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Sample records for tritium oxide release

  1. Release characteristics of tritium from high-purity lithium oxide

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Vogelsang, W.F.

    1985-01-01

    Rates of tritium release from neutronirradiated lithium oxide were determined from isothermal release experiments. High-purity, monocrystalline lithium oxide was purged ex-reactor with helium and helium-hydrogen gas streams. Overall release was found to be controlled by solid-phase diffusion, and was predominantly in the form of condensible species. The result of an independent concentration profile analysis at 923 K was in agreement with the gas release diffusion coefficient. Sweeping the Li 2 O with hydrogen-containing gas was found to enhance tritium removal during the early stage of each run

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

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

    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

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

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

    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 oxide with a helium carrier gas from 300 to 450 C was found to be controlled by the first order surface desorption of mono-tritiated 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. Large single crystals of lithium oxide were fabricated by the vacuum fusion technique. The release rate of tritium from the large single crystals was found to be controlled by diffusion, and the mixed diffusion-desorption controlled release regime

  4. Kinetics that govern the release of tritium from neutron-irradiated lithium oxide

    International Nuclear Information System (INIS)

    Bertone, P.C.

    1986-01-01

    The Lithium Blanket Module (LBM) program being conducted at the Princeton Plasma Physics Laboratory requires that tritium concentrations as low as 0.1 nCi/g, bred in both LBM lithium oxide pellets and gram-size lithium samples, be measured with an uncertainty not exceeding +/-6%. This thesis reports two satisfactory methods of assaying LBM pellets and one satisfactory method of assaying lithium samples. Results of a fundamental kinetic investigation are also reported. The thermally driven release of tritium from neutron-irradiated lithium oxide pellets is studied between the temperatures of 300 and 400 0 C. The observed release clearly obeys first-order kinetics, and the governing activation energy appears to be 28.4 kcal/mole. Finally, a model is presented that may explain the thermally driven release of tritium from a lithium oxide crystal and assemblies thereof. It predicts that under most circumstances the release is controlled by either the diffusion of a tritiated species through the crystal, or by the desorption of tritiated water from it

  5. Oxidative Tritium Decontamination System

    International Nuclear Information System (INIS)

    Gentile, Charles A.; Parker, John J.; Guttadora, Gregory L.; Ciebiera, Lloyd P.

    2002-01-01

    The Princeton Plasma Physics Laboratory, Tritium Systems Group has developed and fabricated an Oxidative Tritium Decontamination System (OTDS), which is designed to reduce tritium surface contamination on various components and items. The system is configured to introduce gaseous ozone into a reaction chamber containing tritiated items that require a reduction in tritium surface contamination. Tritium surface contamination (on components and items in the reaction chamber) is removed by chemically reacting elemental tritium to tritium oxide via oxidation, while purging the reaction chamber effluent to a gas holding tank or negative pressure HVAC system. Implementing specific concentrations of ozone along with catalytic parameters, the system is able to significantly reduce surface tritium contamination on an assortment of expendable and non-expendable items. This paper will present the results of various experimentation involving employment of this system

  6. Tritium oxidation and exchange: preliminary studies

    International Nuclear Information System (INIS)

    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 4 to 10 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 -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

  7. Method and apparatus for controlling accidental releases of tritium

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1980-01-01

    An improvement in a tritium control system based on a catalytic oxidation reactor is provided wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release

  8. Method and apparatus for controlling accidental releases of tritium

    Science.gov (United States)

    Galloway, Terry R. [Berkeley, CA

    1980-04-01

    An improvement in a tritium control system based on a catalytic oxidation reactor wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release.

  9. In-pile test of tritium release from tritium breeding materials (VOM-21H experiment)

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Takeshita, Hidefumi; Watanabe, Hitoshi; Yoshida, Hiroshi.

    1986-10-01

    Material development and blanket design of lithium-based ceramics such as lithium oxide, lithium aluminate, lithium silicate and lithium zirconate have been performed in Japan, United State of America and Europian Communities. Lithium oxide is a most attractive candidate for tritium breeding materials because of its high lithium density, high thermal conductivity and good tritium release performance. This work has been done to clarify the characteristics of tritium release and recovery from Li 2 O by means of in-situ tritium release measurement. The effects of temperature and sweep gas composition on the tritium release were investigated in this VOM-21H Experiment. Good measurement of tritium release was achieved but there were uncertainties in reproduciblity of data. The experimental results show that the role of surface adsorption/desorption makes a significant contribution to the tritium release and tritium inventory. Also, it is necessary to define the rate limiting process either diffusion or surface adsorption/desorption. (author)

  10. Improvement of the model for surface process of tritium release from lithium oxide

    International Nuclear Information System (INIS)

    Yamaki, Daiju; Iwamoto, Akira; Jitsukawa, Shiro

    2000-01-01

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

  11. Environmental aspects of a tritium oxide release from the Savannah River Site on September 2 and 3, 1984

    International Nuclear Information System (INIS)

    Hoel, D.D.; Kurzeja, R.J.; Evans, A.G.

    1990-01-01

    Tritium was released to the atmosphere from the Savannah River Plant during an incident on September 2 and 3, 1984 between 10 PM and 3 AM. During this five hour period, 43,800 Ci of tritium, principally in the form of the oxide (HTO), was released. An additional 14,000, Ci was released during subsequent cleanup operations between September 3 and 7. The total amount released from the incident was 57,800 Ci. The HTO cloud initially moved northward and passed near the towns of New Ellenton and Aiken, SC. Two hours after the release began, the wind shifted and carried the cloud toward Columbia, SC. The cloud moved northeast during the daytime on September 3 over the east-central portion of North Carolina. Environmental sampling teams were dispatched by SRL, SRP, and SCDHEC (South Carolina Department of Health and Environmental Control). SRL collected air and vegetation samples and SRP collected vegetation, water, milk and bioassay samples. SCDHEC collected vegetation, milk, and water samples. The highest activity of HTO measured in vegetation was 501 pCi/mL onsite, 2522 pCi/mL at the plant boundary, and 9859 pCi/mL offsite. These concentrations were approximately 100 times larger than normal values. 13 refs., 7 figs., 10 tabs

  12. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

    Ono, F.; Tanaka, S.; Yamawaki, M.

    1994-01-01

    In a fusion reactor or tritium handling facilities, contamination of concrete by tritium and subsequent release from it to the reactor or experimental rooms is a matter of problem for safety control of tritium and management of operational environment. In order to evaluate these tritium behavior, interaction of tritiated water with concrete or cement should be clarified. In the present study, HTO sorption and subsequent release from cement were studied by combining various experimental methods. From the basic studies on tritium-cement interactions, it has become possible to evaluate tritium uptake by cement or concrete and subsequent tritium release behavior as well as tritium removing methods from them

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    Freund, Jana; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

    2014-01-01

    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

  15. Calculation of tritium release from reactor's stack

    International Nuclear Information System (INIS)

    Akhadi, M.

    1996-01-01

    Method for calculation of tritium release from nuclear to environment has been discussed. Part of gas effluent contain tritium in form of HTO vapor released from reactor's stack was sampled using silica-gel. The silica-gel was put in the water to withdraw HTO vapor absorbed by silica-gel. Tritium concentration in the water was measured by liquid scintillation counter of Aloka LSC-703. Tritium concentration in the gas effluent and total release of tritium from reactor's stack during certain interval time were calculated using simple mathematic formula. This method has examined for calculation of tritium release from JRR-3M's stack of JAERI, Japan. From the calculation it was obtained the value of tritium release as much as 4.63 x 10 11 Bq during one month. (author)

  16. Biological effects of tritium releases from fusion power plants

    International Nuclear Information System (INIS)

    Strand, J.A.; Thompson, R.C.

    1976-09-01

    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

  17. Tritium behavior intentionally released in the room

    International Nuclear Information System (INIS)

    Kobayashi, K.; Hayashi, T.; Iwai, Y.; Yamanishi, T.; Willms, R. S.; Carlson, R. V.

    2008-01-01

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

  18. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

    Ono, F.; Yamawaki, M.

    1995-01-01

    In a fusion reactor or tritium-handling facilities, contamination of concrete by tritium and subsequent release from it to the reator or experimental room is a matter of problem for safe control of tritium and management of operational environment. In order to evaluate this tritium behavior, interaction of tritiated water with concrete or cement should be clarified. In the present study, HTO sorption and subsequent release from cement were experimentally studied.(1)Sorption experiments were conducted using columns packed with cement particles of different sizes. From the analysis of the breakthrough curve, tritium diffusivity in macropores and microparticles were evaluated.(2)From the short-term tritium release experiments, effective desorption rate constants were evaluated and the effects of temperature and moisture were studied.(3)In the long-term tritium release experiments to 6000h, the tritium release mechanism was found to be composed of three kinds of water: initially from capillary water, and in the second stage from gel water and from the water in the cement crystal.(4)Tritium release behavior by heat treatment to 800 C was studied. A high temperature above 600 C was required for the tritium trapped in the crystal water to be released. (orig.)

  19. Procedures for the retention of gaseous tritium released from a tritium enrichment plant

    International Nuclear Information System (INIS)

    Gutowski, H.; Bracha, M.

    1987-01-01

    General aim of the study is the comparison of two alternative processes for the retention of gaseous tritium which is released during normal operation and emergency operation in a tritium-enrichment-plant. Two processes for the retention of tritium were compared: 1. Oxidation-process. The hydrogen-gas containing HT will be burnt on an oxidation catalyst to H 2 O and HTO. In a subsequent step the water will be removed from the process by condensation, freezing and adsorption. 2. TROC-process (Tritium Removal by Organic Compounds). The tritium is added to an organic compound (acid) via catalyst. This reaction is irreversible and leads to solid products. (orig./RB) [de

  20. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.; Kopasz, J.P.; Tam, S.W.

    1994-01-01

    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

  1. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    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)

  2. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

    Engineering data on neutron irradiation performance of tritium breeders are needed to design the breeding blanket of fusion reactor. In this study, tritium release experiments of the breeders were carried out to examine the effects of various parameters (such as sweep gas flow rate, hydrogen content in sweep gas, irradiation temperature and thermal neutron flux) on tritium generation and release behavior. Lithium titanate (Li 2 TiO 3 ) is considered as a candidate tritium breeder in the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to reduce the thermal stress induced in the breeder. Li 2 TiO 3 pebbles of about 170g in total weight and with 0.3 and 2 mm in diameter were manufactured by a wet process, and an assembly packed with the binary Li 2 TiO 3 pebbles was irradiated in Japan Materials Testing Reactor (JMTR). The tritium was generated in the Li 2 TiO 3 pebble bed and released from the pebble bed, and was swept downstream using the sweep gas for on-line analysis of tritium content. Concentration of total tritium and gaseous tritium (HT or T 2 gas) released from the Li 2 TiO 3 pebble bed were measured by ionization chambers, and the ratio of (gaseous tritium)/(total tritium) was evaluated. The sweep gas flow rate was changed from 100 to 900cm 3 /min, and hydrogen content in the sweep gas was changed from 100 to 10000 ppm. Furthermore, thermal neutron flux was changed using a window made of hafnium (Hf) neutron absorber. The irradiation temperature at an outer region of the Li 2 TiO 3 pebble bed was held between 200 and 400degC. The main results of this experiment are summarized as follows. 1) When the temperature at the outside edge of the Li 2 TiO 3 pebble bed exceeded 100degC, the tritium release from the Li 2 TiO 3 pebble bed started. The ratio of the tritium release rate and the tritium generation rate (normalized tritium release rate: R/G) reached

  3. Analysis of in-pile tritium release experiments

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Tam, S.W.; Johnson, C.E.

    1992-01-01

    The objective of this work is to characterize tritium release behavior from lithium ceramics and develop insight into the underlying tritium release mechanisms. Analysis of tritium release data from recent laboratory experiments with lithium aluminate has identified physical processes which were previously unaccounted for in tritium release models. A new model that incorporates the recent data and provides for release from multiple sites rather than only one site was developed. Calculations of tritium release using this model are in excellent agreement with the tritium release behavior reported for the MOZART experiment

  4. Tritium releases, birth defects and infant deaths

    International Nuclear Information System (INIS)

    1991-01-01

    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

  5. Evaluation of tritium release behavior from Li{sub 2}TiO{sub 3} during DT neutron irradiation by use of an improved tritium collection method

    Energy Technology Data Exchange (ETDEWEB)

    Edao, Yuki, E-mail: edao.yuki@jaea.go.jp [Tritium Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kawamura, Yoshinori [Blanket Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Hoshino, Tsuyoshi [Breeding Functional Materials Development Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Ochiai, Kentaro [BA Project Coordination Group, Department of Fusion Power Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-15

    Highlights: • Behavior of tritium released from Li{sub 2}TiO{sub 3} under neutron irradiation was measured. • Tritium collection method with hydrophobic catalyst was demonstrated successfully. • Temperature of Li{sub 2}TiO{sub 3} was dominant to control the chemical form of tritium release. - Abstract: The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li{sub 2}TiO{sub 3} during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li{sub 2}TiO{sub 3}. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li{sub 2}TiO{sub 3} as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li{sub 2}TiO{sub 3} by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li{sub 2}TiO{sub 3} irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li{sub 2}TiO{sub 3} was the dominant parameter to control the chemical form of tritium released from the Li{sub 2}TiO{sub 3}.

  6. Release of gaseous tritium during reprocessing

    International Nuclear Information System (INIS)

    Bruecher, H.; Hartmann, K.

    1983-01-01

    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.) [de

  7. Tritium release of titan-tritium layers in air, aqueous solutions and living organisms of animals

    International Nuclear Information System (INIS)

    Biro, J.; Feher, I.; Mate, L.; Varga, L.

    1978-01-01

    Samples containing 400-1100 MBq (10-30 mCi) tritium were prepared and the effect of storage time on tritium release was followed. In 250 days one thousandth of the tritium was released in aqueous solution; in air the ratio of release per hour fell in the range of 10 -6 -10 -7 . Ti-T plates with different storage times were surgically placed in the abdomen of rats. Their tritium release dropped with time and the activity appearing in the circulation was lower than that of plates with 5-6 orders of magnitude. Checking the tritium incorporation of neutron generator operators it must be held in mind that only a minor part of tritium can be detected by the measurement of the tritium content of urine. (author)

  8. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Kolb, Matthias H.H., E-mail: matthias.kolb@kit.edu; Rolli, Rolf; Knitter, Regina

    2017-06-15

    The tritium loading of current grades of advanced ceramic breeder pebbles with three different lithium orthosilicate (LOS)/lithium metatitanate (LMT) compositions (20–30 mol% LMT in LOS) and pebbles of EU reference material, was performed in a consistent way. The temperature dependent release of the introduced tritium was subsequently investigated by temperature programmed desorption (TPD) experiments to gain insight into the desorption characteristics. The obtained TPD data was decomposed into individual release mechanisms according to well-established desorption kinetics. The analysis showed that the pebble composition of the tested samples does not severely change the release behaviour. Yet, an increased content of lithium metatitanate leads to additional desorption peaks at medium temperatures. The majority of tritium is released by high temperature release mechanisms of chemisorbed tritium, while the release of physisorbed tritium is marginal in comparison. The results allow valuable projections for the tritium release behaviour in a fusion blanket.

  9. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Science.gov (United States)

    Kolb, Matthias H. H.; Rolli, Rolf; Knitter, Regina

    2017-06-01

    The tritium loading of current grades of advanced ceramic breeder pebbles with three different lithium orthosilicate (LOS)/lithium metatitanate (LMT) compositions (20-30 mol% LMT in LOS) and pebbles of EU reference material, was performed in a consistent way. The temperature dependent release of the introduced tritium was subsequently investigated by temperature programmed desorption (TPD) experiments to gain insight into the desorption characteristics. The obtained TPD data was decomposed into individual release mechanisms according to well-established desorption kinetics. The analysis showed that the pebble composition of the tested samples does not severely change the release behaviour. Yet, an increased content of lithium metatitanate leads to additional desorption peaks at medium temperatures. The majority of tritium is released by high temperature release mechanisms of chemisorbed tritium, while the release of physisorbed tritium is marginal in comparison. The results allow valuable projections for the tritium release behaviour in a fusion blanket.

  10. Tritium release experiments with CATS and numerical simulation

    International Nuclear Information System (INIS)

    Munakata, Kenzo; Wajima, Takaaki; Hara, Keisuke; Wada, Kohei; Takeishi, Toshiharu; Shinozaki, Yohei; Mochizuki, Kazuhiro; Katekari, Kenichi; Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko

    2010-01-01

    In D-T fusion power plants, large amounts of tritium would be handled. Tritium is the radioisotope of protium, and is easily taken into the human body, and thus the behavior of tritium accidentally released in fusion power plants should be studied for the safety design and radioprotection of workers. Therefore, it is necessary to investigate the behavior of tritium released into large rooms with objectives, since complex flow fields should exist in such rooms and they could influence the ventilation of the air containing released tritium. Thus, tritium release experiments were conducted using Caisson Assembly for Tritium Safety Study (CATS) in TPL/JAEA. Some data were taken for tritium behavior in the ventilated area and response of tritium monitors. In the experiments, approximately 17 GBq of tritium was released into Caisson with the total volume of 12 m 3 , and the room was ventilated at the rate of 12 m 3 /h after release of tritium. It was found that placement of an objective in the vessel substantially affects decontamination efficiency. With regard to an experimental result, numerical calculation was performed and the experimental result and the result of numerical calculation were compared, which indicates that experimental results are qualitatively reproduced by numerical calculation. However, further R and D needs to be carried out for quantitative reproduction of the experimental results.

  11. Tritium release experiments with CATS and numerical simulation

    Energy Technology Data Exchange (ETDEWEB)

    Munakata, Kenzo, E-mail: kenzo@gipc.akita-u.ac.jp [Faculty of Engineering and Resource Sciences, Akita University, Tegata-gakuen-cho 1-1, Akita 010-8502 (Japan); Wajima, Takaaki; Hara, Keisuke; Wada, Kohei [Faculty of Engineering and Resource Sciences, Akita University, Tegata-gakuen-cho 1-1, Akita 010-8502 (Japan); Takeishi, Toshiharu; Shinozaki, Yohei; Mochizuki, Kazuhiro; Katekari, Kenichi [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581 (Japan); Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko [Tritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2010-12-15

    In D-T fusion power plants, large amounts of tritium would be handled. Tritium is the radioisotope of protium, and is easily taken into the human body, and thus the behavior of tritium accidentally released in fusion power plants should be studied for the safety design and radioprotection of workers. Therefore, it is necessary to investigate the behavior of tritium released into large rooms with objectives, since complex flow fields should exist in such rooms and they could influence the ventilation of the air containing released tritium. Thus, tritium release experiments were conducted using Caisson Assembly for Tritium Safety Study (CATS) in TPL/JAEA. Some data were taken for tritium behavior in the ventilated area and response of tritium monitors. In the experiments, approximately 17 GBq of tritium was released into Caisson with the total volume of 12 m{sup 3}, and the room was ventilated at the rate of 12 m{sup 3}/h after release of tritium. It was found that placement of an objective in the vessel substantially affects decontamination efficiency. With regard to an experimental result, numerical calculation was performed and the experimental result and the result of numerical calculation were compared, which indicates that experimental results are qualitatively reproduced by numerical calculation. However, further R and D needs to be carried out for quantitative reproduction of the experimental results.

  12. Simulation of tritium behavior after intended tritium release in ventilated room

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko; Kobayashi, Kazuhiro; Nishi, Masataka

    2001-01-01

    At the Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI), Caisson Assembly for Tritium Safety study (CATS) with 12 m 3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate tritium behavior in the case where a tritium leak event should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak event should happen in a ventilated room. The RNG model was found to be valid for eddy flow calculation in the 50 m 3 /h ventilated Caisson with acceptable engineering precision. The calculated initial and removal tritium concentration histories after intended tritium release were consistent with the experimental observations in the 50 m 3 /h ventilated Caisson. It is found that the flow near a wall plays an important role for the tritium transport in the ventilated room. On the other hand, tritium behavior intentionally released in the 3,000 m 3 of tritium handling room was investigated experimentally under a US-Japan collaboration. The tritium concentration history calculated with the same method was consistent with the experimental observations, which proves that the present developed method can be applied to the actual scale of tritium handling room. (author)

  13. Lithium ceramics: sol-gel preparation and tritium release

    International Nuclear Information System (INIS)

    Renoult, O.

    1994-04-01

    Ceramics based on lithium aluminate (LiA1O 2 ), lithium zirconate (Li 2 ZrO 3 ) and lithium titanate (Li 2 TiO 3 ) are candidates as tritium breeder blanket materials for forthcoming nuclear fusion reactors. Lithium silico-aluminate Li 4+x A1 4-3x Si 2x O 8 (0 ≤ x ≤ 0,25) powders were synthetized from alkoxyde-hydroxyde sol-gel route. By direct sintering at 850-1100 deg C (without prior calcination), ceramics with controlled stoichiometry and homogenous microstructure were obtained. We have also prepared, using a comparable method, Li 2 Zr 1-x Ti x O 3 (x = 0, x = 0,1 et x = 1) materials. All these ceramics, with different microstructures and compositions, have been tested in out-of-reactor experiments. Concerning lithium aluminate microporous ceramics, the silicon substitution leads to a significant improvement of the tritrium release. Classical models taking into account independent surface mechanisms are not able to describe correctly the observed tritium release kinetics. We show, using a simple model, that the release kinetics is in fact limited by an intergranular diffusion followed by a desorption. The delay in tritium release, which occurs when the ceramic compacity increases, is explained in terms of an enhancement of the ionic T + diffusion path length. The energy required for desorption includes a leading term independent of hydrogen contained in the sweep gas. This term is attributed to the limiting recombination step of T + in molecular species HTO. For similar microstructures, the facility of tritium release for the different studied materials is explained by three properties: the crystal structure of the ceramic, the acidity of oxides and finally the presence of electronic non-stoichiometric defects. (author). 89 refs., 50 figs., 2 tabs., 1 annexe

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

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.

    1990-12-01

    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.) [de

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

    International Nuclear Information System (INIS)

    Duran, J.; Malatova, I.

    2009-01-01

    A computer code PTM H TO 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

  16. Tritium release reduction and radiolysis gas formation

    Energy Technology Data Exchange (ETDEWEB)

    Batifol, G.; Douche, Ch.; Sejournant, Ch. [CEA Valduc, 21 - Is-sur-Tille (France)

    2008-07-15

    At CEA Valduc, the usual tritiated waste container is the steel drum. It allows good release reduction performance for middle activity waste but in some cases tritium outgassing from the waste drums is too high. It was decided to over-package each drum in a tighter container called the over-drum. According to good safety practices it was also decided to measure gas composition evolution into the over-drum in order to defect hydrogen formation over time. After a few months, a significant release reduction was observed. Additionally there followed contamination reduction in the roof storage building rainwater. However hydrogen was also observed in some over-drums, in addition to other radiolysis products. Catalyst will be added to manage the hydrogen risk in the over-drums. (authors)

  17. Environmental contamination due to release of a large amount of tritium

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    1988-01-01

    Tritium release incidents have occurred many times in the Savannah Rever Plant in the U.S. A tritium release incident also took place in the Lawrence Livermore Laboratory. The present article outlines the reports by the plant and laboratory on these incidents and makes some comments on environmental contamination that may results from release of a large amount of tritium from nuclear fusion facilities. Tritium is normally released in the form of a combination of chemical compounds such as HT, DT and T 2 and oxides such as HTO, DTO and T 2 O. The percentage of the oxides is given in the reports by the plant. Oxides, which can be absorbed through the skin, are considered to be nearly a thousand times more toxic than the other type of tritium compounds. The HT type compounds (HT, DT and T 2 ) can be oxidized by microorganisms in soil into oxides (HTO, DTO and T 2 O) and therefore, great care should also given to this type of compounds. After each accidental tritium release, the health physics group of the plant collected various environmental samples, including ground surface water, milk, leaves of plants, soil and human urine, in leeward areas. Results on the contamination of surface water, fish and underground water are outlined and discussed. (Nogami, K.)

  18. The hazard to man of accidental releases of tritium

    International Nuclear Information System (INIS)

    Brearley, I.R.

    1985-03-01

    Some aspects of the atmospheric dispersion of tritium are discussed, followed by consideration of the dosimetric pathways. In order to assess the significance of a tritium release the doses from various pathways are estimated and compared with the doses estimated from a similar release of iodine-131. The major hazard from tritium is the ingestion of contaminated food products. For similar releases of tritium and I 131 the ingestion hazard can be comparable if the release occurs near and before the end of the harvest season. However, in the tritium release case the agricultural season influences the consequences markedly and, at other times during the year, the ingestion hazard from tritium may be approximately 20 times less. The dose from inhalation of tritium is sensitive to its chemical form and for similar releases of tritiated water and tritium gas then the dose from tritiated water is approximately 10 4 greater than the dose from tritium gas. For similar releases of tritiated water and iodine-131 then a comparison of the inhalation shows that the dose from the iodine is approximately 300 times greater. (author)

  19. Release enhancement of tritium from graphite by addition of hydrogen

    International Nuclear Information System (INIS)

    Saeki, Masakatsu; Masaki, N.M.

    1989-01-01

    The release behavior of tritium from graphite was studied in pure He and He + H 2 atmosphere. The release from powdered graphite was significantly enhanced in hydrogen environment. Apparent diffusion coefficients of tritium in graphite also became much higher in an atmosphere containing hydrogen than values obtained in pure helium atmosphere. A careful investigation of the release processes resulted in the conclusion that the most important process of tritium behaviour in graphite was diffusion, but the desorption process of tritium from the surface played a significant role. The enhancement of the desorption process was controlled by atomic hydrogen. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Oxidation of tritium by hopcalite bed

    Energy Technology Data Exchange (ETDEWEB)

    Nishikawa, Masabumi; Shinnai, Kohsuke; Matsunaga, Sohichi; Kinoshita, Yoshihiko

    1984-08-01

    Oxidation by the catalyst bed with a metal oxide and subsequent adsorption to the porous dehydrative reagents is supposed to be effective process for scavenging tritium from an inert atmosphere. Use of spongy copper oxide or wires of copper oxide is not recommended to use as the metal oxide catalyst from the view point of mass transfer because of sintering and of limited effective surface area. Use of hopcalites and copper oxide-kieselguhr are examined in this study and it is concluded that hopcalites are more suitable as the metal oxide catalyst because they not only remain the oxidation power on hydrogen isotopes even at an ambient temperature, but also show a negligible drop in oxidation performances with repeated regeneration. The effective temperature is about 400/sup 0/C for hopcalites and 300-600/sup 0/C for copper oxide-kieselguhr to use as the oxidation bed of tritium.

