficient from diet to animal product is not appropriate for tritium. It is possible to derive concentration ratio values which relate the concentration of tritiated water and organically bound tritium in an animal product to the corresponding concentrations in the animals diet. These concentration ratios are shown to be less subject to

  19. Tritium dynamics in soils and plants at a tritium processing facility in Canada

    The dynamics of tritium released as tritiated water (HTO) have been studied extensively with results incorporated into environmental models such as CSA N288.1 used for regulatory purposes in Canada. The dispersion of tritiated gas (HT) and rates of oxidation to HTO have been studied under controlled conditions, but there are few studies under natural conditions. HT is a major component of the tritium released from a gaseous tritium light manufacturing facility in Canada (CNSC INFO-0798). To support the improvement of models, a garden was set up in one summer near this facility in a spot with tritium in air averaging ∼ 5 Bq/m3 HTO (passive diffusion monitors). Atmospheric stack releases (575 GBq/week) were recorded weekly. HT releases occur mainly during working hours with an HT:HTO ratio of 2.6 as measured at the stack. Soils and plants (leaves/stems and roots/tubers) were sampled for HTO and organically-bound tritium (OBT) weekly. Active day-night monitoring of air was conducted to interpret tritium dynamics relative to weather and solar radiation. The experimental design included a plot of natural grass/soil, contrasted with grass (sod) and Swiss chard, pole beans and potatoes grown in barrels under different irrigation regimes (in local topsoil at 29 Bq/L HTO, 105 Bq/L OBT). All treatments were exposed to rain (80 Bq/L) and atmospheric releases of tritium (weekdays), and reflux of tritium from soils (initial conditions of 284 Bq/L HTO, 3,644 Bq/L OBT) from 20 years of operations. Three irrigation regimes were used for barrel plants to mimic home garden management: rain only, low tritium tap water (5 Bq/L), and high tritium well water (mean 10,013 Bq/L). This design provided a range of plants and starting conditions with contrasts in initial HTO/OBT activity in soils, and major tritium inputs from air versus water. Controls were two home gardens far from any tritium sources. Active air monitoring indicated that the plume was only occasionally present for

  20. Tritium dynamics in soils and plants at a tritium processing facility in Canada

    Mihok, S.; St-Amanat, N.; Kwamena, N.O. [Canadian Nuclear Safety Commission (Canada); Clark, I.; Wilk, M.; Lapp, A. [University of Ottawa (Canada)

    2014-07-01

    The dynamics of tritium released as tritiated water (HTO) have been studied extensively with results incorporated into environmental models such as CSA N288.1 used for regulatory purposes in Canada. The dispersion of tritiated gas (HT) and rates of oxidation to HTO have been studied under controlled conditions, but there are few studies under natural conditions. HT is a major component of the tritium released from a gaseous tritium light manufacturing facility in Canada (CNSC INFO-0798). To support the improvement of models, a garden was set up in one summer near this facility in a spot with tritium in air averaging ∼ 5 Bq/m{sup 3} HTO (passive diffusion monitors). Atmospheric stack releases (575 GBq/week) were recorded weekly. HT releases occur mainly during working hours with an HT:HTO ratio of 2.6 as measured at the stack. Soils and plants (leaves/stems and roots/tubers) were sampled for HTO and organically-bound tritium (OBT) weekly. Active day-night monitoring of air was conducted to interpret tritium dynamics relative to weather and solar radiation. The experimental design included a plot of natural grass/soil, contrasted with grass (sod) and Swiss chard, pole beans and potatoes grown in barrels under different irrigation regimes (in local topsoil at 29 Bq/L HTO, 105 Bq/L OBT). All treatments were exposed to rain (80 Bq/L) and atmospheric releases of tritium (weekdays), and reflux of tritium from soils (initial conditions of 284 Bq/L HTO, 3,644 Bq/L OBT) from 20 years of operations. Three irrigation regimes were used for barrel plants to mimic home garden management: rain only, low tritium tap water (5 Bq/L), and high tritium well water (mean 10,013 Bq/L). This design provided a range of plants and starting conditions with contrasts in initial HTO/OBT activity in soils, and major tritium inputs from air versus water. Controls were two home gardens far from any tritium sources. Active air monitoring indicated that the plume was only occasionally present for

  1. Behaviour of tritium in the vacuum vessel of JT-60U

    Kobayashi, K.; Miya, N.; Ikeda, Y. [JT-60 Safety Assessment Group, JAEA, Mukoyama (Japan); Torikai, Y. [Hydrogen Isotope Research Center, University of Toyama, Gofuku (Japan); Saito, M.; Alimov, V. [ITER Project Management Group, JAEA, Mukoyama (Japan)

    2015-03-15

    The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D{sub 2} (92.8 %) - T{sub 2} (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily.

  2. Behaviour of tritium in the vacuum vessel of JT-60U

    The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D2 (92.8 %) - T2 (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily

  3. Studies on behavior of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

    The confinement and removal of tritium are the key subjects for safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell building, tritium is often released, as vapor and is in contact with the inner walls. Also those of an ITER tritium plant building will be exposed to tritium in an accident. However, the data are scarce, especially on the penetration of tritium into the concrete of the wall materials. The tritium released in the buildings is removed by the Atmosphere Detritiation Systems (ADS), where the tritium is oxidized by catalysts and is removed as water. Special gas of SF6 is used in ITER, and is expected to be released in an accident such as fire. Although the SF6 gas has the potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed yet. Tritiated water is produced in the regeneration process of ADS, and is subsequently processed by the ITER Water Detritiation System (WDS). One of the key components of WDS is an electrolysis cell. The electrolysis cell is made of organic compounds, and there is no data on the durability of the cell exposed to tritium. To overcome these issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R and D task: 1) tritium behavior in concrete; 2) effect of SF6 on performance of ADS; and 3) tritium durability of electrolysis cell of ITER-WDS. 1) The tritiated water vapor penetrated into the concrete up to 2 cm from the surface only in two months' exposure. The penetration rate of tritium in the concrete was thus appreciably large, so that it is required to evaluate the effect of the lining on the penetration rate quantitatively from the actual tritium tests. 2) The SF6 gas decreased the detritiation factor of ADS. Since the effect of the SF6 depends on its concentration closely, the amount of SF6 released into the tritium handling area in an accident should be deduced by some ideas of the arrangement of

  4. Tritium liquid effluents from the Krsko NPP

    In the past, 12-months' fuel cycles in the Krsko NPP had not caused any problems regarding compliance with its Technical Specifications and license limits on liquid tritium releases (20 TBq/year, 8 TBq/three months). The first 18-months' fuel cycle, which was introduced in 2004, required fuel with higher enrichment, higher boron concentration in the primary coolant and more fuel rods with burnable poisons. In 2005, the NPP operated without refueling outage for the whole year and produced the highest amount of energy so far. Due to these facts and a few unplanned shutdowns and power reductions, production of tritium and releases increased strongly in 2005. As a result, the Krsko NPP hardly succeeded to stay within regulatory limits on tritium releases. However, the three-months' limit was exceeded in the first quarter of 2006. On the basis of conclusions acquired from the SNSA's study and practice of other European countries the SNSA considerably increased the annual limit of permitted liquid tritium releases (from 20 TBq to 45 TBq) and abolished the three-months' limit. At the same time, the SNSA reduced the limit of fission and activation products by halves. (author)

  5. A small volume tritium monitor and fill system

    Environmental concerns demand that all tritium releases be minimized. In particular an experiment recently fielded at Los Alamos National Laboratory required a 4 mm3 volume to be dynamically filled to a pressure of 7 MPa, while keeping the total tritium content (and potential release) below 20 Ci (1 Ci=3.7x1010 Bq). We describe the system which was designed and built for this purpose, complete with a low volume fiberoptic monitor and pressure transducers. (orig.)

  6. Tapentadol immediate release: a new treatment option for acute pain management

    Afilalo, Marc

    2010-01-01

    Marc Afilalo1, Jens-Ulrich Stegmann2, David Upmalis31Sir Mortimer B. Davis Jewish General Hospital, Montréal, Canada; 2Global Drug Safety, Grünenthal GmbH, Aachen, Germany; 3Johnson & Johnson Pharmaceutical Research and Development, L.L.C., Raritan, New Jersey, USAAbstract: The undertreatment of acute pain is common in many health care settings. Insufficient management of acute pain may lead to poor patient outcomes and potentially life-threatening complications. O...

  7. Tritium Issues in Next Step Devices

    Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

  8. Tritium: An analysis of key environmental and dosimetric questions

    This document summarizes new theoretical and experimental data that may affect the assessment of environmental releases of tritium and analyzes the significance of this information in terms of the dose to man. Calculated doses resulting from tritium releases to the environment are linearly dependent upon the quality factor chosen for tritium beta radiation. A reevaluation of the tritium quality factor by the ICRP is needed; a value of 1.7 would seem to be more justifiable than the old 1.0 value. A new exposure model is proposed, based primarily upon the approach recommended by the National Council on Radiation Protection and Measurements. Employing a /open quotes/typical/close quotes/ LMFBR reprocessing facility source term, a /open quotes/base case/close quotes/ dose commitment to total body (for a maximally exposed individual) was calculated to be 4.0 /times/ 10/sup /minus/2/ mSv, with 3.2 /times/ 10/sup /minus// mSv of the dose due to intake of tritium. The study analyzes models which exist for evaluating the buildup of global releases of tritium from man-made sources. Scenarios for the release of man-made tritium to the environment and prediction of collective dose commitment to future generations suggest that the dose from nuclear weapons testing will be less than that from nuclear energy even though the weapons source term is greater than that for any of our energy scenarios

  9. Tritium: An analysis of key environmental and dosimetric questions

    Till, J E; Meyer, H R; Etnier, E L; Bomar, E S; Gentry, R D; Killough, G G; Rohwer, P S; Tennery, V J; Travis, C C

    1980-05-01

    This document summarizes new theoretical and experimental data that may affect the assessment of environmental releases of tritium and analyzes the significance of this information in terms of the dose to man. Calculated doses resulting from tritium releases to the environment are linearly dependent upon the quality factor chosen for tritium beta radiation. A reevaluation of the tritium quality factor by the ICRP is needed; a value of 1.7 would seem to be more justifiable than the old 1.0 value. A new exposure model is proposed, based primarily upon the approach recommended by the National Council on Radiation Protection and Measurements. Employing a /open quotes/typical/close quotes/ LMFBR reprocessing facility source term, a /open quotes/base case/close quotes/ dose commitment to total body (for a maximally exposed individual) was calculated to be 4.0 /times/ 10/sup /minus/2/ mSv, with 3.2 /times/ 10/sup /minus// mSv of the dose due to intake of tritium. The study analyzes models which exist for evaluating the buildup of global releases of tritium from man-made sources. Scenarios for the release of man-made tritium to the environment and prediction of collective dose commitment to future generations suggest that the dose from nuclear weapons testing will be less than that from nuclear energy even though the weapons source term is greater than that for any of our energy scenarios.

  10. Tritium Attenuation by Distillation

    The objective of this study was to determine how a 100 Area distillation system could be used to reduce to a satisfactory low value the tritium content of the dilute moderator produced in the 100 Area stills, and whether such a tritium attenuator would have sufficient capacity to process all this material before it is sent to the 400 Area for reprocessing

  11. Atmospheric tritium 1968-1984. Tritium Laboratory data report No. 14

    Tritium in the form of water, HTO, from the atmospheric testing of nuclear devices in the 60s has now mainly disappeared from the atmosphere and entered the ocean. The additions of such tritium from Chinese and French tests in the 70s were observed but did not make a big impression on the diminishing inventory of atmospheric HTO. Tritium in elemental form, HT, went through a maximum in the mid 70s, apparently primarily as a results of some underground testing of large nuclear devices and releases from civilian and military nuclear industry. The mid 70s maximum was 1.3 kg of tritium in this form, and in 1984 0.5 kg remain. The disappearance is slower than the decay rate of tritium, so sources must still have been present during this time. The global distribution shows, not unexpectedly, smaller inventory in the Southern Hemisphere across the equator and thus southward transport of HT. The chemical lifetime of hydrogen gas in the atmosphere, assuming the elemental tritium being in the form of HT, not T2, has been estimated between 6 and 10 years. It is to be expected that increasing activity of nuclear fuel reprocessing would in the near future again increase the global tritium gas inventory. Tritium in the form of light hydrocarbons, primarily methane, has also been measured, and in this form a quantity of 200 g of tritium resided in the global atmosphere 1956 to 1976. By 1982 it had decreased to 50 g. 25 refs., 5 figs., 11 tabs

  12. Role of lysosomal enzymes released by alveolar macrophages in the pathogenesis of the acute phase of hypersensitivity pneumonitis

    J. L. Pérez-Arellano

    1995-01-01

    Full Text Available Hydrolytic enzymes are the major constituents of alveolar macrophages (AM and have been shown to be involved in many aspects of the inflammatory pulmonary response. The aim of this study was to evaluate the role of lysosomal enzymes in the acute phase of hypersensitivity pneumonitis (HPs. An experimental study on AM lysosomal enzymes of an HP-guinea-pig model was performed. The results obtained both in vivo and in vitro suggest that intracellular enzymatic activity decrease is, at least partly, due to release of lysosomal enzymes into the medium. A positive but slight correlation was found between extracellular lysosomal activity and four parameters of lung lesion (lung index, bronchoalveolar fluid total (BALF protein concentration, BALF LDH and BALF alkaline phosphatase activities. All the above findings suggest that the AM release of lysosomal enzymes during HP is a factor involved, although possibly not the only one, in the pulmonary lesions appearing in this disease.

  13. The sythetic endomorphin-1 analog, CYT-1010, inhibits sensory neuropeptide release, acute neurogenic inflammation and heat injury-induced thermal hyperalgesia in rodent models

    Z. Helyes; J. Szolcsanyi; T. Maione

    2011-01-01

    Calcitonin gene-related peptide (CGRP) and substance P released from capsaicin-sensitive afferents induce neurogenic inflammatory and nociceptive actions. Since we have shown that the m opioid receptor agonist endomorphin-1 inhibits sensory neuropeptide outflow, the effects of its synthetic, peptidase-resistant analog, CYT-1010, was studied on CGRP release, acute neurogenic inflammation and thermal hyperalgesia. CGRP release from sensory fibres of isolated rat tracheae was evoked by electrica...

  14. Tritium Systems Test Assembly operator training program

    Proper operator training is needed to help ensure the safe operation of fusion facilities by personnel who are qualified to carry out their assigned responsibilities. Operators control and monitor the Tritium Systems Test Assembly (TSTA) during normal, emergency, and maintenance phases. Their performance is critical both to operational safety, assuring no release of tritium to the atmosphere, and to the successful simulation of the fusion reaction progress. Through proper training we are helping assure that TSTA facility operators perform their assignments in a safe and efficient manner and that the operators maintain high levels of operational proficiency through continuing training, retraining, requalification, and recertification

  15. Studies on the behaviour of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

    Confinement and the removal of tritium are key subjects for the safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell, tritium is often released as vapour and is in contact with the inner walls. The inner walls of the ITER tritium plant building will also be exposed to tritium in an accident. The tritium released in the buildings is removed by the atmosphere detritiation systems (ADS), where the tritium is oxidized by catalysts and is removed as water. A special gas of SF6 is used in ITER and is expected to be released in an accident such as a fire. Although the SF6 gas has potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed as yet. Tritiated water is produced in the regeneration process of ADS and is subsequently processed by the ITER water detritiation system (WDS). One of the key components of the WDS is an electrolysis cell. To overcome the issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R and D task: (1) tritium behaviour in concrete; (2) the effect of SF6 on the performance of ADS and (3) tritium durability of the electrolysis cell of the ITER-WDS. (1) The tritiated water vapour penetrated up to 50 mm into the concrete from the surface in six months' exposure. The penetration rate of tritium in the concrete was thus appreciably first, the isotope exchange capacity of the cement paste plays an important role in tritium trapping and penetration into concrete materials when concrete is exposed to tritiated water vapour. It is required to evaluate the effect of coating on the penetration rate quantitatively from the actual tritium tests. (2) SF6 gas decreased the detritiation factor of ADS. Since the effect of SF6 depends closely on its concentration, the amount of SF6 released into the tritium handling area in an accident should be reduced by some ideas of arrangement of components in the buildings. (3) It was expected

  16. Studies on behavior of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

    The confinement and removal of tritium are the key subjects for safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell, tritium is often released, as vapor and is in contact with the inner walls. Also those of an ITER tritium plant building will be exposed to tritium in an accident. The tritium released in the buildings is removed by the Atmosphere Detritiation Systems (ADS), where the tritium is oxidized by catalysts and is removed as water. Special gas of SF6 is used in ITER, and is expected to be released in an accident such as fire. Although the SF6 gas has the potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed yet. Tritiated water is produced in the regeneration process of ADS, and is subsequently processed by the ITER Water Detritiation System (WDS). One of the key components of WDS is an electrolysis cell. To overcome the issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER R and D task: 1) tritium behavior in concrete; 2) effect of SF6 on performance of ADS; and 3) tritium durability of electrolysis cell of ITER-WDS. 1) The tritiated water vapor penetrated up to 50 mm into the concrete from the surface in six months' exposure. The penetration rate of tritium in the concrete was thus appreciably first, the isotope exchange capacity of cement paste plays an important role at tritium trapping and penetration in concrete materials when concrete is exposed to tritiated water vapor. It is required to evaluate the effect of the lining on the penetration rate quantitatively from the actual tritium tests. 2) The SF6 gas decreased the detritiation factor of ADS. Since the effect of the SF6 depends on its concentration closely, the amount of SF6 released into the tritium handling area in an accident should be deduced by some ideas of the arrangement of components in the buildings. 3) It was expected that the electrolysis cell of ITER

  17. Uptake of tritium through foliage in capsicum fruitescens, L

    Tritium uptake and release patterns throuogh foliage in Capsicum fruitescens, L. were investigated using twelve potted plants, under different conditions of exposure and release. The plants studied belonged to two age groups, 3 months and 5 months. The average half residence time for the species was found to be 42.6 min, on the basis of treating the entire group of plants as a single cluster. The individual release rates showed a variation of up to a factor of two, for half residence time values (Tsub(1/2)). The second component was not easily resolvable in most of the cases. Tissue bound tritium showed interesting uptake patterns. The ratios between tissue bound tritium and tissue free water tritium concentrations indicated regular mode of uptake with well defined rate constants in the case of long exposure periods. (author)

  18. Study on oxidation of hydrogen over commercial catalyst for tritium recovery system

    For the establishment of the D-T fusion reactor technology, recovery of tritium released into the working area of fusion power plants is quite important. When tritium leaks to working areas, the last barrier is the wall of the building. Due to higher diffusion coefficient of tritium, it diffuses through the wall and would be readily liberated to the environment. Thus, the tritium recovery system is indispensable for the D-T fusion reactor. The objective of the present study is to develop the advanced technology of the tritium recovery system. In the near future, deuterium plasma discharge experiments scheduled be conducted with Large Helical Device (LHD) in National Institute for Fusion Science. A small amount of tritium is produced by D-D reaction in LHD. Tritium in plasma exhaust gases and process gas during discharge needs to be recovered, and thus the design and construction of the tritium recovery system used for that purpose is a matter of considerable urgency. The tritium recovery system usually consists of catalysts and adsorbents, which is the most conventional and reliable method for removing tritium that is accidentally released into the working area of these facilities. However, more recent and advanced type of catalysts on the market cannot be directly applied to the design of tritium recovery system, because of paucity of design data for tritium recovery system. In this study, the authors performed oxidation experiments of hydrogen over a catalyst. The experiments were performed by changing various experimental parameters.

  19. Exotic development of ceramic tritium breeding materials

    In the near future fusion reactors will be based on the tritium-deuterium plasma reaction. As such the production of tritium, a non-natural element, becomes of crucial importance in fusion technology. An experimental programme, EXOTIC, is in progress since 1983 in which the laboratories of SNL-Springfields, ECN-Petten, JRC-Petten and SCK-CEN Mol work in close collaboration within the framework of the manufacture, characterization and irradiation of ceramic lithium compounds. This programme must result in the understanding of the tritium release processes and the effect of material characteristics on this release. Up to now three irradiation experiments in the High Flux Reactor-Petten were scheduled (EXOTIC I, II and III). The Annual Progress Report 1985 summarizes information on these experiments during the period of 1985. The reader is also referred to the previous Annual Progress Report 1984. During the EXOTIC I experiment lithium silicate pellets and lithium aluminate pellets were irradiated. The resulting tritium release data are still to be interpreted in full detail. Some preliminary observations are presented in this Report. In the EXOTIC II experiment lithium oxide, lithium aluminate and lithium silicate pellets were used. In the EXOTIC III experiment lithium oxide, lithium zirconate and lithium silicate pellets were used. (Author)

  20. Role of cardiac volume receptors in the control of ADH release during acute simulated weightlessness in man

    Convertino, V. A.; Benjamin, B. A.; Keil, L. C.; Sandler, H.

    1984-01-01

    Hemodynamic responses and antidiuretic hormone (ADH) were measured during body position changes, designed to induce central blood volume shifts in ten cardiac and one heart-lung transplant recipients, to assess the contribution of cardiac volume receptors in the control of ADH release during the initial acute phase of exposure to weightlessness. Each subject underwent 15 min of a sitting-control period (C) followed by 30 min of 6 deg headdown tilt (T) and 30 min of resumed sitting (S). Venous blood samples and cardiac dimensions were taken at 0 and 15 min of C; 5, 15, and 30 min of T; and 5, 15, and 30 min of S. Blood samples were analyzed for hematocrit, plasma osmolality, plasma renin activity (PRA), and ADH. Heart rate and blood pressure were recorded every two min. Plasma osmolality was not altered by posture changes. Mean left ventricular end-diastolic volume increased (P less than 0.05) from 90 ml in C to 106 ml in T and returned to 87 ml in S. Plasma ADH was reduced by 20 percent (P less than 0.05) with T, and returned to control levels with S. These responses were similar in six normal cardiac-innervated control subjects. These data may suggest that cardiac volume receptors are not the primary mechanism for the control of ADH release during acute central volume shifts in man.

  1. Leptin enhances the release of cytokines by peripheral blood mononuclear cells from acute multiple sclerosis patients

    2006-01-01

    Objective To explore the effect of leptin on cytokine production by PBMCs obtained from MS patients either in acute (relapse) or in stable (nonrelapse) phase of disease. Methods PBMCs were collected from 25 untreated acute MS patients, 11 stable MS patients and 20 healthy controls. PBMCs were cultured either with RPMI-1640 alone or with leptin (1.25 nmol/ml), phytohemagglutinin (PHA) ( 100 μg/ml), and leptin + PHA. 72 h later the supernate of the culture medium were collected and stored at -70℃. The pro-inflammatory cytokine (IFN-γ) concentration were determined using an enzyme-linked immunosorbent assay ( ELISA), and the anti-inflammatory cytokine (IL-4) concentration were investigated by radioimmunity methods. Results Our data showed that leptin induced IFN-γproduction by PBMCs of patients in an acute phase of disease but not in a stable phase or in healthy controls. Moreover, we found that PHA induced IL-4 production by PBMCs of patients in an acute phase of disease, but leptin inhibited this ability of PHA. Conclusion Leptin can affect on pro- and anti-inflammatory cytokine production by PBMCs collected from MS patients, may be this connected with leptin increase the susceptiveness of MS.

  2. Modeling of tritium behavior in ceramic breeder materials

    Computer models are being developed to predict tritium release from candidate ceramic breeder materials for fusion reactors. Early models regarded the complex process of tritium release as being rate limited by a single slow step, usually taken to be tritium diffusion. These models were unable to explain much of the experimental data. We have developed a more comprehensive model which considers diffusion and desorption from the grain surface. In developing this model we found that it was necessary to include the details of the surface phenomena in order to explain the results from recent tritium release experiments. A diffusion-desorption model with a desorption activation energy which is dependent on the surface coverage was developed. This model provided excellent agreement with the results from the CRITIC tritium release experiment. Since evidence suggests that other ceramic breeder materials have desorption activation energies which are dependent on surface coverage, it is important that these variations in activation energy be included in a model for tritium release. 17 refs., 12 figs

  3. Interactions of tritium and materials

    Yamawaki, Michio; Yamaguchi, Kenji; Tanaka, Satoru; Ono, Futaba (Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.); Yamamoto, Takuya

    1993-11-01

    In D-T burning fusion reactors, problems related to tritium-material interactions are vitally important. From this point of view, plasma-material interactions, blanket breeder material-tritium interactions, safety aspects of tritium-material interactions and tritium storage materials are reviewed with emphasis on the works going on in the authors' laboratories. (author) 83 refs.

  4. Tritium breeding in fusion reactors

    Abdou, M.A.

    1982-10-01

    Key technological problems that influence tritium breeding in fusion blankets are reviewed. The breeding potential of candidate materials is evaluated and compared to the tritium breeding requirements. The sensitivity of tritium breeding to design and nuclear data parameters is reviewed. A framework for an integrated approach to improve tritium breeding prediction is discussed with emphasis on nuclear data requirements.

  5. A review of tritium conversion reactions

    The chemical processes by which elemental tritium can be converted to tritiated water have been examined by reviewing the available literature on these processes. It would appear that gas phase conversion reactions at room temperature are slow and that they do not contribute significantly to any observed conversion following releases of elemental tritium. The effects of surfaces are not clearly understood. Metals, however, can increase the rate over the gas phase processes, but the magnitude of this increase is not well documented. Further work is necessary to examine the effects of various materials, elevated temperatures, and other parameters on conversion reactions in order to more closely reflect conditions in reactor buildings and other tritium containing facilities

  6. TFTR tritium inventory analysis

    Pontau, A.E.; Brice, D.K.; Buchenauer, D.A.; Causey, R.A.; Doyle, B.L.; Hsu, W.L.; Lee, S.R.; McGrath, R.T.; Mills, B.; Wampler, W.R.; Wilson, K.L. (Sandia National Labs., Livermore, CA (USA); Sandia National Labs., Albuquerque, NM (USA)); Langley, R. (Oak Ridge National Lab., TN (USA)); Dylla, H.F.; Heifetz, D.B.; Kilpatrick, S.; Lamarche, P.H.; Sissingh, R.A.P.; Ulrickson, M. (Princeton Univ., NJ (USA). Plasma Physics Lab.); Brooks, J.N. (Argonne National Lab., IL (USA))

    1989-06-01

    The Tokamak Fusion Test Reactor (TFTR) is scheduled to begin D-T operation in 1990 with the on-site tritium inventory limited to 5 grams. The physics and chemistry of the in-vessel tritium inventory will impact safety concerns, and also the entire operating schedule of the tokamak. We have investigated plasma-material interaction processes that will affect this first tritium-fueled tokamak. Tritium inventory estimates for TFTR are derived from: (1) Laboratory simulation, (2) in-situ plasma measurements, (3) post-run surface analysis, and (4) modeling. This paper presents the results of these investigations, the derivation of a tritium inventory estimate and its uncertainties, and a discussion of its impact. A particular discharge-by-discharge operating schedule has been developed and evaluated. The major source of in-vessel tritium inventory will be codeposition of tritium and eroded carbon onto surfaces. We find that the on-site limit may be approached unless specific inventory reduction techniques are invoked, e.g., discharge cleaning. (orig.).

  7. Tritium technology. A Canadian overview

    An overview of the various tritium research and operational activities in Canada is presented. These activities encompass tritium processing and recovery, tritium interactions with materials, and tritium health and safety. Many of these on-going activities form a sound basis for the tritium use and handling aspects of the ITER project. Tritium management within the CANDU heavy water reactor, associated detritiation facilities, research and development facilities, and commercial industry and improving the understanding of tritium behaviour in humans and the environment remain the focus of a long-standing Canadian interest in tritium. While there have been changes in the application of this knowledge and experience over time, the operating experience and the supporting research and development continue to provide for improved plant and facility operations, an improved understanding of tritium safety issues, and improved products and tools that facilitate tritium management. (author)

  8. Tritium management in PWR fuel reprocessing plants

    Activity, quantity and nature of tritium compounds obtained during head end process (cutting and dissolution) are determined to estimate environmental release hazards in fuel reprocessing plants. Measurements on representative PWR reactor fuels (burnup 33,000 MWdt-1, specific power 30 MW dt-1) show that about 60% of the tritium produced in the reactor diffuses in the cladding where it is fixed. Remaining tritium stays in the irradiated oxide and is found as tritiated water in the solution obtained during fuel dissolution. In the UP3 plant at La Hague (France) tritiated water is disposed into the sea without environmental problems. In the case of a reprocessing plant far from the sea, the PUREX process is slightly modified for concentration of tritium in a limited amount of water (TRILEX process). It is verified experimentally in αβγ lab on actual fuel and by simulation at the pilot seale that the supplementary step ''tritium washing'' of the solvent can be obtained in pulsed columns. 4 tables, 7 figs

  9. Tritium facility at TFTR

    Sissingh, R.A.P. (Canadian Fusion Fuels Technology Project, Mississauga, ON (Canada)); Rossmassler, R.L. (Princeton Univ., NJ (USA). Plasma Physics Lab.)

    1990-06-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton began operation in December 1982. While it has operated successfully with protium and deuterium achieving record energy confinement time and record ion temperatures. TFTR's ultimate goal is to achieve the scientific break even point of Q=1 for which deuterium/tritium injection is needed. This paper will discuss the design parameters resulting from using tritium as a fuel, the design and operating philosophies employed, the additional systems and equipment required, the effect on the heating, ventilating and air conditioning systems, the tritium monitoring system and the personnel training. (orig.).

  10. Tritium powered luminescent signs

    A tritium powered emergency exit sign is provided comprising an elongated tritium powered light tube mounted at the focus of a parabolic reflector coated with a phosphorescent coating. The tube and reflector are mounted in an elongate trough shaped housing, with a translucent cover. In an emergency, e.g. in the event of a power failure in a building or other structure, the phosphorescent coating will reinforce the light emitted by the tritium tube for a short period whilst the eyes adjust to the darkness. (author)

  11. Simulation of physical parameter for in-pipe tritium breeder

    It is necessary to build in-pipe tritium breeder in our country in order to assess breeder material of tritium breeder module (TBM) and to find the release law of tritium. The irradiation vessel is one of the key components of TBM. The physical parameters about in-pipe tritium breeder were simulated with MCNP code. The values of the self-shielding factor, equivalent cross-section, daily production of tritium and total heating power are separately 0.435, 1.09 x 10-22 cm2, 2.8 x 1010 Bq and 8.2 kW. And they would provide necessary data for designing the irradiation vessel. (authors)

  12. Estimation of dose to man from environmental tritium

    Factors important for characterization of tritium in environmental pathways leading to exposure of man are reviewed and quantification of those factors is discussed. Parameters characterizing the behavior of tritium in man are also subjected to review. Factors to be discussed include organic binding, bioaccumulation, quality factor and transmutation. A variety of models are presently in use to estimate dose to man from environmental releases of tritium. Results from four representative models are compared and discussed. Site-specific information is always preferable when parameterizing models to estimate dose to man. There may be significant differences in dose potential among geographic regions due to variable factors. An example of one such factor examined is absolute humidity. It is concluded that adequate methodologies exist for estimation of dose to man from environmental tritium although a number of areas are identified where additional tritium research is desirable

  13. Updating the tritium quality factor: the argument for conservatism

    Estimated doses resulting from tritium releases to the environment are linearly dependent upon the quality factor (Q) chosen for tritium beta radiation. In 1969 the International Commission on Radiological Protection (ICRP) recommended using 1 as the Q for all low energy beta radiation. Considerable improvements have been made in evaluating exposures to tritium at very low dose rates and in refining physiological and biological endpoints since the 1969 ICRP recommendations. This study summarizes recent experiments to determine the relative biological effectiveness of tritium. Based upon our study of published data related to quality factor, its importance in the calculation of dose, and the currently accepted conservative philosophy in radiation protection, it is concluded that a value of 2 would seem to be more defensible for environmental assessments and that a reevaluation of the tritium quality factor by the ICRP is needed

  14. Tritium inventory measurements using calorimetry

    In the past calorimetry has been developed as a powerful tool in radiometrology. Calorimetric methods have been applied for the determination of activities, half lives and mean energies released during the disintegration of radioactive isotopes. The fundamental factors and relations which determine the power output of radioactive samples are presented and some basic calorimeter principles are discussed in this paper. At the Kernforschungszentrum Karlsruhe (KfK) a family of 3 calorimeters has been developed to measure the energy release from radiative waste products arising from reprocessing operations. With these calorimeters, radiative samples with sizes from a few cm3 to 2 ·105 cm3 and heat ratings ranging from a few nW to kW can be measured. After modifications of tits inner part the most sensitive calorimeter among the three calorimeters mentioned above would be best suited for measuring the tritium inventory in T-getters of the Amersham-type

  15. Experience in handling concentrated tritium

    The notes describe the experience in handling concentrated tritium in the hydrogen form accumulated in the Chalk River Nuclear Laboratories Tritium Laboratory. The techniques of box operation, pumping systems, hydriding and dehydriding operations, and analysis of tritium are discussed. Information on the Chalk River Tritium Extraction Plant is included as a collection of reprints of papers presented at the Dayton Meeting on Tritium Technology, 1985 April 30 - May 2

  16. Problems of anthropogenic tritium limitation

    Kochetkov О.A.

    2013-12-01

    Full Text Available This article contains the current situation in respect to the environmental concentrations of anthropogenic and natural tritium. There are presented and analyzed domestic standards for НТО of all Radiation Safety Standards (NRB, as well as the regulations analyzed for tritium in drinking water taken in other countries today. This article deals with the experience of limitation of tritium and focuses on the main problem of rationing of tritium — rationing of organically bound tritium.

  17. Tritium-management requirements for D-T fusion reactors (ETF, INTOR, FED)

    The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design

  18. Tritium interactions with steel and construction materials in fusion devices

    The literature on the interactions of tritium and tritiated water with metals, glasses, ceramics, concrete, paints, polymers and other organic materials is reviewed in this report Some of the processes affecting the amount of tritium found on various materials, such as permeation, sorption and the conversion of tritium found on various materials, such as permeation, sorption and conversion of elemental tritium (T2) to tritiated water (HTO), are also briefly outlined. Tritium permeation in steels is fairly well understood, but effects of surface preparation and coatings on sorption are not yet clear. Permeation of T2 into other metals with cleaned surfaces has been studied thoroughly at high temperature, and the effect of surface oxidation has also been explored. The room-temperature permeation rates of low-permeability metals with cleaned surfaces are much faster than indicated by high-temperature results, because of grain-boundary diffusion. Elastomers have been studied to a certain extent, but some mechanisms of interaction with tritium gas and sorbed tritium are unclear. Ceramics have some of the lowest sorption and permeation rates, but ceramic coatings on stainless steels do not lower permeation or tritium as effectively as coatings obtained by oxidation of the steel, probably because of cracking caused by differences in thermal expansion coefficient. Studies on concrete are in their early stages; they show that sorption of tritiated water on concrete is a major concern in cleanup of releases of elemental tritium into air in tritium handling facilities. Some of the codes for modelling releases and sorption of T2 and HTO contain unproven assumptions about sorption and T2 → HTO conversion. Several experimental programs will be required in order to clear up ambiguities in previous work and to determine parameters for materials which have not yet been investigated. (146 refs., tab.)

  19. Method of removing tritium

    Purpose: To remove trituim in airs simply and reliably in a large amount. Constitution: Tritium contained in air is oxidized in an oxidizing column into water and incorporated in the air. The water-air mixture is caused to flow into and cooled in a first freezing type air drier where almost of tritium water in the air are condensated and separated from the air and, after falling through the drier, recovered by way of a drain tube. The air passing through the freezing type air drier in humidified by a humidifier and then caused to flow into the second freezing drier. Then, a slight amount of tritium water remained in the air is mixed with steams by the humidifier for easier separation, dried in a drier and removed with tritium into cleaned air. After properly humidifying the air in the humidifier, it is flown out through the exit. (Kamimura, M.)

  20. TFTR tritium program

    Sissingh, R.A.P.; Rossmassler, R.L.

    1988-09-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton began operation in December 1982. Since then it has operated successfully with protium and deuterium achieving energy confinement time at peak electron density of 10/sup 19/ m/sup -3/s, with ion temperatures of 20 keV. This paper describes the systems and preparations required for D-T operation, i.e. introducing and operating the tokamak with tritium in order to achieve the scientific break even point of Q=1. These systems include the tritium storage and delivery system, the tritium injection systems, the tritium clean-up systems, and the plasma exhaust and collection systems. It is expected that TFTR will have these systems fully operational, with trained personnel.

  1. Tritium waste package

    Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

    1995-01-01

    A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

  2. Tritium catalyzed deuterium tokamaks

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the 3He from the D(D,n)3He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general)

  3. Tritium protective clothing

    Occupational exposures to radiation from tritium received at present nuclear facilities and potential exposures at future fusion reactor facilities demonstrate the need for improved protective clothing. Important areas relating to increased protection factors of tritium protective ventilation suits are discussed. These areas include permeation processes of tritium through materials, various tests of film permeability, selection and availability of suit materials, suit designs, and administrative procedures. The phenomenological nature of film permeability calls for more standardized and universal test methods, which would increase the amount of directly useful information on impermeable materials. Improvements in suit designs could be expedited and better communicated to the health physics community by centralizing devlopmental equipment, manpower, and expertise in the field of tritium protection to one or two authoritative institutions

  4. Tritium effects on germ cells and fertility

    Primordial oocytes in juvenile mice show acute gamma-ray LD50 as low as 6 rad. This provides opportunities for determining dose-response relations at low doses and chronic exposure in the intact animal - conditions of particular interest for hazard evaluation. Examined in this way, 3HOH in body water is found to kill murine oocytes exponentially with dose, the LD50 level for chronic exposure being only 2μCi/ml (delivering 0.4 rad/day). At very low doses and dose rates, where comparisons between tritium and other radiations are of special significance for radiological protection, the RBE of tritium compared with 60Co gamma radiation reaches approximately 3. Effects on murine fertility from tritium-induced oocyte loss have been quantified by reproductive capacity measurements. Chronic low-level exposure has been examined also in three primate species - squirrel, rhesus, and bonnet monkeys. In squirrel monkeys the ovarian germ-cell supply is 99% destroyed by the time of birth from prenatal exposure to body-water levels of 3HOH (administered in maternal drinking water) of only 3 μCi/ml, the LD50 level being 0.5 μCi/ml (giving 0.1 rad/day), one fourth that in mice. Though not completely ruled out, similar high sensitivity of female germ cells has not been found in macaques; and it probably does not occur in man. The exquisite radiosensitivity of primordial oocytes in mice is apparently due to vulnerability of the plasma membrane (or something of similar geometry and location), not DNA. Evidence for this comes from tritium data as well as neutron studies. Tritium administered as 3HOH, and therefore generally distributed, is much more effective in killing murine oocytes than is tritium administered as 3H-TdR, localized in the nucleus. This situation in the mouse may have implications for estimating radiation genetic risk in the human female

  5. PRODUCTION OF TRITIUM

    Jenks, G.H.; Shapiro, E.M.; Elliott, N.; Cannon, C.V.

    1963-02-26

    This invention relates to a process for the production of tritium by subjecting comminuted solid lithium fluoride containing the lithium isotope of atomic mass number 6 to neutron radiation in a self-sustaining neutronic reactor. The lithium fiuoride is heated to above 450 deg C. in an evacuated vacuum-tight container during radiation. Gaseous radiation products are withdrawn and passed through a palladium barrier to recover tritium. (AEC)

  6. Tritium application: self-luminous glass tube(SLGT)

    Kim, K.; Lee, S.K.; Chung, E.S.; Kim, K.S.; Kim, W.S. [Nuclear Power Lab., Korea Electric Power Research Inst. (KEPRI), Daejeon (Korea); Nam, G.J. [Engineering Information Technology Center, Inst. for Advanced Engineering (IAE), Kyonggi-do (Korea)

    2005-07-01

    To manufacture SLGTs (self-luminous glass tubes), 4 core technologies are needed: coating technology, tritium injection technology, laser sealing/cutting technology and tritium handling technology. The inside of the glass tubes is coated with greenish ZnS phosphor particles with sizes varying from 4{proportional_to}5 [{mu}m], and Cu, and Al as an activator and a co-dopant, respectively. We also found that it would be possible to produce a phosphor coated glass tube for the SLGT using the well established cold cathode fluorescent lamp (CCFL) bulb manufacturing technology. The conceptual design of the main process loop (PL) is almost done. A delicate technique will be needed for the sealing/cutting of the glass tubes. Instead of the existing torch technology, a new technology using a pulse-type laser is under investigation. The design basis of the tritium handling facilities is to minimize the operator's exposure to tritium uptake and the emission of tritium to the environment. To fulfill the requirements, major tritium handling components are located in the secondary containment such as the glove boxes (GBs) and/or the fume hoods. The tritium recovery system (TRS) is connected to a GB and PL to minimize the release of tritium as well as to remove the moisture and oxygen in the GB. (orig.)

  7. Establishment of tritium dating facility for hydrological studies in PNRI

    The release of excess tritium (3H) into the atmosphere from nuclear weapons tests conducted between 1952 and 1963 'tagged' rain water, and thereby all surface waters with 3HHO. Measurement of 3H concentrations in rain, surface water and groundwater is useful index of vulnerability and sustainability of the aquifer to pollution and human exploitation. These determinations are currently being used in the characterization of different environments and in pollution studies, in the framework of research projects, international collaborations and services. Liquid scintillation counting (LSC) was the method of choice for the evaluation of the tritium concentrations in precipitation, groundwater and surface water samples. Prior to counting process, the samples are enriched in tritium by an electrolysis procedure to improve the overall detection limit. Low-level hydrological water samples go through an electrolytic enrichment step, in which tritium concentrations are increased to about seventy-fold through volume reduction. The amount of tritium in water is expressed in tritium units (TU). Water samples taken from selected areas of Bulacan province within the period of 2007 and 2008 were analyzed as part of the current hydrological studies being done by our group in PNRI. The typical tritium values for the rain water, surface water, and groundwater were found to be 1.20±0.11 TU, 1.12±0.11 TU, and 0.40±0.07, respectively. Procedures are now available in our laboratory for measurement of tritium in water samples of different water types. (author)

  8. Tritium in the Savannah River Site environment. Revision 1

    Murphy, C.E. Jr.; Bauer, L.R.; Hayes, D.W.; Marter, W.L.; Zeigler, C.C.; Stephenson, D.E.; Hoel, D.D.; Hamby, D.M.

    1991-05-01

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found.

  9. Tritium Removal by Laser Heating and Its Application to Tokamaks

    A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm2, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed

  10. Tritium stripping in a nitrogen glovebox using SAES St 198

    SAES metal getter material St 198 was chosen for glovebox stripper tests to evaluate its effectiveness of removing tritium from a nitrogen atmosphere. The St 198 material is unique from a number of other metal hydride-based getter materials in that it is relatively inert to nitrogen and can thus be used in nitrogen glovebox atmospheres. Six tritium stripper experiments which mock-up the use of a SAES St 198 stripper bed for a full-scale (10,500 liter) nitrogen glovebox have been completed. Experiments consisted of a release of small quantity of protium/deuterium spiked with tritium which were scaled to simulate tritium releases of 0.1 g., 1.0 g., and 10 g. into the glovebox. The tritium spike allows detection using tritium ion chambers. The St 198 stripper system produced a reduction in tritium activity of approximately two orders of magnitude in 24 hours (6--8 atmosphere turn-overs) of stripper operation

  11. Spatial distribution of tritium in the Rawatbhata Rajasthan site environment

    Tritium is one of the most environmentally mobile radionuclides and hence has high potential for migration into the different compartments of environment. Tritium from nuclear facilities at RAPS site is released into the environment through 93 m and 100 m high stack mainly as tritiated water (HTO). The released tritium undergoes dilution and dispersion and then follows the ecological pathway of water molecule. Environmental Survey Laboratory of Health Physics Division, Bhabha Atomic Research Centre (BARC), located at Rajasthan Atomic Power Station (RAPS) site is continuously monitoring the concentration of tritium in the environment to ensure the public safety. Atmospheric tritium activity during the period (2009-2013) was measured regularly around Rajasthan Atomic Power Station (RAPS). Data collected showed a large variation of H-3 concentration in air fluctuating in the range of 0.43 - 5.80 Bq.m-3 at site boundary of 1.6 km. This paper presents the result of analyses of tritium in atmospheric environment covering an area up to 20 km radius around RAPS site. Large number of air moisture samples were collected around the RAPS site, for estimating tritium in atmospheric environment to ascertain the atmospheric dispersion and computation of radiation dose to the public

  12. LIBRETTO-4: Understanding and modeling tritium transport under irradiation

    LIBRETTO-4 irradiation experiment with in situ T production in eutectic alloy LiPb inside EUROFER97 capsules was performed at the High Flux Reactor (HFR), Petten. Two identical capsules at two temperature ranges (500-550 deg. C and 300-350 deg. C) tested the tritium permeation rate through pre-oxidized ferritic-martensitic EUROFER97 into a He + 1000 H2 vppm cooling channel. A second channel with similar flowing gas composition obtained the tritium released from the LiPb and permeation through the internal wall of austenitic steel AISI316L. A 2-dimensional mathematical model has been developed. The model simulates tritium fluxes in capsule materials for variable experimental conditions (generation rates, gas pressures and flow rates) for a 27 days cycle. Modeling provides a consistent physical picture of release-rate mechanisms within the capsule. Tritium transport data under irradiation in capsule's material have been derived.

  13. The design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

    The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time; unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

  14. Glovebox stripper system tritium capture efficiency-literature review

    James, D. W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Poore, A. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-28

    Glovebox Stripper Systems (GBSS) are intended to minimize tritium emissions from glovebox confinement systems in Tritium facilities. A question was raised to determine if an assumed 99% stripping (decontamination) efficiency in the design of a GBBS was appropriate. A literature review showed the stated 99% tritium capture efficiency used for design of the GBSS is reasonable. Four scenarios were indicated for GBSSs. These include release with a single or dual stage setup which utilizes either single-pass or recirculation for stripping purposes. Examples of single-pass as well as recirculation stripper systems are presented and reviewed in this document.

  15. Tritium analyses of COBRA-1A2 beryllium pebbles

    Baldwin, D.L. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    Selected tritium measurements have been completed for the COBRA-1A2 experiment C03 and D03 beryllium pebbles. The completed results, shown in Tables 1, 2, and 3, include the tritium assay results for the 1-mm and 3-mm C03 pebbles, and the 1-mm D03 pebbles, stepped anneal test results for both types of 1-mm pebbles, and the residual analyses for the stepped-anneal specimens. All results have been reported with date-of-count and are not corrected for decay. Stepped-anneal tritium release response is provided in addenda.

  16. Comparison of tritium production facilities

    Detailed investigation and research on the source of tritium, tritium production facilities and their comparison are presented based on the basic information about tritium. The characteristics of three types of proposed tritium production facilities, i.e., fissile type, accelerator production tritium (APT) and fusion type, are presented. APT shows many advantages except its rather high cost; fusion reactors appear to offer improved safety and environmental impacts, in particular, tritium production based on the fusion-based neutron source costs much lower and directly helps the development of fusion energy source

  17. Recent progress on tritium technology research and development for a fusion reactor in Japan Atomic Energy Agency

    JAEA (Japan Atomic Energy Agency) manages 2 tritium handling laboratories: Tritium Processing Laboratory (TPL) in Tokai and DEMO-RD building in Rokkasho. TPL has been accumulating a gram level tritium safety handling experiences without any accidental tritium release to the environment for more than 25 years. Recently, our activities have focused on 3 categories, as follows. First, the development of a detritiation system for ITER. This task is the demonstration test of a wet Scrubber Column (SC) as a pilot scale (a few hundreds m3/h of processing capacity). Secondly, DEMO-RD tasks are focused on investigating the general issues required for DEMO-RD design, such as structural materials like RAFM (Reduced Activity Ferritic/Martensitic steels) and SiC/SiC, functional materials like tritium breeder and neutron multiplier, and tritium. For the last 4 years, we have spent a lot of time and means to the construction of the DEMO-RD facility and to its licensing, so we have just started the actual research program with tritium and other radioisotopes. This tritium task includes tritium accountancy, tritium basic safety research such as tritium interactions with various materials, which will be used for DEMO-RD and durability. The third category is the recovery work from the Great East Japan earthquake (2011 earthquake). It is worth noting that despite the high magnitude of the earthquake, TPL was able to confine tritium properly without any accidental tritium release

  18. Acute turpentine inflammation and kinin release in rat-paw thermic oedema.

    Limãos, E. A.; Borges, D R; Souza-Pinto, J. C.; Gordon, A. H.; Prado, J. L.