  2. Oxidation of tritium by hopcalite bed

    International Nuclear Information System (INIS)

    Nishikawa, Masabumi; Shinnai, Kohsuke; Matsunaga, Sohichi; Kinoshita, Yoshihiko

    1984-01-01

    Oxidation by the catalyst bed with a metal oxide and subsequent adsorption to the porous dehydrative reagents is supposed to be effective process for scavenging tritium from an inert atmosphere. Use of spongy copper oxide or wires of copper oxide is not recommended to use as the metal oxide catalyst from the view point of mass transfer because of sintering and of limited effective surface area. Use of hopcalites and copper oxide-kieselguhr are examined in this study and it is concluded that hopcalites are more suitable as the metal oxide catalyst because they not only remain the oxidation power on hydrogen isotopes even at an ambient temperature, but also show a negligible drop in oxidation performances with repeated regeneration. The effective temperature is about 400 0 C for hopcalites and 300--600 0 C for copper oxide-kieselguhr to use as the oxidation bed of tritium. (author)

  3. Investigation of tritium release and retention in lithium aluminate

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Tistchenko, S.; Botter, F.

    1991-01-01

    Tritium release from lithium aluminate, although previously investigated by both in-reactor and ex-reactor experiments, remains poorly understood. Agreement between experiments is lacking, and the mechanisms responsible for tritium release from lithium aluminate are under debate. In an effort to improve our understanding of the mechanisms of tritium release from lithium ceramics, we have investigated tritium release from pure lithium aluminate and lithium aluminate doped with impurities. The results of these experiments on large grain size material indicate that after anneals at low temperature, a large fraction of the tritium present before the anneal remains in the sample. We have modeled this behavior based on first-order release from three types of sites. At the lowest temperature, the release is dominated by one site, while the tritium in the other sites is retained in the solid. Adding magnesium dopant to the ceramic appears to alter the distribution of tritium between the sites. This addition decreases the fraction of tritium released at 777 degree C, while increasing the fractions released at 538 and 950 degree C. 11 refs., 8 figs., 1 tab

  4. Tritium release from lithium ceramics at constant temperature

    International Nuclear Information System (INIS)

    Verrall, R.A.; Miller, J.M.

    1992-02-01

    Analytic methods for post-irradiation annealing tests to measure tritium release from lithium ceramics at constant temperature are examined. Modifications to the Bertone (1) relations for distinguishing diffusion-controlled release from desorption-controlled release are shown. The methods are applied to tests on sintered LiA10 2 ; first-order desorption is shown to control tritium release for these tests

  5. Tritium release during inspection of reactor 'RA' at 'Vinca' institute

    International Nuclear Information System (INIS)

    Sipka, V.; Miljevic, N.; Grsic, Z.; Todorovic, D.; Radenkovic, M.

    1997-01-01

    Tritium content in daily precipitation, atmospheric water vapor inside of the reactor hall and around 'Vinca' Institute as well as in soil up to 800 m distance was monitoring during the regular inspection of the fuel channels. Tritium activity in the reactor hall air moisture was in the range from 0.022 to 6.7 MBq/m 3 . Tritium content in soil moisture between 12.7 and 530.9 Bq/l indicate a certain contamination due to tritium release in the environment, depending on the depth and distance from the place of release (author) [sr

  6. Analysis of tritium releases to the atmosphere by a CTR

    International Nuclear Information System (INIS)

    Renne, D.S.; Sandusky, W.F.; Dana, M.T.

    1975-08-01

    Removal by atmospheric processes of routinely and accidentally released tritium from a controlled thermonuclear reactor (CTR) was investigated. Based on previous studies, the assumed form of the tritium for this analysis was HTO or tritiated water vapor. Assuming a CTR operation in Morris, Illinois, surface water and ground-level air concentration values of tritium were computed for three space (or time) scales: local (50 Km of a plant), regional (up to 1000 Km of the plant), and global

  7. Tritium recovery from lithium oxide pellets

    International Nuclear Information System (INIS)

    Bertone, P.C.; Jassby, D.L.

    1984-01-01

    The TFTR Lithium Blanket Module is an assembly containing 650 kg of lithium oxide that will be used to test the ability of neutronics codes to model the tritium breeding characteristics of limited-coverage breeding zones in a tokamak. It is required that tritium concentrations as low as 0.1 nCi/g bred in both metallic lithium samples and lithium oxide pellets be measured with an uncertainty not exceeding +- 6%. A tritium assay technique for the metallic samples which meets this criterion has been developed. Two assay techniques for the lithium oxide pellets are being investigated. In one, the pellets are heated in a flowing stream of hydrogen, while in the other, the pellets are dissolved in 12 M hydrochloric acid

  8. Tritium

    International Nuclear Information System (INIS)

    Fiege, A.

    1992-07-01

    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.) [de

  9. Tritium release from lithium titanate, a low-activation tritium breeding material

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Miller, J.M.; Johnson, C.E.

    1994-01-01

    The goals for fusion power are to produce energy in as safe, economical, and environmentally benign a manner as possible. To ensure environmentally sound operation low-activation materials should be used where feasible. The ARIES Tokamak Reactor Study has based reactor designs on the concept of using low-activation materials throughout the fusion reactor. For the tritium breeding blanket, the choices for low activation tritium breeding materials are limited. Lithium titanate is an alternative low-activation ceramic material for use in the tritium breeding blanket. To date, very little work has been done on characterizing the tritium release for lithium titanate. We have thus performed laboratory studies of tritium release from irradiated lithium titanate. The results indicate that tritium is easily removed from lithium titanate at temperatures as low as 600 K. The method of titanate preparation was found to affect the tritium release, and the addition of 0.1% H 2 to the helium purge gas did not improve tritium recovery. ((orig.))

  10. Tritium release from advanced beryllium materials after loading by tritium/hydrogen gas mixture

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, Vladimir, E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, Rolf; Moeslang, Anton; Kurinskiy, Petr; Vladimirov, Pavel [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Dorn, Christopher [Materion Beryllium & Composites, 6070 Parkland Boulevard, Mayfield Heights, OH 44124-4191 (United States); Kupriyanov, Igor [Bochvar Russian Scientific Research Institute of Inorganic Materials, Rogova str., 5, 123098 Moscow (Russian Federation)

    2016-06-15

    Highlights: • A major tritium release peak for beryllium samples occurs at temperatures higher than 1250 K. • A beryllium grade with comparatively smaller grain size has a comparatively higher tritium release compared to the grade with larger grain size. • The pebbles of irregular shape with the grain size of 10–30 μm produced by the crushing method demonstrate the highest tritium release rate. - Abstract: Comparison of different beryllium samples on tritium release and retention properties after high-temperature loading by tritium/hydrogen gas mixture and following temperature-programmed desorption (TPD) tests has been performed. The I-220-H grade produced by hot isostatic pressing (HIP) having the smallest grain size, the pebbles of irregular shape with the smallest grain size (10–30 μm) produced by the crushing method (CM), and the pebbles with 1 mm diameter produced by the fluoride reduction method (FRM) having a highly developed inherent porosity show the highest release rate. Grain size and porosity are considered as key structural parameters for comparison and ranking of different beryllium materials on tritium release and retention properties.

  11. Development of a tritium monitor combined with an electrochemical tritium pump using a proton conducting oxide

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, M. [National Institute for Fusion Science, Toki, Gifu (Japan); Sugiyama, T. [Nagoya University, Fro-cho, Chikusa-ku, Nagoya (Japan)

    2015-03-15

    The detection of low level tritium is one of the key issues for tritium management in tritium handling facilities. Such a detection can be performed by tritium monitors based on proton conducting oxide technique. We tested a tritium monitoring system composed of a commercial proportional counter combined with an electrochemical hydrogen pump equipped with CaZr{sub 0.9}In{sub 0.1}O{sub 3-α} as proton conducting oxide. The hydrogen pump operated at 973 K under electrolysis conditions using tritiated water vapor (HTO). The proton conducting oxide extracts tritium molecules (HT) from HTO and tritium concentration is measured by the proportional counter. The advantage of the proposed tritium monitoring system is that it is able to convert HTO into molecular hydrogen.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  13. Organically bound tritium (OBT) for various plants in the vicinity of a continuous atmospheric tritium release

    International Nuclear Information System (INIS)

    Vichot, L.; Boyer, C.; Boissieux, T.; Losset, Y.; Pierrat, D.

    2008-01-01

    In order to quantify tritium impact on the environmental, we studied vegetation continuously exposed to a tritiated atmosphere. We chose lichens as bio-indicators, trees for determination of past tritium releases of the Valduc Centre, and lettuce as edible vegetables for dose calculation regarding neighbourhood. The Pasquill and Doury models from the literature were tested to estimate tritium concentration in the air around vegetable for distance from the release point less than 500 m. The results in tree rings show that organically bound tritium (OBT) concentration was strongly correlated with tritium releases. Using the GASCON model, the modelled variation of OBT concentration with distance was correlated with the measurements. Although lichens are recognized as bio-indicators, our experiments show that they were not convenient for environmental surveys because their age is not definitive. Thus, tritium integration time cannot be precisely determined. Furthermore, their biological metabolism is not well known and tritium concentration appears to be largely dependent on species. An average conversion rate of HTO to OBT was determined for lettuce of about 0.20-0.24% h -1 . Nevertheless, even if it is equivalent to values already published in the literature for other vegetation, we have shown that this conversion rate, established by weekly samples, varies by a factor of 10 during the different stages of lettuce development, and that its variation is linked to the biomass derivative

  14. Organically bound tritium (OBT) for various plants in the vicinity of a continuous atmospheric tritium release

    Energy Technology Data Exchange (ETDEWEB)

    Vichot, L. [Commissariat a l' Energie Atomique, CVA/DSTA/SPR/LMSE, 21120 Is-sur-Tille (France)], E-mail: laurent.vichot@cea.fr; Boyer, C.; Boissieux, T.; Losset, Y.; Pierrat, D. [Commissariat a l' Energie Atomique, CVA/DSTA/SPR/LMSE, 21120 Is-sur-Tille (France)

    2008-10-15

    In order to quantify tritium impact on the environmental, we studied vegetation continuously exposed to a tritiated atmosphere. We chose lichens as bio-indicators, trees for determination of past tritium releases of the Valduc Centre, and lettuce as edible vegetables for dose calculation regarding neighbourhood. The Pasquill and Doury models from the literature were tested to estimate tritium concentration in the air around vegetable for distance from the release point less than 500 m. The results in tree rings show that organically bound tritium (OBT) concentration was strongly correlated with tritium releases. Using the GASCON model, the modelled variation of OBT concentration with distance was correlated with the measurements. Although lichens are recognized as bio-indicators, our experiments show that they were not convenient for environmental surveys because their age is not definitive. Thus, tritium integration time cannot be precisely determined. Furthermore, their biological metabolism is not well known and tritium concentration appears to be largely dependent on species. An average conversion rate of HTO to OBT was determined for lettuce of about 0.20-0.24% h{sup -1}. Nevertheless, even if it is equivalent to values already published in the literature for other vegetation, we have shown that this conversion rate, established by weekly samples, varies by a factor of 10 during the different stages of lettuce development, and that its variation is linked to the biomass derivative.

  15. Organically bound tritium (OBT) for various plants in the vicinity of a continuous atmospheric tritium release.

    Science.gov (United States)

    Vichot, L; Boyer, C; Boissieux, T; Losset, Y; Pierrat, D

    2008-10-01

    In order to quantify tritium impact on the environmental, we studied vegetation continuously exposed to a tritiated atmosphere. We chose lichens as bio-indicators, trees for determination of past tritium releases of the Valduc Centre, and lettuce as edible vegetables for dose calculation regarding neighbourhood. The Pasquill and Doury models from the literature were tested to estimate tritium concentration in the air around vegetable for distance from the release point less than 500 m. The results in tree rings show that organically bound tritium (OBT) concentration was strongly correlated with tritium releases. Using the GASCON model, the modelled variation of OBT concentration with distance was correlated with the measurements. Although lichens are recognized as bio-indicators, our experiments show that they were not convenient for environmental surveys because their age is not definitive. Thus, tritium integration time cannot be precisely determined. Furthermore, their biological metabolism is not well known and tritium concentration appears to be largely dependent on species. An average conversion rate of HTO to OBT was determined for lettuce of about 0.20-0.24% h(-1). Nevertheless, even if it is equivalent to values already published in the literature for other vegetation, we have shown that this conversion rate, established by weekly samples, varies by a factor of 10 during the different stages of lettuce development, and that its variation is linked to the biomass derivative.

  16. Dose Assessment Model for Chronic Atmospheric Releases of Tritium

    International Nuclear Information System (INIS)

    Shen Huifang; Yao Rentai

    2010-01-01

    An improved dose assessment model for chronic atmospheric releases of tritium was proposed. The proposed model explicitly considered two chemical forms of tritium.It was based on conservative assumption of transfer of tritiated water (HTO) from air to concentration of HTO and organic beam tritium (OBT) in vegetable and animal products.The concentration of tritium in plant products was calculated based on considering dividedly leafy plant and not leafy plant, meanwhile the concentration contribution of tritium in the different plants from the tritium in soil was taken into account.Calculating the concentration of HTO in animal products, average water fraction of animal products and the average weighted tritium concentration of ingested water based on the fraction of water supplied by each source were considered,including skin absorption, inhalation, drinking water and food.Calculating the annual doses, the ingestion doses were considered, at the same time the contribution of inhalation and skin absorption to the dose was considered. Concentrations in foodstuffs and dose of annual adult calculated with the specific activity model, NEWTRI model and the model proposed by the paper were compared. The results indicate that the model proposed by the paper can predict accurately tritium doses through the food chain from chronic atmospheric releases. (authors)

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

    International Nuclear Information System (INIS)

    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. - Highlights: • Tritium can be incorporated in all the hydrogenated components of milk. • Components' isotopic ratios T/H of chronically exposed milk remain in the same range. • In environmental conditions, distribution of tritium in milk components varies. • Metabolism plays a role in the distribution of tritium in the components of milk. • In environmental conditions, dilution of hydrogen dims possible isotopic effects.

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

    Science.gov (United States)

    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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Dose estimation models for environmental tritium released from fusion facilities

    International Nuclear Information System (INIS)

    Murata, Mikio

    1993-01-01

    Various mathematical models are being developed to predict the behavior of HT released to the natural environment and their consequent impact. This report outlines models and the major findings of HT field release studies in France and Canada. The models are constructed to incorporate the key processes thought to be responsible for the formation of atmospheric HTO from a release of HT. It has been established from the experiments that HT oxidized in surface soil is incorporated almost entirely into soil water as HTO. This tritium may be reemitted to the atmosphere in the form of HTO through exchange of soil and atmospheric moisture as well as through the bulk water mass flux from the soil the atmosphere due to evaporation and transpiration. The direct conversion of HT to HTO in air and direct uptake of HT by vegetation are expected to be negligible for the time and space scales of interest in considering short duration releases. HTO emitted to the atmosphere is can further exchange with soil and vegetation water. Validation of these models against experimental data is conducted to demonstrate their credibility. It may be concluded that further laboratory and field works are needed in order to develop a sufficiently good understanding of the dependence of the key processes on environmental factors (including diurnal cycling and seasonality) to allow the rates of the processes to be predicted from a knowledge of environmental conditions. (author)

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

    International Nuclear Information System (INIS)

    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

  1. Tritium and helium retention and release from irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    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)

  2. Release of tritium from fuel and collection for storage

    International Nuclear Information System (INIS)

    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

  3. Pebble fabrication and tritium release properties of an advanced tritium breeder

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi, E-mail: hoshino.tsuyoshi@jaea.go.jp [Breeding Functional Materials Development Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Obuch, Omotedate, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan); Edao, Yuki [Tritium Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-4 Shirakata, Shirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kawamura, Yoshinori [Blanket Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ochiai, Kentaro [BA Project Coordination Group, Department of Fusion Power Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Obuch, Omotedate, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Li{sub 2}TiO{sub 3} with excess Li (Li{sub 2+x}TiO{sub 3+y}) pebble as an advanced tritium breeders was fabricated using emulsion method. • Grain size of Li{sub 2+x}TiO{sub 3+y} pebbles was controlled to be less than 5 μm. • Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties similar to that of Li{sub 2}TiO{sub 3} pebbles. - Abstract: Li{sub 2}TiO{sub 3} with excess Li (Li{sub 2+x}TiO{sub 3+y}) has been developed as an advanced tritium breeder. With respect to the tritium release characteristics of the blanket, the optimum grain size after sintering was less than 5 μm. Therefore, an emulsion method was developed to fabricate pebbles with this target grain size. The predominant factor affecting grain growth was assumed to be the presence of binder in the gel particles; this remaining binder was hypothesized to react with the excess Li, thereby generating Li{sub 2}CO{sub 3}, which promotes grain growth. To inhibit the generation of Li{sub 2}CO{sub 3}, calcined Li{sub 2+x}TiO{sub 3+y} pebbles were sintered under vacuum and subsequently under a 1% H{sub 2}–He atmosphere. The average grain size of the sintered Li{sub 2+x}TiO{sub 3+y} pebbles was less than 5 μm. Furthermore, the tritium release properties of Li{sub 2+x}TiO{sub 3+y} pebbles were evaluated, and deuterium–tritium (DT) neutron irradiation experiments were performed at the Fusion Neutronics Source facility in the Japan Atomic Energy Agency. To remove the tritium produced by neutron irradiation, 1% H{sub 2}–He purge gas was passed through the Li{sub 2+x}TiO{sub 3+y} pebbles. The Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties, similar to those of Li{sub 2}TiO{sub 3} pebbles. In particular, the released amount of tritiated hydrogen gas for easier tritium handling was greater than the released amount of tritiated water.

  4. Lithium ceramics: sol-gel preparation and tritium release; Ceramiques lithiees: elaboration sol-gel et relachement du tritium

    Energy Technology Data Exchange (ETDEWEB)

    Renoult, O

    1994-04-01

    Ceramics based on lithium aluminate (LiA1O{sub 2}), lithium zirconate (Li{sub 2}ZrO{sub 3}) and lithium titanate (Li{sub 2}TiO{sub 3}) are candidates as tritium breeder blanket materials for forthcoming nuclear fusion reactors. Lithium silico-aluminate Li{sub 4+x}A1{sub 4-3x}Si{sub 2x}O{sub 8} (0 {<=} x {<=} 0,25) powders were synthetized from alkoxyde-hydroxyde sol-gel route. By direct sintering at 850-1100 deg C (without prior calcination), ceramics with controlled stoichiometry and homogenous microstructure were obtained. We have also prepared, using a comparable method, Li{sub 2}Zr{sub 1-x}Ti{sub x}O{sub 3} (x = 0, x = 0,1 et x = 1) materials. All these ceramics, with different microstructures and compositions, have been tested in out-of-reactor experiments. Concerning lithium aluminate microporous ceramics, the silicon substitution leads to a significant improvement of the tritrium release. Classical models taking into account independent surface mechanisms are not able to describe correctly the observed tritium release kinetics. We show, using a simple model, that the release kinetics is in fact limited by an intergranular diffusion followed by a desorption. The delay in tritium release, which occurs when the ceramic compacity increases, is explained in terms of an enhancement of the ionic T{sup +} diffusion path length. The energy required for desorption includes a leading term independent of hydrogen contained in the sweep gas. This term is attributed to the limiting recombination step of T{sup +} in molecular species HTO. For similar microstructures, the facility of tritium release for the different studied materials is explained by three properties: the crystal structure of the ceramic, the acidity of oxides and finally the presence of electronic non-stoichiometric defects. (author). 89 refs., 50 figs., 2 tabs., 1 annexe.

  5. Oxidation of molecular tritium by intact soils

    International Nuclear Information System (INIS)

    Sweet, C.W.; Murphy, C.E. Jr.

    1980-01-01

    The effects of environmental factors on the rate of oxidation of molecular tritium (T 2 ) to tritiated water (HTO) were determined for intact soils during field exposures. Maximum deposition velocities of approximately 0.03 cm/sec were measured for T 2 at low wind speeds for a variety of soils over a wide range of conditions. Deposition velocities were slightly inhibited in wet soils and at 0 0 C. In dry soils, oxidation of T 2 to HTO occurred deeper in the soil profile, but deposition velocities were unaffected

  6. Simulation study of intentional tritium release experiments in the caisson assembly for tritium safety at the TPL/JAERI

    International Nuclear Information System (INIS)

    Iwai, Y.; Hayashi, T.; Kobayashi, K.; Nishi, M.

    2001-01-01

    At the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI), Caisson assembly for tritium safety study (CATS) with 12 m 3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate the tritium behavior in the case, where the tritium leak accident should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak accident should happen in a ventilated room. As for the understanding of initial tritium behavior until the tritium concentration become steady, the precise estimation of local flow rate in a room and time-dependent release behavior from the leak point are essential to predict the tritium behavior by simulation code. The three-dimensional eddy flow model considering, tritium-related phenomena was adopted to estimate the local flow rate in the 50 m 3 /h ventilated Caisson. The time-dependent tritium release behavior from the sample container was calculated by residence time distribution function. The calculated tritium concentrations were in good agreement with the experimental observations. The primary removal tritium behavior was also investigated by another code. Tritium gas concentrations decreased logarithmically to the time by ventilation. These observations were understandable by the reason that the flow in the ventilated Caisson was regarded as the perfectly mixing flow. The concentrations of tritiated water measured, and indications of tritium concentration by tritium monitors became gradually flat. This phenomena called 'tritium soaking effect' was found to be reasonably explained by considering the contribution of the exhaustion velocity by ventilation system, and the adsorption and desorption reaction rate of tritiated water on the wall material which is SUS 304. The calculated tritium concentrations were in good agreement with the experimental observations

  7. In-pile test of tritium recovery from lithium oxide

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Yoshida, Hiroshi; Watanabe, Hitoshi; Takeshita, Hidefumi; Miyauchi, Takejiro; Matsui, Tomoaki

    1984-05-01

    In-situ tritium recovery experiment with sintered lithium oxide pellets was performed under a high neutron fluence in the JRR-2. The irradiation hole VT-10 is the vertical one in the fuel rods region of the reactor, and the neutron flux is as follows: the thermal neutron flux with the epithermal neutron; 1.12 x 10 14 n/cm 2 . sec, the fast neutron flux; 1.0 x 10 12 n/cm 2 . sec. Irradiation material is the four pellets of cylindrical Li 2 O with the size of 11mm-OD, 1.8mm-ID, 10mm-H, and their total weight is 6.67g(the apparent bulk density 86%TD). A sweep gas capsule with a inner heater was constructed for the present study. Irradiation temperatures were regulated in the high temperature range, 470 -- 760 0 C. Four cycles of irradiation tests were carried out from May to August in 1983, and the effective thermal neutron fluence and the burnup of 6 Li were 5.9 x 10 19 nvt and 0.24% of total lithium(natural abundance of Li), respectively. The amount of generated tritium was calculated to be 31.2Ci by using a value of the depression factor of the thermal neutron flux(0.148) and the effective neutron cross section(543b) for the 6 Li(n, α) 3 H reaction. Present report describes the tritium release behavior in the in-situ tritium recovery apparatus and discuss the effects of the moisture, the hydrogen spiking, the irradiation temperature, etc.. Problems relative to a real time measurement of a comparatively high tritium concentration(10 -1 -- 10 2 μCi/cm 3 ) in the helium gas stream were also investigated. (author)

  8. Tritium and helium release from beryllium pebbles neutron-irradiated up to 230appm tritium and 3000appm helium

    Directory of Open Access Journals (Sweden)

    Vladimir Chakin

    2016-12-01

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

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

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.

    1991-01-01

    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 6 Li(n,α) 3 He 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

  10. Tritium release and retention properties of highly neutron-irradiated beryllium pebbles from HIDOBE-01 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A.; Klimenkov, M.; Kolb, M.; Vladimirov, P.; Kurinskiy, P.; Schneider, H.-C. [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Til, S. van; Magielsen, A.J. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-11-15

    The current helium cooled pebble bed (HCPB) tritium breeding blanket concept for fusion reactors includes a bed of 1 mm diameter beryllium pebbles to act as a neutron multiplier. Beryllium pebbles, fabricated by the rotating electrode method, were neutron irradiated in the HFR in Petten within the HIDOBE-01 experiment. This study presents tritium release and retention properties and data on microstructure evolution of beryllium pebbles irradiated at 630, 740, 873, 948 K up to a damage dose of 18 dpa, corresponding to a helium accumulation of about 3000 appm. The measured cumulative released activity from the beryllium pebbles irradiated at 948 K was found to be significantly lower than the calculated value. After irradiation at 873 and 948 K scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses revealed large pores or bubbles in the bulk and oxide films with a thickness of up to 8 μm at the surface of the beryllium pebbles. The radiation-enhanced diffusion of tritium and the formation of open porosity networks accelerate the tritium release from the beryllium pebbles during the high-flux neutron irradiation.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  13. Environmental and radiological impact of accidental tritium release

    Energy Technology Data Exchange (ETDEWEB)

    Guetat, Ph. [CEA Valduc, 21 - Is-sur-Tille (France); Patryl, L. [CEA Bruyeres-le-Chatel, 91 (France)

    2008-07-15

    Within the IAEA program EMRAS, an exercise has been performed by 7 countries to evaluate the consequences of an acute atmospheric release of tritium (10 g). This study aimed at giving practical technical information to decision-makers. Three scenarios with different meteorological conditions were modeled. The objective of this paper is to give the main information about transfer and impact, evaluate uncertainties in models/assessments and so help to set countermeasures. From the results of this exercise, reference activity values for countermeasures can be discussed. All pathways and chemical forms (HTO-HT-OBT) are considered. (authors)

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

    International Nuclear Information System (INIS)

    Furukubo, Y.; Nishikawa, M.; Nishida, Y.; Kinjyo, T.; Tanifuji, Takaaki; Kawamura, Yoshinori; Enoeda, Mikio

    2004-01-01

    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)

  15. Effect of surface water on tritium release behavior from Li4SiO4

    International Nuclear Information System (INIS)

    Hanada, T.; Fukada, S.; Nishikawa, M.; Suematsu, K.; Yamashita, N.; Kanazawa, T.

    2010-01-01

    The tritium release model to represent the release behavior of bred tritium from solid breeder materials has been developed by the blanket group of Kyushu University. It has been found that water is released to the purge gas from solid breeder materials and that this water affects the tritium release behavior. In this study, the amount of surface water released from Li 4 SiO 4 is quantified by the experiment. In addition, the tritium release behavior from Li 4 SiO 4 are estimated based on the tritium release model using parameters obtained in our studies under conditions of commercial reactor operation and ITER test blanket module operation. The effect of the surface water on tritium release behavior is discussed from the obtained results. Moreover, the tritium inventory of Li 4 SiO 4 is discussed based on calculation under the unsteady state condition. Further, the effects of grain size and temperature on distribution of tritium inventory under the steady state condition are evaluated, and the optimal grain size is discussed from the view point of tritium release from Li 4 SiO 4 .

  16. Influence of radiation defects on tritium release parameters from Li2O

    International Nuclear Information System (INIS)

    Grishmanov, V.; Tanaka, S.; Yoneoka, T.

    1998-01-01

    The study of the influence of radiation defects on tritium release behavior from polycrystalline Li 2 O was performed by simultaneous measurements of the luminescence emission and tritium release. It was found that the radiation defects in Li 2 O introduced by electron irradiation cause the retention of tritium. It is thought that the tritium recovery is affected by the formation of a Li-T bond, which is tolerant of high temperatures. The retardation of tritium decreases with increasing absorbed dose in the dose range from 50 to 140 MGy. The aggregation of radiation defects at high irradiation doses is considered to be responsible for the decrease of the interaction of tritium with radiation defects. The mechanism of the interaction of radiation defects with tritium is discussed. (orig.)