    1981-01-01

    Livers from rats at 2-3 days after s.c. injection of turpentine, when perfused, synthesized prekallikrein nearly 3 times faster than did livers from normal rats. On the other hand paw oedema, produced by heating to 46 degrees, in rats injured in this way was less marked. Likewise in such rats the amount of bradykinin release by 50 min. of coaxial perfusion of the paw was only 13.6 +/- 4.6 compared with 63.1 +/- 13.4 ng in normal rats. A possible explanation for the observed reduction in produ...

  19. R and D for tritium removal system using bacteria for safety enhancement of tritium handling in fusion reactor

    In order to safety enhancement of fusion reactor, Japan Atomic Energy Agency (JAEA) investigates effective tritium removal system in case of incidental tritium release in the building. Recently, we paid attention to the bacteria, which had an ability of hydrogen oxidation at room temperature, and a new effective detritiation system using the bacteria has been studied under the collaboration with Ibaraki University, instead of conventional detritiation system using catalytic oxidation reactor of rare metal. (author)

  20. Development of the irradiation assembly used in-pile tritium production

    The irradiation assembly of in-pile tritium production is main part of the first in-pile tritium demonstration apparatus for hybrid reactor in China. The paper describes its principle configuration and specifications. Design calculation are given in which include tritium production calculation, heat transfer calculation and stress calculation. Key technology in development of the assembly is explained. Operation and in-situ tritium release experiments of the assembly show that thermal and nuclear characteristics of the assembly is good. Adjustable temperature range of the tritium breeder is 250∼700 degree C. Radius temperature difference in the tritium breeder is less than 70 degree C. Tritium production efficiency of the assembly under thermal neutron irradiation is 0.183 x 10-7 Bq·cm2. The assembly and its utilization is discussed and reviewed

  1. Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

    The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab

  2. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  3. Modeling acute health risks associated with accidental releases of toxic gases

    Haskin, F.E.; Ding, C.; Summa, K.J. [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Young, M. [Sandia National Labs., Albuquerque, NM (United States). Accident Analysis and Consequence Assessment Dept.

    1996-09-01

    CHEM{_}MACCS has been developed from the radiological accident consequence code, MACCS, to perform probabilistic calculations of potential off-site consequences of the accidental atmospheric release of hazardous chemicals. The principal phenomena considered in CHEM{_}MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways, and early and latent health effects. CHEM{_}MACCS provides the following capabilities: (1) statistical weather sampling data (8,760 hourly data points per year), (2) population dose and health effect risk calculations based on site-specific population data, (3) health effects calculations including the consideration of potential site specific mitigative actions (evacuation and shielding), and (4) modeling of multiple release segments. Three different sample problems are contained in this report to show how to use CHEM{_}MACCS. Three test problems are run to compare CHEM{_}MACCS and D2PC. The doses versus the downwind centerline distances from the source for the given doses are in very close agreement.

  4. Tritium analysis at TFTR

    The tritium analytical system at TFRR is used to determine the purity of tritium bearing gas streams in order to provide inventory and accountability measurements. The system includes a quadrupole mass spectrometer and beta scintillator originally configured at Monsanto Mound Research Laboratory in the late 1970's and early 1980's. The system was commissioned and tested between 1991 and 1992 and is used daily for analysis of calibration standards, incoming tritium shipments, gases evolved from uranium storage beds and measurement of gases returned to gas holding tanks. The low resolution mass spectrometer is enhanced by the use of a metal getter pump to aid in resolving the mass 3 and 4 species. The beta scintillator complements the analysis as it detects tritium bearing species that often are not easily detected by mass spectrometry such as condensable species or hydrocarbons containing tritium. The instruments are controlled by a personal computer with customized software written with a graphical programming system designed for data acquisition and control. A discussion of the instrumentation, control systems, system parameters, procedural methods, algorithms, and operational issues will be presented. Measurements of gas holding tanks and tritiated water waste streams using ion chamber instrumentation are discussed elsewhere

  5. Tritium monitoring in the environment of the French territory

    Leprieur, F.; Roussel-Debet, S.; Pierrard, O.; Tournieux, D.; Boissieux, T.; Caldera-Ideias, P. [Institut de radioprotection et de surete nucleaire (France)

    2014-07-01

    Introduction: Radioactive releases in the environment from civilian and military nuclear facilities have significantly decreased over the last few decades, except for discharges of tritium which are forecast to increase due to changes in the fuel management in power plants and in the longer term by new tritium-emitting units (fusion reactors). In the aim to perform its radiological monitoring mission throughout the French territory, IRSN uses and develops advanced technology equipment to sample and to analyze tritium in the different environmental compartments. Methodology: IRSN uses bubblers to collect both tritium vapour (HTO) and gaseous tritium (mainly HT) in the air. Another method, developed by IRSN, consists in directly sampling the water vapour in the air by condensing in a cold trap and more recently with passive sampler. In continental and marine surface water, samples are usually collected by automatic water samplers. Instantaneous surface water samples are also collected by grab sample devices. In addition, IRSN conducts animal and plant samples near French nuclear facilities. Natural origin and tritium remaining from testing of nuclear weapons In the atmosphere, the background levels of tritium of 1 to 2 Bq/L measured in water vapour, equivalent to an activity of 0.01 to 0.02 Bq/m{sup 3} of air. In fresh waters, the tritium activity currently ranges between 1 and 3 Bq/L of water. In the marine environment, tritium emitted during nuclear weapon tests has been totally 'diluted' in cosmogenic tritium and concentration levels at the surface have remained around 0.1 to 0.2 Bq/L. In biological matrices, total tritium concentration range from 1 to 3 Bq/kg f.w. with a variable proportion of free and organically bounded forms. Tritium around nuclear facilities: Close to facilities releasing more than 2x10{sup 13} Bq/year of gaseous tritium, higher activity levels, ranging from a few tens to a few hundred Bq/L, are observed in the atmospheric and

  6. Efficacy and tolerability of once-daily extended release quetiapine fumarate in acute schizophrenia : A randomized, double-blind, placebo-controlled study

    Kahn, Rene S.; Schulz, S. Charles; Palazov, Veselin D.; Reyes, Efren B.; Brecher, Martin; Svensson, Ola; Andersson, Henrik M.; Meulien, Didier

    2007-01-01

    Objective: To evaluate the efficacy and tolerability of extended release quetiapine fumarate (quetiapine XR) in a 6-week, double-blind, randomized study. Method: Patients with a DSM-IV diagnosis of acute schizophrenia were randomly assigned to fixed-dose quetiapine XR 400, 600, or 800 mg/day (once d

  7. Development of a dynamic compartment model for the prediction of tritium behavior around NPPs

    KAERI has developed a new model to find the relationship between the tritium release rate and tritium concentration in the environment. The model was based upon a dynamic compartment model. In this paper three kinds of global tritium cycling model were compared to estimate the natural background concentration of tritium in Korea. The dry and wet deposition rates were calculated using a computer program called DEPOS to derive a source term. The mechanisms considered for the transfer of tritium between the compartments were evaporation, groundwater flow, infiltration, runoff, and hydrodynamic dispersion. Also, transfer coefficients between the compartments were obtained using realistic geographical data. In order to illustrate the model various release scenarios were developed, and the change of tritium concentration in groundwater and surface water around the nuclear power plants was estimated. (author)

  8. Tritium in the water environment

    Tritium activity concentrations in water environment in China have been summarized. The levels in different water categories can be listed as: precipitation, river, reservoir, tap water, well, sea and spring in the order of decrease of tritium. (5 tabs.)

  9. Carbon monoxide-Releasing Molecule-2 (CORM-2 attenuates acute hepatic ischemia reperfusion injury in rats

    Zhang Weihui

    2010-05-01

    Full Text Available Abstract Background Hepatic ischemia-reperfusion injury (I/Ri is a serious complication occurring during liver surgery that may lead to liver failure. Hepatic I/Ri induces formation of reactive oxygen species, hepatocyte apoptosis, and release of pro-inflammatory cytokines, which together causes liver damage and organ dysfunction. A potential strategy to alleviate hepatic I/Ri is to exploit the potent anti-inflammatory and cytoprotective effects of carbon monoxide (CO by application of so-called CO-releasing molecules (CORMs. Here, we assessed whether CO released from CORM-2 protects against hepatic I/Ri in a rat model. Methods Forty male Wistar rats were randomly assigned into four groups (n = 10. Sham group underwent a sham operation and received saline. I/R group underwent hepatic I/R procedure by partial clamping of portal structures to the left and median lobes with a microvascular clip for 60 minutes, yielding ~70% hepatic ischemia and subsequently received saline. CORM-2 group underwent the same procedure and received 8 mg/kg of CORM-2 at time of reperfusion. iCORM-2 group underwent the same procedure and received iCORM-2 (8 mg/kg, which does not release CO. Therapeutic effects of CORM-2 on hepatic I/Ri was assessed by measuring serum damage markers AST and ALT, liver histology score, TUNEL-scoring of apoptotic cells, NFkB-activity in nuclear liver extracts, serum levels of pro-inflammatory cytokines TNF-α and IL-6, and hepatic neutrophil infiltration. Results A single systemic infusion with CORM-2 protected the liver from I/Ri as evidenced by a reduction in serum AST/ALT levels and an improved liver histology score. Treatment with CORM-2 also up-regulated expression of the anti-apoptotic protein Bcl-2, down-regulated caspase-3 activation, and significantly reduced the levels of apoptosis after I/Ri. Furthermore, treatment with CORM-2 significantly inhibited the activity of the pro-inflammatory transcription factor NF-κB as measured in

  10. Acute turpentine inflammation and kinin release in rat-paw thermic oedema.

    Limãos, E A; Borges, D R; Souza-Pinto, J C; Gordon, A H; Prado, J L

    1981-12-01

    Livers from rats at 2-3 days after s.c. injection of turpentine, when perfused, synthesized prekallikrein nearly 3 times faster than did livers from normal rats. On the other hand paw oedema, produced by heating to 46 degrees, in rats injured in this way was less marked. Likewise in such rats the amount of bradykinin release by 50 min. of coaxial perfusion of the paw was only 13.6 +/- 4.6 compared with 63.1 +/- 13.4 ng in normal rats. A possible explanation for the observed reduction in production of bradykinin may be inhibition of kallikrein due to an increased concentration of alpha 2-macroglobulin. PMID:6173056

  11. Environmental monitoring of molecular tritium

    The oxidation of atmospheric molecular tritium (HT) in vegetation was determined by in vitro experiments for various kinds of woody and herbaceous plant leaves, mosses and lichens taken from a forest and a garden in Ibaraki prefecture and a forest in Gifu prefecture, and comparison of the HT oxidation activity in vegetation was made with those in its neighboring surface soil (0-5cm in depth). The oxidation of HT in woody plant leaves was extremely low, only about 1/10000-1/1000 that in the surface soil as well as herbaceous plant leaves with some exception, whereas HT oxidation in mosses and lichens was 50-500 times that in pine needles. These results suggest the usefulness of mosses and lichens as monitor vegetation for accidental release of HT into the environment. (author)

  12. Tritium - is it underestimated

    Practical experience in the use of the Whitlock Tritium Meter in various laboratories and industrial establishments throughout the world has shown that:-a) Measurements by smear/wipe tests can often be in error by three orders of magnitude or more; b) Sub-visual surface scratches (8μ deep) are radiologically important; c) Volatile forms of tritium exist in 20% to 30% of establishments visited. It is concluded that a) the widespread use of smear/wipe techniques for the assessment of 3H surface contamination based on the assumption that 10% of removable activity is collected by the smear/wipe should be re-examined and b) tritium surface contamination assessed as 'fixed' can contain volatile fractions with a hazard potential which may be considerably greater than the hazard from removable activity at present covered by maximum permissible level recommendations. (H.K.)

  13. Tritium in HTR systems

    Starting from the basis of the radiological properties of tritium, the provisions of present-day radiation protection legislation are discussed in the context of the handling of this radionuclide in HTR plants. Tritium transportation is then followed through from the place of its creation up until the sink, i.e. disposal and/or environmental route, and empirical values obtained in experiments and in plant operation translated into guidelines for plant design and planning. The use of the example of modular HTR plants permits indication that environmental contamination via the 'classical' routes of air and water emissions, and contamination of products, and resulting consumer exposure, are extremely low even on the assumption of extreme conditions. This leads finally to a requirement that the expenditure for implementation of measures for further reduction of tritium activity rates be measured against low radiological effect. (orig.)

  14. Aquatic dispersion modelling of a tritium plume in Lake Ontario

    Approximately 2900 kg of tritiated water, containing 2.3E+15 Bq of tritium, were released to Lake Ontario via the cooling water discharge when a leak developed in a moderator heat exchanger in Unit 1 at the Pickering Nuclear Generating Station (PNGS) on 1992 August 2. The release provided the opportunity to study the dispersion of a tritium plume in the coastal zone of Lake Ontario. Current direction over the two-week period following the release was predominantly parallel to the shore, and elevated tritium concentrations were observed up to 20 km east and 85 km west of the PNGS. Predictions of the tritium plume movement were made using current velocity measurements taken at 8-m depth, 2.5 km offshore from Darlington and using a empirical relationship where alongshore current speed is assumed to be proportional to the alongshore component of the wind speed. The tritium migration was best described using current velocity measurements. The tritium plume dispersion is modelled using the one-dimensional advection-dispersion equation. Transport parameters are the alongshore current speed and longitudinal dispersion coefficient. Longitudinal dispersion coefficients, estimated by fitting the solution of the advection-dispersion equation to measured concentration distance profiles ranged from 3.75 to 10.57 m2s-1. Simulations using the fitted values of the dispersion coefficient were able to describe maximum tritium concentrations measured at water supply plants located within 25 km of Pickering to within a factor of 3. The dispersion coefficient is a function of spatial and temporal variability in current velocity and the fitted dispersion coefficients estimated here may not be suitable for predicting tritium plume dispersion under different current conditions. The sensitivity of the dispersion coefficient to variability in current conditions should be evaluated in further field experiments. (author). 13 refs., 7 tabs., 12 figs

  15. Breeding blanket development; Tritium release from breeder

    土谷 邦彦; 河村 弘; 長尾 美春

    2006-01-01

    核融合炉ブランケットを設計するためには、微小球を用いたブランケット構造体の中性子照射に関する工学的データが必要不可欠である。工学的データのうち、トリチウム生成放出特性は、最も重要なデータの1つである。このため、トリチウム増殖材料の候補材であるチタン酸リチウム(Li2TiO3)微小球からのトリチウム生成放出試験を行い、トリチウム放出特性に対するスイープガス流量,照射温度,スイープガス中の水素添加量,熱中性子束の変化等の効果について調べた。本試験の結果、(1)Li2TiO3微小球充填体の外壁温度が100circC以上になった時、トリチウム放出が観測された。また、充填体の外壁温度が300sim400circCのとき、トリチウム生成・放出率(R/G)は1に到達した。(2)スイープガス流量を100sim900cm3/min(Li2TiO3微小球充填体の空塔速度:0.53sim4.8cm/s)の範囲で変化させても、定常時におけるLi2TiO3微小球充填体からのトリチウム放出に影響はなかった。(3)スイープガス中の水素添加量はトリチウム放出に影響することがわかった。...

  16. Interaction of tritium and helium with lead–lithium eutectic under reactor irradiation

    Highlights: • We studied T and He behavior in Pb-Li eutectic under reactor irradiation. • Temperature dependences of T/He release were obtained for different reactor powers. • We proposed phenomenological models to describe T and He generation and release. • We defined the Arrhenius dependence of constant of tritium capture rate in lithium. - Abstract: This paper describes the study of tritium and helium generation and release from the lead–lithium eutectic under irradiation in the IVG1.M reactor (Institute of Atomic Energy of the Republic of Kazakhstan). The experiments were carried out using the method of mass-spectrometric registration of released gases. Experimental conditions were as follows: the irradiation temperature was from 573 K to 773 K; the reactor power levels were 1, 2 and 6 MW. The study allowed to obtain the temperature dependences of tritium and helium release from the lead–lithium eutectic at different reactor powers (for different neutron fluxes and, respectively, for different rates of helium and tritium generation in material). Phenomenological models were proposed for description of the processes of tritium and helium generation and release from the lead–lithium eutectic. These models allowed us to describe the experimental data very well. Helium release simulation assumed that the flow of helium from the eutectic's surface is linearly dependent on its bulk concentration. For modeling the tritium release the process was divided into two phases: the first one—the yield of tritium atoms on the surface, has been described in the same assumption as for the helium release; and the second phase included a description of the second-order desorption from the surface of the eutectic. All the main parameters of the models, such as the effective release coefficient of tritium and helium atoms on a surface, the effective constant of desorption rate of tritium atoms from the eutectic surface were identified

  17. Interaction of tritium and helium with lead–lithium eutectic under reactor irradiation

    Tazhibayeva, Irina, E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy, National Nuclear Center of the Republic of Kazakhstan, Kurchatov (Kazakhstan); Kulsartov, Timur; Barsukov, Nikolay; Gordienko, Yuri; Ponkratov, Yuri; Zaurbekova, Zhanna; Tulubayev, Eugeniy; Gnyrya, Vyachaslav; Baklanov, Viktor [Institute of Atomic Energy, National Nuclear Center of the Republic of Kazakhstan, Kurchatov (Kazakhstan); Kenzhin, Ergazy [Shakarim Semey State University, Semey (Kazakhstan)

    2014-10-15

    Highlights: • We studied T and He behavior in Pb-Li eutectic under reactor irradiation. • Temperature dependences of T/He release were obtained for different reactor powers. • We proposed phenomenological models to describe T and He generation and release. • We defined the Arrhenius dependence of constant of tritium capture rate in lithium. - Abstract: This paper describes the study of tritium and helium generation and release from the lead–lithium eutectic under irradiation in the IVG1.M reactor (Institute of Atomic Energy of the Republic of Kazakhstan). The experiments were carried out using the method of mass-spectrometric registration of released gases. Experimental conditions were as follows: the irradiation temperature was from 573 K to 773 K; the reactor power levels were 1, 2 and 6 MW. The study allowed to obtain the temperature dependences of tritium and helium release from the lead–lithium eutectic at different reactor powers (for different neutron fluxes and, respectively, for different rates of helium and tritium generation in material). Phenomenological models were proposed for description of the processes of tritium and helium generation and release from the lead–lithium eutectic. These models allowed us to describe the experimental data very well. Helium release simulation assumed that the flow of helium from the eutectic's surface is linearly dependent on its bulk concentration. For modeling the tritium release the process was divided into two phases: the first one—the yield of tritium atoms on the surface, has been described in the same assumption as for the helium release; and the second phase included a description of the second-order desorption from the surface of the eutectic. All the main parameters of the models, such as the effective release coefficient of tritium and helium atoms on a surface, the effective constant of desorption rate of tritium atoms from the eutectic surface were identified.

  18. Production of nuclear fusion reactor fuel by ceramic tritium breeder material

    Fuel tritium is generated from the nuclear reaction between the fusion neutron and the lithium of the breeder material arranged in the blanket that encloses the fusion plasma in the fusion reactor. However, the release process of the generated tritium has not been completely clarified. Recently, Japan Atomic Energy Agency started the tritium generation and recovery experiment in using nuclear fusion neutron source (FNS). In this report, the recent results of study on breeder material and its manufacturing technology is presented. (author)

  19. Environmental Aspects of Tritium Around the Vinca Institute of Nuclear Sciences

    An overview of environmental distribution of tritium around the Institute of Nuclear Sciences Vinca during the period 1988-1994 is presented. Temporal and local variations of the specific tritium variations in precipitation (Usek, Zeleno Brdo), river waters (the Danube, the Sava and Mlaka Creek) as well as atmospheric water vapor are given. Estimates based on precipitation measurements have shown that 6.3 TBq of tritium activity should be released annually into the atmosphere from the Vinca Institute of Nuclear Sciences. (author)

  20. Monitoring of tritium

    Corbett, James A.; Meacham, Sterling A.

    1981-01-01

    The fluid from a breeder nuclear reactor, which may be the sodium cooling fluid or the helium reactor-cover-gas, or the helium coolant of a gas-cooled reactor passes over the portion of the enclosure of a gaseous discharge device which is permeable to hydrogen and its isotopes. The tritium diffused into the discharge device is radioactive producing beta rays which ionize the gas (argon) in the discharge device. The tritium is monitored by measuring the ionization current produced when the sodium phase and the gas phase of the hydrogen isotopes within the enclosure are in equilibrium.

  1. Environmental tritium transport model. Application to the dose evaluation for members of the public

    A model of tritium dispersion and cycling in terrestial ecosystems surrounding an atmospheric source of tritium is presented. The model structure is described. Results of simulations of three release cases are used to illustrate the dynamics of the model. The three cases are: a release of 100% TH, a release of 100% THO, and a release of 50% TH and 50% THO. The TH release is characterized by having a lower total retention of tritium in the ecosystem than the THO case. However, a greater fraction (but lower absolute amount) of the tritium in the TH release is found in the organic component of vegetation and the animals that feed on the vegetation. 24 references, 1 table

  2. Modeling tritium behavior in Li{sub 2}ZrO{sub 3}

    Billone, M.C. [Argonne National Lab., IL (United States). Fusion Power Program

    1998-03-01

    Lithium metazirconate (Li{sub 2}ZrO{sub 3}) is a promising tritium breeder material for fusion reactors because of its excellent tritium release characteristics. In particular, for water-cooled breeding blankets (e.g., ITER), Li{sub 2}ZrO{sub 3} is appealing from a design perspective because of its good tritium release at low operating temperatures. The steady-state and transient tritium release/retention database for Li{sub 2}ZrO{sub 3} is reviewed, along with conventional diffusion and first-order surface desorption models which have been used to match the database. A first-order surface desorption model is recommended in the current work both for best-estimate and conservative (i.e., inventory upper-bound) predictions. Model parameters are determined and validated for both types of predictions, although emphasis is placed on conservative design predictions. The effects on tritium retention of ceramic microstructure, protium partial pressure in the purge gas and purge gas flow rate are discussed, along with other mechanisms for tritium retention which may not be dominant in the experiments, but may be important in blanket design analyses. The proposed tritium retention/release model can be incorporated into a transient thermal performance code to enable whole-blanket predictions of tritium retention/release during cyclic reactor operation. Parameters for the ITER driver breeding blanket are used to generate a numerical set of model predictions for steady-state operation. (author)

  3. Measurement of tritium penetration through concrete material covered by various paints coating

    The present study aims at obtaining fundamental data on tritium migration in porous materials, which include soaking effect, interaction between tritium and cement paste coated with paints and transient tritium sorption in porous cement. The amounts of tritium penetrated into or released from cement paste with epoxy and urethane paint coatings were measured. The tritium penetration amounts were increased with the HTO (tritiated water) exposure time. Time to achieve a saturated value of tritium sorption was more than 60 days for cement paste coated with epoxy paint and with urethane paint, while that for cement paste without any paint coating took 2 days to achieve it. The effect of tritium permeation reduction by the epoxy paint was higher than that of the urethane. Although their paint coatings were effective for reduction of tritium penetration through the cement paste which was exposed to HTO for a short period, it was found that the amount of tritium trapped in the paints became large for a long period. Tritium penetration rates were estimated by an analysis of one-dimensional diffusion in the axial direction of a thickness of a sample. Obtained data were helpful for evaluation of tritium contamination and decontamination. (authors)

  4. Accelerator Production of Tritium Programmatic Environmental Impact Statement Input Submittal

    The Programmatic Environmental Impact Statement for Tritium Supply and Recycling considers several methods for the production of tritium. One of these methods is the Accelerator Production of Tritium. This report summarizes the design characteristics of APT including the accelerator, target/blanket, tritium extraction facility, and the balance of plant. Two spallation targets are considered: (1) a tungsten neutron-source target and (2) a lead neutron-source target. In the tungsten target concept, the neutrons are captured by the circulating He-3, thus producing tritium; in the lead target concept, the tritium is produced by neutron capture by Li-6 in a surrounding lithium-aluminum blanket. This report also provides information to support the PEIS including construction and operational resource needs, waste generation, and potential routine and accidental releases of radioactive material. The focus of the report is on the impacts of a facility that will produce 3/8th of the baseline goal of tritium. However, some information is provided on the impacts of APT facilities that would produce smaller quantities

  5. Accelerator Production of Tritium Programmatic Environmental Impact Statement Input Submittal

    Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Greene, G.A. [Brookhaven National Lab., Upton, NY (United States); Boyack, B.E. [Los Alamos National Lab., NM (United States)

    1996-02-01

    The Programmatic Environmental Impact Statement for Tritium Supply and Recycling considers several methods for the production of tritium. One of these methods is the Accelerator Production of Tritium. This report summarizes the design characteristics of APT including the accelerator, target/blanket, tritium extraction facility, and the balance of plant. Two spallation targets are considered: (1) a tungsten neutron-source target and (2) a lead neutron-source target. In the tungsten target concept, the neutrons are captured by the circulating He-3, thus producing tritium; in the lead target concept, the tritium is produced by neutron capture by Li-6 in a surrounding lithium-aluminum blanket. This report also provides information to support the PEIS including construction and operational resource needs, waste generation, and potential routine and accidental releases of radioactive material. The focus of the report is on the impacts of a facility that will produce 3/8th of the baseline goal of tritium. However, some information is provided on the impacts of APT facilities that would produce smaller quantities.