  17. Effluents and releases of tritium from Novo-Voronezh-5 reactor

    International Nuclear Information System (INIS)

    Babenko, A.G.; Mekhedov, B.N.; Podporinova, L.E.; Popov, S.V.; Shalin, A.N.

    1990-01-01

    Results of systematic measurements of tritium concentration within technological systems of reactor of Novo-Voronezh NPP conducted to evaluate tritium effluents and releases and radiation doses to population from these effluents and releases are given. It is shown that 68% concerning tritium total amount were disposed into sewerage while 17% - through vent tube and 15% - with water and steam from secondary circuit systems. Standartized tritium effluents from WWER-1000 reactor for 5 year run constitute 15±1.9 GBq/MWxyear and it corresponds to mean value of effluents for foreign NPPs. Tritium concentration in the atmosphere constituted according to calculations (4.1-20)x10 -5 Bq/l. Conclusion is made about insignificant dose to population from tritium gaseous effluents. Detail study is necessary for dose connected with tritium contained in water effluents

  18. Factors affecting calculations of dose resulting from a tritium release into the atmosphere

    International Nuclear Information System (INIS)

    Otaduy, P.; Easterly, C.E.; Booth, R.S.; Jacobs, D.G.

    1976-01-01

    Tritium releases in the form of HT represent a lower hazard to man than releases as HTO. However, during movement in the environment, HT is converted into HTO. The effects of the conversion rate on calcultions of dose are described, and a general method is presented for determining the dose from tritium for various conversion rates and relative HTO/HT risk factors

  19. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    International Nuclear Information System (INIS)

    Xiao, Chengjian; Gao, Xiaoling; Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke; Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming; Wang, Xiaolin; Oya, Yasuhisa; Okuno, Kenji

    2013-01-01

    Tritium release kinetics in lithium orthosilicate (Li 4 SiO 4 ) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li 4 SiO 4 pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10 −7.0 exp (−40.3 × 10 3 /RT) cm 2 s −1

  20. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Chengjian; Gao, Xiaoling [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Wang, Xiaolin, E-mail: xlwang@caep.ac.cn [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Oya, Yasuhisa; Okuno, Kenji [Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Shizuoka 422-8529 (Japan)

    2013-07-15

    Tritium release kinetics in lithium orthosilicate (Li{sub 4}SiO{sub 4}) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li{sub 4}SiO{sub 4} pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10{sup −7.0} exp (−40.3 × 10{sup 3}/RT) cm{sup 2} s{sup −1}.

  1. Tritium release from fast neutron irradiated boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1977-01-01

    A high-energy neutron reaction with boron produces tritium. In the LMFBR control material, B 4 C, most of the tritium that is generated remains in the pellets. Potential retention mechanisms are discussed. 5 figures

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Tritium

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    The role played the large amount supply of tritium and its effects are broadly reviewed. This report is divided into four parts. The introductory part includes the history of tritium research. The second part deals with the physicochemical properties of tritium and the compounds containing tritium such as tritium water and labeled compounds, and with the isotope effects and self radiation effects of tritium. The third part deals with the tritium production by artificial reaction. Attention is directed to the future productivity of tritium from B, Be, N, C, O, etc. by using the beams of high energy protons or neutrons. The problems of the accepting market and the accuracy of estimating manufacturing cost are discussed. The expansion of production may bring upon the reduction of cost but also a large possibility of social impact. The irradiation problem and handling problem in view of environmental preservation are discussed. The fourth part deals with the use of tritium as a target, as a source of radiation or light, and its utilization for geochemistry. The future development of the solid tritium target capable of elongating the life of neutron sources is expected. The rust thickness of the surface of iron can be measured with the X-ray of Ti-T or Zr-T. The tritium can substitute self-light emission paint or lamp. The tritium is suitable for tracing the movement of sea water and land surface water because of its long half life. (Iwakiri, K.)

  4. The effect of helium generation and irradiation temperature on tritium release from neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Gorokhov, V.A.; Vlasov, V.V.; Kovalev, A.M.; Chakin, V.P.

    2004-01-01

    The effect of neutron irradiation condition on tritium release from beryllium is described in this paper. Beryllium samples were irradiated in the SM reactor with neutron fluence (E > 0.1 MeV) of (0.37-2.0) x 10 22 cm -2 at 70-100degC and 650-700degC. Mass-spectrometer technique was used in out of tritium release experiments during stepped-temperature anneal within a temperature range from 250 to 1300degC. The total amount of helium accumulated in irradiated beryllium samples varied from 521 appm to 3061 appm. The first signs of tritium release were detected at temperature of 406-553degC. It was shown that irradiation temperature and helium generation level significantly affect the tritium release. A fraction of 44 - 74 % of tritium content in samples irradiated at low temperature (70 - 100degC) is release from beryllium at an annealing temperature below 800degC, whereas for samples after high temperature irradiation (650 - 700 degC) tritium release did not exceed 14 %. Majority of tritium (∼68%) is released within a temperature range from 800 to 920 degC. The increase of helium generation from 521 appm to 3061 appm results in lowering the temperature of maximal tritium release rate and the upper temperature of tritium release from beryllium by 100-130degC and 200-240degC, correspondingly. On the basis of data obtained, the diffusion coefficients of tritium in beryllium were calculated. (author)

  5. Environmental effects of normal and off-normal releases of tritium from CTR systems

    International Nuclear Information System (INIS)

    McKone, T.E.

    1978-08-01

    Near term fusion technology will utilize the deuterium-tritium reaction. To quantify the magnitude of the hazard presented by major tritium release mechanisms, a method is presented for determining doses to the public from releases of tritium as tritiated water vapor or tritiated lithium compounds. Inclusion of this method in a computer model is described. This model uses the Gaussian dispersion method to predict distribution of tritium species in the downwind environment. Movement of tritium into biological systems is determined by treating these systems as a series of interacting water compartments. Dispersion and uptake calculations are applied to two sample sites in order to predict health effects. These effects are compared to the long range effect of introducing tritium into the world water system

  6. Reconstruction of tritium release history from contaminated groundwater using tree ring analysis

    International Nuclear Information System (INIS)

    Kalin, R.M.; Murphy, C.E. Jr.; Hall, G.

    1995-01-01

    The history of tritium releases to the groundwater from buried waste was reconstructed through dendrochronology. Wood from dated tree rings was sectioned from a cross-section of a tree that was thought to tap the groundwater. Cellulose was chemically separated from the wood. The cellulose was combusted and the water of combustion collected for liquid scintillation counting. The tritium concentration in the rings rose rapidly after 1972 which was prior to the first measurements made in this area. Trends in the tritium concentration of water outcropping to the surface are similar to the trends in tritium concentration in tree rings. 14 refs., 3 figs

  7. Environmental transport and long-term exposure for tritium released in the biosphere

    International Nuclear Information System (INIS)

    Bergman, R.; Bergstroem, U.; Evans, S.

    1979-01-01

    Global cycling of tritium is studied with regard to long-term exposure and dose. Dose and dose commitment are calculated for releases at different latitudes to the troposphere, land and upper ocean layer, with particular regard to effects from release into recipients of intermediate size as, for example, the Baltic Sea. The global transport of tritium appears to be governed by first order kinetics. Compartment models based on linear differential equation systems, as used in this study, should therefore be adequate. The realism and applicability of ecological compartment models are analysed with respect to completeness of the systems of reservoirs and pathways as well as accuracy in assumed reservoir sizes and exchange rates. By introducing different biospheric reservoirs and transfer mechanisms, important carriers and recipients are identified for the analysis of tritium released to air, land and water. Terrestrial biota and groundwater are shown to be significant both with regard to reservoir sizes and influence on the land-troposphere and land-sea exchange of tritium. Model studies regarding the conversion of HT to HTO in different biospheric reservoirs indicate that an atmospheric release of HT may yield up to 1.7 times the dose commitment obtained after release of the same amount of tritium as HTO. The global collective dose commitment from a tropospheric release of tritium is 0.002-0.004 man.rem per Ci depending on the latitude at the release point. Release to the surface ocean layers gives a ten times lower collective dose. (author)

  8. Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.

    1996-01-01

    Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig

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

    International Nuclear Information System (INIS)

    Le Dizes, S.; Tamponnet, C.

    2004-01-01

    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)

  10. The releases of krypton-85 and tritium to the environment and tritium to krypton-85 ratios as source indicators

    International Nuclear Information System (INIS)

    Schroeder, K.J.P.; Roether, W.

    1975-01-01

    More than 95% of the krypton-85 that is at present observed in the environment on a global scale originates from the combined releases following nuclear-power generation and plutonium production, while atmospheric fusion-bomb testing accounts for the same percentage of the global tritium. The global inventories at the end of 1973 were 55 MCi (estimated uncertainty +- 10%) of krypton-85, and 2900 MCi (+- 25%) of tritium. From the excess of the global krypton-85 inventory over the amounts accounted for by the sources other than plutonium production, it is concluded that 130 tonnes of plutonium-239 have been produced up to 1970. Pronounced fractionation occurs between atmosphere and hydrosphere, in the sense that the concentration of krypton-85 is lowered, and that of tritium (being overwhelmingly in the form of a water molecule, HTO) is greatly increased, in the hydrosphere. Largely as a consequence of this fractionation, the environmental tritium to krypton-85 ratios cover a wide range, with values lowest in the troposphere (typical order of magnitude 10 -1 Ci/Ci) and highest in continental surface water (of the order of 10 5 ). Preliminary oceanic krypton-85 data are reported, which indicate that the concentration decrease with depth in the ocean is very similar for both nuclides. Effluents from nuclear power reactors and reprocessing plants have tritium to krypton-85 ratios that do not differ markedly from those found at present in the troposphere; on the other hand, the very high ratios that are typical of the hydrosphere may offer opportunities of detecting such effluents in the hydrosphere on the basis of tritium to krypton-85 ratios. Considerable uncertainties in the ratios of the effluents have, however, to be anticipated, arising from varying, and fluctuating, fractionation between the two nuclides in the release processes. (author)

  11. Measuring nanocurie quantities of tritium bred in metallic lithium and lithium oxide samples

    International Nuclear Information System (INIS)

    Bertone, P.C.

    1985-01-01

    The LBM program requires that nanocurie quantities of tritium, bred in both lithium oxide pellets and lithium samples, be measured with an uncertainty not exceeding + or - 6%. Two methods of accurately measuring nanocurie quantities of tritium bred in LBM lithium oxide pellets and one method of accurately measuring nanocurie quantities of tritium bred in lithium samples are described. Potential errors associated with these tritium measurement techniques are also discussed

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  13. Estimation of release of tritium from measurements of air concentrations in reactor building of PHWR

    International Nuclear Information System (INIS)

    Purohit, R.G.; Sarkar, P.K.

    2010-01-01

    In this paper an attempt has been made to estimate the releases from measured air concentrations of tritium at various locations in Reactor Building (RB). Design data of Kaiga Generating Station and sample measurements of tritium concentrations at various locations in RB and discharges for a period of fortnight were used. A comparison has also been made with actual measurements. It has been observed that there is good matching in estimated and actual measurements of tritium release on some days while on some days there is high difference

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

    International Nuclear Information System (INIS)

    Sharma, L.N.; Dube, B.; Varakhedkar, V.K.

    2001-01-01

    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 bar w for NAPS site is estimated as 2.92E-03 m.s - 1. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    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. - 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

  16. Results of dose calculations for NET accidental and normal operation releases of tritium and activation products

    International Nuclear Information System (INIS)

    Raskob, W.; Hasemann, I.

    1992-08-01

    This report documents conditions, data and results of dose calculations for accidental and normal operation releases of tritium and activation products, performed within the NET subtask SEP2.2 ('NET-Benchmark') of the European Fusion Technology Programme. For accidental releases, the computer codes UFOTRI and COSYMA for assessing the radiological consequences, have been applied for both deterministic and probabilistic calculations. The influence on dose estimates of different release times (2 minutes / 1 hour), two release heights (10 m / 150 m), two chemical forms of tritium (HT/HTO), and two different model approaches for the deposition velocity of HTO on soil was investigated. The dose calculations for normal operation effluents were performed using the tritium model of the German regulatory guidelines, parts of the advanced dose assessment model NORMTRI still under development, and the statistical atmospheric dispersion model ISOLA. Accidental and normal operation source terms were defined as follows: 10g (3.7 10 15 Bq) for accidental tritium releases, 10 Ci/day (3.7 10 11 Bq/day) for tritium releases during normal operation and unit releases of 10 9 Bq for accidental releases of activation products and fission products. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Billone, M.C.; Tanaka, S.

    1994-01-01

    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

  18. Kinetics of the gas-phase tritium oxidation reaction

    International Nuclear Information System (INIS)

    Failor, R.A.

    1989-01-01

    Homogeneous gas-phase kinetics of tritium oxidation (2T 2 + O 2 →2T 2 O) have been studied with a model that accounts explicitly for radiolysis of the major species and the kinetics of the subsequent reactions of ionic, excited-state, and neutral species. Results from model calculations are given for 10 -4 -1.0 mol% T 2 in O 2 (298 K, 1 atm). As the reaction evolves three different mechanisms control T 2 O production, each with a different overall rate expression and a different order with respect to the T 2 concentration. The effects of self-radiolysis of pure T 2 on the tritium oxidation reaction were calculated. Tritium atoms, the primary product of T 2 self-radiolysis, altered the oxidation mechanism only during the first few seconds following the initiation of the T 2 -O 2 reaction. Ozone, an important intermediate in T 2 oxidation, was monitored in-situ by U.V. absorption spectroscopy for 0.01-1.0 mol% T 2 an 1 atm O 2 . The shape of the experimental ozone time profile agreed with the model predictions. As predicted, the measured initial rate of ozone production varied linearly with initial T 2 concentration ([T 2 ] 0.6 o ), but at an initial rate one-third the predicted value. The steady-state ozone concentration ([O 3 ]ss) was predicted to be dependent on [T 2 ] 0.3 o , but the measured value was [T 2 ] 0.6 o , resulting in four times higher [O 3 ]ss than predicted for a 1.0% T 2 -O 2 mixture. Adding H 2 to the T 2 -O 2 mixture, to provide insight into the differences between the radiolytic and chemical behavior of the tritium, produced a greater decrease in [O 3 ]ss than predicted. Adjusting the reaction cell surface-to-volume ratio showed implications of minor surface removal of ozone

  19. Tritium recapture behavior at a nuclear power reactor due to airborne releases.

    Science.gov (United States)

    Harris, Jason T; Miller, David W; Foster, Doug W

    2008-08-01

    This paper describes the initiatives taken by Cook Nuclear Plant to study the on-site behavior of recaptured tritium released in its airborne effluents. Recapture is the process where a released radioactive effluent, in this case tritium, is brought back on-site through some mechanism. Precipitation, shifts in wind direction, or anthropogenic structures that restrict or alter effluent movement can all lead to recapture. The investigation was started after tritium was detected in the north storm drain outfall. Recent inadvertent tritium releases by several other nuclear power plants, many of which entered the groundwater, have led to increased surveillance and scrutiny by regulatory authorities and the general public. To determine the source of tritium in the outfall, an on-site surface water, well water, rainwater and air-conditioning condensate monitoring program was begun. Washout coefficients were also determined to compare with results reported by other nuclear power plants. Program monitoring revealed detectable tritium concentrations in several precipitation sample locations downwind of the two monitored containment building release vents. Tritium was found in higher concentrations in air-conditioning condensate, with a mean value of 528 Bq L(-1) (14,300 pCi L(-1)). The condensate, and to a lesser extent rainwater, were contributing to the tritium found in the north storm drain outfall. Maximum concentration values for each sample type were used to estimate the most conservative dose. A maximum dose of 1.1 x 10(-10) mSv (1.1 x 10(-8) mrem) total body was calculated to determine the health impact of the tritium detected.

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

    CERN Document Server

    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...

  1. Kinetics of catalyzed tritium oxidation in air at ambient temperature

    International Nuclear Information System (INIS)

    Sherwood, A.E.

    1980-01-01

    Tritium/air oxidation kinetic data are derived from measurements carried out with three catalysts. All experiments were carried out at room temperature - a regime that provides a severe test for catalyst effectiveness. Each catalyst consists of a high-surface-area substrate in pelletized form, onto which precious metal has been dispersed. The metal/substrate combinations investigated are: platinum/alumina, palladium/kaolin, and paladium/zeolite. Each of the dispersed-metal catalysts is extremely effective in promoting tritium oxidation in comparison with self-catalyzed atmospheric conversion; equivalent first-order rate constants are higher by roughly nine orders of magnitude. Electron-microprobe scans reveal that the dispersed metal is deposited near the outer surface of the catalyst, with metal concentration decreasing exponentially from the pellet surface. The platinum-based catalyst is more effective than the palladium catalysts on a surface-area basis by about a factor of three. Rate coefficients are determined from concentration decay following a spike injection of tritium into an air-filled enclosure processed by recirculation through an oxidation/adsorption system. The catalytic reaction is first-order in tritium concentration in the range 10 to 10 5 μCi/m 3 (4 ppt-40 ppB). Addition of hydrogen carrier gas is unnecessary. Catalytic activity for all three catalysts declines with time of exposure to air after activation, following a power-law decay with an exponent of -1/2. Reactivation with hot hydrogen gas effectively restores initial catalytic activity

  2. Effect of excess ozone on UV-stimulated tritium oxidation

    International Nuclear Information System (INIS)

    Hasegawa, Kiyoshi; Horii, Kazuhiro; Matsuyama, Masao; Watanabe, Kuniaki.

    1995-01-01

    The authors have reported that the oxidation of tritium is considerably accelerated by irradiating a mixture gas of HT(H 2 )-O 2 with UV-photons, and this UV-stimulated HT oxidation is mainly due to the formation of intermediates such as ozone and activated oxygen species. This suggests that the oxidation will be much more enhanced in the presence of excess ozone in the reaction system. To examine this possibility, effects of the excess ozone on the UV-stimulated HT oxidation was experimentally studied on the one hand, and reaction mechanisms were investigated by developing a computer simulation program applicable to the three-component system of HT(H 2 )-O 2 -O 3 . The formation rate of HTO was measured for gas mixtures consisting of O 2 (75.5 Torr), O 3 (0.5-2% of O 2 ), H 2 (0.1-3% of O 2 ) and HT(H 2 /HT=12000). The experiments showed considerable enhancement of the HTO production rate in the presence of excess ozone by UV-photons from a low pressure mercury lamp(5W). The time course of the reaction was reproduced quite well by computer simulation, indicating that the assumed reaction mechanism is valid. This is also supported by observations that computer simulation reproduced the experimentally observed dependence of ozone decomposition rate on ozone and hydrogen pressures under the UV-irradiation. Those results showed that UV-stimulated HT oxidation was accelerated by about 14000 times in the presence of excess ozone. It strongly suggests that the UV-stimulated oxidation in the presence of excess ozone will be applicable to tritium handling systems as a non-catalytic tritium removal method. (author)

  3. Determination of tritium generation and release parameters at lithium CPS under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ponkratov, Yuriy, E-mail: ponkratov@nnc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Baklanov, Viktor; Skakov, Mazhyn; Kulsartov, Timur; Tazhibayeva, Irina; Gordienko, Yuriy; Zaurbekova, Zhanna; Tulubayev, Yevgeniy [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Chikhray, Yevgeniy [Institute of Experimental and Theoretical Physics of Kazakh National University, Almaty (Kazakhstan); Lyublinski, Igor [JSC “Star”, Moscow (Russian Federation); NRNU “MEPhI”, Moscow (Russian Federation); Vertkov, Alexey [JSC “Star”, Moscow (Russian Federation)

    2016-11-01

    Highlights: • The main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1 M research reactor is described in paper. • In the experiments a very small tritium release was fixed likely due to its high solubility in liquid lithium. • If the lithium CPS will be used as a plasma facing material in temperature range up to 773 K under neutron irradiation only helium will release from lithium CPS into a vacuum chamber. - Abstract: This paper describes the main parameters of tritium generation and release from lithium capillary-porous system (CPS) under neutron irradiation at the IVG.1 M research reactor. The experiments were carried out using the method of mass-spectrometric registration of released gases and using a specially constructed ampoule device. Irradiation was carried out at different reactor thermal powers (1, 2 and 6 MW) and sample temperatures from 473 to 773 K. In the experiments a very small tritium release was detected likely due to its high solubility in liquid lithium. It can be caused by formation of lithium tritide during tritium diffusion to the lithium surface.

  4. Measured tritium in groundwater related to atmospheric releases from the Marcoule nuclear site

    International Nuclear Information System (INIS)

    Levy, F.; Clech, A.; Crochet, P.

    1996-01-01

    Tritium is released into the atmosphere during normal operation from the industrial facilities operated by COGEMA at Marcoule; over a 1 5-year period covered by this study (1979-1994) the quantities ranged from 4940 to 520 TBq·yr -1 . Atmospheric release in rainy weather results in tritium migration into the ground water by a series of mechanisms associated with the water cycle. COGEMA monitors the ground water by means of bore holes. Atmospheric monitoring is also routinely performed; data on the tritium activity concentration in the air and rainwater are available for the same time period. A simplified observation suggests a relation between the atmospheric tritium release and the ground water radioactivity. In 1994, the activity ranged from 100 to 200 Bq·l -1 in the boreholes located 1 km and 2 km downwind from the point of release, diminishing with the distance to less than 20 Bq·l -1 at about 3 km. The authors attempted to model two types of transfers: atmospheric transfer from the release chimney to the borehole, and transfer in the alluvial ground water. The aquifer comprises the alluvial deposits forming the Codolet plain extending to the south of Marcoule, downwind from the point of atmospheric tritium release. The hydrogeology of the entire Marcoule site has been described in previous studies by the French bureau of geological and mineralogical research (BRGM) and ANTEA. (author)

  5. Effect of electromagnetic radiation on the release of tritium from a fusion reactor

    International Nuclear Information System (INIS)

    Causey, R.A.

    1982-01-01

    Electromagnetic radiation may play an important role in determining the amount of tritium that passes through the first wall of a fusion reactor. Photons enhance the outgassing of hydrogen from metal surfaces and therefore lower the surface concentration. The diffusion coefficient for hydrogen in metals is also increased by photon interactions. The importance of these processes on the tritium release depends on the actual conditions that will exist in the fusion reactor. Because electromagnetic radiation stimulates the release of tritium from traps, it could also affect the tritium inventory in the first wall. The effects of electromagnetic radiation on hydrogen in metals have been reviewed. Because electrons are produced by high energy photons, the effect of electrons on hydrogen has also been included

  6. Individual and population dose to users of the Savannah River following K-Reactor tritium release

    International Nuclear Information System (INIS)

    Carlton, W.H.; Hamby, D.M.

    1992-01-01

    Approximately 5700 curies of tritium were released to Pen Branch between December 22, 1991 and December 25, 1991. As expected, the tritiated water traveled through the Savannah River swamp to Steel Creek and the Savannah River. Elevated tritium concentrations in the river at Becks Ferry (Beaufort-Jasper) and Abercorn Creek (Port Wentworth) has caused some concern among downstream water users as to the amount of tritium available for uptake through the domestic drinking water supplies at the Beaufort-Jasper and Port Wentworth water treatment facilities. Radiation dose to the downstream drinking water population is estimated in this report

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

    Science.gov (United States)

    Jeffers, Rebecca S; Parker, Geoffrey T

    2014-01-01

    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. Copyright © 2013 Elsevier Ltd. All rights reserved.

  8. Tritium oxide uptake and desorption kinetics in a primary producer: chlorella pyrenoidosa

    International Nuclear Information System (INIS)

    Dunstall, T.G.

    1983-01-01

    The alga Chlorella pyrenoidosa grown in batch culture under chronic tritium oxide exposure was used to model behavior of tritium at the primary producer level of an aquatic food chain. The specific activity ratio of organically bound tritium to medium tritium increased during initial growth stages, then reached an asymptotic steady state value of 0.59 after approximately seven cell doublings. The intracellular to extracellular concentrations of tritium oxide appeared to be in equilibrium. Loss of previously formed organically bound tritium in cells transferred to tritium-free media averaged less than 5 % for exponential growth phase cultures which had undergone more than three cell doublings. Over a comparable time period, a greater loss of organically bound tritium from stationary cells (average 13.4 %) was attributed to increased degradative metabolism in senescent cultures. Concentration of tritium in organically bound form may exceed environmental tritium oxide levels under dynamic conditions in which a pulse of tritium oxide to the environment is dissipated over time

  9. Experimental investigation on tritium release from lithium titanate pebble under high temperature of 1073 K

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Kentaro, E-mail: howartre@onid.oregonstate.edu [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Edao, Yuki; Kawamura, Yoshinori [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Ohta, Masayuki; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan)

    2015-10-15

    Highlights: • We have performed the tritium recovery experiment with the DT neutron source at 1073 K. • The tritium recovery corresponded with the calculated tritium production. • The chemical form of recovered tritium is affected by the temperature and kind of sweep gas. • The recovered HT increases at higher temperature and dry hydrogen circumstance. - Abstract: The temperature of Li{sub 2}TiO{sub 3} pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li{sub 2}TiO{sub 3} pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li{sub 2}TiO{sub 3} pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li{sub 2}TiO{sub 3} layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H{sub 2}/He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H{sub 2}O/He gas (H{sub 2}O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H{sub 2}, it is indicated that the increasing tendency of HT release is similar to that of the dry H{sub 2}/He.

  10. Tritium release from Li{sub 4}SiO{sub 4} ceramic pebbles in high magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Ran, Guangming [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Xiao, Chengjian; Chen, Xiaojun; Gong, Yu; Zhao, Linjie [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Wang, Xiaolin [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); China Academy of Engineering Physics, Mianyang 621900 (China)

    2015-12-15

    Highlights: • Behavior of tritium release from Li{sub 4}SiO{sub 4} pebbles in MF up to 7 T was investigated. • The MF effect on tritium release is not significant according to the TPD results. • Tritium release from the BIG sample is much easier than that from the SMALL sample. • The MF effect on tritium diffusion was probably weakened by surface desorption. - Abstract: The behavior of tritium release from Li{sub 4}SiO{sub 4} ceramic pebbles in high magnetic field (MF) was investigated by temperature programmed desorption (TPD). Two batches of Li{sub 4}SiO{sub 4} pebbles produced by wet method were used as the experimental samples, one batch with an average pebble diameter of 0.8 mm (the SMALL samples), and the other 1.2 mm (the BIG samples). A superconducting magnet was applied to generate MF up to 7 T in the sample area during annealing. For both batches of samples, the tritium release curves within and without MF showed very similar characteristics, indicating that the effect of high MF on tritium release behavior is not significant. The tritium release peaks for the BIG samples were observed at much lower temperatures than that for the SMALL samples, even though the grain sizes of the BIG samples are much bigger than that of the SMALL samples. It is considered that surface desorption process dominates the overall tritium release behavior in this work, which probably weakened the MF effect.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    Saito, Masahiro; Raskob, Wolfgang

    2003-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Le Dizes, S.; Maro, D.; Rozet, M.; Hebert, D.; Solier, L.; Nicoulaud, V. [Institut de radioportection et de surete nucleaire - IRSN (France); Vermorel, F.; Aulagnier, C. [Electricite de France - EDF (France)

    2014-07-01

    Tritium ({sup 3}H) 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 CO{sub 2}, H{sub 2}O fluxes between soil and air compartments will be carried out. Then, the TOCATTA-c model previously developed to simulate {sup 14}C 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

  14. Tritium labeling of gonadotropin releasing hormone in its proline and histidine residues

    International Nuclear Information System (INIS)

    Klauschenz, E.; Bienert, M.; Egler, H.; Pleiss, U.; Niedrich, H.; Nikolics, K.

    1981-01-01

    3,4-dehydroproline9-GnRH prepared by solid phase peptide synthesis was tritiated catalytically under various conditions yielding 3H-GnRH with specific radioactivities in the range from 35-60 Ci/mmol and full LH releasing activity in vitro. Using palladium/alumina catalyst, the tritiation of the double bond occurs within ten minutes. Investigation of the tritium distribution between the amino acid residues showed a remarkably high incorporation of tritium into the histidine residue (11 to 37%). On the basis of this observation, the tritium labeling of GnRH and angiotensin I by direct catalytic hydrogen-tritium exchange was found to be useful for the labeling of these peptides at remarkably high specific radioactivity

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-01-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 x 10 -3 person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 x 10 -2 person-Sv/Pbq (2.4 x 10 -4 person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river

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

    Science.gov (United States)

    Korolevych, V Y; Kim, S B

    2013-04-01

    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. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

  18. Tritium release from neutron-irradiated Li2O: Transport in porous sintered pellets

    International Nuclear Information System (INIS)

    Tanifuji, Takaaki; Yamaki, Daiju; Jitsukawa, Shiro

    2006-01-01

    The tritium release behavior from Li 2 O sintered pellets (81-88% T.D.) is examined by isothermal heating tests. (1) For the 88% T.D. specimens, the fraction of residual tritium is found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (2) For the 81% T.D. specimens, which are annealed after irradiation at 630 K for 4 h, the fraction of residual tritium is also found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (3) For the 81% T.D. specimens as irradiated, the tritium release rate is found to follow the square-root law of the annealing time. The rate-determining process is controlled by Kohlrauch stretched exponential form. Tritium trapped in irradiation defects released with recovering the defects by isothermal heating

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

    International Nuclear Information System (INIS)

    Jensen, G.A.; Martin, J.B.

    1988-04-01

    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

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

    Science.gov (United States)

    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. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Gautam, Y.P.; Sharma, Saivajay; Rao, K.S.; Singh, Bhikam; Kumar, Avinash; Ravi, P.M.

    2012-01-01

    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, c 0 (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 V w for NAPS site is estimated as 3.17E-03 m.s -1 . (author)

  2. Projected tritium releases from F ampersand H Area Seepage Basins and the Solid Waste Disposal Facilities to Fourmile Branch

    International Nuclear Information System (INIS)

    Looney, B.B.; Haselow, J.S.; Lewis, C.M.; Harris, M.K.; Wyatt, D.E.; Hetrick, C.S.

    1993-01-01

    A large percentage of the radioactivity released to the environment by operations at the Savannah River Site (SRS) is due to tritium. Because of the relative importance of the releases of tritium from SRS facilities through the groundwater to the environment, periodic evaluation and documentation of the facility operational status, proposed corrective actions, and projected changes/reductions in tritium releases are justified. Past, current, and projected tritium releases from the F and H Area Seepage Basins and the Solid Waste Disposal Facilities (SWDF) to Fourmile Branch are described. Each section provides a brief operational history along with the current status and proposed corrective actions. A conceptual model and quantitative estimates of tritium release from the facilities into the groundwater and the environment are developed. Tritium releases from the F and H Area Seepage Basins are declining and will be further reduced by the implementation of a groundwater corrective action required by the Resource Conservation and Recovery Act (RCRA). Tritium releases from the SWDF have been relatively stable over the past 10 years. It is anticipated that SWDF tritium releases to Fourmile Branch will remain approximately at current levels for at least 10--20 years. Specific characterization activities are recommended to allow an improved projection of tritium flux and to assist in developing plans for plume mitigation. SRS and the South Carolina Department of Health and Environmental Control are developing groundwater corrective action plans for the SWDF. Portions of the SWDF are also regulated under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). Reduction of tritium flux is one of the factors considered in the development of the RCRA/CERCLA groundwater corrective action. The final section of the document presents the sum of the projected tritium fluxes from these facilities to Fourmile Branch

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

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, Sumi; Noguchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ryufuku, Susumu; Sasaki, Toshihisa; Kurosawa, Naohiro [Visible Information Center, Inc., Tokai, Ibaraki (Japan)

    2002-11-01

    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: inhalation from a primary plume (HT and/or HTO) released from the facilities and inhalation from a secondary plume (HTO) reemitted from the ground following deposition of HT and HTO. This report describes an outline of the ACUTRI code, a user guide and the results of test calculation. (author)

  4. Oxidation of tritium in packed bed of noble metal catalyst for detritiation from system gases

    International Nuclear Information System (INIS)

    Nishikawa, Masabumi; Takeishi, Toshiharu; Munakata, Kenzo; Kotoh, Kenji; Enoeda, Mikio

    1985-01-01

    Catalytic oxidation rates of tritium in the bed of the noble metal catalysts are obtained and compared with the oxidation rates observed for the packed bed of spongy copper oxide or hopcalites. Use of Pt- or Pd-aluminia catalysts is recommended in this study because they give effective oxidation rates of tritium in the ambient temperature range. The adsorption performance of tritiated water in the catalyst bed is also discussed. (orig.)

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

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Miki, Nobuharu; Akiba, Masato

    2003-06-01

    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 Li 2 TiO 3 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 Li 2 TiO 3 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 328

  6. Reduction of tritium permeation through Inconel 718 and Incoloy 800 HT by means of natural oxides

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, A., E-mail: antonio.aiello@enea.it [ENEA C.R. Brasimone, I-40032 Camugnano (Italy); Utili, M.; Ciampichetti, A. [ENEA C.R. Brasimone, I-40032 Camugnano (Italy)

    2011-10-01

    Chronical releases of tritium from the helium primary coolant into the water secondary coolant is a fundamental safety issue in the design of a fusion reactor steam generator. It is well known that the steam/water circuit of a fusion reactor would be considered not relevant from a radiological point of view, while if a strong permeation of tritium will be present it will be released together with incondensable gases in the condenser. The permeation of hydrogen isotopes through candidate steam generator materials in different conditions was studied in the past. Further experiments demonstrated that nickel alloys of nuclear interest are always covered by a thin and adherent oxide layer able to reduce permeation of orders of magnitude. The major objective of this work is the evaluation of the permeated flux through nickel alloys, when exposed to pure hydrogen and to an oxidant gas stream, to verify the real permeability of these materials in conditions close to those foreseen in the helium side of the steam generator.

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

    International Nuclear Information System (INIS)

    Saito, Masahiro

    2005-01-01

    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)

  8. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

    It is well known that tritium is used as the fuel for nuclear fusion reactors. The neutrons produced by the nuclear fusion reaction of deuterium and tritium react with lithium in blankets, and tritium is produced. The blankets reproduce the tritium consumed in the D-T reaction. Tritium circulates through the main cooling system and the fuel supply and evacuation system, and is accumulated. Tritium is a radioactive substance emitting β-ray with 12.6 year half-life, and harmful to human bodies. It is an isotope of hydrogen, and apt to diffuse and leak. Especially at high temperature, it permeates through materials, therefore it is important to evaluate the release of tritium into environment, to treat leaked tritium to reduce its release, and to select the method of containing tritium. The permeability of tritium and its solubility in structural materials are discussed. The typical blanket-cooling systems of nuclear fusion reactors are shown, and the tungsten coating of steam generator tubes and tritium recovery system are adopted for reducing tritium leak. In case of the Tokamak type reactor of JAERI, the tritium recovery system is installed, in which the tritium gas produced in blankets is converted to tritium steam with a Pd-Pt catalytic oxidation tower, and it is dehydrated and eliminated with a molecular sieve tower, then purified and recovered. (Kako, I.)

  9. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO/sub 2/) and lithium silicate (Li/sub 2/SiO/sub 3/) by the reaction: Li/sup 6/ + n ..-->.. /sup 4/He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100/sup 0/C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T/sub 2/), while in laboratory extractions (300-1300/sup 0/C), the tritium appeared primarily in the condensible form (HTO and T/sub 2/O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H/sub 2/O, CO/sub 2/, CO, O/sub 2/, H/sub 2/, NO, SO/sub 2/, SiF/sub 4/ and traces of hydrocarbons.

  10. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

  11. Tritium safety issues for TFCX

    International Nuclear Information System (INIS)

    Reilly, H.J.; Piet, S.J.; Merrill, B.J.

    1985-01-01

    Estimated tritium releases from the Tokamak Fusion Core Experiment are compared to the expected limits. A reaction kinetics model is described that predicts the conversion of tritium to the oxide form in free space. An analysis of the required capacity of the Emergency Tritium Cleanup System is also presented. The conclusions of this work are expected to be applicable to other experimental fusion devices that are now being considered

  12. Fabrication and tritium release property of Li2TiO3-Li4SiO4 biphasic ceramics

    Science.gov (United States)

    Yang, Mao; Ran, Guangming; Wang, Hailiang; Dang, Chen; Huang, Zhangyi; Chen, Xiaojun; Lu, Tiecheng; Xiao, Chengjian

    2018-05-01

    Li2TiO3-Li4SiO4 biphasic ceramic pebbles have been developed as an advanced tritium breeder due to the potential to combine the advantages of both Li2TiO3 and Li4SiO4. Wet method was developed for the pebble fabrication and Li2TiO3-Li4SiO4 biphasic ceramic pebbles were successfully prepared by wet method using the powders synthesized by hydrothermal method. The tritium release properties of the Li2TiO3-Li4SiO4 biphasic ceramic pebbles were evaluated. The biphasic pebbles exhibited good tritium release property at low temperatures and the tritium release temperature was around 470 °C. Because of the isotope exchange reaction between H2 and tritium, the addition of 0.1%H2 to purge gas He could significantly enhance the tritium gas release and the fraction of molecular form of tritium increased from 28% to 55%. The results indicate that the Li2TiO3-Li4SiO4 biphasic ceramic pebbles fabricated by wet method exhibit good tritium release property and hold promising potential as advanced breeder pebbles.

  13. Quantitative determination of tritium in metals and oxides

    International Nuclear Information System (INIS)

    Vance, D.E.; Smith, M.E.; Waterbury, G.R.

    1979-04-01

    Metallic samples are analyzed for tritium by heating the sample at 1225 K in a moist oxygen stream. The volatile products are trapped and the tritium is quantitatively determined by scintillation spectroscopy. The method is used to determine less than 1 ppb of tritium in 100-mg samples of lithium, iron, nickel, cerium, plutonium, and plutonium dioxide. Analysis of 18 cuts of a tritium-zirconium, copper foil standard over a 3-yr period showed a tritium content of 45 ppM and a standard deviation of 6 ppM

  14. Transfer of tritium released by nuclear facilities to the food supply

    International Nuclear Information System (INIS)

    Bovard, P.; Delmas, J.; Belot, Y.; Camus, H.; Grauby, A.; Hoek, J. van den

    1979-01-01

    The use for agricultural purposes of river waters receiving releases or discharges of tritium results in contamination of irrigated crops and of animals given such water to drink or consuming the contaminated crops. It therefore seemed of importance to assess the part played by tritium in the contamination of the food chain, together with its possible effects on organisms. With this in mind, French, Belgian and Netherlands laboratories have joined forces to study, more especially, the relationship between environmental contamination rates and those of produce harvested in the Mediterranean region and in a humid temperate climate, the transfer process in the chain: water - fodder - bovines - dairy produce, and the role of technology in the contamination of the food chain. The present status of research undertaken jointly by organizations in the three countries is reviewed. In the Atlantic environment the experiments involved four annual crops consumed on a large scale: potatoes, sugar beet, carrots and peas, and in the Mediterranean environment several perennial species such as vine, olive, orange and apple were studied. The results obtained relate to the residence time for tritium in the various organs of each species, the part played by evapotranspiration and the physiological functions of the different parts of the plants, the uptake of tritium by tissue water and organic matter, and the distribution of tritium in the soil profile. (author)

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

    International Nuclear Information System (INIS)

    Iseli, M.; Esser, B.

    1989-01-01

    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

  16. Environmental aspects of a tritium release from the Savannah River Plant on March 23, 1984

    International Nuclear Information System (INIS)

    Evans, A.G.; Hoel, D.D.; Kantelo, M.V.

    1985-01-01

    The environmental impact of a tritium release from the Savannah River Plant (SRP) on March 23, 1984, was assessed by using both predictive and measurement techniques. Prediction of the onsite and offsite consequences by the WIND emergency response system agreed with results determined from environmental samples. The maximum calculated radiation dose to an individual at the SRP boundary was 0.06 mrem. The maximum radiation dose to a hypothetical individual consuming milk collected at the point of highest airborne tritium concentration was calculated to be 0.17 mrem. The maximum measured (by urinalysis) dose to offsite individuals in the release trajectory was 0.02 mrem. The 0.17 mrem dose corresponds to 0.09% of the 189 mrem dose that persons who live in the vicinity of SRP receive annually from natural radiation, prorated medical radiation, and fallout. Thus the public health consequences of the release were insignificant. 11 references, 15 figures, 12 tables

  17. Influence of additives and impurities in sweep gas and solid tritium release behaviour from lithium ceramics (review)

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1991-01-01

    Tritium release from solid breeding material is affected by small amounts of additives or impurities in the sweep gas or solid itself. Addition of hydrogen or water vapor to the sweep gas is reported to enhance the surface reaction of tritium release. Doping to solid breeder with elements of different valence from lithium has a possibility to improve tritium diffusion in the solid. Surface reaction and migration behavior in bulk are believed to be also affected by impurities in the sweep gas and in the solid. In order to model tritium release behavior in the blanket of fusion reactor, the mechanism of interaction with these additives or impurities must be quantitatively formulated. However, the mechanism of these remains to be elucidated. In this paper effects of these additives and impurities on tritium migration are reviewed. The mechanism of surface reaction for He+H 2 sweep gas is also discussed. (orig.)

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

    International Nuclear Information System (INIS)

    Velarde, M.; Perlado, J.M.; Sedano, L.

    2006-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Shu, W.M. E-mail: shu@tpl.tokai.jaeri.go.jp; Gentile, C.A.; Skinner, C.H.; Langish, S.; Nishi, M.F

    2002-11-01

    A series of tests involving 10 h baking under the current ITER design conditions (240 deg. C with 933 Pa O{sub 2}) 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 {mu}m/h near the surface and 0.9 {mu}m/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 O{sub 2} 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%O{sub 2}. 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.

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

    International Nuclear Information System (INIS)

    Shu, W.M.; Gentile, C.A.; Skinner, C.H.; Langish, S.; Nishi, M.F.

    2002-01-01

    A series of tests involving 10 h baking under the current ITER design conditions (240 deg. C with 933 Pa O 2 ) 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 O 2 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%O 2 . 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

  1. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

    2016-08-31

    Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  2. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

    2016-08-31

    This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  3. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

    The beryllium neutron multiplier in the ITER breeding blanket will generate tritium through transmutations. That tritium constitutes a safety hazard. Experiments evaluating tritium storage and release mechanisms have shown that most of the tritium comes out in a burst during thermal ramping. A small fraction of retained tritium is released by thermally activated processes. Analysis of recent experimental data shows that most of the tritium resides in helium bubbles. That tritium is released when the bubbles undergo swelling sufficient to develop porosity that connects with the surface. That appears to occur when swelling reaches about 10--15%. Other tritium appears to be stored chemically at oxide inclusions, probably as Be(OT) 2 . That component is released by thermal activation. There is considerable variation in published values for tritium diffusion through the beryllium and solubility in it. Data from experiments using highly irradiated beryllium from the Idaho National Engineering Laboratory showed diffusivity generally in line with the most commonly accepted values for fully dense material. Lower density material, planned for use in the ITER blanket may have very short diffusion times because of the open structure. The beryllium multiplier of the ITER breeding blanket was analyzed for tritium release characteristics using temperature and helium production figures at the midplane generated in support of the ITER Summer Workshop, 1990 in Garching. Ordinary operation, either in Physics or Technology phases, should not result in the release of tritium trapped in the helium bubbles. Temperature excursions above 600 degree C result in large-scale release of that tritium. 29 refs., 10 figs., 3 tabs

  4. Effluent Treatment Facility tritium emissions monitoring

    International Nuclear Information System (INIS)

    Dunn, D.L.

    1991-01-01

    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

  5. In-Pile Assemblies for Investigation of Tritium Release from Li2TiO3 Lithium Ceramic

    International Nuclear Information System (INIS)

    Shestakov, V.; Tazhibayeva, I.; Kawamura, H.; Kenzhin, Y.; Kulsartov, T.; Chikhray, Y.; Kolbaenkov, A.; Arinkin, F.; Gizatulin, Sh.; Chakrov, P.

    2005-01-01

    The description of algorithm to design in-pipe experimental ampoule devices (IPAD) is presented here, including description of IPAD design for irradiation tests of highly enriched lithium ceramics at WWR-K reactor. The description of the system for registration of tritium release from ceramics during irradiation is presented as well. Typical curve of tritium release from the IPAD during irradiation under various temperatures of the samples is shown here

  6. Two investigations concerning the release of tritium. I. Tritium leakage from 3H(Sc) EC-detectors

    International Nuclear Information System (INIS)

    Bergman, C.; Wesslen, E.

    1977-01-01

    Recently the manufacturers of EC-detectors for gas chromatographs introduced a new type of 3 H EC-detector where the tritium is bound to scandium instead of to titanium and has an activity up to 1 Ci. It is expected that the scandium-based detector will take a great part of the Swedish EC-detector market. The Swedish National Institute of Radiation Protection is anxious to make sure that the introduction of the new detector, which will be used at higher temperature, will not give rise to any increased risk of tritium intake to the personnel handling the chromatographs. The leakage of tritium from commercially available 3 H(Sc) EC-detectors containing 1 Ci of tritium was measured as a function of the detector temperature. Tritium appears both in the form of tritium gas dissolved in the scandium and in the form of tritide. The gas evaporates rather easily with increasing temperature while the dissociation of the tritide is a slower process. The evaporation of tritium due to the dissociation of the tritide was found to be negligible, less than 0.2 μCi/h at temperatures less than 100 0 C, but rises rapidly with temperature. The study also showed that even when the detector is stored at room temperature, a re-distribution of the tritium occures, from the tritide to the dissolved tritium gas, which then easily evaporates even at moderately elevated temperatures

  7. 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)

    International Nuclear Information System (INIS)

    Kalyanam, K.M.

    1996-06-01

    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

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

    International Nuclear Information System (INIS)

    Soares, Abner Duarte; Simoes Filho, Francisco Fernando Lamego; Cunha, Tatiana Santos da; Aguiar, Andre Silva de; Lapa, Celso Marcelo Franklin

    2011-01-01

    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 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 3 ) during the first 14 days after the accident. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  10. An assessment of the tritium inventory in, permeation through and releases from the NET plasma facing materials

    International Nuclear Information System (INIS)

    Wu, C.H.

    1986-01-01

    The tritium retention, permeation and release characteristics of D-T tokamaks are extremely important from both an environmental and a plasma physics point of view. Tokamak measurements have demonstrated that release of retained hydrogen isotopes by plasma-wall interactions play a dominant role in fuel recycling during a discharge. In addition, retained tritium in the plasma facing materials may contribute substantially to the on-site tritium inventory of D-T devices. Austenitic and martensitic steels are being considered as first wall materials. Tungsten and molybdenum will be possibly used as divertor armour materials for NET. By using a computer code, the tritium inventory in, permeation through and release from these materials have been calculated as functions of material thickness, temperature and impinging fluxes. It is shown that the tritium inventory in the first wall will be strongly affected by the temperature gradient in the materials. It is evident, that the tritium permeation as well as the tritium inventory can be reduced appropriately by controlling the temperatures at the plasma and cooling sides of the first wall. The results are discussed and the possible consequences are analysed. (author)

  11. Tritium in the environment. Knowledge synthesis

    International Nuclear Information System (INIS)

    2009-01-01

    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

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

    International Nuclear Information System (INIS)

    Jacobs, D.G.; Easterly, C.E.; Phillips, J.E.

    1979-01-01

    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)

  13. Calculation of tritium inventory and permeation in an INTOR like tokamak device and of its release after shut down

    International Nuclear Information System (INIS)

    Wienhold, P.; Waelbroeck, F.s; Winter, J.; Ali-Khan, I.

    1980-12-01

    The tritium inventory in the wall, its escape via permeation and its release in general are evaluated during the operation and later shut-down phases of a tokamak like INTOR by means of the PERI code. No real tritium problem arises after shut-down: simple outgasing techniques at wall temperatures Tsub(W) of approx. 500 0 C should decrease the tritium release rate down to a sufficient low level. Very serious problems are on the other hand expected to arise in the areas of inventory and permeation of tritium during the operation phase: one has the choice between e.g. operating at Tsub(W) = 350 0 C with a quasi-stationary tritium inventory of some 100 grams and a permeation flux phisub(p) approx. 1 gram/day or increasing Tsub(W) to 500 0 C, reducing the inventory down to some grams, but have then a leakage rate by permeation of 35 grams/day. For Tsub(W) = 100 0 C the tritium release into the torus will be intolerable. These figures are confirmed by preliminary measurements on Inconel 600. A suggestion is made as to how to alleviate the problem. (orig.) [de

  14. Non-labile tritium in Savannah River Plant pine trees

    International Nuclear Information System (INIS)

    Sanders, S.M. Jr.

    1976-06-01

    Non-labile tritium bound in cellulose of pine trees was measured to learn about the effects and fate of tritium contributed to the environment by the Savannah River Plant (SRP). An estimation of the regional inventory and the distance tritium can be observed from SRP was desired because tritium is a major component of the radioactivity released by SRP, and as the oxide, it readily disperses in the environment

  15. Catalytic oxidation efficiencies for tritium and tritiated methane in a mature, industrial-scale decontamination system

    International Nuclear Information System (INIS)

    Mintz, J.M.; Gildea, P.D.

    1981-01-01

    Almost all tritium decontamination systems proposed for fusion facilities employ catalytic oxidation to water, followed by drying, to remove tritium and tritiated hydrocarbons from gas streams. One such large-scale system, the gas purification system (GPS), has been operating in the Tritium Research Laboratory (TRL) at Sandia National Laboratories, Livermore, CA, since October 1977. A series of experiments have recently been conducted there to assesss the current operating characteristics of the GPS catalyst. The experiments used tritium and tritiated methane and covered a range of temperatures, flow rates, and concentration levels. When contrasted with 1977 data, the results indicate that no measurable degradation of catalyst function had occurred. However, some reduction in active metal surface area, as indicated by B.E.T. surface area measurements (approx. 100 → 90m 2 /g) and AES scans (approx. 1.4 → 0.9 at. % Pt), had occurred. Kinetic rate coefficients were also derived and a rough temperature dependence obtained

  16. Catalytic oxidation efficiencies for tritium and tritiated methane in a mature, industrial-scale decontamination system

    International Nuclear Information System (INIS)

    Mintz, J.M.; Gildea, P.D.

    1980-10-01

    Almost all tritium decontamination systems proposed for fusion facilities employ catalytic oxidation to water, followed by drying, to remove tritium and tritiated hydrocarbons from gas streams. One such large-scale system, the gas purification system (GPS), has been operating in the Tritium Research Laboratory (TRL) at Sandia National Laboratories, Livermore, CA, since October 1977. A series of experiments have recently been conducted there to assess the current operating characteristics of the GPS catalyst. The experiments used tritium and tritiated methane and covered a range of temperatures, flow rates, and concentration levels. When contrasted with 1977 data, the results indicate that no measurable degradation of catalyst function had occurred. However, some reduction in active metal surface area, as indicated by B.E.T. surface area measurements (approx. 100 → 90 m 2 /g) and AES scans (approx. 1.4 → 0.9 at% Pt), had occurred. Kinetic rate coefficients were also derived and a rough temperature dependence obtained

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

    International Nuclear Information System (INIS)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Lapa, Celso Marcelo Franklin; Alvim, Antonio Carlos Marques; Soares, Abner Duarte

    2013-01-01

    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. It was postulated, then, the LOCA - Loss of Coolant Accident -, accident in the emergency cooling system of the nucleus ( without fusion), where was lost 66m 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 or operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1MBq/m 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 9 to 5x10 5 Bq/m 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 ( 3 ). (author)

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

    International Nuclear Information System (INIS)

    Soares, Abner Duarte

    2010-01-01

    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 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 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 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 9 to 5x10 5 Bq/m 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 ( 3 ). (author)

  19. 3He release characteristics of metal tritides and scandium--tritium solid solutions

    International Nuclear Information System (INIS)

    Perkins, W.G.; Kass, W.J.; Beavis, L.C.

    1975-01-01

    Tritides of such metals as scandium, titanium, and erbium are useful materials for determining the effects of helium accumulation in metallic solids, for example, CTR first wall materials. Such effects include lattice strain and gross deformation, as reported elsewhere, which are related to 3 He retention and ultimate release. Long term gas release studies have indicated that, during the early life of a metal ditritide, a large fraction of the 3 He is retained in the solid. At more advanced ages (2 to 4 years, depending on the parent metal), the 3 He release rate becomes comparable to the generation rate. Statistical analysis of the data indicates that the acceleration in 3 He release rate depends on accumulated 3 He concentration rather than strictly on age. 3 He outgassing results are presented for thin films of ScT 2 , TiT 2 , and ErT 2 , and the critical 3 He concentrations are discussed in terms of a percolation model. Phase transformations which occur on tritide formation cast some doubt on the validity of extrapolating results obtained for metal tritides to predictions regarding the accumulation of helium in metals. Scandium is unique among the early transition and rare-earth metals in that the metal exhibits a very high room temperature tritium solubility (T/Sc = 0.4) with no phase transformation. Indeed, even the lattice parameters of the hcp scandium lattice are only minimally changed by tritium solution, and we have succeeded in obtaining single crystal ScT 0 . 3 samples in two crystallographic orientations. Using a very sensitive technique, we have measured 3 He emission from both these samples, as well as from fine-grained thin film scandium-tritium solid solution samples (ScT 0 . 3 - 0 . 4 ). The fine-grained film samples release 3 He at 2 to 3 percent of the generation rate, while the emission rate from the single-crystal samples is approximately 0.05 percent of the generation rate, indicating a strong grain size effect

  20. State of dog's metabolism in the remote period after the oxide tritium influence

    International Nuclear Information System (INIS)

    Kalistratova, V.S.; Tishchenko, G.S.; Bortnik, L.A.; Nisimov, P.G.; Romanova, I.B.