  6. Statistical correlation of environmental tritium values at Trombay

    Tritium releases from a 40 MW(th) D2O moderated reactor (Cirus) at Trombay and the environmental tritium concentrations (ambient air and vegetation samples) at different ground stations have been measured over a period of five years and the statistical correlation among the different sets of values were assessed. Distribution patterns proved to be gaussian in nature with identifiable skewness, caused by occasional larger release. The effect of the mean, median and the most probable values on dilution rate factors were found to be significant. The correlation among the sets of data showing tritium concentrations in release air, the ambient air and vegetation samples at different stations were found to be statistically good, and very nearly perfect. (U.K.)

  7. Prediction and distribution of tritium levels in primary sodium in PFBR

    Tritium is a radioactive isotope of hydrogen with half life of 12.3 years. It is a beta emitting gas with an average energy of 5.7 keV. In FBRs tritium originates principally by ternary fission in fuel and by neutron reaction with B4C in control rods and shields. Additional sources of tritium arise due to neutron reaction with lithium and boron impurities in the fuel and coolant. Due to chemical and physical trapping only 13% of the tritium is released to the coolant sodium from boron carbide where as 99% of the tritium produced in the fuel escapes out of the cladding and mixes to coolant. Tritium release to environment is of concern. Hence its distribution in the reactor and release into atmosphere is important. Tritium is removed from the primary sodium by cold traps and diffusion through the reactor vessel. Accumulation of tritium in primary and secondary cold traps during the purification of the sodium results in a significant radioactive source and its estimation is relevant to regeneration aspects of the cold traps. In this report, equilibrium level of tritium in primary and secondary sodium is estimated

  8. A uranium bed with ceramic body for tritium storage

    Khapov, A.S.; Grishechkin, S.K.; Kiselev, V.G. [' All Russia Research Institute of Automatics' - FSUE VNIIA, Moscow (Russian Federation)

    2015-03-15

    It is widely recognized that ceramic coatings provide an attractive solution to lower tritium permeation in structural materials. Alumina based ceramic coatings have the highest permeation reduction factor for hydrogen. For this reason an attempt was made to apply crack-free low porous ceramics as a structural material of a bed body for tritium storage in a setup used for hydrogenating neutron tube targets at VNIIA. The present article introduces the design of the bed. This bed possesses essentially a lower hydrogen permeation factor than traditionally beds with stainless steel body. Bed heating in order to recover hydrogen from the bed is suggested to be implemented by high frequency induction means. Inductive heating allows decreasing the time necessary for tritium release from the bed as well as power consumption. Both of these factors mean less thermal power release into glove box where a setup for tritium handling is installed and thus causes fewer problems with pressure regulations inside the glove box. Inductive heating allows raising tritium sorbent material temperature up to melting point. The latter allows achieving nearly full tritium recovery.

  9. A uranium bed with ceramic body for tritium storage

    It is widely recognized that ceramic coatings provide an attractive solution to lower tritium permeation in structural materials. Alumina based ceramic coatings have the highest permeation reduction factor for hydrogen. For this reason an attempt was made to apply crack-free low porous ceramics as a structural material of a bed body for tritium storage in a setup used for hydrogenating neutron tube targets at VNIIA. The present article introduces the design of the bed. This bed possesses essentially a lower hydrogen permeation factor than traditionally beds with stainless steel body. Bed heating in order to recover hydrogen from the bed is suggested to be implemented by high frequency induction means. Inductive heating allows decreasing the time necessary for tritium release from the bed as well as power consumption. Both of these factors mean less thermal power release into glove box where a setup for tritium handling is installed and thus causes fewer problems with pressure regulations inside the glove box. Inductive heating allows raising tritium sorbent material temperature up to melting point. The latter allows achieving nearly full tritium recovery

  10. ARIES-I tritium system

    A key safety concern in a D-T fusion reactor is the tritium inventory. There are three components in a fusion reactor with potentially large inventories, i.e., the blanket, the fuel processing system and the plasma facing components. The ARIES team selected the material combinations, decided the operating conditions and refined the processing systems, with the aiming of minimizing the tritium inventories and leakage. The total tritium inventory for the ARIES-I reactor is only 700 g. This paper discussed the calculations and assumptions we made for the low tritium inventory. We also addressed the uncertainties about the tritium inventory. 13 refs., 2 figs., 3 tabs

  11. Natural attentuation of tritium in vadose zone moisture and ground water at a Lawrence Livermore site in Northern California, USA

    Tritium used in explosives experiments and buried in unlined landfills at a remote Lawrence Livermore National Laboratory (LLNL) site has resulted in three ground water tritium plumes. Using an innovative approach, we determined that despite ground water tritium activities of up to 1.5 million picoCuries per liter (pCi/L) in some locations, natural attenuation processes are significantly limiting the migration of tritium to environmental receptors. We used soil vapor and moisture tritium activity measurements to calculate the source inventory of tritium in the vadose zone. We determined the 12 year annual inventory of tritium in ground water, using objective tritium activity contours and the highly variable saturated thickness of the aquifer. Our analysis indicates that despite seasonal slug releases of tritium, the two plumes emanating from two landfills are stable, with the 1,000 and 20,000 pCi/L contours essentially fixed in space. The third plume emanates continuously from an explosives testing platform; the 1,000 pCi/L contour is translating slightly, but the 20,000 pCi/L contour is retreating upgradient towards the source. Additionally, the long-term trend in total tritium activity for each plume is decreasing. Three processes account for the attenuation of tritium observed: 1) radioactive decay, 2) hydrodynamic dispersion, and 3) dwindling tritium sources. In preparation for the possibility that remediation may be required anyway, we have evaluated innovative remediation technologies for tritium at this site

  12. Tritium breeding materials

    Tritium breeding materials are essential to the operation of D-T fusion facilities. Both of the present options - solid ceramic breeding materials and liquid metal materials are reviewed with emphasis not only on their attractive features but also on critical materials issues which must be resolved

  13. Tritium retention in TFTR

    This report discusses the materials physics related to D-T operation in TFTR. Research activities are described pertaining to basic studies of hydrogenic retention in graphite, hydrogen recycling phenomena, first-wall and limiter conditioning, surface analysis of TFTR first-wall components, and estimates of the tritium inventory

  14. Modeling of tritium transport in lithium aluminate fusion solid breeders

    Billone, M.C.; Clemmer, R.G.

    1985-02-01

    Lithium aluminate is a candidate tritium-breeding material for fusion reactor blankets. One of the concerns with using LiAlO/sub 2/ is tritium recovery from this material, particularly at low operating temperatures and high fluences. The data from various tritium release experiments with ..gamma..-LiAlO/sub 2/ and related materials are reviewed and analyzed to determine under what conditions bulk diffusion is the rate-limiting mechanism for tritium transport and what the effective bulk diffusion coefficient should be. Steady-state and transient models based on bulk diffusion are developed and used to interpret the data. Design calculations are then performed with the verified models to determine the steady-state inventory and time to reach equilibrium for a full-scale fusion blanket.

  15. The use of passive detectors to monitor tritium on surfaces

    Gammage, R.B.; Meyer, K.E.; Brock, J.L. [Oak Ridge National Lab., TN (United States)

    1996-12-31

    Commercially available BeO exoelectron dosemeters and electret ion chambers (EICs) are being adapted and applied to in situ field monitoring of tritium on surfaces. Thin layer BeO on a conductive graphite substrate is of the order of 50 times more sensitive to tritium than the EIC. At the US department of Energy release limit for fixed surface tritium of 5000 dpm per 100 cm{sup 2}, the exposure time for quantification with the exoelectron dosemeter is of the order of one hour. A multipoint Geiger counter was used for reading exoelectron emission. An alternative ceramic BeO dosemeter (Thermalox 995) has low electrical conductivity and will require a different reader to overcome problems of surface charging during exoemission. The electret is very easy to use and read. Its practical use will be for surfaces with relatively high levels of tritium contamination. (author).

  16. Assessment of tritium in the Savannah River Site environment

    This report is the first revision to a series of reports on radionuclides inn the SRS environment. Tritium was chosen as the first radionuclide in the series because the calculations used to assess the dose to the offsite population from SRS releases indicate that the dose due to tritium, through of small consequence, is one of the most important the radionuclides. This was recognized early in the site operation, and extensive measurements of tritium in the atmosphere, surface water, and ground water exist due to the effort of the Environmental Monitoring Section. In addition, research into the transport and fate of tritium in the environment has been supported at the SRS by both the local Department of Energy (DOE) Office and DOE's Office of Health and Environmental Research

  17. Tritium Systems Test Assembly: design for major device fabrication review

    This document has been prepared for the Major Device Fabrication Review for the Tritium Systems Test Assembly (TSTA). The TSTA is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for fusion reactor systems. The principal objectives for TSTA are: (a) demonstrate the fuel cycle for fusion reactor systems; (b) develop test and qualify equipment for tritium service in the fusion program; (c) develop and test environmental and personnel protective systems; (d) evaluate long-term reliability of components; (e) demonstrate long-term safe handling of tritium with no major releases or incidents; and (f) investigate and evaluate the response of the fuel cycle and environmental packages to normal, off-normal, and emergency situations. This document presents the current status of a conceptual design and cost estimate for TSTA. The total cost to design, construct, and operate TSTA through FY-1981 is estimated to be approximately $12.2 M

  18. Assessment of tritium in the Savannah River Site environment

    Carlton, W.H.; Murphy, C.E. Jr.; Bauer, L.R. [and others

    1993-10-01

    This report is the first revision to a series of reports on radionuclides inn the SRS environment. Tritium was chosen as the first radionuclide in the series because the calculations used to assess the dose to the offsite population from SRS releases indicate that the dose due to tritium, through of small consequence, is one of the most important the radionuclides. This was recognized early in the site operation, and extensive measurements of tritium in the atmosphere, surface water, and ground water exist due to the effort of the Environmental Monitoring Section. In addition, research into the transport and fate of tritium in the environment has been supported at the SRS by both the local Department of Energy (DOE) Office and DOE`s Office of Health and Environmental Research.

  19. Tritium contaminated water from Fukushima Daiichi Nuclear Power Plants

    About 400 tons of cooling water per day is used at the site of Fukushima nuclear power plants which inevitably produce radioactivity-contaminated water as radioactive wastes. The used cooling water is stored in a waste-storage tank and transmitted to a Cs absorbing apparatus through which the concentration of Cs 134 and Cs 137 are decreased to one hundred thousand's while the other radioactive nuclides unabsorbed. This paper points out especially the risk of tritium for human health. It explains the properties of radiations from tritium, its risk toward biological effectiveness and then strongly stresses the necessity of removal of tritium from contaminated water referring the case of childhood leukemia around Canadian nuclear facilities followed by tritium release from the Pickering Nuclear Generating Station and the birth defects and infant mortality in near by communities. (S. Ohno)

  20. Tritium Exchange in Biological Systems

    Whenever tritium-labelled water is employed as a test solute or tracer in biological systems, an appreciable exchange between tritium and labile hydrogen atoms occurs that frequently affects the nature and interpretation of experimental results. The studies reported here are concerned with the magnitude of the effect that tritium exchange introduces into measurements of total body water and water metabolism in animals and humans. Direct measurements of exchange were made in rats, guinea pigs, pigeons, and rabbits. Tritium-labelled water was administered intravenously or by mouth, and tritium space and turnover determined from the concentration of tritium in blood. The animals were then desiccated to constant weight in vacuo. The specific activity of water collected periodically during desiccation increased by 50% as a result of isotope effects. Water from combustion of dried rabbit tissues contained about 2% of the tritium originally given to the animal. Adipose tissue alone contained little or no exchange tritium. The dried tissues of the other animals were rehydrated with inactive water and the appearance of tritium in the water observed. The specific activity of the water increased in exponential fashion, i.e., 1-exp. (kt), with about 90% of exchange occurring with a half-time of 1 h, and the remaining 10% with a half-time of 10 h. The total tritium extracted accounted for 1.5 to 3.5% of the dose given to the animal, which agrees with the difference between the tritium space and total body water determined by desiccation. An indirect estimate of exchange in humans was derived from concurrent measurements of tritium and antipyrene spaces. The average difference of about 2% in water volume agrees with the direct estimates of exchanges in animals. It is evident that tritium space should be reduced by about 2% to identify it with total body water. The magnitude and relatively slow rate of exchange may also influence the interpretation of metabolic studies with

  1. Measurement and modelling of tritium dispersion in vicinity of nuclear fusion facilities

    To construct and validate models for the assessment of the impact of tritium gas releases from future nuclear fusion facilities, it is essential to investigate the fate of tritium gas after release to the environment. JAERI's experimental results of a tritium gas (HT) field release experiment and modelling of environmental tritium dispersion in the vicinity of facilities are described in this paper. In the HT gas field release experiment, air, soil and pine needle samples were collected for analysis during the extended period of 5 days after the HT release. Efforts were directed toward quantifying: direct oxidation rate of HT to HTO in the atmosphere; rate of appearance of atmospheric HTO from a release of HT; deposition velocities of HT and HTO to soil; loss rate of HTO from soil; deposition velocity of HTO to pine needles; diffusion coefficient and oxidation rate constant of HT gas in soil. A model for cycling of tritium in the environment near fusion facilities was developed considering possible items relating to tritium behavior. And it was applied to the experimental data to determine diffusion coefficients and oxidation rate constants of HT in soil. These parameter values and the model were used for prediction of environmental tritium concentrations. (author)

  2. Tritium system for compact high field devices

    Some theoretical results and the current status of the work on a prototype plant for the Tritium cycle of compact high-field tokamaks (such as, Ignitor, CIT, etc.), using the SAES Getter St 707 getter material, are described in this report. The schematics and present status of the main subplants of the cycle are reported together with some experimental results demostrating the possibility of utilizing the St 707 material to purify the inert atmosphere of the glove-boxes and the secondary containment of the double-containment metal canalization which is to eventually house the various parts of the plant. Finally, as an example, the FTU machine, under construction at ENEA Frascati, has been taken as a reference, and theoretical evaluations are given for the inventory, permeation and release of the Tritium from the first wall and the thermal shieldes of such a tokamak

  3. Tritium distribution on plasma facing graphite tiles of JT-60U

    Tritium distributions on the graphite divertor tiles, the dome units and the baffle plates of JT-60U were successfully measured. Poloidally, the highest tritium level was found at the dome top tiles and the outer baffle plates, where the plasma did not hit directly. On the other hand, although the toroidal tritium profiles on each tiles appeared uniform, detailed profiles in full toroidal direction clearly showed a periodic variation corresponding to the position of the magnetic field coils, indicating the ripple loss of high energy tritons as suggested by the OFMC code. Finally, the temperature increase owing to the plasma heat load was found to release the once retained tritium. (author)

  4. A model of water and tritium desorption from lithium ortho-silicate by TPD method

    In the present paper a model proposed, on the basis of the large amount of experimental findings, of water and tritium desorption from lithium silicate irradiated in HFR reactor at Petten. The materials have been prepared in pellets and pebbles forms. Major experimental finding is on the tritium retention time V.S. operating ambient temperature. However the model of a tritium recovery process for lithium silicates is not established yet. Another finding concerned with moisture desorption and tritium release. The moisture concentration in the specimens are obtained as a function of partial pressure of water vapor. (author)

  5. Two-compartment environmental transport model for tritium

    The radiological doses to the public result from both natural and man-made radioactivity. By far, the greatest part of radiation received by the population comes from natural sources. In addition to this natural or background radiation there are the man-made radiological doses due to the radionuclides released from nuclear facilities. The release of any potential radioactive pollutant to the environment during routine operation of a Nuclear Power Plant should be the subject of appropriate controls and assessments. CANDU reactors are both moderated and cooled by heavy water (D2O). Tritium is produced in CANDU reactors by neutron reactions with deuterium, boron, and lithium and by ternary fission. Very small amounts of tritiated heavy water DTO may escape from moderator and heat transport systems of CANDU reactors during maintenance and normal operation. DTO is rapidly transferred from air to human body by inhalation and through the skin. Special dryers were designed and used to remove moisture from different ventilation systems of the reactor building in order to maintain the tritium doses well below the limits established by the national authorities. The purpose of the present paper is to present a two-compartment model to estimate the equilibrium distribution of tritiated water between air and soil media, that would result from gaseous tritium emission from Cernavoda NPP. Using the principle of mass conservation, the model provides an algorithm for predicting equilibrium specific activities of tritiated water in the soil and air compartments. The following results are given: Landscape proprieties for the Cernavoda NPP environment; Annual gaseous tritium activity releases at Cernavoda NPP; Tritium turnover rates for the air and soil compartments; Tritium concentrations and inventories and a comparison between estimated and measured tritium levels in air and soil, respectively. (authors)

  6. Measurement of spread of tritium using the tritium labeled compound

    It is known that a radioisotope disperses in the air from a radioisotope labeled compound in an aqueous solution via the isotopic exchange reaction. In this research, in order to examine the dispersion mechanism of tritium in the air from a tritium labeled compound, the model room which imitated a working room in the controlled area was made, and the radioactivity of the tritium contained in the air of the model room was measured by sampling the air in the model room. It was found that the dispersion rate of tritium in the air increased with the passage time from its purchase. The dispersion rate of tritium from 3H-ATP changed from 0.10% to 0.76% after 17.7 months. Furthermore, the two-dimensional distribution of tritium on the surface of the whole walls in the model room was obtained using an imaging plate technique. (author)

  7. Tritium transport vessel using depleted uranium

    Heung, L.K.

    1995-01-01

    A tritium transport vessel using depleted uranium was tested in the laboratory using deuterium and protium. The vessel contains 0.5 kg of depleted uranium and can hold up to 18 grams of tritium. The conditions for activation, tritium loading and tritium unloading were defined. The safety aspects that included air-ingress, tritium diffusion, temperature and pressure potentials were evaluated.

  8. Considerations on techniques for improving tritium confinement in helium-cooled ceramic breeder blankets

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic BIT blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. Our results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (orig.)

  9. Visualization of hydrogen distribution around blisters by tritium radio-luminography

    Hydrogen distribution around blisters on aluminum (Al) and molybdenum (Mo) was examined by tritium radio-luminography, i.e. tritium autoradiography (TARG) and an imaging plate technique. Tritium accumulated in the blisters on Al surface was successfully visualized at the first time. The tritium density in the blisters was found to increase with their radius to the power of 2.3. This supports the blister mechanism of bubble coalescence but the blister shape was flattened along the surface with increasing their size. For Mo, tritium distribution was not well correlated with blisters, and the bubbles coalescence was not clearly observed, too. But the erosion or exfoliation of thick layers with wider area than blisters were observed and hydrogen was released by the exfoliation of the thick surface layers, remaining not tritium on the exfoliated surface. Such exfoliation is very likely caused by mechanical stress given by accumulated hydrogen at trapping site such as grain boundaries, intrinsic defect, or self trapping.

  10. Considerations on techniques for improving tritium confinement in helium-cooled ceramic breeder blankets

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (author) 8 refs.; 2 figs

  11. Tritium processing for the European test blanket systems: current status of the design and development strategy

    Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)

  12. Tritium concentration reducing method in atmosphere in nuclear reactor containment facility

    A portion of water content in an atmosphere is condensed by a condensation/evaporation device disposed in a nuclear reactor containment building and then a portion of the condensed water is evaporated in the atmosphere. A portion of hydrogen nuclides constituting the evaporated water content is subjected to isotopic exchange with tritium nuclides in the atmosphere. A portion of water content in the atmosphere applied with the isotopic exchange is condensed in the condensation/evaporation device. That is, the hydrogen nuclides in steams are applied with isotopic exchange with tritium nuclides, and steams incorporating tritium nuclides are condensed again in the condensation/evaporation device, to transfer the tritium nuclides in the atmosphere to condensed water. The condensed water is recovered without releasing the tritium nuclides to the outside of the reactor containment building, thereby enabling to reduce the tritium concentration in the atmosphere. (N.H.)

  13. Dismantling of the PETRA glove box: tritium contamination and inventory assessment

    The PETRA facility is the first installation in which experiments with tritium were carried out at the Tritium Laboratory Karlsruhe. After completion of two main experimental programs, the decommissioning of PETRA was initiated with the aim to reuse the glove box and its main still valuable components. A decommissioning plan was engaged to: -) identify the source of tritium release in the glove box, -) clarify the status of the main components, -) assess residual tritium inventories, and -) de-tritiate the components to be disposed of as waste. Several analytical techniques - calorimetry on small solid samples, wipe test followed by liquid scintillation counting for surface contamination assessment, gas chromatography on gaseous samples - were deployed and cross-checked to assess the remaining tritium inventories and initiate the decommissioning process. The methodology and the main outcomes of the numerous different tritium measurements are presented and discussed. (authors)

  14. Tritium retention properties of tungsten, graphite and co-deposited carbon film

    DT+ ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both the amount of retained tritium and the reduction of retained tritium after preservation in vacuum were investigated using an IP technique and BIXS. In addition, the relationship between the retention properties of tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. The amount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1 keV of DT+ irradiation, graphite showed no reduction of the amount of retained tritium after six months preservation while that of carbon film decreased by approximately 20% after 40 days preservation. It was suggested that this difference might be associated with differences in the microstructure between graphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately half after 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decrease in tritium retention during preservation became slower. Tungsten's properties of releasing tritium while preserved in vacuum would be a useful tool for the reduction/removal of retained tritium

  15. Tritium neutrino mass experiments

    The current status of the experimental search for neutrino mass is reviewed, with emphasis on direct kinematic methods, such as the beta decay of tritium. The situation concerning the electron neutrino mass as measured in tritium beta decay is essentially unchanged from a year ago, although a great deal of experimental work is in progress. The ITEP group continues to find evidence for a nonzero mass, now slightly revised to 26(5) eV. After correcting for recently discovered errors in the energy loss distribution and source thickness, however, the Z/umlt u/rich group still claims and upper limit of 18 eV. There may be evidence for neutrino mass and mixing in the SN1987a data, in the same range suggested by the ITEP experiment. 42 refs., 3 figs

  16. Environmental monitoring for tritium at tritium separation facility

    The Cryogenic Pilot is an experimental project in the nuclear energy national research program, which has the aim of developing technologies for tritium and deuterium separation by cryogenic distillation. The experimental installation is located 15 km near the highest city of the area and 1 km near Olt River. An important chemical activity is developed in the area and the Experimental Cryogenic Pilot's, almost the entire neighborhood are chemical plants. It is necessary to emphasize this aspect because the sewerage system is connected with the other three chemical plants from the neighborhood. This is the reason that we progressively established elements of an environmental monitoring program well in advance of tritium operation in order to determine baseline levels. The first step was the tritium level monitoring in environmental water and waste water of industrial activity from neighborhood. In this work, a low background liquid scintillation is used to determine tritium activity concentration according to ISO 9698/1998. We measured drinking water, precipitation, river water, underground water and waste water. The tritium level was between 10 TU and 27 TU that indicates there is no source of tritium contamination in the neighborhood of Cryogenic Pilot. In order to determine baseline levels we decide to monitories monthly each location. In this paper a standard method is presented which it is used for tritium determination in water sample, the precautions needed in order to achieve reliable results, and the evolution of tritium level in different location near the Experimental Pilot Tritium and Deuterium Cryogenic Separation.(author)

  17. Tritium implantation in the accelerator production of tritium device

    We briefly describe the methods we have developed to compute the magnitude and spatial distribution of born and implanted tritons and protons in the Accelerator Production of Tritium (AFT) device. The methods are verified against experimental measurements and then used to predict that 16% of the tritium is implanted in the walls of the APT distribution tubes. The methods are also used to estimate the spatial distribution of implanted tritium, which will be required for determining the possible diffusion of tritium out of the walls and back into the gas stream

  18. Tritium monitoring : present status

    The report summarizes the present status of techniques employed for the monitoring of tritium in water, air and other samples. A brief mention of the work done by numerous workers in the field, critical comments about the work and a fairly exhaustive list of references about the work done during the last 4 decades has been presented. On-line monitoring on real time basis in nuclear reactors is also discussed. (author). 83 refs., 10 refs., 2 tabs

  19. Tritium in atmospheric hydrogen

    Martin, J. David; Hackett, Joseph P.

    2011-01-01

    The radioactivity of tritiated hydrogen (HT) in the atmosphere in Westwood, New Jersey was measured at approximately weekly intervals from August 1971 to August 1973. The background level remained constant at approximately 80 tritium atoms per milligram of air. Frequent increases in the activity level of up to an order of magnitude were observed until January 1973. The source(s) of HT which was responsible for the frequent increases apparently ceased as a tropospheric source in January 1973. ...

  20. Muon capture by tritium

    The muon capture rate is computed with realistic wave function for the initial tritium nuclei (Faddeev equations on configuration space with realistic potentials), and plane wave approximation for the final three neutrons, with the effective Hamiltonian of Fujii and Primakoff for muon capture and via a non energy weighted sum rule. Such a forbidden transition is hoped to be a probe for exchange current contributions

  1. Bioaccumulation of tritiated water in phytoplankton and trophic transfer of organically bound tritium to the blue mussel, Mytilus edulis

    Large releases of tritium are currently permitted in coastal areas due to assumptions that it rapidly disperses in the water and has a low toxicity due to its low energy emissions. This paper presents a laboratory experiment developed to identify previously untested scenarios where tritium may concentrate or transfer in biota relevant to Baltic coastal communities. Phytoplankton populations of Dunaliella tertiolecta and Nodularia spumigena were exposed at different growth-stages, to tritiated water (HTO; 10 MBq l−1). Tritiated D. tertiolecta was then fed to mussels, Mytilus edulis, regularly over a period of three weeks. Activity concentrations of phytoplankton and various tissues from the mussel were determined. Both phytoplankton species transformed HTO into organically-bound tritium (OBT) in their tissues. D. tertiolecta accumulated significantly more tritium when allowed to grow exponentially in HTO than if it had already reached the stationary growth phase; both treatments accumulated significantly more than the corresponding treatments of N. spumigena. No effect of growth phase on bioaccumulation of tritium was detectable in N. spumigena following exposure. After mussels were given 3 feeds of tritiated D. tertiolecta, significant levels of tritium were detected in the tissues. Incorporation into most mussel tissues appeared to follow a linear relationship with number of tritiated phytoplankton feeds with no equilibrium, highlighting the potential for biomagnification. Different rates of incorporation in species from a similar functional group highlight the difficulties in using a ‘representative’ species for modelling the transfer and impact of tritium. Accumulations of organic tritium into the mussel tissues from tritiated-phytoplankton demonstrate an environmentally relevant transfer pathway of tritium even when water-concentrations are reduced, adding weight to the assertion that organically bound tritium acts as a persistent organic pollutant. The

  2. A personal tritium monitor

    A tritium monitor, similar in size to a normal gamma survey meter, is being developed to improve the measurement of tritiated water vapour (HTO) near workers in Candu nuclear power plants. Methods are available for sampling and monitoring on-line from work areas; the instrument described here is intended to complement such monitoring by allowing on-the-spot individual assessment of tritium hazards. Size, mass and cost are more important than sensitivity in an instrument of this kind than in a central monitor. Accordingly, only inexpensive, readily obtainable mechanical and electrical components have been used in a simple assembly needing little machining. The tritium detector is an ionization chamber. A signal proportional to the concentration of HTO in air is obtained as the difference between the currents from two 90 cm3 ionization chambers. Sample air flows directly through one chamber and through the other after being dried by passing through a replaceable desiccant cartridge. This technique reduces the unwanted signals from gamma radiation and radioactive noble gases. The electronics comprise a MOSFET, single chip amplifier and a liquid crystal digital display that indicates concentrations in the range 1-1999 (MPC)sub(a). The mass of the instrument is 2 kg. (H.K.)

  3. Perturbations in Effort-Related Decision-Making Driven by Acute Stress and Corticotropin-Releasing Factor.

    Bryce, Courtney A; Floresco, Stan B

    2016-07-01

    Acute stress activates numerous systems in a coordinated effort to promote homeostasis, and can exert differential effects on mnemonic and cognitive functions depending on a myriad of factors. Stress can alter different forms of cost/benefit decision-making, yet the mechanisms that drive these effects, remain unclear. In the present study, we probed how corticotropin-releasing factor (CRF) may contribute to stress-induced alterations in cost/benefit decision-making, using an task where well-trained rats chose between a low effort/low reward lever (LR; two pellets) and a high effort/high reward lever (HR; four pellets), with the effort requirement increasing over a session (2, 5, 10, and 20 presses). One-hour restraint stress markedly reduced preference for the HR option, but this effect was attenuated by infusions of the CRF antagonist, alpha-helical CRF. Conversely, central CRF infusion mimicked the effect of stress on decision-making, as well as increased decision latencies and reduced response vigor. CRF infusions did not alter preference for larger vs smaller rewards, but did reduce responding for food delivered on a progressive ratio, suggesting that these treatments may amplify perceived effort costs that may be required to obtain rewards. CRF infusions into the ventral tegmental area recapitulated the effect of central CRF treatment and restraint on choice behavior, suggesting that these effects may be mediated by perturbations in dopamine transmission. These findings highlight the involvement of CRF in regulating effort-related decisions and suggest that increased CRF activity may contribute to motivational impairments and abnormal decision-making associated with stress-related psychiatric disorders such as depression. PMID:26830960

  4. Analysis on tritium management in FLiBe blanket for force-free helical reactor FFHR2

    In FFHR2 design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extract tritium from breeder and control the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The factors which affected tritium extraction and permeation were calculated and evaluated, such as the heat exchanger material, tritium permeation reduction factor (TPRF) in blanket, proportion of FLiBe flow in tritium recover system (TRS) and efficiency of TRS etc. The results of the analysis showed that further R and D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. (author)

  5. Effect of hydrophobic paints coating for tritium reduction in concrete materials

    Highlights: ► Effects of hydrophobic paint coating in tritium transport are investigated. ► Two kinds of paints, acrylic-silicon resin and epoxy resin are used. ► The hydrophobic paints are effective to reduce tritium permeation. ► The effect of tritium reduction of epoxy paint is higher than that of silicon. - Abstract: The effects of hydrophobic paint coating on a concrete material of cement paste on the tritium transport are investigated. The cement paste is coated with two kinds of paints, acrylic-silicon resin paint and epoxy resin paint. We investigated the amount of tritium trapped in the samples exposed to tritiated water vapor by means of sorption and release. It was found that both the hydrophobic paints could reduce effectively tritium permeation during 50 days exposure of tritiated water vapor. The effect of tritium reduction of the epoxy paint was higher than that of silicon while the amount of tritium trapped in the epoxy paint was larger than that of silicon due to difference of the structure. Based on an analysis of a diffusion model, the rate-determining step of tritium migration through cement paste coated with the paints is diffusion through the paints respectively. It was found that tritium was easy to penetrate through silicon because there were many pores or voids in the silicon comparatively. In the case of tritium released from the epoxy paint, it is considered that tritium diffusion in epoxy is slow due to retardation by isotope exchange reaction to water included in epoxy paint.

  6. Tritium source and behavior of tritiated water vapor in KUR

    Total tritium release rate from all of nuclear facilities, including thermonuclear experimental facilities, to environment as a gaseous form (HT) or a liquid form (HTO) reaches recently to the production rate of tritium in atmosphere by cosmic ray, although tritium effluent amounts from a single reactor facility is a few in comparison with nuclear explosion experiments in atmosphere. Tritium is produced by the nuclear reaction, D(n,γ)T. Tritium sources in KUR are divided into three pathways as follows: (1) heavy water contained in the primary light water cooling system, (2) heavy were as a moderator in preliminary tank which is installed in the reactor side, (3) liquid deuterium (∼4 L) in a cold neutron source facility (CNS). The tritiated water (HTO) in air-condensed water during stack exhaustion is sampled, and it's concentration is measured by a liquid scintillation method. The HTO concentration of reactor room during ventilation changes at every repair work of heavy water-transport pipe system. An ionization chamber (∼1.5 L) for continuous monitoring is installed near the reactor side. The ionization monitoring system, however, didn't work during reactor operation because of the radioactivity of Ar-41. The behavior of HTO concentration in the decay tank of heavy water exhaustion system is measured, and a tritium gas leakage from the system is detected. The behaviors of tritiated water vapor in KUR containment building are discussed in details. (Suetake, M.)

  7. Tritium handling in the Mirror Fusion Hybrid Reactor

    A reference design study for a Mirror Fusion Hybrid Reactor has been completed which examines the tritium handling problems. Breeding pins composed of aluminum alloys contain lithium hydride with a four-year residence time for tritium production. The slip-stream helium tritium capture system is designed to handle a 0.1 percent pin failure and will reduce environmental losses to below 3 Ci/day. The neutral beam injectors and direct converters utilize small, thin electrode tubes at 7000C for accelerating the deuterium or tritium, and they will by triton implantation permeate about 3 x 105 Ci/day into the internal helium coolant flow. A capture system will reduce these losses to 6 Ci/day, combined. The reactor hall is designed with a low humidity, air atmosphere which is continuously processed in order to handle leakage and permeability losses from the nuclear island at 180 Ci/day while still maintaining levels of tritium below MPC. The precessor is also able to handle severe accidental releases of tritium at the 26 kilogram level and permit worker re-entry (with ventilated suits) in a matter of about one week. These approaches to fusion power plant are found to be technically feasible today and economically attractive

  8. An overview of tritium production

    The characteristics of three types of proposed tritium production facilities, fissile type, accelerator production tritium (APT), and fusion type, are presented. The fissile reactors, especially commercial light water reactor, use comparatively mature technology and are designed to meet current safety and environmental guidelines. Conversely, APT shows many advantages except its rather high cost, while fusion reactors appear to offer improved safety and environmental impact, in particular, tritium production based on the fusion-based neutron source. However, its cost keeps unknown

  9. Tritium gas transfer pump development

    Non-lubricated, hermetically sealed pumps for tritium service have been selected to replace Sprengel pumps in the existing Tritium Facility. These pumps will be the primary gas-transfer pumps in the planned Replacement Tritium Facility. The selected pumps are Metal Bellows Corporation's bellows pumps and Normetex scroll pumps. Pumping range for a Normetex/Metal Bellows system is from 0.01 torr suction to 2300 torr discharge. Performance characteristics of both pumps are presented. 10 figs

  10. Tritium stripping in a nitrogen glove box using palladium/zeolite and SAES St 198 trademark

    Glove box clean-up experiments were conducted in a nitrogen glove box using palladium deposited on zeolite (Pd/z) and a SAES St 198 trademark getter as tritium stripping materials. Protium/deuterium samples spiked with tritium were released into a 620 liter glove box to simulate tritium releases in a 10,500 liter glove box. The Pd/z and the SAES St 198 trademark stripper beds produced a reduction in tritium activity of approximately two to three orders of magnitude and glove box clean-up was limited by a persistent background tritium activity level. Attempts to significantly reduce the glove box activity to lower levels without purging were unsuccessful

  11. Derived Intervention Levels for Tritium Based on Food and Drug Administration Methodology

    In 1998, the FDA released it recommendations for age-dependent derived intervention levels for several radionuclides involved in nuclear accidents. One radionuclide that is not included in that document is tritium

  12. Tritium-assisted fusion breeders

    This report undertakes a preliminary assessment of the prospects of tritium-assisted D-D fuel cycle fusion breeders. Two well documented fusion power reactor designs - the STARFIRE (D-T fuel cycle) and the WILDCAT (Cat-D fuel cycle) tokamaks - are converted into fusion breeders by replacing the fusion electric blankets with 233U producing fission suppressed blankets; changing the Cat-D fuel cycle mode of operation by one of the several tritium-assisted D-D-based modes of operation considered; adjusting the reactor power level; and modifying the resulting plant cost to account for the design changes. Three sources of tritium are considered for assisting the D-D fuel cycle: tritium produced in the blankets from lithium or from 3He and tritium produced in the client fission reactors. The D-D-based fusion breeders using tritium assistance are found to be the most promising economically, especially the Tritium Catalyzed Deuterium mode of operation in which the 3He exhausted from the plasma is converted, by neutron capture in the blanket, into tritium which is in turn fed back to the plasma. The number of fission reactors of equal thermal power supported by Tritium Catalyzed Deuterium fusion breeders is about 50% higher than that of D-T fusion breeders, and the profitability is found to be slightly lower than that of the D-T fusion breeders

  13. Study of Traces of Tritium at the World Trade Center

    Semkow, T M; Hafner, S R; Parekh, P P; Wozniak, G J; Haines, D K; Husain, L; Rabun, R L; Williams, P G

    2002-10-01

    Traces of tritiated water (HTO) were detected at the World Trade Center (WTC) ground zero after the 9/11/01 terrorist attack. A water sample from the WTC sewer, collected on 9/13/01, contained 0.164 {+-} 0.074 (2 {sigma}) nCi/L of HTO. A split water sample, collected on 9/21/01 from the basement of WTC Building 6, contained 3.53 {+-} 0.17 and 2.83 {+-} 0.15 nCi/L, respectively. These results are well below the levels of concern to human exposure. Several water and vegetation samples were analyzed from sites outside ground zero, located in Manhattan, Brooklyn, Queens, and the Kensico and Croton Reservoirs. No HTO above the background was found in those samples. Tritium radioluminescent (RL) devices were investigated as possible sources of the traces of tritium at ground zero. It was determined that the two Boeing 767 aircraft that hit the Twin Towers contained a combined 34 Ci of tritium at the time of impact in their emergency exit signs. There is also evidence that many weapons from law enforcement were present and destroyed at WTC. Such weaponry contains by design tritium sights. The fate and removal of tritium from ground zero were investigated, taking into consideration tritium chemistry and water flow originating from the fire fighting, rain, as well as leaks from the Hudson River and broken mains. A box model was developed to describe the above scenario. The model is consistent with instantaneous oxidation of the airplane tritium in the jet-fuel explosion, deposition of a small fraction of HTO at ground zero, and water-flow controlled removal of HTO from the debris. The model also suggests that tritium from the weapons would be released and oxidized to HTO at a much slower rate in the lingering fires at ground zero.

  14. The transport, dispersion, and cycling of tritium in the environment. [Contains Bibliography

    Murphy, C.E. Jr.

    1990-01-01

    The processes which determine transport, dispersion, and cycling of tritium are identified for atmospheric, terrestrial, aquatic, and groundwater systems. The processes are discussed in terms of the storage capacity for tritium in each component of each system and ranges of residence times are derived. The residence times of each component of the systems are discussed in terms of the residence time of the whole system for transient releases of tritium into different components of the systems. The role of the ocean as a sink for tritium is described. The concentration of tritium in the system at steady state is described in terms of the inputs and outputs to the components of the systems. The analysis indicates that the key residence time for a specific release of tritium into the environment is dependent on both the residence time of the components and the means of introduction into the environment. The initial concentration ad residence time of tritium in the terrestrial system after an exposure to tritiated water vapor are determined by the atmospheric and vegetative conditions at the time of the release. The dominant residence time is that of the vegetation. On the other hand, the initial concentration and residence time of tritium in the terrestrial system after an exposure to tritiated hydrogen are determined by the atmospheric and soil conditions at the time of the release. The dominant residence time is that of the soil. The initial concentration and residence time after a liquid release to the soil surface are determined by the diluting soil water content and the residence time for water in the rooting zone of the soil. Little tritium enters the organic fraction of terrestrial systems from transient releases of gases or liquid water. 102 refs., 19 figs., 2 tabs.

  15. Technology developments for improved tritium management

    Tritium technology developments have been an integral part of the advancement of CANDU reactor technology. An understanding of tritium behaviour within the heavy-water systems has led to improvements in tritium recovery processes, tritium measurement techniques and overall tritium control. Detritiation technology has been put in place as part of heavy water and tritium management practices. The advances made in these technologies are summarized. (author). 20 refs., 5 figs

  16. A posteriori verification and validation of a tritium dispersion and consequence model

    An American National Standards Institute (ANSI) posteriori (backfit) process, available to provide software quality assurance (SQA) for software developed outside of required qualification protocol, has been applied to the special-purpose, versatile tritium dispersion and consequence model, UFOTRI, a computer model developed at the German Karlsruhe laboratory. UFOTRI was chosen because of its strengths in initial tritium-related consequence analyses and its potential for application in a Department of Energy accident analysis context. The six-task process met key ANSI requirements and was performed during a several-month level of effort. Included project deliverables were Assessment, Test Plan, Configuration Procedure, Error Notification Procedure, Comprehensive Technical Report, and SQA Qualification Report documentation. Comparison to acute release conditions is still in progress, but results to date indicate satisfactory, bounding predictions can be achieved with UFOTRI relative to measurements. Results of this compact effort appear to identify UFOTRI as a suitable candidate for a software toolkit, i.e., minimum verification and validation (V and V) requirements are satisfied, and a configuration controlled version is deemed appropriate for use in a DOE accident analysis context

  17. Tritium in waste as a tracer of landfill leachate in surface and ground water in South Africa

    Tritium concentrations higher than expected environmental levels were first discovered in South Africa in association with landfill sites in 1995. Since then, a survey has identified numerous landfill sites of various types which produce leachate containing elevated levels of tritium, in one case 105 TU, approaching the limit for drinking water. Tritium, a low energy (18 keV max) β emitter, half-life 12.43 years, is a conservative tracer of water in the oxidised state. It exchanges readily between different chemical compounds, such as tritium-containing material and moisture moving through a landfill. In southern Africa at present the range of tritium values in rain is close to the assumed natural level (about 5 TU), about one order of magnitude lower than contemporary values in e.g. western Europe with much more extensive industrial use and release of tritium. This renders very sensitive the tracing of leachate containing waste tritium in the southern African environment

  18. Improving Tritium Exposure Reconstructions Using Accelerator Mass Spectrometry

    Love, A; Hunt, J; Knezovich, J

    2003-06-01

    Exposure reconstructions for radionuclides are inherently difficult. As a result, most reconstructions are based primarily on mathematical models of environmental fate and transport. These models can have large uncertainties, as important site-specific information is unknown, missing, or crudely estimated. Alternatively, surrogate environmental measurements of exposure can be used for site-specific reconstructions. In cases where environmental transport processes are complex, well-chosen environmental surrogates can have smaller exposure uncertainty than mathematical models. Because existing methodologies have significant limitations, the development or improvement of methodologies for reconstructing exposure from environmental measurements would provide important additional tools in assessing the health effects of chronic exposure. As an example, the direct measurement of tritium atoms by accelerator mass spectrometry (AMS) enables rapid low-activity tritium measurements from milligram-sized samples, which permit greater ease of sample collection, faster throughput, and increased spatial and/or temporal resolution. Tritium AMS was previously demonstrated for a tree growing on known levels of tritiated water and for trees exposed to atmospheric releases of tritiated water vapor. In these analyses, tritium levels were measured from milligram-sized samples with sample preparation times of a few days. Hundreds of samples were analyzed within a few months of sample collection and resulted in the reconstruction of spatial and temporal exposure from tritium releases.

  19. Tritium uptake by SS316 and its decontamination

    Torikai, Y.; Penzhorn, R.-D.; Matsuyama, M.; Watanabe, K.

    2004-08-01

    As-received and highly polished SS316 specimens were loaded with HT at 473-573 K. The uptake by polished samples was found to be up to five times that of as-received ones, when loading was performed immediately after polishing. This disparity vanished when polished specimens were subjected to a prolonged exposure to air prior to loading. The tritium loss from tritium-loaded SS316 specimens was examined by chemical etching and by thermal release in a flow system using various carrier gases at several temperatures. While at moderate temperatures the type of carrier has an impact on the tritium release rate, at higher ones this effect disappears. Moisture in the carrier gas has little influence on the loss rate of bulk tritium. Etching depth profiles of specimens previously heat-treated in the presence of air or Ar + H 2 and of untreated specimens are given. Evidence for chronic tritium liberation from SS316 at 298 K is provided.

  20. Design of the tritium scrubber system for Omega Upgrade

    The Omega Upgrade at the Laboratory for Laser Energetics will commence deuterium-tritium (DT) shots in 1995. Omega Upgrade utilizes a 350 nm, 30 kJ laser to energize DT filled targets contained within an evacuated target chamber (TC). Up to 10 DT shots per day are planned, 5 days a week, with each target containing 20 mCi of tritium. Cryopumps, used to achieve high vacuum in the TC and to collect unburnt gases following DT shots, are regenerated on a monthly basis. The Tritium Scrubber System (TSS) has been designed to capture tritium released from the cryopumps, vacuum pump exhaust streams, outgassing from internal surfaces, and in the event of target rupture. A large quantity of low activity gas needs to be processed by the system; the target chamber and diagnostic antechambers will be evacuated up to 8 and 2000 times each month, respectively. Modifications are required to the facility to enable the capture of tritiated species from exhaust streams before being released to the atmosphere. The targeted annual tritium emission for the facility is less than 10 mCi. The TSS utilizes zirconium alloy hydride beds and molecular sieve beds to detritiate exhaust streams prior to stacking to satisfy this stringent emission target. 8 refs., 3 figs., 1 tab

  1. Tritium management in the EPR™ and ATMEA1™ reactors: An AREVA computer tool AbsoluT-3

    Tritium, in NPP, is mainly produced by neutron activation reactions with lithium and boron isotopes dissolved in the primary coolant as well as with naturally-occurring deuterium in the primary coolant. The use of enriched boric acid, the increase in lithium concentration to control the pH and the new fuel management increase the tritium production in the primary coolant and thus the tritium releases in the environment. In this context AREVA developed a computational tool for management of the tritium activity: AbsoluT-3. AbsoluT-3 allows estimating the tritium activity in the primary coolant taking into account the mode of operation (with or without load follow), water movements and many other parameters. This activity can be estimated for many cycles in power operation and during shutdown. If the tritium releases are regulatory imposed, AbsoluT-3 allows determining maximum tritium concentration to be reached in the primary coolant and the total tritium releases to respect the discharge authorization. AbsoluT-3 allows anticipating tritium management issues according to the design as well as operational constraints. These constraints are mainly twofold: radiation protection during shutdown and releases into the environment. (author)

  2. In vivo changes in plasma acute phase protein levels in the rat induced by slow release of IL-1, IL-6 and TNF

    E. J. Lewis

    1992-01-01

    Full Text Available Administration of large doses of cytokines by injection is required to induce changes in acute phase protein levels. Comparisons were made in the rat of the effects of administering recombinant human cytokines by injection with continuous release from implanted osmotic minipumps. Continuous release of interleukin-1β (0.2–2.1 ng h-1 induced dose-related changes in the plasma levels of albumin, seromucoid proteins, haptoglobin and caeruloplasmin; interleukin-1α had similar effects but required higher doses (2–21 ng h-1. Tumour necrosis factor α (50 ng h-1 only significantly increased seromucoid levels, whereas IL-6 (3–30 ng h-1 induced haptoglobin and caeruloplassynthesis. This method provides a better technique for studying the in rive effects of cytokines which may be relevant to the release mechanisms in inflammation.