    2000-01-01

    Influence of tritium oxide on the metabolism by some indices of lipid metabolism (common lipids, β-lipoproteins, cholesterin), protein metabolism (cholinesterase) and carbohydrate metabolism (blood sugar) was studied. It was established that the introduction into organism of tritium oxide in the quantities, which could form lethal and sublethal doses of internal radiation, provoked the main changes of values of mentioned indices of metabolism. The character of metabolism changes in the remote period allows to judge about the development of sclerosis processes which can be the result of radiation-stipulated acceleration of organism aging [ru

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

    International Nuclear Information System (INIS)

    Fast, J.D.; O'Steen, B.L.; Addis, R.P.

    1992-01-01

    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

  2. Fate of gaseous tritium and carbon-14 released from buried low-level radioactive waste

    International Nuclear Information System (INIS)

    Striegl, R.G.

    1988-01-01

    Microbial decomposition, chemical degradation, and volatilization of buried low-level radioactive waste results in the release of gases containing tritium ( 3 H) and carbon-14 ( 14 C) to the surrounding environment. Water vapor, carbon dioxide, and methane that contain 3 H or 14 C are primary products of microbial decomposition of the waste. Depending on the composition of the waste source, chemical degradation and volatilization of waste also may result in the production of a variety of radioactive gases and organic vapors. Movement of the gases in materials that surround waste trenches is affected by physical, geochemical, and biological mechanisms including sorption, gas-water-mineral reactions, isotopic dilution, microbial consumption, and bioaccumulation. These mechanisms either may transfer 3 H and 14 C to solids and infiltrating water or may result in the accumulation of the radionuclides in plant or animal tissue. Gaseous 3 H or 14 C that is not transferred to other forms is ultimately released to the atmosphere

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  6. Modelization of tritium transfer into the organic compartments of algae

    International Nuclear Information System (INIS)

    Bonotto, S.; Gerber, G.B.; Arapis, G.; Kirchmann, R.

    1982-01-01

    Uptake of tritium oxide and its conversion into organic tritium was studied in four different types of algae with widely varying size and growth characteristics (Acetabularia acetabulum, Boergesenia forbesii, two strains of Chlamydomonas and Dunaliella bioculata). Water in the cell and the vacuales equilibrates rapidly with external tritium water. Tritium is actively incorporated into organically bound form as the organisms grow. During the stationary phase, incorporation of tritium is slow. There exists a discrimination against the incorporation of tritium into organically bound form. A model has been elaborated taking in account these different factors. It appears that transfer of organic tritium by algae growing near the sites of release would be significant only for actively growing algae. Algae growing slowly may, however, be useful as cumulative indicators of discontinuous tritium release. (author)

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

    International Nuclear Information System (INIS)

    Thompson, P.A.; Kwamena, N.-O.A.; Ilin, M.; Wilk, M.; Clark, I.D.

    2015-01-01

    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

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

    International Nuclear Information System (INIS)

    Snyder, S.F.; Hwang, S.T.

    1994-03-01

    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

  9. Tritium removal from air streams by catalytic oxidation and water adsorption

    International Nuclear Information System (INIS)

    Sherwood, A.E.

    1976-06-01

    An effective method of capturing tritium from air streams is by catalytic oxidation followed by water adsorption on a microporous solid adsorbent. Performance of a burner/dryer combination is illustrated by overall mass balance equations. Engineering design methods for packed bed reactors and adsorbers are reviewed, emphasizing the experimental data needed for design and the effect of operating conditions on system performance

  10. Isothermal release of tritium from neutron-irradiated Li/sub 2/O pellets

    Energy Technology Data Exchange (ETDEWEB)

    O' Hira, Shigeru; Nagao, Hiroshi; Fujii, Yasuhiko; Okamoto, Makoto

    1986-04-01

    Li/sub 2/O pellets irradiated with thermal neutrons were isothermally annealed to release tritium in a helium atmosphere at temperatures ranging from 673 to 1073 K. The release rates were found to significantly increase at elevated temperatures and to depend on the density of the Li/sub 2/O pellet. The overall diffusion coefficients of the release process were calculated using the cylindrical geometry model for the pellets as D(cm/sup 2/ s/sup -1/)=1.02 x 10/sup -3/ exp(-51.0 kJ mol/sup -1//RT)(90% theoretical density pellets), and D (cm/sup 2/ s/sup -1/)=2.64 x 10/sup -3/ exp(-46.5 kJ mol/sup -1//RT)(ca. 80% T.D. pellets) over the region 773 <= T <= 1073/sup 0/K. The result of the release experiment at 673/sup 0/K sugested that the diffusion rate was controlled by the decomposition of lithium hydroxide on the surface of Li/sub 2/O grains.

  11. The behaviour of ceramic breeder materials with respect to tritium release and pellet/pebble mechanical integrity

    Science.gov (United States)

    Kwast, H.; Conrad, R.; May, R.; Casadio, S.; Roux, N.; Werle, H.

    1994-09-01

    In situ tritium release experiments from several candidate fusion blanket ceramic breeder materials have been performed in the High Flux Reactor (HFR) at Petten over the last few years. The sixth experiment, EXOTIC-6, contained pellets of LiAlO 2, Li 2XrO 3, Li 6Xr 2O 7 and Li 8ZrO 6 and pebbles of Li 4SiO 4 and Li 2ZrO 3 which were irradiated up to a lithium burnup of 3%. A large number of temperature transients and purge gas composition changes were performed. From the temperature transients tritium residence times have been determined. Some preliminary results were presented at the 17th Symposium on Fusion Technology (SOFT) held in Rome in 1992. In the present paper results of a further analysis of the residence times are presented together with some postirradiation examination results. The LiAlO 2 pellets showed a better mechanical stability than the Li-zirconates pellets. The pebbels remained intact. The tritium residence times determined from the tritium inventories were in good agreement with those previously determined from temperature transients. The tritium release characteristics of the materials investigated remain substantially unchanged up to the maximum lithium burnup achieved in this experiment.

  12. Kinetic isotope effects in reaction of ferment oxidation of tritium-labelled D-galactosamine

    International Nuclear Information System (INIS)

    Akulov, G.P.; Korsakova, N.A.

    1992-01-01

    Primary, secondary and intramolecular kinetic isotopic effects in reaction of ferment oxidation of D-galactosamine labelled by tritium in position 6, were measured. When comparing values of the effects with previously obtained results for similar reaction D-[6- 3 H]galactose, it was ascertained that the presence of aminogroup in galactopyranosyl mainly affects kinetics of substrate-ferment complex formation stage. The possibility to use kinetic isotope effects for increase in molar activity of D-galactosamine, labelled by tritium in position 6, is shown

  13. Tritium effluent removal system

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Gibbs, G.E.

    1978-01-01

    An air detritiation system has been developed and is in routine use for removing tritium and tritiated compounds from glovebox effluent streams before they are released to the atmosphere. The system is also used, in combination with temporary enclosures, to contain and decontaminate airborne releases resulting from the opening of tritium containment systems during maintenance and repair operations. This detritiation system, which services all the tritium handling areas at Mound Facility, has played an important role in reducing effluents and maintaining them at 2 percent of the level of 8 y ago. The system has a capacity of 1.7 m 3 /min and has operated around the clock for several years. A refrigerated in-line filtration system removes water, mercury, or pump oil and other organics from gaseous waste streams. The filtered waste stream is then heated and passed through two different types of oxidizing beds; the resulting tritiated water is collected on molecular sieve dryer beds. Liquids obtained from regenerating the dryers and from the refrigerated filtration system are collected and transferred to a waste solidification and packaging station. Component redundancy and by-pass capabilities ensure uninterrupted system operation during maintenance. When processing capacity is exceeded, an evacuated storage tank of 45 m 3 is automatically opened to the inlet side of the system. The gaseous effluent from the system is monitored for tritium content and recycled or released directly to the stack. The average release is less than 1 Ci/day. The tritium effluent can be reduced by isotopically swamping the tritium; this is accomplished by adding hydrogen prior to the oxidizer beds, or by adding water to the stream between the two final dryer beds

  14. Investigation of the tritium release from Building 324 in which the stack tritium sampler was off, April 14 through 17, 1998

    International Nuclear Information System (INIS)

    Brown, D.H.

    1998-01-01

    On April 14, 1998, a Pacific Northwest National Laboratory (PNNL) researcher performing work in the Building 324 facility approached facility management and asked if facility management could turn off the tritium sampler in the main exhaust stack. The researcher was demonstrating the feasibility of treating components from dismantled nuclear weapons in a device called a plasma arc furnace and was concerned that the sampler would compromise classified information. B and W Hanford Company (BWHC) operated the facility, and PNNL conducted research as a tenant in the facility. The treatment of 200 components in the furnace would result in the release of up to about 20 curies of tritium through the facility stack. The exact quantity of tritium was calculated from the manufacturing data for the weapons components and was known to be less than 20 curies. The Notice of Construction (NOC) approved by the Washington State Department of Health (WDOH) had been modified to allow releasing 20 curies of tritium through the stack in support of this research. However, there were irregularities in the way the NOC modification was processed. The researcher was concerned that data performed on the sampler could be used to back-calculate the tritium content of the components, revealing classified information about the design of nuclear weapons. He had discussed this with the PNNZ security organization, and they had told him that data from the sampler would be classified. He was also concerned that if he could not proceed with operation of the plasma arc furnace, the furnace would be damaged. The researcher told BWHC management that the last time the furnace was shut down and restarted it had cost $0.5 million and caused a six month delay in the project's schedule. He had already begun heating up the furnace before recognizing the security problem and was concerned that stopping the heatup could damage the furnace. The NOC that allowed the research did not have an explicit requirement to

  15. Assessment of radiological releases to the environment from a fusion reactor power plant

    International Nuclear Information System (INIS)

    Shank, K.E.; Oakes, T.W.; Easterly, C.E.

    1978-05-01

    This report summarizes the expected tritium and activation-product inventories and presents an assessment of the potential radiological releases from a fusion reactor power plant, hypothetically located at the Oak Ridge National Laboratory. Routine tritium releases and the resulting dose assessment are discussed. Uncertainties associated with the conversion of tritium gas to tritium oxide and the global tritium cycling are evaluated. The difficulties of estimating releases of activated materials and the subsequent dose commitment are reviewed

  16. A study of tritium behavior in lithium oxide by ion conductivity measurements

    International Nuclear Information System (INIS)

    Noda, Kenji; Ishii, Yoshinobu; Ohno, Hideo; Watanabe, Hitoshi

    1989-01-01

    Ion conductivity of lithium oxide (Li 2 O) irradiated with oxygen ions was measured to obtain information about the effects of irradiation on the behavior of lithium ions and tritium. The conductivity around 490 K decreased with the ion fluence, while around 440 K it increased. The decrease around 490 K and the increase around 440 K were assumed to be attributed to the F + centers and the unspecified radiation defects, respectively. From the point of view that the rate determinant in the mechanism of diffusion of lithium ions in Li 2 O leading to the ion conductivity is the same as that of tritium, the diffusivity of tritium is assumed to be as follows: the diffusivity of tritium is decreased by the F + centers in the range from 490 K to the temperature at which almost all of F + centers are recovered, while it is increased around 440 K by the unspecified radiation defects. In addition, effects of the irradiation on valence states of tritium (i.e., T + , T - ) were discussed in terms of the radiation defects. (orig.)

  17. A study of wet deposition of atmospheric tritium releases at the Ontario Power Generation, Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Crooks, G.; DeWilde, J.; Yu, L.

    2001-01-01

    The Ontario Power Generation,Pickering Nuclear Generating Station (PNGS) has been investigating deposition of atmospheric releases of tritium on their site. This study has included numerical dispersion modelling studies conducted over the past three years, as well as an ongoing field monitoring study. The following paper will present results of the field monitoring study and make comparisons to the numerical modelling. The results of this study could be of potential use to nuclear stations in quantifying tritium deposition in near field regions where building wake effects dominate pollutant dispersion

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

    Science.gov (United States)

    Thompson, P A; Kwamena, N-O A; Ilin, M; Wilk, M; Clark, I D

    2015-02-01

    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. Crown Copyright © 2014. Published by Elsevier Ltd. All rights reserved.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

    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

  1. Spherical diffusion of tritium from a point of release in a uniform unsaturated soil

    International Nuclear Information System (INIS)

    Smiles, D.E.; Gardner, W.R.; Schulz, R.K.

    1993-10-01

    Tritium (Tr), when released as tritiated water at a point in a uniform and relatively dry soil, redistributes in both the liquid and vapor phases. The flux density of Tr in the liquid will exceed that in the vapor phase provided the water content is greater than approximately 15% of the total soil porosity. Thus Tr redistribution must be modeled recognizing transfer ''in parallel'' in both phases. The authors use the diffusion equation cast in spherical coordinates to analyze this problem in order to provide a basis for design of field experiments, and to offer observations on the long term behavior of such systems. The solution of the diffusion equation permits calculation of the evolution of profiles of Tr concentration, within and external to the sphere of released solution, assuming the initial concentration within this sphere to be uniform. The authors also predict the rate of advance of the maximum of Tr as it advances, and attenuates, in the soil. Calculations for the case of 1 million Curies of Tr diluted in 1 liter of water and released at a depth of 20 meters, and 200 meters above the water table, are demonstrated. If the soil has an initial water volume fraction of 0.06 and total porosity of 0.3 they show, for example, that at 5 meters from the point of discharge, the Tr concentration increases to a maximum in 24 years and then slowly declines. That maximum is 1 Curie/liter. The concentration in the gas phase will be 5 orders-of-magnitude less than this. At 60 meters the maximum ever reached in the liquid phase is ca 10 -21 Ci/liter; that maximum will be achieved after 408 years. The authors discuss the effects of variation in the volume fractions of water and air originally present in the soil on the effective diffusion coefficient of Tr in soil, consider the effects of a net flux of water in the system, and identify questions to be answered to achieve safe systematic disposal of tritium in the deep unsaturated zone of desert soil

  2. Conceptual design of an emergency tritium clean-up system

    International Nuclear Information System (INIS)

    Muller, M.E.

    1978-01-01

    The Los Alamos Scientific Laboratory (LASL) has been selected by the Department of Energy (DOE) to design, build, and operate a facility to demonstrate the operability of the tritium-related subsystems that would be required to successfully develop fusion reactor systems. An emergency tritium clean-up subsystem (ETC) for this facility will be designed to remove tritium from the cell atmosphere if an accident causes the primary and secondary tritium containment to be breached. Conceptually, the ETC will process cell air at the rate of 0.65 actual m 3 /s and will achieve an overall decontamination factor of 10 6 per tritium oxide (T 2 O). Following the maximum credible release of 100 g of tritium, the ETC will restore the cell to opertional status within 24 h without a significant release of tritium to the environment

  3. Sources of tritium

    International Nuclear Information System (INIS)

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

    1980-12-01

    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

  4. In vitro determination of oxidation of atmospheric tritium gas in vegetation and soil in Ibaraki and Gifu, Japan

    International Nuclear Information System (INIS)

    Ichimasa, Michiko; Ichimasa, Yusuke; Suzuki, Masatomo; Obayashi, Haruo; Sakuma, Youichi

    1999-01-01

    To quantify the rate of oxidation of tritium gas (referred to as HT) to tritiated water in the environment, various woody and herbaceous plant leaves and roots, mosses and lichens taken from a forest and fields in Ibaraki prefecture, and a forest in Toki, Gifu prefecture, were investigated as to their ability to oxidize atmospheric HT in vitro experiments. The HT oxidation activity in vegetation was compared with that in the surrounding surface soil (0-5 cm in depth). The rate of oxidation of HT in woody plant leaves including pine needles was extremely low, only about 1/10000-1/1000 that in the surface soil, as well as in herbaceous plant leaves with some exceptions (Phalaris arundinacea and Vaccinium smallii), whereas the rate in mosses and lichens was 50-500 times that in pine needles. The HT oxidation activity in roots of several plants including Phalaris arundinacea, Pieris japonica and Lespedeza homoloba was quite high and comparable to that in the surrounding surface soil. These results suggest that mosses, lichens and the leaves or roots of particular plants with high HT oxidation activity can be used to monitor the accidental release of HT into the environment. (author)

  5. A database on tritium behavior in the chronic HT release experiment. 1. Meteorological data and tritium concentrations in air and soil

    Energy Technology Data Exchange (ETDEWEB)

    Noguchi, Hiroshi; Yokoyama, Sumi; Kinouchi, Nobuyuki; Murata, Mikio; Amano, Hikaru; Ando, Mariko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Fukutani, Satoshi

    1999-03-01

    This report comprises a database that can be used to develop and validate tritium models to assess doses to the general public due to HT continuously released from fusion facilities into the atmosphere. The data was collected in the 1994 chronic HT release experiment carried out at the Chalk River Laboratories in Canada. The data set include meteorological conditions such as solar radiation, net solar radiation, wind speed, air temperature and humidity, soil temperature and soil heat flux; soil conditions such as bulk density, water content and free pore volume fraction; HT and HTO concentrations in air, HTO concentrations in soil moisture and HTO deposition to water surface. Evapo-transpiration rates and turbulent diffusivity are estimated and tabulated. The report also contains experimental methods to observe meteorological conditions and take air and soil samples. (author)

  6. Transfer of tritium released into the marine environment by French nuclear facilities bordering the English Channel.

    Science.gov (United States)

    Fiévet, Bruno; Pommier, Julien; Voiseux, Claire; Bailly du Bois, Pascal; Laguionie, Philippe; Cossonnet, Catherine; Solier, Luc

    2013-06-18

    Controlled amounts of liquid tritium are discharged as tritiated water (HTO) by the nuclear industry into the English Channel. Because the isotopic discrimination between 3H and H is small, organically bound tritium (OBT) and HTO should show the same T/H ratio under steady-state conditions. We report data collected from the environment in the English Channel. Tritium concentrations measured in seawater HTO, as well as in biota HTO and OBT, confirm that tritium transfers from HTO to OBT result in conservation of the T/H ratio (ca. 1 × 10(-16)). The kinetics of the turnover of tritium between seawater HTO, biota HTO, and OBT was investigated. HTO in two algae and a mollusk is shown to exchange rapidly with seawater HTO. However, the overall tritium turnover between HTO and the whole-organism OBT is a slow process with a tritium biological half-life on the order of months. Nonsteady-state conditions exist where there are sharp changes in seawater HTO. As a consequence, for kinetic reasons, the T/H ratio in OBT may deviate transiently from that observed in HTO of samples from the marine ecosystem. Dynamic modeling is thus more realistic for predicting tritium transfers to biota OBT under nonsteady-state conditions.

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

    International Nuclear Information System (INIS)

    Jain, Narendra; Bhatia, A.L.

    2008-01-01

    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 16 th 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

  8. Tritium dosimetry and standardization

    International Nuclear Information System (INIS)

    Balonov, M.I.

    1983-01-01

    Actual problem of radiation hygiene such as an evaluation of human irradiation hazard due to a contact with tritium compounds both in industrial and public spheres is under discussion. Sources of tritium release to environment are characterized. Methods of tritium radiation monitoring are discussed. Methods of dosimetry of internal human exposure resulted from tritium compounds are developed on the base of modern representations on metbolism and tritium radiobiological effect. A system of standardization of permissible intake of tritium compounds for personnel and persons of population is grounded. Some protection measures are proposed as applied to tritium overdosage

  9. Change in activity of catalysts for the oxidation of tritium during a fire event

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Sato, Katsumi; Yamanishi, Toshihiko

    2012-01-01

    Highlights: ► We experimentally demonstrated the influence of produced gases from burned low-halogen cable on the activity of catalysts for tritium oxidation. ► At 423 K, no considerable decrease in catalytic activity was observed. ► At 293 K, considerable increase in catalytic activity was initially observed due to the effect of produced hydrogen. Then the temporary decrease was observed due mainly to the effect of produced moisture, however the activity was gradually recovered. - Abstract: The catalytic performance should be maintained in any off normal events. Fire accident is the typical off normal event. In the fusion plant, typical combustibles are evaluated to be polymeric low-halogen cables. Produced gases from burned low-halogen cable may affect the activity of catalysts for the oxidation of tritium. We experimentally demonstrated the influence of produced gases from burned low-halogen cable on the activity of catalyst using tritium gas. Our evaluation showed that ethylene, methane and benzene were major produced gases. The activity of catalysts for the oxidation of tritium during a fire event was evaluated using a commercial hydrophilic Pt/Al 2 O 3 catalyst and a commercial hydrophobic Pt-catalyst. The temperature of catalytic reactor was selected to be 423 and 293 K. At 423 K, no considerable decrease in catalytic activity was observed for both catalysts even in the presence of produced gases from burned low-halogen cable. At 293 K, considerable increase in catalytic activity was initially observed for both catalysts due to the effect of produced hydrogen. Then the temporary decrease was observed, however the catalytic activity was gradually recovered to be the original activity. Consequently, the irreversible decrease in activity of the catalysts during a fire event was not observed.

  10. Oxidation of atmospheric molecular tritium in plant leaves, lichens and mosses

    Energy Technology Data Exchange (ETDEWEB)

    Ichimasa, Michiko; Ichimasa, Yusuke; Yagi, Yoshimi; Ko, Rinkei [Ibaraki Univ., Mito (Japan). Faculty of Science; Suzuki, Masatomo; Akita, Yasukazu

    1989-12-01

    The oxidation of atmospheric molecular tritium (HT) in vegetation was determined by in vitro experiments for pine needles, pine bark, lichens attached to pine trees, taken from a coastal pine forest in Ibaraki prefecture and comparison of such measurements was made with those in soil. The oxidation of HT in pine needles was extremely low, being only about 1/40000 that in the surface soil of a pine forest, whereas its oxidation in pine bark with a lichen was almost 1000-7000 times higher than that in pine needles. HT oxidation in pine bark, a lichen and a moss was determined in each case under light and dark conditions and was found to be essentially the same. All mosses and lichens examined in the present study were found to have unusually high levels of HT oxidation whether their habitat was tree or ground surface. (author).

  11. Oxidation of atmospheric molecular tritium in plant leaves, lichens and mosses

    International Nuclear Information System (INIS)

    Ichimasa, Michiko; Ichimasa, Yusuke; Yagi, Yoshimi; Ko, Rinkei; Suzuki, Masatomo; Akita, Yasukazu.

    1989-01-01

    The oxidation of atmospheric molecular tritium (HT) in vegetation was determined by in vitro experiments for pine needles, pine bark, lichens attached to pine trees, taken from a coastal pine forest in Ibaraki prefecture and comparison of such measurements was made with those in soil. The oxidation of HT in pine needles was extremely low, being only about 1/40000 that in the surface soil of a pine forest, whereas its oxidation in pine bark with a lichen was almost 1000-7000 times higher than that in pine needles. HT oxidation in pine bark, a lichen and a moss was determined in each case under light and dark conditions and was found to be essentially the same. All mosses and lichens examined in the present study were found to have unusually high levels of HT oxidation whether their habitat was tree or ground surface. (author)

  12. EXOTIC: Development of ceramic tritium breeding materials for fusion reactor blankets. The behaviour of tritium in: lithium aluminate, lithium oxide, lithium silicates, lithium zirconates

    Energy Technology Data Exchange (ETDEWEB)

    Kwast, H [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Stijkel, H [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Muis, R [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Conrad, R [Commission of the European Communities, Petten (Netherlands). Joint Reseach Centre

    1995-12-01

    This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 6}Zr{sub 2}O{sub 7}, Li{sub 8}ZrO{sub 6}, Li{sub 2}O and Li{sub 2}SiO{sub 3} 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.).

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

    International Nuclear Information System (INIS)

    Johnson, K.C.; Rouleau, J.

    1991-10-01

    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. Basic study of influence of radiation defects on tritium release processes from lithium silicates

    Energy Technology Data Exchange (ETDEWEB)

    Abramenkovs, A.; Tiliks, J.; Kizane, G.; Supe, A. [Latvia Univ., Riga (Latvia). Dept. of Chem.; Grishmanovs, V. [Department of Quantum Engineering and System Science, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113 (Japan)

    1997-09-01

    The radiolysis of Li{sub 2}SiO{sub 3} and Li{sub 4}SiO{sub 4} was studied using the chemical scavengers method (CSM), thermoluminescence, lyoluminescence, electron spin resonance and spectrometric methods. The influence of the absorbed dose and many another parameters such as: irradiation conditions, sample preparation conditions and concentration of impurities on the accumulation rate of each type RD and RP were studied. Several possibilities for reducing the radiolysis of silicates were discussed. It has been found that tritium localization on the surface and in grains proceed by two different mechanisms. Tritium thermoextraction from the surface proceeds as chemidesorption of tritiated water, but from the bulk as diffusion. The tritium retention processes were studied. It has been found that tritium retention depends on irradiation conditions. Tritium retention is due to the formation of chemical bonds Li-T and thermal stable {identical_to}Si-T bonds. The accumulation of colloidal silicon and lithium can increase the tritium retention up to 25-35%. (orig.).

  15. Properties of tritium and its compounds

    International Nuclear Information System (INIS)

    Belovodskij, L.F.; Gaevoj, V.K.; Grishmanovskij, V.I.

    1985-01-01

    Ways of tritium preparation and different aspects of its application are considered. Physicochemical properties of this isotope and some compounds of it - tritium oxides, lithium, titanium, zirconium, uranium tritides, tritium organic compounds - are discussed. In particular, diffusion of tritium and its oxide through different materials, tritium oxidation processes, decomposition of tritium-containing compounds under the action of self-radiation are considered. Main radiobiological tritium properties are described

  16. Tritium release from beryllium pebbles after high temperature irradiation up to 3000 appm He in the HIDOBE-01 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Til, S. van, E-mail: vantil@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Fedorov, A.V.; Stijkel, M.P.; Cobussen, H.L.; Mutnuru, R.K.; Idsert, P. van der [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and The Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-11-15

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

  17. Large-scale distribution of tritium in a commercial product

    International Nuclear Information System (INIS)

    Combs, F.; Doda, R.J.

    1979-01-01

    Tritium enters the environment from various sources including nuclear reactor operations, weapons testing, natural production, and from the manufacture, use and ultimate disposal of commercial products containing tritium. A recent commercial application of tritium in the United States of America involves the backlighting of liquid crystal displays (LCD) in digital electronic watches. These watches are distributed through normal commercial channels to the general public. One million curies (MCi) of tritium were distributed in 1977 in this product. This is a significant quantity of tritium compared with power reactor-produced tritium (3MCi yearly) or with naturally produced tritium (6MCi yearly). This is the single largest commercial application involving tritium to date. The final disposition of tritium from large quantities of this product, after its useful life, must be estimated by considering the means of disposal and the possibility of dispersal of tritium concurrent with disposal. The most likely method of final disposition of this product will be disposal in solid refuse; this includes burial in land fills and incineration. Burial in land fills will probably contain the tritium for its effective lifetime, whereas incineration will release all the tritium gas (as the oxide) to the atmosphere. The use and disposal of this product will be studied as part of an environmental study that is at present being prepared for the U.S. Nuclear Regulatory Commission. (author)

  18. Excretion of organic and inorganic tritiated compounds in cow's milk after ingestion of tritium oxide

    International Nuclear Information System (INIS)

    Van den Hoek, J.; Gerber, G.B.; Kirchmann, R.