  3. Handling of tritium-contaminated effluents and wastes. Part of a coordinated programme on handling tritium-contaminated effluents and wastes

    The work was carried out on: (i) Applicability of cotton, woodpulp, sawdust, and certain cellulosic derivatives for the removal of tritium from aqueous medium. (ii) Containment and fixation of tritiated water in non-leachable matrices. The absorption studies on cotton, woodpulp, sawdust, and cellulose acetates were carried out with a view to assessing their potentialities as concentration media and also to choose a matrix which can concentrate tritium to the maximum extent possible. The experiments on water hyacinth plants were designed to see the applicability of concentrating tritium and also for providing a via medium for slow release of tritium into the atmosphere. The immobilisation studies on tritiated water in cement matrices were aimed at maximum retention of tritium

  4. Modeling of the environmental behavior of tritium around the nuclear power plants

    The relationship between the tritium release rate from the nuclear power plant and tritium concentration in the environment around the Kori site was modeled. The tritium concentration in the atmosphere was calculated by multiplying the release rates and χ/Q values, and the dry deposition rate at each sector according to the direction and the distance was obtained using a dry deposition velocity. The area around Kori site was divided into 6 zones according to the deposition rate. The six zones were divided into 14 compartments for the numerical simulation. Transfer coefficients between the compartments were derived using site characterization data. Source terms were calculated from the dry deposition rates. Tritium concentration in surface soil water and groundwater was calculated based upon a compartment model. The semi-analytical solution of the compartment model was obtained with a computer program, AMBER. The results showed that most of tritium deposited onto the land released into the atmosphere and the sea. Also, the estimated concentration in the top soil agreed well to that measured. Using the model, tritium concentration was predicated in the case that the tritium release rates were doubled

  5. Tritium accountancy in fusion systems

    The US Department of Energy (DOE) has clearly defined requirements for nuclear material control and accountability (MCA) of tritium whereas the International Atomic Energy Agency (IAEA) does not since tritium is not a fissile material. MCA requirements are expected for tritium fusion machines and will be dictated by the host country or regulatory body where the machine is operated. Material Balance Areas (MBA) are defined to aid in the tracking and reporting of nuclear material movements and inventories. Material sub-accounts (MSA) are established along with key measurement points (KMP) to further subdivide a MBA to localize and minimize uncertainties in the inventory difference (ID) calculations for tritium accountancy. Fusion systems try to minimize tritium inventory which may require continuous movement of material through the MSA. The ability of making meaningful measurements of these material transfers is described in terms of establishing the MSA structure to perform and reconcile ID calculations. For fusion machines, changes to the traditional ID equation will be discussed which includes breeding, burn-up, and retention of tritium in the fusion device. The concept of 'net' tritium quantities consumed or lost in fusion devices is described in terms of inventory taking strategies and how it is used to track the accumulation of tritium in components or fusion machines. (authors)

  6. Evaluation of environmental tritium level in the area of Cernavoda Nuclear Power Plant

    This work is a continuation of the study entitled Evaluation of environmental tritium levels before and after start of a new tritium source for model verification' that commenced two years ago as part of an European research contract having as objective the 'Investigations and Modelling of the Dynamics of Environmental HT/HTO/OBT Levels Resulting from the Tritium Releases'. The aim of this study was to evaluate the environmental tritium level in pre-operational stage of Cernavoda Nuclear Power Plant and the dynamics of variation of tritium concentration in the first operational years of NPP, for determination of site-specific transfer parameters and model validation. Representative samples for Cernavoda area were analyzed; - air humidity; - water from Danube River, Danube-Black Sea Canal and lakes; - drinking, rain and snow water; - HTO concentration in soils at different depths; - tissue free water tritium in vegetal and animal foodstuffs relevant for human diets like cereals (wheat, maize, barley). vegetables (potatoes, tomato, cabbage, onion, bean), grapes and wine. The mean concentration of tritium obtained in 1996 for these types of measurements is presented. The values of tritium concentration in air, water, soil and plants are as low as for areas without contaminating sources. The operation in 1996 of the Cernavoda NPP did not modify the tritium environmental level. (authors)

  7. The tritium operations experience on TFTR

    The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grains of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes- the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the US Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described

  8. The tritium operations experience on TFTR

    von Halle, A.; Gentile, C. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Anderson, J.L. [Los Alamos National Lab., NM (United States)] [and others

    1994-09-01

    The Tokamak Fusion Test Reactor (TFTR) tritium gas system is administratively limited to 5 grains of tritium and provides the feedstock gas for the neutral beam and torus injection systems. Tritium operations on TFTR began with leak checking of gas handling systems, qualification of the gas injection systems, and high power plasma operations using using trace amounts of tritium in deuterium feedstock gas. Full tritium operation commenced with four highly diagnosed neutral beam pulses into a beamline calorimeter to verify planned tritium beam operating routines and to demonstrate the deuterium to tritium beam isotope exchange. Since that time, TFTR has successfully operated each of the twelve neutral beam ion sources in tritium during hundreds of tritium beam pulses and torus gas injections. This paper describes- the TFTR tritium gas handling systems and TFTR tritium operations of the gas injection systems and the neutral beam ion sources. Tritium accounting and accountability is discussed, including tritium retention issues of the torus limiters and beam impinged surfaces of the beamline components. Also included is tritium beam velocity analysis that compares the neutral beam extracted ion species composition for deuterium and tritium and that determines the extent of beam isotope exchange on subsequent deuterium and tritium beam pulses. The required modifications to TFTR operating routines to meet the US Department of Energy regulations for a low hazard nuclear facility and the problems encountered during initial tritium operations are described.

  9. Tritium clouds environmental impact in air into the Western Mediterranean Basin evaluation

    Castro, P., E-mail: paloma.castro@ciemat.es [EURATOM-CIEMAT Association, LNF Fusion National Laboratory, BBTU, Avda Complutense,40 28040 Madrid (Spain); Velarde, M. [ETSII Nuclear Fusion Institute: DENIM, Madrid (Spain); Ardao, J. [AEMET, Environmental Applications Service, 28040 Madrid (Spain); Perlado, J.M. [ETSII Nuclear Fusion Institute: DENIM, Madrid (Spain); Sedano, L. [EURATOM-CIEMAT Association, LNF Fusion National Laboratory, BBTU, Avda Complutense,40 28040 Madrid (Spain)

    2012-08-15

    The paper considers short-term releases of tritium (mainly but not only tritium hydride (HT)) to the atmosphere from a potential ITER-like fusion reactor located in the Mediterranean Basin and explores if the short range legal exposure limits are exceeded (both locally and downwind). For this, a coupled Lagrangian ECMWF/FLEXPART model has been used to follow real time releases of tritium. This tool was analyzed for nominal tritium operational conditions under selected incidental conditions to determine resultant local and Western Mediterranean effects, together with hourly observations of wind, to provide a short-range approximation of tritium cloud behavior. Since our results cannot be compared with radiological station measurements of tritium in air, we use the NORMTRI Gaussian model. We demonstrate an overestimation of the sequence of tritium concentrations in the atmosphere, close to the reactor, estimated with this model when compared with ECMWF/FLEXPART results. A Gaussian 'mesoscale' qualification tool has been used to validate the ECMWF/FLEXPART for winter 2010/spring 2011 with a database of the HT plumes. It is considered that NORMTRI allows evaluation of tritium-in-air-plume patterns and its contribution to doses.

  10. Evaluation of Tritium Behavior in the Epoxy Painted Concrete Wall of ITER Hot Cell

    Tritium behavior released in the ITER hot cell has been investigated numerically using a combined analytical methods of a tritium transport analysis in the multi-layer wall (concrete and epoxy paint) with the one dimensional diffusion model and a tritium concentration analysis in the hot cell with the complete mixing model by the ventilation. As the results, it is revealed that tritium concentration decay and permeation issues are not serious problem in a viewpoint of safety, since it is expected that tritium concentration in the hot cell decrease rapidly within several days just after removing the tritium release source, and tritium permeation through the epoxy painted concrete wall will be negligible as long as the averaged realistic diffusion coefficient is ensured in the concrete wall. It is also revealed that the epoxy paint on the concrete wall prevents the tritium inventory increase in the concrete wall greatly (two orders of magnitudes), but still, the inventory in the wall is estimated to reach about 0.1 PBq for 20 years operation

  11. A Study on the Tritium Behavior in the Rice Plant after a Short-Term Exposure of HTO

    Yook, D-S.; Lee, K. J.; Choi, Y-H.

    2002-02-26

    In many Asian countries including Korea, rice is a very important food crop. Its grain is consumed by humans and its straw is used to feed animals. In Korea, there are four CANDU type reactors that release relatively large amounts of tritium into the environment. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out the experimental studies to obtain domestic data on various parameters concerning the direct contamination of plant. In this study, the behavior of tritium in the rice plant is predicted and compared with the measurement performed at KAERI. Using the conceptual model of the soil-plant-atmosphere tritiated water transport system which was suggested by Charles E. Murphy, tritium concentrations in the soil and in leaves to time were derived. If the effect of tritium concentration in the soil is considered, the tritium concentration in leaves is described as a double exponential model. On the other hand if the tritium concentration in the soil is disregarded, the tritium concentration in leaves is described by a single exponential term as other models (e.g. Belot's or STAR-H3 model). Also concentration of organically bound tritium in the seed is predicted and compared with measurements. The results can be used to predict the tritium concentration in the rice plant at a field around the site and the ingestion dose following the release of tritium to the environment.

  12. Analysis on tritium management in FLiBe blanket for LHD-type helical reactor FFHR2

    In FFHR2 (LHD-type helical reactor) design, FLiBe has been selected as a self-cooling tritium breeder for low reactivity with oxygen and water and lower conductivity. Considering the fugacity of the tritium, particular care and adequate mitigation measures should be applied for the effectively extracting tritium from breeder and controlling the tritium release to the environment. In this paper, a tritium analysis model of the FLiBe blanket system was developed and the preliminary analysis on tritium permeation and extraction for FLiBe blanket system were done. The results of the analysis showed that it was reasonable to select W alloy as heat exchanger (HX) material, the proportion of FLiBe flow in tritium recover system (TRS) was 0.2, the efficiency of TRS was 0.85 and tritium permeation reduction factor (TPRF) was 20 in blanket etc.. In addition, further R and D efforts were required for FFHR2 tritium system to guarantee the tritium self-sufficient and safety, for example reasonable quality of tritium permeation barriers on blanket, requirement for the TRS and fabrication technology of the heat exchanger etc.. (author)

  13. Tritium recovery and separation from CTR plasma exhausts and secondary containment atmospheres

    Recent experimental successes have generated increased interest in the development of thermonuclear reactors as power sources for the future. This paper examines tritium containment problems posed by an operating CTR and sets forth some processing schemes currently being evaluated at the Oak Ridge National Laboratory. An appreciation of the CTR tritium management problem can best be realized by recalling that tritium production rates for various fission reactors range from 2 x 104 to 9 x 105 Ci/yr per 1000 MW(e). Present estimates of tritium production in a CTR blanket exceed 109 Ci/yr for the same level of power generation, and tritium process systems may handle 10 to 20 times that amount. Tritium's high permeability through most materials of construction at high temperatures makes secondary containment mandatory for most piping. Processing of these containment atmospheres will probably involve conversion of the tritium to a nonpermeating form (T2O) followed by trapping on conventional beds of desiccant material. In a similar fashion, all purge streams and process fluid vent gases will be subjected to tritium recovery prior to atmospheric release. Two tritium process systems will be required, one to recover tritium produced by breeding in the blanket and another to recover unburned tritium in the plasma exhaust. Plasma exhaust processing will be unconventional since the exhaust gas pressure will lie between 10-3 and 10-6 torr. Treatment of this gas stream will entail the removal of small quantities of protium and helium from a much larger deuterium-tritium mixture which will be recycled. (U.S.)

  14. Tritium-surface interactions

    The report deals broadly with tritium-surface interactions as they relate to a fusion power reactor enterprise, viz., the vacuum chamber, first wall, peripherals, pumping, fuel recycling, isotope separation, repair and maintenance, decontamination and safety. The main emphasis is on plasma-surface interactions and the selection of materials for fusion chamber duty. A comprehensive review of the international (particularly U.S.) research and development is presented based upon a literature review (about 1 000 reports and papers) and upon visits to key laboratories, Sandia, Albuquerque, Sandia, Livermore and EGβG Idaho. An inventory of Canadian expertise and facilities for RβD on tritium-surface interactions is also presented. A number of proposals are made for the direction of an optimal Canadian RβD program, emphasizing the importance of building on strength in both the technological and fundamental areas. A compendium of specific projects and project areas is presented dealing primarily with plasma-wall interactions and permeation, anti-permeation materials and surfaces and health, safety and environmental considerations. Potential areas of industrial spinoff are identified

  15. Knockdown of p54nrb inhibits migration, invasion and TNF-α release of human acute monocytic leukemia THP1 cells.

    Zhang, Xiujuan; Wu, Changli; Xiong, Wei; Chen, Chunling; Li, Rong; Zhou, Guangji

    2016-06-01

    54 kDa nuclear RNA- and DNA-binding protein (p54nrb) which is also called non-POU domain-containing octamer-binding protein (NONO) is known to be multifunctional involved in many nuclear processes. It was shown that p54nrb/NONO was closely related to the occurrence of erythroleukemia. Whether p54nrb/NONO plays a role in progress of human acute monocytic leukemia remains unknown. In the present study, we examined the effects of p54nrb/NONO silencing on the biological characteristics of human acute monocytic leukemia THP1 cells. The results showed that p54nrb was strongly expressed in THP1 cells, and knockdown of p54nrb slightly promoted proliferation and strongly inhibited motility and invasion of THP1 cells. Moreover, knockdown of p54nrb strongly decreased the release of TNF-α from THP1 cells by inhibiting certain process of TNF-α secretion, specially for the release of TNF-α induced by lipopolysaccharide (LPS). Notably, the infection of negative control shRNA-containing lentiviruses promoted the migration and the release of TNF-α induced by LPS in THP1 cells. All the above results demonstrated that p54nrb slightly inhibited THP1 cell proliferation, but significantly promoted migration, invasion and release of TNF-α induced by LPS in THP1 cells. The present study indicates that p54nrb is a powerful molecule involved in the regulation of cell motility and promotes the migration and invasion of THP1 cells, and it is more likely to be involved in the release of inflammatory mediators and the motility of inflammatory cells. PMID:27108701

  16. Tritium migration in the materials proposed for fusion reactors: Li{sub 2}TiO{sub 3} and beryllium

    Kulsartov, T.V., E-mail: kulsartov@nnc.kz [Institute of Atomic Energy NNC RK, 071100, Krasnoarmeiskay St., 10, Kurchatov (Kazakhstan); Gordienko, Yu.N.; Tazhibayeva, I.L. [Institute of Atomic Energy NNC RK, 071100, Krasnoarmeiskay St., 10, Kurchatov (Kazakhstan); Kenzhin, E.A. [Shakarim Semey State University, 071412, Glinka St., 20b, Semey (Kazakhstan); Barsukov, N.I.; Sadvakasova, A.O. [Institute of Atomic Energy NNC RK, 071100, Krasnoarmeiskay St., 10, Kurchatov (Kazakhstan); Kulsartova, A.V. [Nuclear Technology Safety Center, 050020, L. Chaikina 4, Almaty (Kazakhstan); Zaurbekova, Zh.A. [Institute of Atomic Energy NNC RK, 071100, Krasnoarmeiskay St., 10, Kurchatov (Kazakhstan)

    2013-11-15

    The results of tritium and helium gas release from lithium ceramics samples Li{sub 2}TiO{sub 3} irradiated at the WWR-K reactor (Almaty, Kazakhstan) and from beryllium samples irradiated at the BN-350 reactor (Aktau, Kazakhstan) and the IVG.1M reactor (Kurchatov, Kazakhstan) are presented. Experimentally obtained thermal desorption (TDS) spectra have shown that the dependence of tritium release from lithium ceramics has a complicated behavior and to a large extent depends on lithium ceramics type. Nevertheless, it was found that the total amount of tritium released from all types of lithium ceramics has the same order of magnitude, equal to about 10{sup 11} Bq/kg. It was found that in the temperature range from 523 K to 1373 K the process of tritium release from lithium ceramics involves volume diffusion and thermoactivated tritium release from the accumulation centers generated under irradiation. TDS of beryllium samples enables us to obtain characteristics of tritium and helium release during linear heating, to determine integrated quantities of generated helium and tritium, and to determine parameters of release processes.

  17. Preliminary analysis of the safety and environmental impact of the Tritium Systems Test Assembly

    The Tritium Systems Test Assembly (TSTA) is a facility dedicated to the development of technologies associated with the D-T fuel cycle of future fusion reactors while demonstrating that TSTA can be operated safely with no significant losses to the environment. During the initial design stage of TSTA, a safety analysis was performed which investigated the effects of major subsystem component failure, the meteorology and seismicity of the site and their possible effect on the facility, and accident scenarios which result in tritium releases. Major releases of tritium to the environment are considered highly improbable since they require a compound failure of primary and secondary containment, along with either a breach of the building or a failure of the Emergency Tritium Cleanup system. Accidental releases caused by natural phenomena (earthquake, tornado, etc.) are considered highly improbable

  18. Tritium waste control: July--September 1978

    The combined Electrolysis Catalytic Exchange system was modified to allow better control of experimental conditions and to prevent the overflow of water into the air detritation system. A program designed to regenerate the activity of the hydrophobic catalyst was also completed. Slight differences in the release rate of high specific activity tritiated liquid wastes from the drums are now beginning to appear. The three drums with the highest fractional permeation rate had the least amount of tritium when packaged. The fractional permeation rate of the two octane drums appears to have leveled off at about the same rate as the oil and water drums. Tests continued on samples of cement and cement-plaster mixtures which were injected with 386 Ci of tritiated water, cured, and then impregnated with catalyzed styrene monomer. After polymerization, the samples were put into uncontaminated water and the tritium concentration was monitored. No significant differences were noted except in two cases when the polyethylene bottle had been removed, which resulted in 35 to 80 times more tritium being released into the surrounding water. Full scale (cold) waste drum No. 5 was polymerized with excellent results. Pressure increase and gas composition were measured over (1) tritiated water without fixation, (2) polymer-impregnated concrete, and (3) nonpolymer concrete. Activities for all samples were 10 Ci/m3. Pressure buildup results are essentially the same for concrete made with tritiated distilled water and tritiated waste water. However, the pressure buildup rate is slightly higher for the polymer impregnated concrete than for the nonpolymer concrete. Mass analysis of the cover gas over tritiated water without fixation and over the polymer and nonpolymer concrete samples made with tritiated waste water show that hydrogen represents about 85% of the gas generated

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

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

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

    Penzhorn, R.-D.; Torikai, Y.; Watanabe, K.; Matsuyama, M.; Perevezentsev, A.

    2012-10-01

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

  1. Reducing the tritium inventory in waste produced by fusion devices

    Pamela, J., E-mail: jerome.pamela@cea.fr [CEA, Agence ITER-France, F-13108 Saint-Paul-lez-Durance (France); Decanis, C. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Canas, D. [CEA, DEN/DADN, Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Liger, K.; Gaune, F. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2015-04-15

    Highlights: • Fusion devices including ITER will generate tritiated waste, some of which will need to be detritiated before disposal. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Incineration is very attractive for VLLW and possibly SL-LILW soft housekeeping waste, since it offers higher tritium and waste volume reduction than the alternative thermal treatment technique. • For metallic waste, further R&D efforts should be made to optimize tritium release management and minimize the need for interim storage. - Abstract: The specific issues raised by tritiated waste resulting from fusion machines are described. Of the several categories of tritium contaminated waste produced during the entire lifespan of a fusion facility, i.e. operating phase and dismantling phase, only two categories are considered here: metal components and solid combustible waste, especially soft housekeeping materials. Some of these are expected to contain a high level of tritium, and may therefore need to be processed using a detritiation technique before disposal or interim storage. The reference solution for tritiated waste management in France is a 50-year temporary storage for tritium decay, with options for reducing the tritium content as alternatives or complement. An overview of the strategic issues related to tritium reduction techniques is proposed for each radiological category of waste for both metallic and soft housekeeping waste. For this latter category, several options of detritiation techniques by thermal treatment like heating up or incineration are described. A comparison has been made between these various technical options based on several criteria: environment, safety, technical feasibility and costs. For soft housekeeping waste, incineration is very attractive for VLLW and possibly SL-LILW. For metallic waste, further R&D efforts should be conducted.

  2. Reducing the tritium inventory in waste produced by fusion devices

    Highlights: • Fusion devices including ITER will generate tritiated waste, some of which will need to be detritiated before disposal. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Incineration is very attractive for VLLW and possibly SL-LILW soft housekeeping waste, since it offers higher tritium and waste volume reduction than the alternative thermal treatment technique. • For metallic waste, further R&D efforts should be made to optimize tritium release management and minimize the need for interim storage. - Abstract: The specific issues raised by tritiated waste resulting from fusion machines are described. Of the several categories of tritium contaminated waste produced during the entire lifespan of a fusion facility, i.e. operating phase and dismantling phase, only two categories are considered here: metal components and solid combustible waste, especially soft housekeeping materials. Some of these are expected to contain a high level of tritium, and may therefore need to be processed using a detritiation technique before disposal or interim storage. The reference solution for tritiated waste management in France is a 50-year temporary storage for tritium decay, with options for reducing the tritium content as alternatives or complement. An overview of the strategic issues related to tritium reduction techniques is proposed for each radiological category of waste for both metallic and soft housekeeping waste. For this latter category, several options of detritiation techniques by thermal treatment like heating up or incineration are described. A comparison has been made between these various technical options based on several criteria: environment, safety, technical feasibility and costs. For soft housekeeping waste, incineration is very attractive for VLLW and possibly SL-LILW. For metallic waste, further R&D efforts should be conducted

  3. Limitation of tritium outgassing from tritiated solid waste drums

    In the framework of the development of fusion thermonuclear reactors, tritiated solid waste is foreseen and will have to be managed. The management of tritiated waste implies limitations in terms of activity and tritium degassing. The degassing tritium can be under the form of tritiated hydrogen, tritiated water and, in some specific cases, negligible amount of tritiated volatile organic compound. Hence, considering the major forms of degassing tritium, CEA has developed a mixed-compound dedicated to tritium trapping in drums. Based on several experiments, the foreseen mixed compound is composed of MnO2, Ag2O, Pt and molecular sieve, the three first species having the ability to convert tritiated hydrogen into tritiated water and the last one acting as a trap for tritiated water. To assess the performance of the trapping mixture, experimental tests were performed at room temperature on tritiated dust composed of beryllium and carbon. It was shown that the metallic oxides mixture used for tritiated hydrogen conversion is efficient and that tritiated water adsorption was limited due to an inefficient regeneration of the molecular sieve prior to its use. Apart from this point, the tritium release from waste was reduced by a factor of 5.5, which can be improved up to 87 if the adsorption step is efficient

  4. Limitation of tritium outgassing from tritiated solid waste drums

    Liger, K.; Trabuc, P.; Lefebvre, X.; Troulay, M.; Perrais, C. [CEA, Centre de Cadarache, DEN/DTN/STPA/LIPC, Saint-Paul-lez-Durance (France)

    2015-03-15

    In the framework of the development of fusion thermonuclear reactors, tritiated solid waste is foreseen and will have to be managed. The management of tritiated waste implies limitations in terms of activity and tritium degassing. The degassing tritium can be under the form of tritiated hydrogen, tritiated water and, in some specific cases, negligible amount of tritiated volatile organic compound. Hence, considering the major forms of degassing tritium, CEA has developed a mixed-compound dedicated to tritium trapping in drums. Based on several experiments, the foreseen mixed compound is composed of MnO{sub 2}, Ag{sub 2}O, Pt and molecular sieve, the three first species having the ability to convert tritiated hydrogen into tritiated water and the last one acting as a trap for tritiated water. To assess the performance of the trapping mixture, experimental tests were performed at room temperature on tritiated dust composed of beryllium and carbon. It was shown that the metallic oxides mixture used for tritiated hydrogen conversion is efficient and that tritiated water adsorption was limited due to an inefficient regeneration of the molecular sieve prior to its use. Apart from this point, the tritium release from waste was reduced by a factor of 5.5, which can be improved up to 87 if the adsorption step is efficient.

  5. RADIOLYTIC GAS PRODUCTION RATES OF POLYMERS EXPOSED TO TRITIUM GAS

    Clark, E.