    1980-01-01

    The secretion of tritium in milk as water, casein and lipids was studied in cows given tritiated water to drink for 25 days. The parameters of the exponential functions describing the secretion of tritium as water, casein and lipid are determined during this 'loading phase' and also for a 'decay period' of 75 days after the tritiated water administration had been terminated. Secretion started after a short delay (0.5 day). Turnover of all three compounds was short (half time of about 5 days). Calculations of the dilution of the ingested water during metabolism showed that about 83% of the milk water secreted is found to originate from drinking water; the rest comes from water in food and from that formed in metabolism. The parameters of the decay phase indicated that the principle components of water, casein and lipids had about the same turnover rates as during the loading phase. Using the loading and decay phase data, calculations were performed to determine how much activity in each compound and in each metabolic component would be excreted after a single and after a continuous application of tritium oxide when integrated over infinite times. (UK)

  19. Recovery of humoral immunity parameters in mice under a long-term action of tritium oxide

    International Nuclear Information System (INIS)

    Kirillova, E.N.; Man'ko, V.M.; Muksinova, K.N.

    1986-01-01

    Using the mice-males of the CBA line at the age of 10-12 weeks and body mass of 20-23 g the recovery value of quantitative and qualitative factors of humoral immunity under a long-term action of tritium oxide which has been injected during 6 months in the quantity of 370 kBq per 1g of body mass (cumulative dose 8.73 Gy). The long-term internal mice irradiation with tritium oxide resulted in marked devastation of central and peripheral organs of immune system. An earlier and complete recovery of cells quantity in the bone marrow and spleen, recover up to 50% in lymphnodes and minimum repopulation (from 10 to 20%) in thymus as compared with tested animals of the same age is pointed out. In experimental mice CFU 5 pool decrease in bone marrow and spleen is found. CFUs content in the spleen recovered up to the norm, whereas in the bone marrow it constituted not more than 55% of the control. Deep function injury of V-lymphocyte and T - helper precursors the activity of which has not recovered during the whole observation period. The long-term tritium oxide intake lead to antibodies production suppression (by 30-50%), the tendency to the decrease of antibody formation of these animals has been conserved up to the end of life. The functional activity of T - suppressors in humoral response to thymus-dependent antigen during the remote periods upon long-term irradiation decreased more than twice

  20. The study of stability of Pt-SDB in isotopic exchange between tritium oxide and hydrogen

    International Nuclear Information System (INIS)

    Dan Guiping

    2008-06-01

    The stability of Pt-SDB on its catalytic activity and the Pt contained in Pt-SDB have been studied during three years in isotopic exchange between tritium oxide and hydrogen. The convert efficiency of the catalyst declined 10% in first two years and then it is 2% after another two years, when it is used interruptedly. The losing of Pt on Pt-SDB is separately 9.09% and 2.31% after it is washed 2000 hours by water and immersed in water 525 days. (authors)

  1. Tritium distribution between the fuel can and the oxide of fuel elements of light-water reactors

    International Nuclear Information System (INIS)

    Masson, M.

    1986-12-01

    The study on the measurement of tritium and other radionuclide contained in zircaloy fuel cans of the water cooled reactor fuel elements had two aims: the first was to estimate with accuracy the distribution of tritium in a fuel element (can + oxide). The measurement of tritium in the zircaloy fuel cans of the BORSSELE fuel elements associated with the measurement of tritium in the oxide allowed the establishment of a complete tritium balance on an industrial spent fuel element. This result has been compared to the values calculated by the code CEA/SEN and will allow to validate or adjust this calculation. The second aim delt with the characterization of the other radionuclides gaseous (Kr85) or not (Cs 134 and 137) contained in the solid zircaloy wastes (hulls) coming from the industrial reprocessing of ''water cooled'' fuel elements. These activity measurements in the hulls allowed to estimate the residual content of tritium, Kr 85 and other radionuclides which may be found in these solid wastes (high-level βγ radioactive wastes). Original experimental methods have been developed to reach these aims (dissolution in ammonium bifluoride medium and quantitative recovery of gases produced, radiochromatography, and liquid scintillation after double distillation). One tries to explain the presence of Kr 85 in the irradiated can [fr

  2. The nucleic acid metabolism in rat liver after single and long-term administration of tritium oxide

    International Nuclear Information System (INIS)

    Shorokhova, V.B.

    1984-01-01

    It was shown that after a single administration of tritiUm oxide in a dose of 22.2 MBq/g body mass the liver mass increased, the concentration of nucleic acids decreased and the biosynthesjs rate increased dUring a one-month observation. By the end of the observation period (the first year) the parameters under study were normalized. The long-term administration of tritium oxide in daily doses of 0.37, 0.925 and 1.85 MBq/g body mass caused changes in the nucleac acid metabolism which were less manifest (at early times), than in the case of a single injection. At the same time, the long-term administration of tritium oxide in the dose of 0.925 MBq/g caused a substantial disturbance of the nucleic acid metabolism at later times (after 2-9 months)

  3. Comparison of the regulatory models assessing off-site radiological dose due to the routine releases of tritium

    International Nuclear Information System (INIS)

    Hwang, W. T.; Kim, E. H.; Han, M. H.; Choi, Y. H.; Lee, H. S.; Lee, C. W.

    2005-01-01

    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, NEWTRIT model may be have a meaning in the understanding of phenomena for the behavior of HTO released into the environment

  4. Thermal release of {sup 3}He from tritium aged LaNi{sub 4.25}Al{sub 0.75} hydride

    Energy Technology Data Exchange (ETDEWEB)

    Staack, G.C.; Crowder, M.L.; Klein, J.E. [Savannah River National Laboratory, Aiken, SC (United States)

    2015-03-15

    The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 (LaNi{sub 4.25}Al{sub 0.75})in the tritium process to store hydrogen isotopes. The vast majority of {sup 3}He 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 {sup 3}He and residual hydrogen isotopes, the second was primarily {sup 3}He. The bulk of the gas was released by 600 Celsius degrees. (author)

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

    International Nuclear Information System (INIS)

    Raskob, W.

    1994-10-01

    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.)

  6. ZEPHYR tritium system

    International Nuclear Information System (INIS)

    Swansiger, W.; Andelfinger, C.; Buchelt, E.; Fink, J.; Sandmann, W.; Stimmelmayr, A.; Wegmann, H.G.; Weichselgartner, H.

    1982-04-01

    The ignition experiment ZEPHYR will need tritium as an essential component of the fuel. The ZEPHYR Tritium Systems are designed as to recycle the fuel directly at the experiment. An amount of tritium, which is significantly below the total throughput, for example 10 5 Ci will be stored in uranium getters and introduced into the torus by a specially designed injection system. The torus vacuum system operates with tritium-tight turbomolecular pumps and multi-stage roots pumps in order to extract and store the spent fuel in intermediate storage tanks at atmospheric pressure. A second high vacuum system, similar in design, serves as to evacuate the huge containments of the neutral injection system. The spent fuel will be purified and subsequently processed by an isotope separation system in which the species D 2 , DT and T 2 will be recovered for further use. This isotope separation will be achieved by a preparative gaschromatographic process. All components of the tritium systems will be installed within gloveboxes which are located in a special tritium handling room. The atmospheres of the gloveboxes and of the tritium rooms are controlled by a tritium monitor system. In the case of a tritium release - during normal operation as well as during an accident - these atmospheres become processed by efficient tritium absorption systems. All ZEPHYR tritium handling systems are designed as to minimize the quantity of tritium released to the environment, so that the stringent German laws on radiological protection are satisfied. (orig.)

  7. Determination of organically bound Tritium in environmental samples by application of the oxidizing plasma technique

    International Nuclear Information System (INIS)

    Strack, S.; Koenig, L.A.

    1981-12-01

    The low-temperature oxidizing plasma technique with a suitable system for trapping the water formed in the oxidation process can be used to determine T bound organically in low-level samples. First, the samples are freeze-dried and the tissue water obtained in this way is measured, after distillation, in a liquid scintillation spectrometer. The residual dry matter is ashed in the reactor chamber of the plasma system. Oxidation takes place at temperatures not exceeding 200 0 C in an oxygen flow of about 40 ml/min. The water of oxidation is collected in a cold trap installed behind the reactor chamber. A volume of about 10 ml of water is sufficient to measure the tritium activity without enrichment. The oxidation behavior of various organic materials has been tested. Some first results of T concentrations in tissue water and the organic dry matter from food and plant samples collected in the vicinity of the Nuclear Research Center are presented. The method has the advantage that a commercially available instrument can be used requiring only little additional equipment. Handling is much less dangerous and contamination effects by atmospheric T can be easily kept at a minimum. (orig./HP) [de

  8. Changes of medullary hemopoiesis produced by chronic exposure to tritium oxide and external γ-radiation

    International Nuclear Information System (INIS)

    Murzina, L.D.; Muksinova, K.N.

    1982-01-01

    A comparative study of a chronic effect of tritium oxide ( 3 HOH) and external γ-radiation by 137 Cs on medullary hemopoiesis was conducted in experiments on Wistar rats. 3 HOH was administered for 3mos., 37x10 4 Bk per lg per of body mass daily (the absorbed dose 10.8 Gy), external irradiation was given in correlated values of dose rates and integral doses. Bone marrow depopulation was 1.9 times as deeper in rats exposed to 3 HOH as compared to that in irradiated rats. This difference is caused by early and stable inhibition of erythropoiesis with the administration of the radionuclide. The integral index showing the injuring effect of tritium on erythropoiesis was 4 times as high as compared to that of external γ-irradiation by 137 Cs. The time course of value of the proliferative pool of bone marrow granulocytes with the exposure to 2 types of radiation was monotypic. Differences in maturing and functioning granulocytic pools were marked in early time of the experiment

  9. Results for SEAFP-subtask A 10: Assessments of individual and collective doses to the public for routine and accidental releases of tritium and activation products

    International Nuclear Information System (INIS)

    Raskob, W.

    1995-04-01

    Dose calculations have been performed for accidental releases of tritium and activation products. Unit releases of 1 GBq per nuclide have been investigated for 31 fusion relevant nuclides. Additionally, unit releases of 1g of tritium and several activated materials have been studied. Under normal operation conditions, dose calculations have been performed for real source terms of tritium and activated materials. The individual dose values at the fence of the site (1 km) as well as the collective dose to the public (from 1 km to 100 km) have been obtained. As site specific parameters are still missing, different so called ''worst case'' release conditions have been applied. To have a first guess of the influence of the release duration on the dose to the Most Exposed Individual (MEI) in the vicinity of a reactor, different release durations, ranging from 1 hour up to 168 hours have been investigated, too. Finally, dose calculations have been performed for mobilisation source terms which take account of deposition and retention in the plant. This has been done for several RPM and APM source terms. The dose values of these final source terms seem to be less than every criteria to start emergency actions, however, some problems e.g. the behaviour of tritium in the plant, remain unsolved. (orig.)

  10. Environmental aspects of tritium

    International Nuclear Information System (INIS)

    Quisenberry, D.R.

    1979-01-01

    The potential radiological implications of environmental tritium releases must be determined in order to develop a programme for dealing with the tritium inventory predicted for the nuclear power industry which, though still in its infancy, produces tritium in megacurie quantities annually. Should the development of fusion power generation become a reality, it will create a potential source for large releases of tritium, much of it in the gaseous state. At present about 90% of the tritium produced enters the environment through gaseous and liquid effluents and is deposited in the hydrosphere as tritiated water. Tritium can be assimilated by plants and animals and organically bound, regardless of the exposure pathway. However, there appears to be no concentration factor relative to hydrogen at any level of food chains analysed to date. The body burden, for man, is dependent on the exposure pathway and tissue-bound fractions are primarily the result of organically bound tritium in food. (author)

  11. Tritium Storage Material

    International Nuclear Information System (INIS)

    Cowgill, Donald F.; Luo, Weifang; Smugeresky, John E.; Robinson, David B.; Fares, Stephen James; Ong, Markus D.; Arslan, Ilke; Tran, Kim L.; McCarty, Kevin F.; Sartor, George B.; Stewart, Kenneth D.; Clift, W. Miles

    2008-01-01

    Nano-structured palladium is examined as a tritium storage material with the potential to release beta-decay-generated helium at the generation rate, thereby mitigating the aging effects produced by enlarging He bubbles. Helium retention in proposed structures is modeled by adapting the Sandia Bubble Evolution model to nano-dimensional material. The model shows that even with ligament dimensions of 6-12 nm, elevated temperatures will be required for low He retention. Two nanomaterial synthesis pathways were explored: de-alloying and surfactant templating. For de-alloying, PdAg alloys with piranha etchants appeared likely to generate the desired morphology with some additional development effort. Nano-structured 50 nm Pd particles with 2-3 nm pores were successfully produced by surfactant templating using PdCl salts and an oligo(ethylene oxide) hexadecyl ether surfactant. Tests were performed on this material to investigate processes for removing residual pore fluids and to examine the thermal stability of pores. A tritium manifold was fabricated to measure the early He release behavior of this and Pd black material and is installed in the Tritium Science Station glove box at LLNL. Pressure-composition isotherms and particle sizes of a commercial Pd black were measured.

  12. TFTR tritium handling concepts

    International Nuclear Information System (INIS)

    Garber, H.J.

    1976-01-01

    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

  13. Irradiaiton facilities for testing solid and liquid blanket breeder materials with in-situ tritium release measurements in the HFR Petten

    International Nuclear Information System (INIS)

    Conrad, R.; Debarberis, L.

    1991-01-01

    Lithium-based tritium breeder materials for solid and liquid fusion reactor blanket concepts are being tested in the High Flux Reactor (HFR) Petten with in-situ tritium release measurements since 1985, within the European Fusion Technology Programme and the BEATRIX-I programme. Ceramic breeder materials are being tested in the EXOTIC and COMPLIMENT experimental programmes and the liquid breeder material, Pb-17Li, is being tested in the LIBRETTO experimental programme. The in-pile experiments are performed with irradiation facilities developed by the Joint Research Centre (JRC) Petten. The irradiation vehicles are multi-channel rigs. The sample holders consist of independent, fully instrumented and triple contained capsules. The out-of-pile experimental equipment consist of twelve independent circuits for on-line tritium release and tritium permeation measurements and eight independent circuits for temperature control. The experimental achievements obtained so far contribute to the selection of candidate tritium breeder materials for blanket concepts of near future machines like NET, ITER and DEMO. (orig.)

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

    International Nuclear Information System (INIS)

    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

  15. Tritium fuel cycle modeling and tritium breeding analysis for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Hongli; Pan, Lei; Lv, Zhongliang; Li, Wei; Zeng, Qin, E-mail: zengqin@ustc.edu.cn

    2016-05-15

    Highlights: • A modified tritium fuel cycle model with more detailed subsystems was developed. • The mean residence time method applied to tritium fuel cycle calculation was updated. • Tritium fuel cycle analysis for CFETR was carried out. - Abstract: Attaining tritium self-sufficiency is a critical goal for fusion reactor operated on the D–T fuel cycle. The tritium fuel cycle models were developed to describe the characteristic parameters of the various elements of the tritium cycle as a tool for evaluating the tritium breeding requirements. In this paper, a modified tritium fuel cycle model with more detailed subsystems and an updated mean residence time calculation method was developed based on ITER tritium model. The tritium inventory in fueling system and in plasma, supposed to be important for part of the initial startup tritium inventory, was considered in the updated mean residence time method. Based on the model, the tritium fuel cycle analysis of CFETR (Chinese Fusion Engineering Testing Reactor) was carried out. The most important two parameters, the minimum initial startup tritium inventory (I{sub m}) and the minimum tritium breeding ratio (TBR{sub req}) were calculated. The tritium inventories in steady state and tritium release of subsystems were obtained.

  16. Estimation of accumulated individual doses from waterborne pulse releases of tritium and activation products into the Baltic Sea from a hypothetical fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Edlund, O

    1995-04-01

    The committed effective dose to a critical group during 50 years attributed to pulse unit releases of tritium and activation products (in total 66 nuclides) from a thought postulated fusion power plant into Tvaeren Bay, a part of the Baltic Sea with brackish water, outside Studsvik, Sweden, is calculated. The purpose of this work is to obtain (simple) relationships between release and dose for relevant nuclides for use in the future when realistic source terms are known. 26 refs, 8 tabs, 4 figs.

  17. Estimation of accumulated individual doses from waterborne pulse releases of tritium and activation products into the Baltic Sea from a hypothetical fusion reactor

    International Nuclear Information System (INIS)

    Edlund, O.

    1995-04-01

    The committed effective dose to a critical group during 50 years attributed to pulse unit releases of tritium and activation products (in total 66 nuclides) from a thought postulated fusion power plant into Tvaeren Bay, a part of the Baltic Sea with brackish water, outside Studsvik, Sweden, is calculated. The purpose of this work is to obtain (simple) relationships between release and dose for relevant nuclides for use in the future when realistic source terms are known. 26 refs, 8 tabs, 4 figs

  18. Excretion of organic and inorganic tritiated compounds in cow's milk after ingestion of tritium oxide

    International Nuclear Information System (INIS)

    Van den Hoek, J.; Gerber, G.B.; Kirchmann, R.

    1980-01-01

    The secretion in milk of tritium as water and organic molecules (casein and lipids) was studied after giving cows tritiated water for a period of 25 days. The data indicate that after a single application, tritium is mainly excreted as tritium water of rapid turnover. After continuous application, molecules with long half life, particularly lipids, become more important and may represent somewhat more than 4% of the total. (H.K.)

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

    International Nuclear Information System (INIS)

    Fuetterer, M.A.; Raepsaet, X.; Proust, E.

    1994-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fuetterer, M A; Raepsaet, X; Proust, E

    1994-12-31

    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.

  1. RECOMMENDED TRITIUM OXIDE DEPOSITION VELOCITY FOR USE IN SAVANNAH RIVER SITE SAFETY ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Lee, P.; Murphy, C.; Viner, B.; Hunter, C.; Jannik, T.

    2012-04-03

    The Defense Nuclear Facilities Safety Board (DNFSB) has recently questioned the appropriate value for tritium deposition velocity used in the MELCOR Accident Consequence Code System Ver. 2 (Chanin and Young 1998) code when estimating bounding dose (95th percentile) for safety analysis (DNFSB 2011). The purpose of this paper is to provide appropriate, defensible values of the tritium deposition velocity for use in Savannah River Site (SRS) safety analyses. To accomplish this, consideration must be given to the re-emission of tritium after deposition. Approximately 85% of the surface area of the SRS is forested. The majority of the forests are pine plantations, 68%. The remaining forest area is 6% mixed pine and hardwood and 26% swamp hardwood. Most of the path from potential release points to the site boundary is through forested land. A search of published studies indicate daylight, tritiated water (HTO) vapor deposition velocities in forest vegetation can range from 0.07 to 2.8 cm/s. Analysis of the results of studies done on an SRS pine plantation and climatological data from the SRS meteorological network indicate that the average deposition velocity during daylight periods is around 0.42 cm/s. The minimum deposition velocity was determined to be about 0.1 cm/s, which is the recommended bounding value. Deposition velocity and residence time (half-life) of HTO in vegetation are related by the leaf area and leaf water volume in the forest. For the characteristics of the pine plantation at SRS the residence time corresponding to the average, daylight deposition velocity is 0.4 hours. The residence time corresponding to the night-time deposition velocity of 0.1 cm/s is around 2 hours. A simple dispersion model which accounts for deposition and re-emission of HTO vapor was used to evaluate the impact on exposure to the maximally exposed offsite individual (MOI) at the SRS boundary (Viner 2012). Under conditions that produce the bounding, 95th percentile MOI exposure

  2. Tritium sampling and measurement

    International Nuclear Information System (INIS)

    Wood, M.J.; McElroy, R.G.; Surette, R.A.; Brown, R.M.

    1993-01-01

    Current methods for sampling and measuring tritium are described. Although the basic techniques have not changed significantly over the last 10 y, there have been several notable improvements in tritium measurement instrumentation. The design and quality of commercial ion-chamber-based and gas-flow-proportional-counter-based tritium monitors for tritium-in-air have improved, an indirect result of fusion-related research in the 1980s. For tritium-in-water analysis, commercial low-level liquid scintillation spectrometers capable of detecting tritium-in-water concentrations as low as 0.65 Bq L-1 for counting times of 500 min are available. The most sensitive method for tritium-in-water analysis is still 3He mass spectrometry. Concentrations as low as 0.35 mBq L-1 can be detected with current equipment. Passive tritium-oxide-in-air samplers are now being used for workplace monitoring and even in some environmental sampling applications. The reliability, convenience, and low cost of passive tritium-oxide-in-air samplers make them attractive options for many monitoring applications. Airflow proportional counters currently under development look promising for measuring tritium-in-air in the presence of high gamma and/or noble gas backgrounds. However, these detectors are currently limited by their poor performance in humidities over 30%. 133 refs

  3. Overview of tritium fast-fission yields

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1981-03-01

    Tritium production rates are very important to the development of fast reactors because tritium may be produced at a greater rate in fast reactors than in light water reactors. This report focuses on tritium production and does not evaluate the transport and eventual release of the tritium in a fast reactor system. However, if an order-of-magnitude increase in fast fission yields for tritium is confirmed, fission will become the dominant production source of tritium in fast reactors

  4. 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

    International Nuclear Information System (INIS)

    Vende, L.

    2012-01-01

    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 ( 14 CO 2 , HT...) or in solution ( 14 CO 3 2- , 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)

  5. Use of tritium and sources

    International Nuclear Information System (INIS)

    Noguchi, Hiroshi

    1997-01-01

    There are many kinds of tritium, sources in the environment. The maximum inventory of them is the nuclear tests, although the atmospheric nuclear test has not been carried out since 1981. So that the inventory originated from them will decrease. By the latest data in 1989, the total amount of released tritium was about 24 PBq/yr by the use of atomic energy in the world. The maximum source was the heavy water moderated reactors, for example, CANDU reactor. In the future, large amount of tritium inventory may be the fusion reactor. The test of JET (Joint European Torus) released about 600 GBq of tritium until March in 1992. 80-90% of them were tritium water (HTO). The amount of tritium released from industries and medicine are limited. Although ITER has a large amount of tritium inventory, the amount of release is seemed not to be larger than other nuclear power facility. (S.Y.)

  6. Handling of tritium at TFTR

    International Nuclear Information System (INIS)

    Pierce, C.W.; Howe, H.J.; Yemin, L.; Lind, K.

    1977-01-01

    Some of the engineering approaches taken at TFTR for the tritium control systems are discussed as the requirements being placed on the tritium systems by the operating scenarios of the Tokamak. The tritium control systems presently being designed for TFTR will limit the annual release to the environment to less than 100 curies

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

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Guttadora, G.; Carpe, A.; Langish, S.; Young, K.M.; Nishi, M.; Shu, W.

    2001-01-01

    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/cm 2 , 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

  8. Organically bound tritium

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.

    1993-01-01

    Tritium released into the environment may be incorporated into organic matter. Organically bound tritium in that case will show retention times in organisms that are considerably longer than those of tritiated water which has significant consequences on dose estimates. This article reviews the most important processes of organically bound tritium production and transport through food networks. Metabolic reactions in plant and animal organisms with tritiated water as a reaction partner are of great importance in this respect. The most important production process, in quantitative terms, is photosynthesis in green plants. The translocation of organically bound tritium from the leaves to edible parts of crop plants should be considered in models of organically bound tritium behavior. Organically bound tritium enters the human body on several pathways, either from the primary producers (vegetable food) or at a higher tropic level (animal food). Animal experiments have shown that the dose due to ingestion of organically bound tritium can be up to twice as high as a comparable intake of tritiated water in gaseous or liquid form. In the environment, organically bound tritium in plants and animals is often found to have higher specific tritium concentrations than tissue water. This is not due to some tritium enrichment effects but to the fact that no equilibrium conditions are reached under natural conditions. 66 refs

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

    International Nuclear Information System (INIS)

    Dickson, R.S.

    1990-11-01

    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 (T 2 ) 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 T 2 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 T 2 and HTO contain unproven assumptions about sorption and T 2 → 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.)

  10. 3He release characteristics of metal tritides and scandium--tritium solid solutions

    International Nuclear Information System (INIS)

    Perkins, W.G.; Kass, W.J.; Beavis, L.C.

    1976-01-01

    Tritides of such metals as Sc, Ti, and Er are useful materials for determining the effects of He accumulation in metallic solids, for example, CTR first wall materials. Such effects include lattice strain and gross deformation which are related to 3 He retention and ultimate release. Long term gas release studies have indicated that, during the early life of a metal ditritide, a large fraction of the 3 He is retained in the solid. At more advanced ages, the 3 He release rate becomes comparable to the generation rate. Statistical analysis of the data indicates that the acceleration in 3 He release rate depends on accumulated 3 He concentration rather than strictly on age. 3 He outgassing results are presented for thin films of ScT 2 , TiT 2 , and ErT 2 , and the critical 3 He concentrations are discussed in terms of a percolation model. Phase transformations which occur on tritide formation cast some doubt on the validity of extrapolating results obtained for metal tritides to predictions regarding the accumulation of helium in metals. Sc is unique among the early transition and rare-earth metals in that the metal exhibits a very high room temperature T solubility (T/Sc = 0.4) with no phase transformation. Indeed, even the lattice parameters of the hcp Sc lattice are only minimally changed by T solution. Single crystal ScT/sub 0.3/ samples in two crystallographic orientations were obtained. Using a very sensitive technique, 3 He emission was measured from both these samples, as well as from fine-grained thin film Sc--T solid solution samples (ScT/sub 0.3-0.4/). The fine-grained film samples release 3 He at 2-3 percent of the generation rate, while the emission rate from the single-crystal samples is approximately 0.05 percent of the generation rate, indicating a strong grain size effect

  11. Tritium emissions reduction facility (TERF)

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Hedley, W.H.

    1993-01-01

    Tritium handling operations at Mound include production of tritium-containing devices, evaluation of the stability of tritium devices, tritium recovery and enrichment, tritium process development, and research. In doing this work, gaseous process effluents containing 400,000 to 1,000,000 curies per year of tritium are generated. These gases must be decontaminated before they can be discharged to the atmosphere. They contain tritium as elemental hydrogen, as tritium oxide, and as tritium-containing organic compounds at low concentrations (typically near one ppm). The rate at which these gases is generated is highly variable. Some tritium-containing gas is generated at all times. The systems used at Mound for capturing tritium from process effluents have always been based on the open-quotes oxidize and dryclose quotes concept. They have had the ability to remove tritium, regardless of the form it was in. The current system, with a capacity of 1.0 cubic meter of gas per minute, can effectively remove tritium down to part-per-billion levels

  12. Tritium behavior on a cultivated plot in the 1994 chronic HT release experiment at Chalk River

    International Nuclear Information System (INIS)

    Noguchi, H.; Yokoyama, S.; Kinouchi, N.; Murata, M.; Amano, H.; Atarashi, M.; Ichimasa, Y.; Ichimasa, M.