    2013-08-31

    Data from previous reports on studies of polymers exposed to tritium gas is further analyzed to estimate rates of radiolytic gas production. Also, graphs of gas release during tritium exposure from ultrahigh molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, a trade name is Teflon®), and Vespel® polyimide are re-plotted as moles of gas as a function of time, which is consistent with a later study of tritium effects on various formulations of the elastomer ethylene-propylene-diene monomer (EPDM). These gas production rate estimates may be useful while considering using these polymers in tritium processing systems. These rates are valid at least for the longest exposure times for each material, two years for UHMW-PE, PTFE, and Vespel®, and fourteen months for filled and unfilled EPDM. Note that the production “rate” for Vespel® is a quantity of H{sub 2} produced during a single exposure to tritium, independent of length of time. The larger production rate per unit mass for unfilled EPDM results from the lack of filler- the carbon black in filled EPDM does not produce H{sub 2} or HT. This is one aspect of how inert fillers reduce the effects of ionizing radiation on polymers.

  6. Utility considerations for tritium production

    The Southern Nuclear Company has a long-standing commitment to nuclear power and is interested in pursuing the possible use of one of its existing commercial reactors as an alternative for reinitiating the production of tritium to support the nation's defense program requirements. We understand that Congress and the administration agree on the need to replenish the nation's supply of tritium and that a number of production options are under consideration. This paper discusses the financial considerations, legal and regulatory considerations for the production of tritium utilizing a commercial power reactor

  7. Tritium monitor for fusion reactors

    This report describes the design, operation, and performance of a flow-through ion-chamber instrument designed to measure tritium concentrations in air containing 13N, 16N, and 41Ar produced by neutrons generated by D-T fusion devices. The instrument employs a chamber assembly consisting of two coaxial ionization chambers. The inner chamber is the flow-through measuring chamber and the outer chamber is used for current subtraction. A thin wall common to both chambers is opaque to the tritium betas. Currents produced in the two chambers by higher energy radiation are automatically subtracted, leaving only the current due to tritium

  8. Evaluation of retention and disposal options for tritium in fuel reprocessing

    Five options were evaluated as means of retaining tritium released from light-water reactor or fast breeder reactor fuel during the head-end steps of a typical Purex reprocessing scheme. Cost estimates for these options were compared with a base case in which no retention of tritium within the facility was obtained. Costs were also estimated for a variety of disposal methods of the retained tritium. The disposal costs were combined with the retention costs to yield total costs (capital plus operating) for retention and disposal of tritium under the conditions envisioned. The above costs were converted to an annual basis and to a dollars per curie retained basis. This then was used to estimate the cost in dollars per man-rem saved by retaining the tritium. Only the options that used the least expensive disposal costs could approach the $1000/man-rem cost used as a guide by the Nuclear Regulatory Commission

  9. Process for conditioning tritium for final storage

    The process for conditioning tritium for final storage a) in which the tritium is introduced into a zeolite matrix, b) the zeolite matrix is dehydrated at temperatures above 4000C and c) the tritium in vapour form is brought into contact with the zeolite matrix is characterized by the fact that after saturation of the zeolite matrix with tritium, the tritium in the zeolite matrix is enclosed by means of microwave irradiation. (orig.)

  10. Tritium production, recovery and application in Korea.

    Son, Soon-Hwan; Lee, Sook-Kyung; Kim, Kwang-Sin

    2009-01-01

    Four CANDU reactors have been operating at the site of Wolsong Nuclear Power Generation in Korea. The Wolsong tritium removal facility was constructed to reduce the tritium levels in heavy water systems. This facility was designed to process 100kg/h of tritiated heavy water feed and to produce 99% pure T(2). This recovered tritium will be made available for commercial applications. The initial phases on the tritium applications are made to establish the infrastructure and the tritium controls. PMID:19307127

  11. Tritium Plasma Experiment Upgrade for Fusion Tritium and Nuclear Sciences

    Shimada, Masashi; Taylor, Chase N.; Kolasinski, Robert D.; Buchenauer, Dean A.

    2015-11-01

    The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to directly study tritium retention and permeation in neutron-irradiated materials [M. Shimada et.al., Rev. Sci. Instru. 82 (2011) 083503 and and M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. Recently the TPE has undergone major upgrades in its electrical and control systems. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium. We discuss the electrical upgrade, enhanced operational safety, improved plasma performance, and development of tritium plasma-driven permeation and optical spectrometer system. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, Fusion Nuclear Science Facility (FNSF), and Demonstration reactor (DEMO). This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.

  12. Tritium distribution in the environment in the vicinity of a chronic atmospheric source-assessment of the steady state hypothesis

    The Savannah River Site (SRS) is a major radionuclide production center. Tritium has been released to the atmosphere over the 36 year period of operation. The tritiated water concentration of the atmosphere, rain, vegetation and food have been routinely monitored during this period. Special studies have been made of tritium in soils and in the organic fractions of these same materials. The available data suggest that the average tritium concentration in the components of the terrestrial environment have approached a steady state with the two main sources of tritium, rainfall and atmospheric water vapor

  13. Organically bound tritium (OBT) in soil at different depths around Chalk River Laboratories (CRL), Canada

    Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) is a large nuclear research and test establishment with nuclear and non-nuclear facilities located in Chalk River, Ontario. The CRL Environmental Monitoring Program is designed to demonstrate that radiological exposure resulting from releases from the CRL site remain below the public dose limit specified in the regulations (1 mSv/year). This study was conducted to consolidate environmental effects following a continuous atmospheric tritium release observed at CRL. Soil samples were collected at depths of up to 20 cm using soil probes at the CRL site and surrounding areas. The samples were sectioned at 5 cm intervals, and HTO and OBT concentrations were measured in the samples. Prevailing winds at CRL are from NW and SE, which was suggested to be in close relationship with tritium distribution in environmental samples such as soils and plant leaves. The HTO concentration was the highest in surface soil water and plant leaves at a given sampling point. This result suggests that the concentration of tritium in surface soil water and in plants tissue free water essentially reflects the surrounding atmospheric tritium concentration. OBT concentrations in soil were measured at the historical HT release site, Plant Road, Mattawa Road and three background sites near CRL. The top layer of soil generally had the highest OBT concentration among collected soil samples. This result suggests that OBT concentrations are different from HTO concentrations at the same site and can be representative of previously released environmental tritium at the sampling point. The relationship between the OBT concentration in soil and the amount of tritium released into the environment will be useful for the evaluation of environmental tritium effects and the fate of tritium in the terrestrial ecosystem. The study points out that HTO shows shorter-term dynamic conditions, whereas OBT shows longer-term steady-state conditions

  14. Measurement of Tritium Activity in Plants by Ice Extraction Method

    Tritium is produced primarily by interactions of cosmic rays with the atmosphere. However, nuclear installations may add significantly tritium to the surroundings, increasing its concentration. The main sources of tritium released by man are linked to the nuclear power cycle: nuclear power stations, nuclear fuel reprocessing plants or tritium production plants. Tritium is found in the environment mainly as tritiated water, in gaseous or liquid form (HTO, T2O), in the surrounding air and in soil. It accumulates in plants, which may use as a measure to the level of tritium concentration in the environment. The most common routes of tritium uptake from the environment in plants are from atmospheric humidity and by precipitation water which entered the soil. The fraction of tritium bound to the plant tissue is small compared to that present as tritiated water in the plant (from 0.06% to 0.3% for growing crops). The tritiated water uptake is through the roots, as tritiated water from the soil follows a pathway similar to that of ordinary water. As most tritium in plants consists of tritiated water, the measurements of only the tissue free water tritium concentration (as HTO or T2O) gives an accurate estimate of the tritium content in the plant. Analyzing free tritium in biological matrices usually requires using the freeze-drying method to extract the water from the sample, and then measure the water collected in a cold trap with a Liquid Scintillation Counter (LSC). The 'freeze-drying' occurs because of the sublimation of the frozen water inside the plant, that takes place when the temperature is beneath the triple point and the vapour pressure is low. In the temperature range of -5° to -10° C the mechanism that plants use to avoid freezing is drawing of water from the cell protoplasm into the intercellular spaces. Changes in cell membrane permeability allow water to leave the cell and enter the spaces between the cells where it freezes instead of freezing within the

  15. Tritium transport around nuclear faciliteis

    The transport and cycling of tritium around nuclear facilities is reviewed with special emphasis on studies at the Savannah River Laboratory, Aiken, South Carolina. These studies have shown that the rate of deposition from the atmosphere, the site of deposition, and the subsequent cycling are strongly influenced by the compound with which the tritium is associated. Tritiated hydrogen is largely deposited in the soil, while tritiated water is deposited in the greatest quantity in the vegetation. Tritiated hydrogen is converted in the soil to tritiated water that leaves the soil slowly, through drainage and transpiration. Tritiated water deposited directly to the vegetation leaves the vegetation more rapidly after exposure. Only a small part of the tritium entering the vegetation becomes bound in organic molecules. However, it appears that the existence of soil organic compounds with tritium concentrations greater than the equilibrium concentration in the associated water can be explained by direct metabolism of tritiated hydrogen in vegetation. (J.P.N.)

  16. High-concentration tritium sensor

    Paglieri, S. N. (Stephen N.); Richmond, S. (Scott); Snow, R. C. (Ronny C.); Morris, J. S. (John S.); Tuggle, D. G. (Dale Glenn)

    2004-01-01

    A bi-layer device was fabricated and tested for the direct collection of electrons emitted by tritium beta decay. The sensor functions at high pressures and concentrations where previously no simple and cost effective direct measurement technique existed for tritium. A polished KOVAR{trademark} (Fe-Ni-Co alloy) rod was coated with a 1-{mu}m thick insulating layer of alumina using electron-beam evaporation, physical vapor deposition (PVD) of aluminum with oxygen dosing. The alumina deposition process was optimized to minimize pinholes and obtain a stable coating with high resistivity. The detector exhibited a nanoampere electrical response over a few decades of tritium concentration, up to pure tritium at 200 kPa. The sensor has been in service for several months now without showing signs of degradation and no discernible physical damage or change in efficiency or linearity has been observed.

  17. Tritium in fusion reactor components

    When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

  18. Tritium transport around nuclear facilities

    The transport and cycling of tritium around nuclear facilities is reviewed with special emphasis on studies at the Savannah River Laboratory, Aiken, South Carolina. These studies have shown that the rate of deposition from the atmosphere, the site of deposition, and the subsequent cycling are strongly influenced by the compound with which the tritium is associated. Tritiated hydrogen is largely deposited in the soil, while tritiated water is deposited in the greatest quantity in the vegetation. Tritiated hydrogen is converted in the soil to tritiated water that leaves the soil slowly, through drainage and transpiration. Tritiated water deposited directly to the vegetation leaves the vegetation more rapidly after exposure. Only a small part of the tritium entering the vegetation becomes bound in organic molecules. However, it appears tht the existence of soil organic compounds with tritium concentrations greater than the equilibrium concentration in the associated water can be explained by direct metabolism of tritiated hydrogen in vegetation

  19. High-concentration tritium sensor

    A bi-layer device was fabricated and tested for the direct collection of electrons emitted by tritium beta decay. The sensor functions at high pressures and concentrations where previously no simple and cost effective direct measurement technique existed for tritium. A polished KOVAR(trademark) (Fe-Ni-Co alloy) rod was coated with a 1-μm thick insulating layer of alumina using electron-beam evaporation, physical vapor deposition (PVD) of aluminum with oxygen dosing. The alumina deposition process was optimized to minimize pinholes and obtain a stable coating with high resistivity. The detector exhibited a nanoampere electrical response over a few decades of tritium concentration, up to pure tritium at 200 kPa. The sensor has been in service for several months now without showing signs of degradation and no discernible physical damage or change in efficiency or linearity has been observed.

  20. Tritium pellet injector for TFTR

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellet. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  1. Tritium pellet injector for TFTR

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single- stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. A new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellets. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  2. Correlation of rates of tritium migration through porous concrete

    In a nuclear facility when tritium leaks from a glovebox to room accidentally, an atmosphere detritiation system (ADS) starts operating, and HTO released is recovered by ADS. ADS starts when tritium activity in air becomes higher than its controlled level. Before ADS operates, the laboratory walls are the final enclosure facing tritium and are usually made of porous concrete coated with a hydrophobic paint. In the present study, previous data on the diffusivity and adsorption coefficient of concrete and paints are reviewed. Tritium penetrates and migrates into concrete by following 3 ways. First, gaseous HT or T2 easily penetrates into porous concrete. Its diffusivity is almost equal to that of H2. When a gaseous molecule diffuses through pores with a smaller diameter than a mean free path, its migration rate is described by the Knudsen diffusion formula. The second mechanism is H2O vapor diffusion in pores. Concrete holds a lot of structural water. Therefore, H2O or HTO vapor can diffuse inside concrete pores along with adsorption-desorption and isotopic exchange with structural water, which is the third mechanism. Literature shows that the diffusivity of HTO through the epoxy-resin paint is determined as D(HTO)=1.0*10-16 m2/s. We have used this data to set a model and we have applied it to estimate residual tritium in laboratory walls. We have considered 2 accidental cases and a normal case: first, ADS starts operating 1 hour after 100 Ci HTO is released in the room, secondly, ADS starts 24 hours after 100 Ci HTO release and thirdly, when the walls are exposed to HTO for 10 years of normal operation. It appears that the immediate start up of ADS is indispensable for safety

  3. Correlation of rates of tritium migration through porous concrete

    Fukada, S.; Katayama, K.; Takeishi, T. [Kyushu University, Fukuoka (Japan); Edao, Y.; Kawamura, Y.; Hayashi, T.; Yamanishi, T. [JAEA-TPL, Muramatsu, Tokai-mura (Japan)

    2015-03-15

    In a nuclear facility when tritium leaks from a glovebox to room accidentally, an atmosphere detritiation system (ADS) starts operating, and HTO released is recovered by ADS. ADS starts when tritium activity in air becomes higher than its controlled level. Before ADS operates, the laboratory walls are the final enclosure facing tritium and are usually made of porous concrete coated with a hydrophobic paint. In the present study, previous data on the diffusivity and adsorption coefficient of concrete and paints are reviewed. Tritium penetrates and migrates into concrete by following 3 ways. First, gaseous HT or T{sub 2} easily penetrates into porous concrete. Its diffusivity is almost equal to that of H{sub 2}. When a gaseous molecule diffuses through pores with a smaller diameter than a mean free path, its migration rate is described by the Knudsen diffusion formula. The second mechanism is H{sub 2}O vapor diffusion in pores. Concrete holds a lot of structural water. Therefore, H{sub 2}O or HTO vapor can diffuse inside concrete pores along with adsorption-desorption and isotopic exchange with structural water, which is the third mechanism. Literature shows that the diffusivity of HTO through the epoxy-resin paint is determined as D(HTO)=1.0*10{sup -16} m{sup 2}/s. We have used this data to set a model and we have applied it to estimate residual tritium in laboratory walls. We have considered 2 accidental cases and a normal case: first, ADS starts operating 1 hour after 100 Ci HTO is released in the room, secondly, ADS starts 24 hours after 100 Ci HTO release and thirdly, when the walls are exposed to HTO for 10 years of normal operation. It appears that the immediate start up of ADS is indispensable for safety.

  4. Development of an on-line tritium monitor with gamma-ray rejection and energy discrimination

    With the prospect of large fusion facilities coming on-line in the not-too-distant future, it is becoming increasingly important that an on-line tritium-monitoring system be developed which is capable of detecting small amounts of released tritium. Since tritium oxide is some 400 times as hazardous as elemental tritium, it is necessary to distinguish between the two in order to properly evaluate the hazard. Presently available on-line instrumentation has marginal sensitivity, is unable to distinguish between the two forms of tritium, and has poor discrimination against background gamma radiation and air activation products. The objective of our program is to develop a monitoring system with the capability of distinguishing between the two forms of tritium, detecting tritium with a sensitivity of a fraction of an MPC/sub a/ (1 MPC/sub a/ = 5. x 10-6 Ci/M3) for the oxide, and discriminating against gamma activity and airborne activity other than tritium

  5. Tritium levels in leachates and condensates from domestic wastes in landfill sites

    An extensive review study of leachate composition, from landfills which have received mainly domestic wastes in the UK, has been carried out by Aspinwall and Company on behalf of the Department of the Environment. This work has provided detailed information on concentrations of a wide range of determinands. One set of analyses which has produced particularly interesting results comprises the determination of tritium. Tritium is a natural isotope of hydrogen (3H), with a half-life of 12.43 years, which decays radioactively to form stable 3He atoms, releasing beta particles. Although tritium concentrations can be measured accurately as tritium units (TU), these represent extremely low levels of activity, and concentrations of tritium well in excess of 100 000 TU would be needed before acceptable levels of radioactivity in drinking water were exceeded. Background levels of tritium in UK rainfall rarely now exceed 50 TU although, in the 1960s, values of up to 3000 TU in rainfall were associated with atmospheric testing of atomic weapons. It was somewhat surprising, therefore, to discover that relatively high concentrations of tritium (>10 000 TU) were widespread in leachates from landfills which had received primarily domestic wastes. This paper presents results from samples taken at 30 representative UK sites, and discusses potential sources of tritium. (UK)

  6. Tritium permeation in EUROFER97 steel in EXOTIC-9/1 irradiation experiment

    This paper presents the results of the tritium permeation study in EUROFER97 carried out within the EXOTIC (EXtraction Of Tritium In Ceramics) irradiation experiment. In the EXOTIC 9/1 experiment, a pebble bed assembly containing Lithium Titanate (Li2TiO3) pebbles is irradiated for 300 days in the High Flux Reactor (HFR), in the temperature range between 340 and 580 °C, reaching a lithium burn up of 3.5% and 1.2 dpa of damage in steel. The primary objective of this experiment was to measure the in-pile tritium release characteristics of Li2TiO3 pebbles. Additionally tritium permeation through the EUROFER97 pebble bed wall was measured on line. The permeation of tritium was studied at steady state conditions, during temperature transients, and at different hydrogen concentrations in the helium purge gas flow. The model used in the analysis of the experimental data which account for co-permeation of tritium and hydrogen is presented. It has been demonstrated that the permeation of tritium under experiment conditions proceeds in the diffusion limited regime. From the analysis of the experimental data the permeability and diffusivity of tritium in EUROFER97 is determined

  7. Tritium levels in leachates and condensates from domestic wastes in landfill sites

    Robinson, H.D. [Aspinwall and Co., Shrewsbury (United Kingdom); Gronow, J.R. [Environment Agency, London (United Kingdom). Waste Regulation Policy Group

    1996-12-01

    An extensive review study of leachate composition, from landfills which have received mainly domestic wastes in the UK, has been carried out by Aspinwall and Company on behalf of the Department of the Environment. This work has provided detailed information on concentrations of a wide range of determinands. One set of analyses which has produced particularly interesting results comprises the determination of tritium. Tritium is a natural isotope of hydrogen ({sup 3}H), with a half-life of 12.43 years, which decays radioactively to form stable {sup 3}He atoms, releasing beta particles. Although tritium concentrations can be measured accurately as tritium units (TU), these represent extremely low levels of activity, and concentrations of tritium well in excess of 100 000 TU would be needed before acceptable levels of radioactivity in drinking water were exceeded. Background levels of tritium in UK rainfall rarely now exceed 50 TU although, in the 1960s, values of up to 3000 TU in rainfall were associated with atmospheric testing of atomic weapons. It was somewhat surprising, therefore, to discover that relatively high concentrations of tritium (>10 000 TU) were widespread in leachates from landfills which had received primarily domestic wastes. This paper presents results from samples taken at 30 representative UK sites, and discusses potential sources of tritium. (UK).

  8. Trace tritium recovery from the residue of liquid Li17Pb83 alloy

    2010-01-01

    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.

  9. Tritium permeation and transport in the gasoline production system coupled with high temperature gas-cooled reactors (HTGRs)

    This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05*10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit which is 3.7*10-3 Bq/cm3 [STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasoline product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced. (authors)

  10. Development of tritium plant system for fusion reactors. Achievements in the 14-year US-Japan collaboration

    Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safe-operation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore, distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work. (author)

  11. Tests on technology of tritium production under neutron irradiation of liquid metal Pb-17Li

    Economic aspects and environmental safety of fusion reactor safety of fusion reactor depend on tritium production and retention technologies. Tests on technology of tritium production and permeability through austenite steel were performed on the breeder alloy Pb-17Li (600-800 g, 30% enrichment of Li-6) under irradiation of thermal neutron flux density 2 x 1017 m-2s-1 and temperature 540 K (for two containments) and 670 K (for one containment). Tritium release and permeability kinetics was monitored open-quotes on-lineclose quotes during liquid metal irradiation. Dependencies of tritium release and permeability under Pb-17Li irradiation on time (up to 2000 h) and irradiation temperature were obtained

  12. In situ tritium recovery from LiAlO2 pellets

    TEQUILA-1 is the first phase of an in situ tritium release experiment performed in MESULINE reactor, at CEN Grenoble, with almost the same facilities and irradiation conditions as used for the LISA series experiments. Three couples of LiAlO2 specimens with the same density (80% of TD) but quite different microstructures (grain size) were tested in six vented capsules by thermal step cycling in the temperature range 400 C - 700 C. The reference He + 0.1% H2 pruging gas was used, but an oxidizing mixture containing moisture, of composition He + O2 (37 vpm) + H2O (115 vpm), was also successfully tested to extract the tritium from the ceramic breeders. The best tritium releasing performance was achieved by the P-type ceramic breeder, with the smalles grain size (0.3 Lm), for which a Tritium retention time less than one day could be measured at 450 C

  13. The french experiment on environmental tritium behaviour -October 15, 1986

    An experimental study of the behaviour of HT (tritiated hydrogen) released to the atmosphere was carried out in France on October 15, 1986. A total of 2.6 x 1014 Bq of tritium was released over 2 min from a 40 m stack. Measurements of tritium in air, soil and vegetation were made out to 2.5 km downwind under well characterized meteorological conditions. This report summarizes the overall experimental results and details the Canadian contribution which included participation in the planning meetings, measurement of air HT and HTO (tritiated water) concentrations during the release experiment, and model calculations with the Ontario Hydro Tritium Dispersion Code (OHTDC). No evidence was found of rapid conversion of HT to atmospheric HTO, consistent with the results of parallel Canadian experiments carried out in August 1986 and June 1987, and in contrast to a much earlier experiment in France. The HTO concentrations in air during and after the release resulted partially from oxidation of HT to HTO in surface soils followed by emission of this HTO to the atmosphere and partially from the HTO impurity in the source. OHTDC model simulations of the experiment showed reasonable agreement with measured concentrations of HT and HTO in air once the potential HTO contamination in the source gas (approximately 0.02%) was considered

  14. Blockade of peroxynitrite-induced neural stem cell death in the acutely injured spinal cord by drug-releasing polymer

    YU, DOU; Neeley, William L.; Pritchard, Christopher D.; Slotkin, Jonathan R.; Woodard, Eric J.; Langer, Robert; Teng, Yang D.

    2009-01-01

    Therapeutic impact of neural stem cells (NSCs) for acute spinal cord injury (SCI) has been limited by the rapid loss of donor cells. Neuroinflammation is likely the cause. Since there are close temporal-spatial correlations between the inducible nitric oxide (NO) synthase expression and the donor NSC death after neurotrauma, we reasoned that NO-associated radical species might be the inflammatory effectors which eliminate NSC grafts and kill host neurons. To test this hypothesis, human NSCs (...