    1995-01-01

    The behavior of HT and HTO in air and surface soil has been studied extensively in the chronic HT release experiment carried out at Chalk River during the summer of 1994. HTO concentrations in air moisture and soil water collected in a cultivated plot showed similar time-variations, increasing rapidly during the first and second days and becoming gradual after the first 3-4 days. The air HTO concentration decreased during and following rainfall but recovered within a day. The rainfall reduced the HTO concentrations in ridge soil water but little in furrows. Time histories of HTO concentrations in air moisture and soil water suggest that the system was near steady-state within a continuous HT release period of 12 days, in spite of the presence of rain during the period. The air HTO concentrations on clear days showed diurnal cycles that were higher during daytime than at night. The experimental field had a very complex soil regime with respect to HT deposits. The deposits to soil surface varied depending on soil conditions. 12 refs., 5 figs

  13. Tritium Issues in Next Step Devices

    International Nuclear Information System (INIS)

    C.H. Skinner; G. Federici

    2001-01-01

    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

  14. Tritium Issues in Next Step Devices

    Energy Technology Data Exchange (ETDEWEB)

    C.H. Skinner; G. Federici

    2001-09-05

    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

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

    International Nuclear Information System (INIS)

    Murata, Mikio; Noguchi, Hiroshi; Yokoyama, Sumi

    2000-11-01

    A computer code (TRIDOSE) was developed to assess the environmental impact of atmospheric released tritium gas (T 2 ) from nuclear fusion related facilities. The TRIDOSE simulates dispersion of T 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  17. Tritium gas and tritiated water vapour behaviour in the environment from releases into the atmosphere from fusion reactors

    International Nuclear Information System (INIS)

    Velarde, Marta; Perlado, Manuel

    2001-01-01

    The diffusion of tritium from fusion reactors follows different ways according to the present chemical form, tritium gas or tritiated water vapour. The atmospheric conditions, speed and direction of the wind, rain intensity or stability class, are key factors in the dry and wet deposition. The obtained results demonstrate that the wet deposition is critical for the incorporation of the tritiated water vapour to the natural biological chain. However, the dry deposition is the factor that influences in the tritium gas form. The conversion of HT into HTO in the soil is rapid (1-7 days), and 20% of HT deposited in the soil is reemitted to the atmosphere in the form HTO, while the rest incorporates into the biological cycle. The rain factor accelerates the incorporation of tritium to the ground, the superficial waters and the underground waters

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. Tritium formation and elimination in light-water electronuclear plants

    International Nuclear Information System (INIS)

    Dolle, L.; Bazin, J.

    1977-01-01

    In light-water reactors, the tritium balance should be considered from both the working constraint and environmental pollution aspects. In light-water electronuclear stations with pressurized reactors using boric acid in solution for reactivity control, the amounts of tritium formed in the primary circuit are worthy of note. The estimations concerning the tritium production in a hypothetical 1000 MWe reactor are discussed. In the tritium build-up, the part which takes the tritium formed by fission in the fuel, owing to diffusion through cladding, is still difficult to estimate. The tritium balance in different working nuclear power stations are consequently of interest. But the tritium produced by ternary fission in the fuel is always much more abundant, and remains almost entirely confined in the uranium oxide if the fuel is clad with zircaloy. The annual quantity stored in the fuel elements is more than 20 times larger than that of the built up free tritium in the primary circuit water of a reactor. It reaches about 12,400 Ci in the hypothetical reactor. In the presently operated reprocessing plants, tritium is all going over in the effluents, and is almost entirely released in the environment. Taking into account the increasing quantities of high irradiated fuel to be reprocessed, it seems necessary to develop separation processes. Development work and tests have been achieved jointly by CEA and SAINT-GOBAIN TECHNIQUES NOUVELLES in order to: contain the tritium in the high activity part of the plant; and keep small the tritiated effluent volume, about 300 liters per ton of reprocessed uranium. It is then possible to envisage a storage for decay of isotopic separation processes. Such separation processes have been estimated by CEA assuming a daily output of 1500 liters of water containing 2,3 Ci.1 -1 of tritium, the desired decontamination factor being 100 [fr

  20. Japanese university program on tritium radiobiology and environmental tritium

    International Nuclear Information System (INIS)

    Okada, Shigefumi

    1989-01-01

    The university program of the tritium study in the Special Research Project of Nuclear Fusion (1980-1989) is now on its 9th year. The study's aim is to assess tritium risk on man and environment for development of Japanese Nuclear Fusion Program. The tritium study begun by establishing various tritium safe-handling devices and methods to protect scientists from tritium contamination. Then, the tritium studies were initiated in three areas: The first was the studies on biological effects of tritiated water, where their RBE values, their modifying factors and mechanisms were investigated. Also, several human monitoring systems for detection of tritium-induced damage were developed. The second was the metabolic studies of tritium, including a daily tritium monitoring system, methods to enhance excretion of tritiated water from body and means to prevent oxidation of tritium gas in the body. The third was the study of environmental tritium. Tritium levels in environmental waters of various types were estimated all-over in Japan and their seasonal or regional variation were analyzed. Last two years, the studies were extended to estimate tritium activities of plants, foods and man in Japan. (author)

  1. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  2. Conceptual design of an emergency tritium clean-up system

    International Nuclear Information System (INIS)

    Muller, M.E.

    1978-01-01

    The Los Alamos Scientific Laboratory (LASL) has been selected to design, build, and operate a facility to demonstrate the operability of the tritium-related subsystems that would be required to successfully develop fusion reactor systems. Basically, these subsystems would consist of the deuterium-tritium fuel cycle and associated environmental control systems. An emergency tritium clean-up subsystem (ETC) for this facility will be designed to remove tritium from the cell atmosphere if an accident causes the primary and secondary tritium containment to be breached. Conceptually, the ETC will process cell air at the rate of 0.65 actual m 3 /s (1385 ACFM) and will achieve an overall decontamination factor of 10 6 for tritium oxide (T 2 O). Following the maximum credible release of 100 g of tritium, the ETC will restore the cell to operational status within 24 h without a significant release of tritium to the environment. The basic process will include compression of the air to 0.35 MPa (3.5 atm) in a reciprocating compressor followed by oxidation of the tritium to T 2 O in a catalytic reactor. The air will be cooled to 275 K (350 0 F) to remove most of the moisture, including T 2 O, as a condensate. The remaining moisture will be removed by molecular sieve dryer beds that incorporate a water-swamping step between beds, allowing greater T 2 O removal. A portion of the detritiated air will be recirculated to the cell; the remainder will be exhausted to the building ventilation stack to maintain a slight negative pressure in the cell. The ETC will be designed for maximum flexibility so that studies can be performed that involve various aspects of room air detritiation

  3. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1990-01-01

    This document represents a synthesis relative to tritium storage. After indicating the main storage particularities as regards tritium, storages under gaseous and solid form are after examined before establishing choices as a function of the main criteria. Finally, tritium storage is discussed regarding tritium devices associated to Fusion Reactors and regarding smaller devices [fr

  4. Tritium release in Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Linjie, E-mail: zhaolinjie1989@163.com; Long, Xinggui, E-mail: xingguil@caep.cn; Peng, Shuming, E-mail: pengshuming@caep.cn; Chen, Xiaojun; Xiao, Chengjian; Ran, Guangming; Li, Jiamao

    2016-12-15

    Li{sub 4+x}Si{sub 1−x}Al{sub x}O{sub 4} solid solution materials, which were designed as the advanced tritium breeders, were obtained by indirect solid state reactions. The behaviors of tritium release from Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} powders were investigated by temperature programmed desorption. The tritium release curves show different characteristics for the Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} ceramics. The main tritium release peak in the Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} powders is at approximately 600 °C after a high dose irradiation. Moreover, the temperature of the tritium release from Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} was lower than that of the release from Li{sub 4}SiO{sub 4}. This suggests a possible advantage to using the solid solutions as the advanced tritium breeding materials.

  5. Ammonia release method for depositing metal oxides

    Energy Technology Data Exchange (ETDEWEB)

    Silver, G.L.; Martin, F.S.

    1993-12-31

    A method of depositing metal oxides on substrates which is indifferent to the electrochemical properties of the substrates and which comprises forming ammine complexes containing metal ions and thereafter effecting removal of ammonia from the ammine complexes so as to permit slow precipitation and deposition of metal oxide on the substrates.

  6. Tritium monitor with improved gamma-ray discrimination

    Science.gov (United States)

    Cox, Samson A.; Bennett, Edgar F.; Yule, Thomas J.

    1985-01-01

    Apparatus and method for selective measurement of tritium oxide in an environment which may include other radioactive components and gamma radiation, the measurement including the selective separation of tritium oxide from a sample gas through a membrane into a counting gas, the generation of electrical pulses individually representative by rise times of tritium oxide and other radioactivity in the counting gas, separation of the pulses by rise times, and counting of those pulses representative of tritium oxide. The invention further includes the separate measurement of any tritium in the sample gas by oxidizing the tritium to tritium oxide and carrying out a second separation and analysis procedure as described above.

  7. Diffusion of gases in solids: rare gas diffusion in solids; tritium diffusion in fission and fusion reactor metals. Final report

    International Nuclear Information System (INIS)

    Abraham, P.M.; Chandra, D.; Mintz, J.M.; Elleman, T.S.; Verghese, K.

    1976-01-01

    Major results of tritium and rare gas diffusion research conducted under the contract are summarized. The materials studied were austenitic stainless steels, Zircaloy, and niobium. In all three of the metal systems investigated, tritium release rates were found to be inhibited by surface oxide films. The effective diffusion coefficients that control tritium release from surface films on Zircaloy and niobium were determined to be eight to ten orders of magnitude lower than the bulk diffusion coefficients. A rapid component of diffusion due to grain boundaries was identified in stainless steels. The grain boundary diffusion coefficient was determined to be about six orders of magnitude greater than the bulk diffusion coefficient for tritium in stainless steel. In Zircaloy clad fuel pins, the permeation rate of tritium through the cladding is rate-limited by the extremely slow diffusion rate in the surface films. Tritium diffusion rates through surface oxide films on niobium appear to be controlled by cracks in the surface films at temperatures up to 600 0 C. Beyond 600 0 C, the cracks appear to heal, thereby increasing the activation energy for diffusion through the oxide film. The steady-state diffusion of tritium in a fusion reactor blanket has been evaluated in order to calculate the equilibrium tritium transport rate, approximate time to equilibrium, and tritium inventory in various regions of the reactor blanket as a function of selected blanket parameters. Values for these quantities have been tabulated

  8. Environmental monitoring of molecular tritium

    Energy Technology Data Exchange (ETDEWEB)

    Ichimasa, M.; Ichimasa, Y.; Akita, Y. (Ibaraki Univ., Mito (Japan). Faculty of Science); Suzuki, M.; Obayashi, H.; Sakuma, Y.

    1992-01-01

    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).

  9. The ITER tritium systems

    International Nuclear Information System (INIS)

    Glugla, M.; Antipenkov, A.; Beloglazov, S.; Caldwell-Nichols, C.; Cristescu, I.R.; Cristescu, I.; Day, C.; Doerr, L.; Girard, J.-P.; Tada, E.

    2007-01-01

    ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task

  10. Predicting tritium movement and inventory in fusion reactor subsystems using the TMAP code

    International Nuclear Information System (INIS)

    Jones, J.L.; Merrill, B.J.; Holland, D.F.

    1985-01-01

    The Fusion Safety Program of EG and G Idaho, Inc. at the Idaho National Engineering Laboratory (INEL) is developing a safety analysis code called TMAP (Tritium Migration Analysis Program) to analyze tritium loss from fusion systems during normal and off-normal conditions. TMAP is a one-dimensional code that calculated tritium movement and inventories in a system of interconnected enclosures and wall structures. These wall structures can include composite materials with bulk trapping of the permeating tritium on impurities or radiation induced dislocations within the material. The thermal response of a structure can be modeled to provide temperature information required for tritium movement calculations. Chemical reactions and hydrogen isotope movement can also be included in the calculations. TWAP was used to analyze the movement of tritium implanted into a proposed limiter/first wall structure design. This structure was composed of composite layers of vanadium and stainless steel. Included in these calculations was the effect of contrasting material tritium solubility at the composite interface. In addition, TMAP was used to investigate the rate of tritium cleanup after an accidental release into the atmosphere of a reactor building. Tritium retention and release from surfaces and conversion to the oxide form was predicted

  11. Tritium in the aquatic environment

    International Nuclear Information System (INIS)

    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

  12. Tritium monitor calibration at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Bjork, C.J.; Aikin, D.J.; Houlton, T.W.

    1997-08-01

    Tritium in air is monitored at Los Alamos National Laboratory (LANL) with air breathing instruments based on ionization chambers. Stack emissions are continuously monitored from sample tubes which each connect to a Tritium bubble which differentially collects HTO and HT. A set of glass vials of glycol capture the HTO. The HT is oxidized with a palladium catalyst and the resultant HTO is captured in a second set of vials of glycol. The glycol is counted with a liquid scintillation counter. All calibrations are performed with tritium containing gas. The Radiation Instrumentation and Calibration (RIC) Team has constructed and maintains two closed loop gas handling systems based on femto TECH model U24 tritium ion chamber monitors: a fixed system housed in a fume hood and a portable system mounted on two two wheeled hand trucks. The U24 monitors are calibrated against tritium in nitrogen gas standards. They are used as standard transfer instruments to calibrate other ion chamber monitors with tritium in nitrogen, diluted with air. The gas handling systems include a circulation pump which permits a closed circulation loop to be established among the U24 monitor and typically two to four other monitors of a given model during calibration. Fixed and portable monitors can be calibrated. The stack bubblers are calibrated in the field by: blending a known concentration of tritium in air within the known volume of the two portable carts, coupled into a common loop; releasing that gas mixture into a ventilation intake to the stack; collecting oxidized tritium in the bubbler; counting the glycol; and using the stack and bubbler flow rates, computing the bubbler's efficiency. Gas calibration has become a convenient and quality tool in maintaining the tritium monitors at LANL

  13. Thermosensitive liposomes entrapping iron oxide nanoparticles for controllable drug release

    International Nuclear Information System (INIS)

    Tai, L-A; Wang, Y-C; Wang, Y-J; Yang, C-S; Tsai, P-J; Lo, L-W

    2009-01-01

    Iron oxide nanoparticles can serve as a heating source upon alternative magnetic field (AMF) exposure. Iron oxide nanoparticles can be mixed with thermosensitive nanovehicles for hyperthermia-induced drug release, yet such a design and mechanism may not be suitable for controllable drug release applications in which the tissues are susceptible to environmental temperature change such as brain tissue. In the present study, iron oxide nanoparticles were entrapped inside of thermosensitive liposomes for AMF-induced drug release while the environmental temperature was maintained at a constant level. Carboxyfluorescein was co-entrapped with the iron oxide nanoparticles in the liposomes as a model compound for monitoring drug release and environmental temperature was maintained with a water circulator jacket. These experiments have been successfully performed in solution, in phantom and in anesthetized animals. Furthermore, the thermosensitive liposomes were administered into rat forearm skeletal muscle, and the release of carboxylfluorescein triggered by the external alternative magnetic field was monitored by an implanted microdialysis perfusion probe with an on-line laser-induced fluorescence detector. In the future such a device could be applied to simultaneous magnetic resonance imaging and non-invasive drug release in temperature-sensitive applications.

  14. Tritons and tritides as the solute and diffusing species in ceramic tritium breeders

    International Nuclear Information System (INIS)

    Fischer, A.K.; Johnson, C.E.

    1987-01-01

    Intragranular diffusion of tritium is an inherent participant in the process of releasing tritium from lithium-containing ceramics that are used to breed tritium in a fusion reactor. The nature of this transport is reviewed in terms of the understanding established for the mechanism of hydrogen migration in other oxides, namely, that the diffusing species is the proton and that it moves from oxide ion to oxide ion, thereby giving rise to apparent hydroxide migration. Analogously, the triton, transiently bonded to successive oxides and forming successive tritoxides, is taken to be the dominant migrating species in ceramic breeders. In addition, tritide becomes a significant participant at low oxygen activity. The relationship of tritons and tritides as the migrating species to the observed release of both reduced and oxidized forms can be understood in terms of the thermodynamic conditions that prevail. Mechanisms exist that can be proposed to rationalize the participation of these species

  15. Tritium pollution in the Swiss luminous compound industry

    International Nuclear Information System (INIS)

    Krejci, K.; Zeller, Jr.

    1979-01-01

    The Swiss luminous compound industry is an important consumer of tritium. About 350kCi go into production of tritium gas-filled light sources and 40kCi into production of tritium luminous compound annually. To illustrate the pollution problem, a factory is mentioned that handles 200kCi annually and a chain of luminizers, processing 20kCi over the same period as tritium luminous compound. This material is manufactured by coating phosphors with tritiated polystyrene having a specific activity up to 200Ci/g. Because of the high specific activity, the radiation damage produces an average activity release of 5.2% annually, which is one of the main reasons for public and occupational exposure. The processing of large quantities of tritium gas requires special equipment, such as units made entirely of stainless steel for purification and hydrogenation, oxidation systems for highly contaminated air, glove boxes, ventilation and monitoring systems. Nevertheless, contamination of air, surfaces, water and workers cannot be avoided. Only in a few cases were MPC-values for tritium content in urine of workers exceeded. From these results, biological half-lives between 5-15 days were estimated. Regular medical examinations showed no significant influence in blood picture parameters, except in one single case with a tritium concentration in urine of 2.8mCi/litre. Entirely different problems arise in most luminizing factories where luminous paint is processed as an open radioactive source. (author)

  16. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  17. Tritium inventory prediction in a CANDU plant

    International Nuclear Information System (INIS)

    Song, M.J.; Son, S.H.; Jang, C.H.

    1995-01-01

    The flow of tritium in a CANDU nuclear power plant was modeled to predict tritium activity build-up. Predictions were generally in good agreement with field measurements for the period 1983--1994. Fractional contributions of coolant and moderator systems to the environmental tritium release were calculated by least square analysis using field data from the Wolsong plant. From the analysis, it was found that: (1) about 94% of tritiated heavy water loss came from the coolant system; (2) however, about 64% of environmental tritium release came from the moderator system. Predictions of environmental tritium release were also in good agreement with field data from a few other CANDU plants. The model was used to calculate future tritium build-up and environmental tritium release at Wolsong site, Korea, where one unit is operating and three more units are under construction. The model predicts the tritium inventory at Wolsong site to increase steadily until it reaches the maximum of 66.3 MCi in the year 2026. The model also predicts the tritium release rate to reach a maximum of 79 KCi/yr in the year 2012. To reduce the tritium inventory at Wolsong site, construction of a tritium removal facility (TRF) is under consideration. The maximum needed TRF capacity of 8.7 MCi/yr was calculated to maintain tritium concentration effectively in CANDU reactors

  18. Tritium in plants

    International Nuclear Information System (INIS)

    Vichot, L.; Losset, Y.

    2009-01-01

    The presence of tritium in the environment stems from its natural production by cosmic rays, from the fallout of the nuclear weapon tests between 1953 and 1964, and locally from nuclear industry activities. A part of the tritiated water contained in the foliage of plants is turned into organically bound tritium (OBT) by photosynthesis. The tritium of OBT, that is not exchangeable and then piles up in the plant, can be used as a marker of the past. It has been shown that the quantity of OBT contained in the age-rings of an oak that grew near the CEA center of Valduc was directly correlated with the tritium releases of the center. (A.C.)

  19. Tritium accounting for PHWR plants

    International Nuclear Information System (INIS)

    Nair, P.S.; Duraisamy, S.

    2012-01-01

    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 D 2 O 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)

  20. Tritium conference days; Journees tritium

    Energy Technology Data Exchange (ETDEWEB)

    Garnier-Laplace, J.; Lebaron-Jacobs, L.; Sene, M.; Devin, P.; Chretien, V.; Le Guen, B.; Guetat, Ph.; Baglan, N.; Ansoborlo, E.; Boyer, C.; Masson, M.; Bailly-Du-Bois, P.; Jenkinson, St.; Wakeford, R.; Saintigny, Y.; Romeo, P.H.; Thompson, P.; Leterq, D.; Chastagner, F.; Cortes, P.; Philippe, M.; Paquet, F.; Fournier, M.

    2009-07-01

    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/HTO{sub air} and OBT/HTO{sub free} (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.)

  1. Tritium waste package

    Science.gov (United States)

    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. Fission product release by fuel oxidation after water ingress

    International Nuclear Information System (INIS)

    Schreiber.

    1990-01-01

    On the basis of data obtained by a literature search, a computer code has been established for the calculation of the degree of oxidation of the fuel in the damaged fuel particles, and hence of the fission product release as a function of the time period of steam ingress. (orig.) [de

  3. Development of an on-line tritium monitor with gamma-ray rejection and energy discrimination

    International Nuclear Information System (INIS)

    Cox, S.A.; Yule, T.J.; Bennett, E.F.

    1981-01-01

    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/M 3 ) for the oxide, and discriminating against gamma activity and airborne activity other than tritium

  4. Influence of silicon on hot-dip aluminizing process and subsequent oxidation for preparing hydrogen/tritium permeation barrier

    Energy Technology Data Exchange (ETDEWEB)

    Han, Shilei; Li, Hualing; Wang, Shumao; Jiang, Lijun; Liu, Xiaopeng [Energy Materials and Technology Research Institute, General Research Institute for Nonferrous Metals, Beijing 100088 (China)

    2010-04-15

    The development of the International Thermonuclear Experimental Reactor (ITER) requires the production of a material capable of acting as a hydrogen/tritium permeation barrier on low activation steel. It is well known that thin alumina layer can reduce the hydrogen permeation rate by several orders of magnitude. A technology is introduced here to form a ductile Fe/Al intermetallic layer on the steel with an alumina over-layer. This technology, consisting of two main steps, hot-dip aluminizing (HDA) and subsequent oxidation behavior, seems to be a promising coating method to fulfill the required goals. According to the experiments that have been done in pure Al, the coatings were inhomogeneous and too thick. Additionally, a large number of cracks and porous band could be observed. In order to solve these problems, the element silicon was added to the aluminum melt with a nominal composition. The influence of silicon on the aluminizing and following oxidation process was investigated. With the addition of silicon into the aluminum melt, the coating became thinner and more homogeneous. The effort of the silicon on the oxidation behavior was observed as well concerning the suppression of porous band and cracks. (author)

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

    International Nuclear Information System (INIS)

    Gazal, S.; Amiard, J.C.; Caussade, Bernard; Chenal, Christian; Hubert, Francoise; Sene, Monique

    2010-01-01

    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.)

  6. A prototype wearable tritium monitor

    International Nuclear Information System (INIS)

    Surette, R. A.; Dubeau, J.

    2008-01-01

    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)

  7. Study and application of hydrophobic catalyst in treating tritium waste

    International Nuclear Information System (INIS)

    Dan, Gui-ping; Zhang, Dong; Qiu, Yong-mei; Yuan, Guo-Qi

    2008-01-01

    Tritium decontamination from tritium waste is important for the management of tritium waste. Tritium removal from waste tritium oxide can not only get tritium, but also reduce the amount of waste tritium. At the meantime, by cleaning the tritium pollution gas can also reduce the tritium exhausting from tritium facility. At present, the process of hydrogen isotopic exchange in tritium removal from waste tritium oxide and coordination oxidisation-adsorption in tritium cleaning from waste tritium gas are the mainly methods. In these methods, hydrophobic catalysts which can be used in these process are the key technology. There are many references about their preparing and applying, but few on the estimation about their performance changing during their applying. However, their performance stability on isotopic catalytic exchange and catalytic oxidisation will affect their using in reaction. Hydrophobic catalyst Pt-SDB which can be used in tritium isotopic exchange between tritium oxide and hydrogen and the cleaning of tritium pollution gas have been prepared in our laboratory in early days. In order to estimating their performance stability during their using, this work will investigate their stability on their catalytic activity and their radiation-resistance tritium. (author)

  8. Metabolism of organically bound tritium

    International Nuclear Information System (INIS)

    Travis, C.C.

    1984-01-01

    The classic methodology for estimating dose to man from environmental tritium ignores the fact that organically bound tritium in foodstuffs may be directly assimilated in the bound compartment of tissues without previous oxidation. We propose a four-compartment model consisting of a free body water compartment, two organic compartments, and a small, rapidly metabolizing compartment. The utility of this model lies in the ability to input organically bound tritium in foodstuffs directly into the organic compartments of the model. We found that organically bound tritium in foodstuffs can increase cumulative total body dose by a factor of 1.7 to 4.5 times the free body water dose alone, depending on the bound-to-loose ratio of tritium in the diet. Model predictions are compared with empirical measurements of tritium in human urine and tissue samples, and appear to be in close agreement. 10 references, 4 figures, 3 tables

  9. Tritium sources

    International Nuclear Information System (INIS)

    Glodic, S.; Boreli, F.

    1993-01-01

    Tritium is the only radioactive isotope of hydrogen. It directly follows the metabolism of water and it can be bound into genetic material, so it is very important to control levels of contamination. In order to define the state of contamination it is necessary to establish 'zero level', i.e. actual global inventory. The importance of tritium contamination monitoring increases with the development of fusion power installations. Different sources of tritium are analyzed and summarized in this paper. (author)

  10. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1989-01-01

    A general synthesis about tritium storage is achieved in this paper and a particular attention is given to practical application in the Fusion Technology Program. Tritium, storage under gaseous form and solid form are discussed (characteristics, advantages, disadvantages and equipments). The way of tritium storage is then discussed and a choice established as a function of a logic which takes into account the main working parameters

  11. 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

    International Nuclear Information System (INIS)

    2014-07-01

    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

  12. Tritium emissions from a detritiation facility

    International Nuclear Information System (INIS)

    Rodrigo, L.; El-Behairy, O.; Boniface, H.; Hotrum, C.; McCrimmon, K.

    2010-01-01

    Tritium is produced in heavy-water reactors through neutron capture by the deuterium atom. Annual production of tritium in a CANDU reactor is typically 52-74 TBq/MW(e). Some CANDU reactor operators have implemented detritiation technology to reduce both tritium emissions and dose to workers and the public from reactor operations. However, tritium removal facilities also have the potential to emit both elemental tritium and tritiated water vapor during operation. Authorized releases to the environment, in Canada, are governed by Derived Release Limits (DRLs). DRLs represent an estimate of a release that could result in a dose of 1 mSv to an exposed member of the public. For the Darlington Nuclear Generating Station, the DRLs for airborne elemental tritium and tritiated water emissions are ~15.6 PBq/week and ~825 TBq/week respectively. The actual tritium emissions from Darlington Tritium Removal Facility (DTRF) are below 0.1% of the DRL for elemental tritium and below 0.2% of the DRL for tritiated water vapor. As part of an ongoing effort to further reduce tritium emissions from the DTRF, we have undertaken a review and assessment of the systems design, operating performance, and tritium control methods in effect at the DTRF on tritium emissions. This paper discusses the results of this study. (author)

  13. Tritium concentrations in tree ring cellulose

    International Nuclear Information System (INIS)

    Kaji, Toshio; Momoshima, Noriyuki; Takashima, Yoshimasa.

    1989-01-01

    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.)