  15. Heat treatment of graphite and resulting tritium emissions

    Pile I still retains significant amounts of Wigner Energy within graphite bricks not affected by the 1957 fire. A recently undertaken survey could clarify the status of the graphite with respect to stored energy. Based on the history of Pile I it is not possible to present a unique set of release curves valid for all investigated channels. Furthermore peripheric blocks not submitted to any survey, due to the lack of being accessible by trepanning machinery are unknown with respect to possible stored energy. Graphite samples of several blocks show release behaviour that exceeds the heat capacity of unirradiated graphite. Statistically this characteristic correspondence is given only for less than one quarter of the material. Three quarters do not represent any risk in the sense that either handling or later encapsulation of the graphite will result in unforeseeable energy release. For the purpose of planning all stages of the graphite-disposal process the question is addressed whether an inadvertent release of stored energy can occur during handling and storage and whether deliberate or partial annealing of the material is a requirement. On the basis of a multi-activation energy concept it could be cleared that annealing between 200 and 250 deg. C is able to clear all sites eventually being activated by the grouting process. This temperature range was examined with respect to the efficiency of a deliberate annealing procedure. 80 to 90% of the stored energy releasable tip to 500 deg. C can be released if the annealing procedure is executed between 250 and 300 deg. C. As far as tritium release from the graphite during an annealing procedure is concerned measurements of the desorption of tritium as effect of heating of grounded Pile I graphite material were executed. Apparently less than 0.5% of the total tritium content are released at the proposed annealing temperatures. (author)

  16. In-situ tritium recovery from Li2O irradiated in fast neutron flux - Beatrix-II temperature change specimen

    The Beatrix-II irradiation experiment is an in-situ tritium release experiment to evaluate the stability and tritium release characteristics of Li2O under fast neutron irradiation to extended burnups. A thin annular ring specimen capable of temperature changes was irradiated in Phase I of the experiment to a lithium burnup of 5%. The primary emphasis of the test plan was to determine the effect and interrelationship of gas composition and temperature on the tritium inventory with increasing temperature and a series of specific temperature changes were carried out at intervals throughout the experiment to characterize the effect of burnup. Decreasing the amount of hydrogen in the sweep gas resulted in an increase in the tritium inventory in the Li2O specimen. The tritium recovery during startup and shutdown was observed to be strongly influenced by the composition of the sweep gas

  17. Progresses in tritium accident modelling in the frame of IAEA EMRAS II

    The assessment of the environmental impact of tritium release from nuclear facilities is a topic of interest in many countries. In the IAEA's Environmental Modelling for Radiation Safety (EMRAS I) programme, progresses for routine releases were done and in the EMRAS II programme a dedicated working group (WG 7 - Tritium Accidents) focused on the potential accidental releases (liquid and atmospheric pathways). The progresses achieved in WG 7 were included in a complex report - a technical document of IAEA covering both liquid and atmospheric accidental release consequences. A brief description of the progresses achieved in the frame of EMRAS II WG 7 is presented. Important results have been obtained concerning washout rate, the deposition on the soil of HTO and HT, the HTO uptake by leaves and the subsequent conversion to OBT (organically bound tritium) during daylight. Further needs of the processes understanding and the experimental efforts are emphasised

  18. Tritium issues for realization of a DT fusion reactor

    A trend of studies of production and consumption of tritium is described. Realization of DT fusion reactor is discussed by tritium balance obtained from the above studies. It consists of introduction, tritium introduced into plasma vessel, tritium inventory in plasma vessel, tritium loss at fueling cycle system, tritium breeding and loss in blanket system, tritium balance in DT fusion reactor and summary. Investigation of development of external tritium resources has to be started. Tritium flow in DT fusion reactor, comparison of tritium inventory in fusion reactor, schematic diagram of tritium behavior in plasma vessel, change of overall burning efficiency and overall plasma generation rate, tritium inventory in re-deposition layer, effects of recovery efficiency of tritium from re-deposition layer, various breeding efficiencies in solid blanket, tritium flow in inertial confinement reactor with first wall, a tabular comparison of tritium balance calculation values, and comparison between tritium production methods are illustrated. (S.Y.)

  19. EXOTIC: Development of ceramic tritium breeding materials for fusion reactor blankets. The behaviour of tritium in: lithium aluminate, lithium oxide, lithium silicates, lithium zirconates

    This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO2, Li2ZrO3, Li4SiO4, Li6Zr2O7, Li8ZrO6, Li2O and Li2SiO3 have been irradiated at different temperature levels and up to a maximum lithium burnup of about 3%. Tritium residence times of the various breeding materials have been determined from temperature transients performed during irradiation. After irradiation the tritium inventory has been determined from small samples of the various materials. From the out-of-pile tritium release experiments activation energies were determined. These activities have been performed at ECN within the framework of the European Fusion Technology Programme on Breeding Blankets. (orig.)

  20. Tritium Permeability of Incoloy 800H and Inconel 617

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2012-07-01

    Design of the Next Generation Nuclear Plant (NGNP) reactor and its high-temperature components requires information regarding the permeation of fission generated tritium and hydrogen product through candidate heat exchanger alloys. Release of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system represent safety basis and product contamination issues. Of the three potential candidates for high-temperature components of the NGNP reactor design, only permeability for Incoloy 800H has been well documented. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. To support engineering design of the NGNP reactor components, the tritium permeability of Inconel 617 and Incoloy 800H was determined using a measurement system designed and fabricated at Idaho National Laboratory. The tritium permeability of Incoloy 800H and Inconel 617, was measured in the temperature range 650 to 950°C and at primary concentrations of 1.5 to 6 parts per million volume tritium in helium. (partial pressures of 10-6 atm)—three orders of magnitude lower partial pressures than used in the hydrogen permeation testing. The measured tritium permeability of Incoloy 800H and Inconel 617 deviated substantially from the values measured for hydrogen. This may be due to instrument offset, system absorption, presence of competing quantities of hydrogen, surface oxides, or other phenomena. Due to the challenge of determining the chemical composition of a mixture with such a low hydrogen isotope concentration, no categorical explanation of this offset has been developed.

  1. Tritium Permeability of Incoloy 800H and Inconel 617

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2011-09-01

    Design of the Next Generation Nuclear Plant (NGNP) reactor and its high-temperature components requires information regarding the permeation of fission generated tritium and hydrogen product through candidate heat exchanger alloys. Release of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system represent safety basis and product contamination issues. Of the three potential candidates for high-temperature components of the NGNP reactor design, only permeability for Incoloy 800H has been well documented. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. To support engineering design of the NGNP reactor components, the tritium permeability of Inconel 617 and Incoloy 800H was determined using a measurement system designed and fabricated at Idaho National Laboratory. The tritium permeability of Incoloy 800H and Inconel 617, was measured in the temperature range 650 to 950 C and at primary concentrations of 1.5 to 6 parts per million volume tritium in helium. (partial pressures of 10-6 atm) - three orders of magnitude lower partial pressures than used in the hydrogen permeation testing. The measured tritium permeability of Incoloy 800H and Inconel 617 deviated substantially from the values measured for hydrogen. This may be due to instrument offset, system absorption, presence of competing quantities of hydrogen, surface oxides, or other phenomena. Due to the challenge of determining the chemical composition of a mixture with such a low hydrogen isotope concentration, no categorical explanation of this offset has been developed.

  2. Design of the OMEGA Laser Target Chamber Tritium Removal System

    Preparations are currently underway at the OMEGA laser at the University of Rochester Laboratory for Laser Energetics (UR/LLE) to conduct direct drive laser implosion campaigns with inertial confinement fusion targets containing deuterium-tritium (DT) cryogenic ice layers. The OMEGA Cryogenic Target Handling System will fill plastic targets with high-pressure DT (150 MPa) at 300 to 500 K, cool them down to cryogenic temperature (<25 K), form the DT ice layer, and transport the targets to the OMEGA laser target chamber. Targets will then be shot with the 60-beam 30-kJ OMEGA laser. A tritium removal system has been designed to remove tritium from effluents associated with operation of the target chamber and its associated diagnostic antechambers, vacuum pumping systems, and target insertion systems. The design of the target chamber tritium removal system (TCTRS) is based on catalytic oxidation of DT and tritiated methane to tritiated water (DTO), followed by immobilization of DTO on molecular sieves. The design of the TCTRS presented a challenge due to the low tritium release limits dictated by the tritium license at UR/LLE. Aspen Plus, a commercial software package intended for the simulation and design of chemical processing systems operating at steady state, was used to simulate and design the TCTRS. A second commercial software package, Aspen ADSIM, was used to simulate and design the TCTRS molecular sieve beds, which operate at unsteady state. In this paper, we describe the design of the TCTRS and the benefits that were realized by use of the Aspen Plus and Aspen ADSIM software packages

  3. Metabolic models for tritium dosimetry

    Tritium (3H or T) is the radioactive isotope of hydrogen, which is produced by both natural, and man made sources. Tritium has a small relative natural abundance compared to hydrogen and deuterium (D). As was assessed by the United Nations scientific committee the contribution of cosmogenic tritium to annual effective dose in human is very small, only about 0.01 μSv (UNSCEAR, 2000). In case of heavy water reactors annual tritium doses for critical groups are also very small but theoretically could reach values of several μSv. In this case professionally exposed personnel could be exposed to tritium doses of some mSv. According to these considerations environmental and dosimetric aspects of this radionuclide are of special concern for health physicists. Tritium dose assessment methodologies have special particularities because hydrogen is a chemical element with an important metabolic role in the human body. Operating experience to date of CANDU reactors has indicated that the major contributor to the internal dose of professionally exposed people is the tritiated heavy water (DTO). DTO, like the tritiated water HTO, is assumed to be uniformly mixed with body water pool and reaching equilibrium immediately after the intake. All the statements in this paper related to HTO dosimetry are also considered valid in case of DTO. The results of the computations performed with different retention functions corresponding to different compartment models are presented. The differences between the models due to OBT (Organically Bound Tritium) contribution are 5.6% for two- compartment model and 13% for three-compartment Dunford - Johnson model, respectively. In practice the contribution of OBT is considered to be about 10%. (authors)

  4. Tritiated water retention on maize and beans after an acute contamination

    Although tritium is released in large quantities into environment by nuclear industries and peaceful radioisotope utilization, its behavior is not well known. The International Atomic Energy Agency is sponsoring an international study group to obtain more information about tritiated water (HTO) behavior in different ecological systems. This paper presents the studies made on corn and beans after an acute application of tritiated water during their early stages of growth on an experimental field. Sampling and radiochemical analytical methods of tritium and its behaviour on that plants during their growth cycle are outlined. It is shown that the tritiated water retention plot has at least two components, with effective half lifes of about 10 and 100 days for corn, and 8 and 40 days for beans. (author)

  5. Tritium accident containment within a large fusion enclosure: cost, benefit, and risk considerations

    Containment of a tritium accident within a large fusion device building will be difficult and costly. Complete containment is impossible, and with this fact in mind, the global dispersion and health effects of tritium are reviewed. Atmospheric tritium emissions lead to an estimated population dose to the Northern Hemisphere of 5.6 x 10-3 man-rem/Ci, which may also be interpreted as 1 cancer fatality per MCi. Updating the NRC $1000 per man-rem criterion to 1982 costs gives 9.5 $/y per Ci/y as the unit annual health benefit rate from averting tritium release at a continuous rate. Present worth considerations lead to an estimate of $100 per Ci/y for the maximum capital investment justified per expected curie per year of tritium release averted. A simplified enclosure model is used to explore the trade-off between processing capacity and recycle time with the health cost of residual tritium release included in the analysis

  6. Neutron activation, gamma spectrometry and tritium measurements on Italian lithium aluminate and zirconate, as selection means of candidate breeders for fusion reactors

    This paper discusses measurements of impurities and tritium releasing characteristics of Lithium Aluminate and zirconate, prepared by ENEA in the frame of the European Program on Fusion Technology, performed, respectively, by neutron activation analysis (NAA) and out of pile annealing. The resulting tritium removing rate from the ceramics was interpreted in terms of surface desorption kinetics. With reference purge gas (He + 0.1% H2), the predominant form of tritium, released by lithium aluminate is HT/T2, HTO/T2O by lithium zirconate. The latter was found to have a better performance in tritium release than aluminate. The presence of moisture was found to catalyze the tritium release at lower temperatures

  7. Applications developed for byproduct 85Kr and tritium

    The radionuclides, krypton-85 and tritium, both of which are gases under ordinary conditions, are used in many applications in industries and by the military forces. Krypton-85 is produced during the fissioning of uranium and is released during the dissolution of spent-fuel elements. It is a chemically inert gas that emits 0.695-MeV beta rays and a small yield of 0.54-MeV gammas over a half life of 10.3 years. Much of the 85Kr currently produced is released to the atmosphere; however, large-scale reprocessing of fuel will require collection of the gas and storage as a waste product. An alternative to storage is utilization, and since the chemical and radiation characteristics of 85Kr make this radionuclide a relatively low hazard from the standpoint of contamination and biological significance, a number of uses have been developed. Tritium is produced as a byproduct of the nuclear-weapons program, and it has a half life of 12.33 years. It has a 0.01861-MeV beta emission and no gamma emission. The absence of a gamma-ray energy eliminates the need for external shielding of the devices utilizing tritium, thus making them easily transportable. Many of the applications require only small quantities of 85Kr or tritium; however, these uses are important to the technology base of the nation. A significant development that has the potential for beneficial utilization of large quantities of 85Kr and of tritium involves their use in the production of low-level lighting devices. Since these lights are free from external fuel supplies, have a long half life (> 10 years), are maintenance-free, reliable, and easily deployed, both military and civilian airfield-lighting applications are being studied

  8. Apparent enrichment of organically bound tritium in rivers explained by the heritage of our past

    The global inventory of naturally produced tritium (3H) is estimated at 2.65 kg, whereas more than 600 kg have been released during atmospheric nuclear tests (NCRP, 1979; UNSCEAR, 2000) constituting the main source of artificial tritium throughout the Anthropocene. The behaviour of this radioactive isotope in the environment has been widely studied since the 1950s, both through laboratory experiments and, more recently, through field observations (e.g., Cline, 1953; Kirchmann et al., 1979; Daillant et al., 2004; McCubbin et al., 2001; Kim et al., 2012). In its “free” forms, [i.e. 3H gas or 3H hydride (HT); methyl 3H gas (CH3T); tritiated H2O or 3H-oxide (HTO); and Tissue Free Water 3H (TFWT)], tritium closely follows the water cycle. However, 3H bound with organic compounds, mainly during the basic stages of photosynthesis or through weak hydrogen links, is less exchangeable with water, which explains its persistence in the carbon cycle as re underlined recently by Baglan et al. (2013), Jean-Batiste and Fourré (2013), Kim et al. (2013a,b). In this paper, we demonstrate that terrestrial biomass pools, historically contaminated by global atmospheric fallout from nuclear testing, have constituted a significant delayed source of organically bound tritium (OBT) for aquatic systems, resulting in an apparent enrichment of OBT as compared to HTO. This finding helps to explain concentration factors (tritium concentration in biota/concentration in water) greater than 1 observed in areas that are not directly affected by industrial radioactive wastes, and thus sheds light on the controversies regarding tritium ‘bioaccumulation’. Such apparent enrichment of OBT is expected to be more pronounced in the Northern Hemisphere where fallout was most significant, depending on the nature and biodegradability of terrestrial biomass at the regional scale. We further believe that OBT transfers from the continent to oceans have been sufficient to affect tritium concentrations in

  9. Distillation and measurement of two forms of tritium

    Two forms of tritium of HTO, free water tritium and bound tritium, exist in the environment. A method was introduced to acquire free water tritium by 8 hour's distilling at 130 degree C and collect bound tritium by oxidative 4 minute's burning at 850 degree C. Methods of measuring two forms of tritium by liquid scintillometer were also discussed

  10. Prediction of tritium behavior in the atmosphere and assessment of radiological impact

    We have developed a computer program ATDC (Atmospheric Tritium Dispersion Code) to predict downwind tritium air concentrations in the environment around nuclear power plants on the basis of the Gaussian plume model. The tritium air concentrations by field measurement (measured tritium air concentrations in the areas adjacent to Wolsong Unit 1, a pressurized heavy water reactor ; PHWR) were compared with that by calculation to validate the program. The predicted and measured atmospheric tritium concentrations were quite consistent in trend and magnitude. This computer program will be useful in reviewing and evaluating environmental radiological impacts for PHWRs. Especially, it will be of great help to predict the behavior of tritium in the atmospheric environment around nuclear power plants. We also have studied the distribution of environmental tritium and the correlation coefficients between tritium concentrations in several environmental samples and the emissions of tritiated water vapor from Wolsong Unit 1. The annual mean concentration of atmospheric HTO were in the range of 1.31 ∼ 29.2 Bq/m3 and the long-term atmospheric dilution factors were in the range of 10-7 ∼ 10-6 s/m3. Annual mean concentrations of tritium in groundwater were in the range of 19.2 ∼ 27.9 Bq/liter at N-1 and 64.1 ∼ 189 Bq/liter at S-2, and were generally less than 0.2 % of MPCw (222 kBq/liter). The concentrations of tritium in precipitations exponentially decreased with the distance from Wolsong Unit 1, falling to current global levels at about 25 km off-site. The highest concentration of tritium in soil moisture was observed in May and June, when the relative humidity was high. The concentrations of tritium in soil moisture were higher than those in precipitations. The results obtained in this work can be used as valuable baseline data for tritium levels around Wolsong site, and would be eventually useful to assess any environmental impact of tritium released from Wolsong Unit 2

  11. Recommended radiological controls for tritium operations

    This informal report presents recommendations for an adequate radiological protection program for tritium operations. Topics include hazards analysis, facility design, personnel protection equipment, training, operational procedures, radiation monitoring, to include surface and airborne tritium contamination, and program management

  12. Assessment of tritium using two-compartment environmental transport model

    The models used to assess chronic atmospheric releases of radioactivity generate deterministic dose estimates by using standard assumptions about exposure conditions and environmental transport mechanisms. The most significant radionuclide released in gaseous effluences of Cernavoda NPP is tritium, mostly as tritiated water, which represents about 50% of the total radioactivity release in gaseous effluents. The chemical and physical properties of HTO include its complete miscibility in water. Other important properties of tritium like diffusion coefficient and vapor pressure are similar to water properties. No preferential uptake of tritiated water biota or other component of environment was observed. The tritiated water has an uniform distribution among the aqueous phases of all interacting compartments. These properties imply that the state and transport tendencies of HTO are best predicted by modeling it as having infinite affinity for the water phase and negligible affinity for other phases (organic phase of biota or mineral phase of soil). The purpose of the present paper is to present a two-component model to estimate the equilibrium distribution of tritiated water between air and soil media, that would result from gaseous tritium emissions from Cernavoda NPP. Using the principle of mass conservation the model provides an algorithm for predicting equilibrium specific activities of tritiated water in the soil and air components. (authors)

  13. Linear accelerator for tritium production

    For many years now, Los Alamos National Laboratory has been working to develop a conceptual design of a facility for accelerator production of tritium (APT). The APT accelerator will produce high energy protons which will bombard a heavy metal target, resulting in the production of large numbers of spallation neutrons. These neutrons will be captured by a low-Z target to produce tritium. This paper describes the latest design of a room-temperature, 1.0 GeV, 100 mA, cw proton accelerator for tritium production. The potential advantages of using superconducting cavities in the high-energy section of the linac are also discussed and a comparison is made with the baseline room-temperature accelerator. copyright 1996 American Institute of Physics

  14. Linear accelerator for tritium production

    For many years now, Los Alamos National Laboratory has been working to develop a conceptual design of a facility for accelerator production of tritium (API). The APT accelerator will produce high energy protons which will bombard a heavy metal target, resulting in the production of large numbers of spallation neutrons. These neutrons will be captured by a low-Z target to produce tritium. This paper describes the latest design of a room-temperature, 1.0 GeV, 100 mA, cw proton accelerator for tritium production. The potential advantages of using superconducting cavities in the high-energy section of the linac are also discussed and a comparison is made with the baseline room-temperature accelerator

  15. Tritium processing using metal hydrides

    E.I. duPont de Nemours and Company is commissioned by the US Department of Energy to operate the Savannah River Plant and Laboratory. The primary purpose of the plant is to produce radioactive materials for national defense. In keeping with current technology, new processes for the production of tritium are being developed. Three main objectives of this new technology are to ease the processing of, ease the storage of, and to reduce the operating costs of the tritium production facility. Research has indicated that the use of metal hydrides offers a viable solution towards satisfying these objectives. The Hydrogen and Fuels Technology Division has the responsibility to conduct research in support of the tritium production process. Metal hydride technology and its use in the storage and transportation of hydrogen will be reviewed

  16. Tritium calorimeter setup and operation

    Rodgers, D E

    2002-01-01

    The LBNL tritium calorimeter is a stable instrument capable of measuring tritium with a sensitivity of 25 Ci. Measurement times range from 8-hr to 7-days depending on the thermal conductivity and mass of the material being measured. The instrument allows accurate tritium measurements without requiring that the sample be opened and subsampled, thus reducing personnel exposure and radioactive waste generation. The sensitivity limit is primarily due to response shifts caused by temperature fluctuation in the water bath. The fluctuations are most likely a combination of insufficient insulation from ambient air and precision limitations in the temperature controller. The sensitivity could probably be reduced to below 5 Ci if the following improvements were made: (1) Extend the external insulation to cover the entire bath and increase the top insulation. (2) Improve the seal between the air space above the bath and the outside air to reduce evaporation. This will limit the response drift as the water level drops. (...

  17. Modeling Tritium Life cycle in Nuclear Plants

    The mathematical development of a tritium model for nuclear power plants is presented. The model requires that the water and tritium material balance be satisfied throughout normal operations and shutdown. The model results obtained at the time of publishing include the system definitions and comparison of the model predictions of tritium generations compared to the observed plant data of the Braidwood station. A scenario that models using ion exchange resin to remove coolant boron demonstrates the tritium concentration levels are manageable. (authors)

  18. Conceptual Design of Tritium Extraction System

    Miral Thakker, Prof.Amar vaghela

    2012-01-01

    The first generation of fusion reactors will use deuterium and tritium as fuel. Since tritium is not available in nature, it must be produced in the fusion reactor blanket which surrounds the plasma zone. Tritium extraction facility has been designed and fabricated. Calibration procedure has been performed to determine tritium losses, if any during the extraction. Lithium compounds were irradiated using Am-Be neutron source. Out of pile extraction from neutron irradiated lithium compounds was...

  19. Development of a tritium transport analysis code for the LMFBR system

    A tritium transport analysis code for the LMFBR system, TTT code, has been developed and validated using data from a power rising test conducted at Monju in 1995. The behavior of tritium during future long-term full power operation of Monju has been estimated. The TTT code was created from the tritium and hydrogen transport model devised by R. Kumar and ANL. Actual data from some plants has been used to improve the code. In this study, we used data from Monju to increase the accuracy of the calculated to measured ratio, the C/E ratio. As a result of the study, we were able to: 1. show that the calculated tritium concentration distribution and the change in the primary and secondary sodium, steam and water correlated sufficiently closely with the measured, C/E ratio of 1.1; 2. propose a transport model between sodium and the cover gas system taking into account the mechanisms affecting the partial pressure difference and the isotopic exchange of H and H3; 3. examine the considerable effect of the hydrogen source within the sodium cooling system of Monju on tritium behavior and clarify the characteristics at the initial stage of plant; 4. estimate the tritium transport and distribution for the long-term full power operation of Monju. The tritium release from the core will be 7,400 TBq during 30 years of operation. The primary and secondary cold trap will capture 99% of this and 1% or less will be released to the environment as gaseous radioactive waste from stack and its drainage water from SG; and 5. compare the best fitted tritium source rates from cores in Phenix and Monju and estimate the major release from Monju's helium bond closed type control rods. (author)

  20. Effects of acute treadmill running at different intensities on activities of serotonin and corticotropin-releasing factor neurons, and anxiety- and depressive-like behaviors in rats.

    Otsuka, Tomomi; Nishii, Ayu; Amemiya, Seiichiro; Kubota, Natsuko; Nishijima, Takeshi; Kita, Ichiro

    2016-02-01

    Accumulating evidence suggests that physical exercise can reduce and prevent the incidence of stress-related psychiatric disorders, including depression and anxiety. Activation of serotonin (5-HT) neurons in the dorsal raphe nucleus (DRN) is implicated in antidepressant/anxiolytic properties. In addition, the incidence and symptoms of these disorders may involve dysregulation of the hypothalamic-pituitary-adrenal axis that is initiated by corticotropin-releasing factor (CRF) neurons in the hypothalamic paraventricular nucleus (PVN). Thus, it is possible that physical exercise produces its antidepressant/anxiolytic effects by affecting these neuronal activities. However, the effects of acute physical exercise at different intensities on these neuronal activation and behavioral changes are still unclear. Here, we examined the activities of 5-HT neurons in the DRN and CRF neurons in the PVN during 30 min of treadmill running at different speeds (high speed, 25 m/min; low speed, 15m/min; control, only sitting on the treadmill) in male Wistar rats, using c-Fos/5-HT or CRF immunohistochemistry. We also performed the elevated plus maze test and the forced swim test to assess anxiety- and depressive-like behaviors, respectively. Acute treadmill running at low speed, but not high speed, significantly increased c-Fos expression in 5-HT neurons in the DRN compared to the control, whereas high-speed running significantly enhanced c-Fos expression in CRF neurons in the PVN compared with the control and low-speed running. Furthermore, low-speed running resulted in decreased anxiety- and depressive-like behaviors compared with high-speed running. These results suggest that acute physical exercise with mild and low stress can efficiently induce optimal neuronal activation that is involved in the antidepressant/anxiolytic effects. PMID:26542811