  14. Tritium Removal from Codeposits on Carbon Tiles by a Scanning Laser

    International Nuclear Information System (INIS)

    C.H. Skinner; C.A. Gentile; A. Carpe; G. Guttadora; S. Langish; K.M. Young; W.M. Shu; H. Nakamura

    2001-01-01

    A novel method for tritium release has been demonstrated on codeposited layers on graphite and carbon-fiber-composite tiles from the Tokamak Fusion Test Reactor (TFTR). A scanning continuous wave Nd laser beam heated the codeposits to a temperature of 1200-2300 degrees C for 10 to 200 milliseconds in an argon atmosphere. The temperature rise of the codeposit was significantly higher than that of the manufactured tile material (e.g., 1770 degrees C cf. 1080 degrees C). A major fraction of tritium was thermally desorbed with minimal change to the surface appearance at a laser intensity of 8 kW/cm(superscript ''2''), peak temperatures above 1230 degrees C and heating duration 10-20 milliseconds. In two experiments, 46% and 84% of the total tritium was released during the laser scan. The application of this method for tritium removal from a tokamak reactor appears promising and has significant advantages over oxidative techniques

  15. Development of a tritium dispersion code

    International Nuclear Information System (INIS)

    Bell, R.P.; Davis, M.W.; Joseph, S.; Wong, K.Y.

    1985-01-01

    This paper describes the development and verification of a computer code designed to calculate the radiation dose to man following acute or chronic atmospheric releases of tritium gas and oxide from a point source. The Ontario Hydro Tritium Dispersion Code calculates tritium concentrations in air, soil, and vegetation and doses to man resulting from inhalation/immersion and ingestion of food, milk meat and water. The deposition of HT to soil, conversion of HT to HTO by soil enzymes and resuspension of HTO to air have been incorporated into the terrestrial compartment model and are unique features of the code. Sensitivity analysis has identified the HT deposition velocity and the equivalent water depth of the vegetation compartment as two parameters which have a strong influence on dose calculations. Tritium concentrations in vegetation and soil calculated by the code were in reasonable agreement with experimental results. The radiological significance of including the mechanisms of HT to HTO conversion and resuspension of HTO to air is illustrated

  16. Synthesis of tritium labelled 7-dehydrocholesterol 5β, 6β-oxide

    International Nuclear Information System (INIS)

    Michaud, D.P.; Nashed, N.T.; Jerina, D.M.

    1986-01-01

    The synthesis of tritiated 7-dehydrocholesterol 5β, 6β-oxide in high specific activity is reported. Oxidation of 7α-bromo-cholesterol 5β, 6β-oxide to the 3-keto epoxide followed by borotritide reduction, in a special buffer-organic solvent system to minimize undesired rearrangement, regenerated the 3β-hydroxyl group. Base-assisted elimination produced the title compound. (author)

  17. Tritium trick

    Science.gov (United States)

    Green, W. V.; Zukas, E. G.; Eash, D. T.

    1971-01-01

    Large controlled amounts of helium in uniform concentration in thick samples can be obtained through the radioactive decay of dissolved tritium gas to He3. The term, tritium trick, applies to the case when helium, added by this method, is used to simulate (n,alpha) production of helium in simulated hard flux radiation damage studies.

  18. Tritium monitor and collection system

    Science.gov (United States)

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

    1992-01-14

    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. 7 figs.

  19. 13C-Tracer Experiments in DIII-D Preliminary to Thermal Oxidation Experiments to Understand Tritium Recovery in DIII-D, JET, C-Mod, and MAST

    International Nuclear Information System (INIS)

    Stangeby, P.; Allen, S.; Bekris, N.; Brooks, N.; Christie, K.; Chrobak, C.; Coad, J.; Counsell, G.; Davis, J.; Elder, J.; Fenstermacher, M.; Groth, M.; Haasz, A.; Likonen, J.; Lipschultz, B.; McLean, A.; Philipps, V.; Porter, G.; Rudakov, D.; Shea, J.; Wampler, W.; Watkins, J.; West, W.; Whyte, D.

    2006-01-01

    Retention of tritium in carbon co-deposits is a serious concern for ITER. Developing a reliable in-situ removal method of the co-deposited tritium would allow the use of carbon plasma-facing components which have proven reliable in high heat flux conditions and compatible with high performance plasmas. Thermal oxidation is a potential solution, capable of reaching even hidden locations. It is necessary to establish the least severe conditions to achieve adequate tritium recovery, minimizing damage and reconditioning time. The first step in this multi-machine project is 13 C-tracer experiments in DIII-D, JET, C-Mod and MAST. In DIII-D and JET, 13 CH 4 has been (and in C-Mod and MAST, will be) injected toroidally symmetrically, facilitating quantification and interpretation of the results. Tiles have been removed, analyzed for 13 C content and will next be evaluated in a thermal oxidation test facility in Toronto with regard to the ability of different severities of oxidation exposure to remove the different types of (known and measured) 13 C co-deposit. Removal of D/T from B on Mo tiles from C-Mod will also be tested. OEDGE interpretive code analysis of the 13 C deposition patterns is used to generate the understanding needed to apply findings to ITER. First results are reported here for the 13 C injection experiments IN DIII-D

  20. A comparison of individual doses for continuous annual unit releases of tritium and activation products into brackish water and lake-river ecosystems

    Energy Technology Data Exchange (ETDEWEB)

    Edlund, O.; Aquilonius, K.

    1995-12-31

    The annual effective doses to critical group from potential unit releases of tritium and activation products (32 nuclides) from a hypothetical fusion reactor into two aquatic environments, one with brackish water and the other with fresh water, are assessed. Unit continuous releases (1 Bq/year during 50 years) for each relevant activation product are analyzed, and the effective dose rate is calculated for each nuclide. The transfer of released activity is simulated by compartment models using first-order linear differential equations for the transport. The rate constants for the brackish-water ecosystem are based on measurements. Four exposure pathways are considered in the brackish water system, the Tvaeren Bay, (a) consumption of fish, (b) consumption of milk, (c) consumption of meat, and (d) exposure from swimming. For the freshwater system, five additional pathways are considered, namely consumption of (e) water, (f) vegetables, (g) cereals, and (h) root vegetables and (i) external exposure from contaminated ground. The paper presents the compartment models used and a description of how the exposure pathways are treated, especially the pathways via food consumption. The dominating exposure pathways are for most of the nuclides consumption of fish and water. For Ag-isotopes other exposure pathways, such as ground-shine, cereals and meat, are of importance. The results of this study show that individual annual effective doses attributed to unit releases of most of the nuclides to the lake-river system become 1.3-60 times lower than those released to the brackish-water system. The niobium isotopes, however, give a factor 2.5-4.8 higher dose. The reason to that is that the values of the bioaccumulation factor for these isotopes are higher in fresh water than in marine water. An uncertainty analysis is performed on each ecosystem and the results are obtained in the form of distributions. 38 refs, 29 tabs.

  1. A comparison of individual doses for continuous annual unit releases of tritium and activation products into brackish water and lake-river ecosystems

    International Nuclear Information System (INIS)

    Edlund, O.; Aquilonius, K.

    1995-01-01

    The annual effective doses to critical group from potential unit releases of tritium and activation products (32 nuclides) from a hypothetical fusion reactor into two aquatic environments, one with brackish water and the other with fresh water, are assessed. Unit continuous releases (1 Bq/year during 50 years) for each relevant activation product are analyzed, and the effective dose rate is calculated for each nuclide. The transfer of released activity is simulated by compartment models using first-order linear differential equations for the transport. The rate constants for the brackish-water ecosystem are based on measurements. Four exposure pathways are considered in the brackish water system, the Tvaeren Bay, (a) consumption of fish, (b) consumption of milk, (c) consumption of meat, and (d) exposure from swimming. For the freshwater system, five additional pathways are considered, namely consumption of e) water, f) vegetables, g) cereals, and h) root vegetables and i) external exposure from contaminated ground. The paper presents the compartment models used and a description of how the exposure pathways are treated, especially the pathways via food consumption. The dominating exposure pathways are for most of the nuclides consumption of fish and water. For Ag-isotopes other exposure pathways, such as ground-shine, cereals and meat, are of importance. The results of this study show that individual annual effective doses attributed to unit releases of most of the nuclides to the lake-river system become 1.3-60 times lower than those released to the brackish-water system. The niobium isotopes, however, give a factor 2.5-4.8 higher dose. The reason to that is that the values of the bioaccumulation factor for these isotopes are higher in fresh water than in marine water. An uncertainty analysis is performed on each ecosystem and the results are obtained in the form of distributions. 38 refs, 29 tabs

  2. A comparison of individual doses for continuous annual unit releases of tritium and activation products into brackish water and lake-river ecosystems

    Energy Technology Data Exchange (ETDEWEB)

    Edlund, O; Aquilonius, K

    1996-12-31

    The annual effective doses to critical group from potential unit releases of tritium and activation products (32 nuclides) from a hypothetical fusion reactor into two aquatic environments, one with brackish water and the other with fresh water, are assessed. Unit continuous releases (1 Bq/year during 50 years) for each relevant activation product are analyzed, and the effective dose rate is calculated for each nuclide. The transfer of released activity is simulated by compartment models using first-order linear differential equations for the transport. The rate constants for the brackish-water ecosystem are based on measurements. Four exposure pathways are considered in the brackish water system, the Tvaeren Bay, (a) consumption of fish, (b) consumption of milk, (c) consumption of meat, and (d) exposure from swimming. For the freshwater system, five additional pathways are considered, namely consumption of (e) water, (f) vegetables, (g) cereals, and (h) root vegetables and (i) external exposure from contaminated ground. The paper presents the compartment models used and a description of how the exposure pathways are treated, especially the pathways via food consumption. The dominating exposure pathways are for most of the nuclides consumption of fish and water. For Ag-isotopes other exposure pathways, such as ground-shine, cereals and meat, are of importance. The results of this study show that individual annual effective doses attributed to unit releases of most of the nuclides to the lake-river system become 1.3-60 times lower than those released to the brackish-water system. The niobium isotopes, however, give a factor 2.5-4.8 higher dose. The reason to that is that the values of the bioaccumulation factor for these isotopes are higher in fresh water than in marine water. An uncertainty analysis is performed on each ecosystem and the results are obtained in the form of distributions. 38 refs, 29 tabs.

  3. Tritium migration in nuclear desalination plants

    International Nuclear Information System (INIS)

    Muralev, E.D.

    2003-01-01

    Tritium transport, as one of important items of radiation safety assessment, should be taken into consideration before construction of a Nuclear Desalination Plant (NDP). The influence of tritium internal exposition to the human body is very dangerous because of 3 H associations with water molecules. The problem of tritium in nuclear engineering is connected to its high penetration ability (through fuel element cans and other construction materials of a reactor), with the difficulty of extracting tritium from process liquids and gases. Sources of tritium generation in NDP are: nuclear fuel, boron in control rods, and deuterium in heat carrier. Tritium passes easily through the walls of a reactor vessel, intermediate heat exchangers, steam generators and other technological equipment, through the walls of heat carrier pipelines. The release of tritium and its transport could be assessed, using mathematical models, based on the assumption that steady state equilibrium has been attained between the sources of tritium, produced water and release to the environment. Analysis of the model shows the tritium concentration dependence in potable water on design features of NDP. The calculations obtained and analysis results for NDP with BN-350 reactor give good convergence. According to the available data, tritium concentration in potable water is less than the statutory maximum concentration limit. The design of a NDP requires elaboration of technical solutions, capable of minimising the release of tritium to potable water produced. (author)

  4. Nitric oxide-induced calcium release: activation of type 1 ryanodine receptor by endogenous nitric oxide.

    Science.gov (United States)

    Kakizawa, Sho; Yamazawa, Toshiko; Iino, Masamitsu

    2013-01-01

    Ryanodine receptors (RyRs), located in the sarcoplasmic/endoplasmic reticulum (SR/ER) membrane, are required for intracellular Ca2+ release that is involved in a wide range of cellular functions. In addition to Ca2+-induced Ca2+ release in cardiac cells and voltage-induced Ca2+ release in skeletal muscle cells, we recently identified another mode of intracellular Ca2+ mobilization mediated by RyR, i.e., nitric oxide-induced Ca2+ release (NICR), in cerebellar Purkinje cells. NICR is evoked by neuronal activity, is dependent on S-nitrosylation of type 1 RyR (RyR1) and is involved in the induction of long-term potentiation (LTP) of cerebellar synapses. In this addendum, we examined whether peroxynitrite, which is produced by the reaction of nitric oxide with superoxide, may also have an effect on the Ca2+ release via RyR1 and the cerebellar LTP. We found that scavengers of peroxynitrite have no significant effect either on the Ca2+ release via RyR1 or on the cerebellar LTP. We also found that an application of a high concentration of peroxynitrite does not reproduce neuronal activity-dependent Ca2+ release in Purkinje cells. These results support that NICR is induced by endogenous nitric oxide produced by neuronal activity through S-nitrosylation of RyR1.

  5. Tritium practices past and present

    International Nuclear Information System (INIS)

    Gede, V.P.; Gildea, P.D.

    1980-01-01

    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

  6. Study on tritium recovery from breeder materials

    International Nuclear Information System (INIS)

    Moriyama, H.; Moritani, K.

    1997-01-01

    For the development of fusion reactor blanket systems, some of the key issues on the tritium recovery performance of solid and liquid breeder materials were studied. In the case of solid breeder materials, a special attention was focussed on the effects of irradiation on the tritium recovery performance, and tritium release experiments, luminescence measurements of irradiation defects and modeling studies were systematically performed. For liquid breeder materials, tritium recovery experiments from molten salt and liquid lithium were performed, and the technical feasibility of tritium recovery methods was discussed. (author)

  7. 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

    International Nuclear Information System (INIS)

    Little, C.A.; Cotter, S.J.

    1980-01-01

    Concentrations of tritium, 90 Sr, 106 Ru, and 137 Cs 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 90 Sr. The largest dose delivered by 90 Sr was to the bone of the subject individual or community

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

    International Nuclear Information System (INIS)

    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 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 3 ) and, if possible, actual facility buildings

  9. Separation of tritium from reprocessing effluents

    International Nuclear Information System (INIS)

    Bruggeman, A.; Doyen, W.; Harnie, R.; Leysen, R.; Meynendonckx, L.; Monsecour, M.; Goossens, W.R.A.; Baetsle, L.H.

    1980-01-01

    For several years tritium retention has been studied at the Belgian Nuclear Research Centre, SCK/CEN; initially attention was focused on the removal of tritium from gaseous reprocessing effluents. If tritium can be released from the spent fuel into the gaseous phase before any aqueous operation, adsorption on molecular sieves after some isotopic dilution with hydrogen and after complete conversion to (tritiated) water is the most practical collection method. A once-through 15 m 3 .h -1 oxidation-adsorption unit with a closed regeneration system and with a decontamination factor of 1000 at total (tritiated) hydrogen and water inlet concentrations down to 1000 vpm (parts per million by volume) has been constructed and tested at SCK/CEN and it is described in the text. If no special head-end treatment is used an appropriate liquid management inside the reprocessing plant restricts the volume of tritiated aqueous effluents to about 3 m 3 per tonne of LWR fuel processed. However, for further reduction an isotope separation process becomes necessary. SCK/CEN is developing the ELEX process, which is a combination of water ELectrolysis and tritium EXchange between hydrogen and water, the exchange being promoted by a hydrophobic catalyst. For electrolysis under normal conditions an elementary tritium separation factor of 11.6 with a standard deviation of 6% was obtained. As concerns the exchange step a hydrophobic catalyst has been developed which yields for the flow rates used at atmospheric pressure and at 20 0 C an overall exchange rate constant of 9 mol.s -1 .m -3 in a countercurrent trickle-bed reactor. At present an integrated bench scale de-tritiation unit is being built for further tests and for a dynamic demonstration of the ELEX process

  10. Tritium monitoring at the Sandia Tritium Research Laboratory

    International Nuclear Information System (INIS)

    Devlin, T.K.

    1978-10-01

    Sandia Laboratories at Livermore, California, is presently beginning operation of a Tritium Research Laboratory (TRL). The laboratory incorporates containment and cleanup facilities such that any unscheduled tritium release is captured rather than vented to the atmosphere. A sophisticated tritium monitoring system is in use at the TRL to protect operating personnel and the environment, as well as ensure the safe and effective operation of the TRL decontamination systems. Each monitoring system has, in addition to a local display, a display in a centralized control room which, when coupled room which, when coupled with the TRL control computer, automatically provides an immediate assessment of the status of the entire facility. The computer controls a complex alarm array status of the entire facility. The computer controls a complex alarm array and integrates and records all operational and unscheduled tritium releases

  11. Tritium autoradiography

    International Nuclear Information System (INIS)

    Caskey, G.R. Jr.

    1981-01-01

    Hydrogen distribution and diffusion within many materials may be investigated by autoradiography if the radioactive isotope tritium is used in the study. Tritium is unstable and decays to helium-3 by emission of a low energy (18 keV) beta particle which may be detected photographically. The basic principles of tritium autoradiography will be discussed. Limitations are imposed on the technique by: (1) the low energy of the beta particles; (2) the solubility and diffusivity of hydrogen in materials; and (3) the response of the photographic emulsion to beta particles. These factors control the possible range of application of tritium autoradiography. The technique has been applied successfully to studies of hydrogen solubility and distribution in materials and to studies of hydrogen damage

  12. Experimental determination of the kinetic conversion rate of gaseous tritium into HTO

    International Nuclear Information System (INIS)

    Gulden, W.; Guenot, J.; Djerassi, H.; Clerc, H.

    1986-09-01

    To assess the impact of atmospheric tritium releases from a fusion plant to the environment, the dose to man usually is determined under the conservative assumption that all tritium is released in the more dangerous oxide form. To quantify this overprediction, experiments are presently being performed by CEA. Oxidation of tritium gas by soil and subsequent resuspension of HTO from soil to atmosphere is one of them. First results have been obtained by CEA on the kinetics of HTO resuspension from contaminated soils. Immediately after contamination, the fraction of the deposited activity which is resuspended to atmosphere is in the range of 1% to 5% per hour. This resuspension rate then decreases more or less slowly with time, depending on specific conditions

  13. Review of tritium behavior in HTGR systems

    International Nuclear Information System (INIS)

    Gainey, B.W.

    1976-01-01

    The available experimental evidence from laboratory and reactor studies pertaining to tritium production, capture, release, and transport within an HTGR leading to release to the environment is reviewed. Possible mechanisms for release, capture, and transport are considered and a simple model was used to calculate the expected tritium release from HTGRs. Comparison with Federal regulations governing tritium release confirm that expected HTGR releases will be well within the allowable release limits. Releases from HTGRs are expected to be somewhat less than from LWRs based on the published LWR operating data. Areas of research deserving further study are defined but it is concluded that a tritium surveillance at Fort St. Vrain is the most immediate need

  14. 51Cr release and oxidative stress in the lens

    International Nuclear Information System (INIS)

    Stewart-DeHaan, P.J.; Sanwal, M.; Creighton, M.O.; Inch, W.R.; Trevithick, J.R.

    1989-01-01

    Examination of the opaque areas of human cortical cataracts has shown that a large portion of the opacity could be attributed to the globules found there. We tested models involving globule formation as a result of oxidative damage to rat lens cells in culture and whole chick embryo lenses. When cell monolayers from a lens cell line were exposed to oxidizing conditions they developed globules on the cell surface. The cells were protected from damage by the addition of glutathione and vitamin C. Thirteen-day chick embryo lenses were also incubated in oxidizing conditions and the amount of cellular damage was assessed using a chromium-51 release assay we have developed. After 24 hr the percent 51Cr in the medium increased by an average of 20% as a result of 10 mM hydrogen peroxide treatment. The addition of the 10 mM vitamin C to the hydrogen peroxide significantly reduced the 51Cr leakage to the control level. Light microscopy of sections of the lens showed a breakdown of the equatorial fibre arrangement in the presence of H2O2, while addition of vitamin C restored the fibre organization to almost normal. The findings suggest that oxidative stress is an important step in cataractogenesis and point towards the use of water soluble antioxidants as protective agents

  15. Study of Oxidizing Agents for Tritium Removal in ITER -Compatible Conditions: Alternatives to Oxygen and Ozone

    International Nuclear Information System (INIS)

    Tabares, F. L.; Tafalla, D.; Ferreira, J. A.; Gomez-Aleixandre, C.; Maria Albella, J.; Soria, J.; Rodriguez-Ramos, I.

    2007-01-01

    In the present report, the studies of tritiated carbon-film removal by oxidizing agents other than Oxygen and Ozone in ITER are described. Exposure of laboratory produced a-C:H/D films and tokamak flakes (Asdex Upgrade and Textor) to nitric oxide, water and hydrogen peroxide has been carried out. Temperatures of exposure up to 350 degree centigree were used, and thermal desorption of the samples at temperatures up to 750 degree centigree was performed for sample characterization prior to and after the treatment. Elastic Recoil Detection Analysis (ERDA), Infrared Spectroscopy, XPS and Nano indentation hardness analysis were applied to the characterization of the physical and chemical changes of the samples. This work was done under the EFDA Task 04-1175. (Author) 8 refs

  16. Study of Oxidizing Agents for Tritium Removal in ITER -Compatible Conditions: Alternatives to Oxygen and Ozone

    Energy Technology Data Exchange (ETDEWEB)

    Tabares, F. L.; Tafalla, D.; Ferreira, J. A.; Gomez-Aleixandre, C.; Maria Albella, J.; Soria, J.; Rodriguez-Ramos, I.

    2007-07-20

    In the present report, the studies of tritiated carbon-film removal by oxidizing agents other than Oxygen and Ozone in ITER are described. Exposure of laboratory produced a-C:H/D films and tokamak flakes (Asdex Upgrade and Textor) to nitric oxide, water and hydrogen peroxide has been carried out. Temperatures of exposure up to 350 degree centigree were used, and thermal desorption of the samples at temperatures up to 750 degree centigree was performed for sample characterization prior to and after the treatment. Elastic Recoil Detection Analysis (ERDA), Infrared Spectroscopy, XPS and Nano indentation hardness analysis were applied to the characterization of the physical and chemical changes of the samples. This work was done under the EFDA Task 04-1175. (Author) 8 refs.

  17. New arrangement for the air cleanup system to recover tritium

    International Nuclear Information System (INIS)

    Nishikawa, Masabumi; Takahashi, Kohsaku; Munakata, Kenzo; Fukada, Satoshi; Kotoh, Kenji; Takeishi, Toshiharu

    1997-01-01

    At present, the standard arrangement of the air cleanup system responsible for emergency tritium recovery from room air is a catalytic oxidation bed with a heater followed by an adsorption bed with a cooler. One disadvantage of this arrangement is that trouble with the heater or the cooler could result in a loss of capacity to recover tritium. Another disadvantage of the catalyst-adsorption-bed arrangement is that tritiated water must be recovered with a high decontamination factor after dilution with a large amount of water vapor in the working atmosphere. The performance of a new arrangement for the air cleanup system, which consists of a precious metal catalyst bed preceded by an adsorption bed without heating equipment, is discussed. According to calculations, most of the tritium released to the room air is recovered in the catalyst bed through oxidation, adsorption, and isotope exchange reaction when the new arrangement is applied. The adsorption bed placed before the catalyst bed dehumidifies the process gas to such a degree that the oxidation reaction of tritium in the catalyst bed is not hindered by water vapor. 15 refs., 6 figs., 6 tabs

  18. A critical analysis of the impact assessment of environmental tritium

    International Nuclear Information System (INIS)

    Jain, Narendra; Bhatia, Arvind

    2013-01-01

    Tritium, a radionuclide of hydrogen has longer life and gets more rapidly dispersed, but before becoming globally distributed, it represents a significant radiobiological risk to the local population exposed. It is produced naturally in the upper atmosphere by the interaction of cosmic rays with nitrogen and hydrogen. The tritons in the upper atmosphere are oxidized to tritiated water (HTO) and mix with the hydrosphere generally through the movement of air masses and precipitation. Terrestrially, tritium may be formed by the action of lithium on neutrons. There is apprehension that the recent controversy concerning the health and environmental impact of tritium may end up as worldwide contaminants in the final analysis. From many varied reports from different laboratories, it appears that projected levels for fusion reactors may also produce deleterious and detectable effects. The degree of concern over tritium problem is evidenced by a rapid increase in publications on the health implications of environmental tritium. The present issues of controversy will be intensified as the fusion reactor technology approaches the door step of public and the possible health detriment from its radioactive emissions arouse concern. The current project has been planned keeping some such points in view. It reviews the work on the behavior of tritium in its various forms in the environment with an emphasis on the release from various sources, its world inventories at present level sand its transfer into the various compartments of ecosystems. Besides this, its metabolism in biosystem and the possible implications of low doses of tritium in present and future generations have also been discussed. (author)

  19. Antimicrobial Activity of Nitric Oxide-Releasing Ti-6Al-4V Metal Oxide

    Science.gov (United States)

    Reger, Nina A.; Meng, Wilson S.; Gawalt, Ellen S.

    2017-01-01

    Titanium and titanium alloy materials are commonly used in joint replacements, due to the high strength of the materials. Pathogenic microorganisms can easily adhere to the surface of the metal implant, leading to an increased potential for implant failure. The surface of a titanium-aluminum-vanadium (Ti-6Al-4V) metal oxide implant material was functionalized to deliver an small antibacterial molecule, nitric oxide. S-nitroso-penicillamine, a S-nitrosothiol nitric oxide donor, was covalently immobilized on the metal oxide surface using self-assembled monolayers. Infrared spectroscopy was used to confirm the attachment of the S-nitrosothiol donor to the Ti-Al-4V surface. Attachment of S-nitroso-penicillamine resulted in a nitric oxide (NO) release of 89.6 ± 4.8 nmol/cm2 under physiological conditions. This low concentration of nitric oxide reduced Escherichia coli and Staphylococcus epidermidis growth by 41.5 ± 1.2% and 25.3 ± 0.6%, respectively. Combining the S-nitrosothiol releasing Ti-6Al-4V with tetracycline, a commonly-prescribed antibiotic, increased the effectiveness of the antibiotic by 35.4 ± 1.3%, which allows for lower doses of antibiotics to be used. A synergistic effect of ampicillin with S-nitroso-penicillamine-modified Ti-6Al-4V against S. epidermidis was not observed. The functionalized Ti-6Al-4V surface was not cytotoxic to mouse fibroblasts. PMID:28635681

  20. Tritium sample analyses in the Savannah River and associated waterways following the K-reactor release of December 1991

    International Nuclear Information System (INIS)

    Beals, D.M.; Dunn, D.L.; Hall, G.; Kantelo, M.V.

    1992-01-01

    An unplanned release of tritiated water occurred at K reactor on SRS between 22-December and 25-December 1991. This water moved down through the effluent canal, Pen Branch, Steel Creek and finally to the Savannah River. Samples were collected in the Savannah River and associated waterways over a period of a month. The Environmental Technology Section (ETS) of the Savannah River Laboratory performed liquid scintillation analyses to monitor the passage of the tritiated water from SRS to the Atlantic Ocean

  1. Tritium environmental transport studies at TFTR

    International Nuclear Information System (INIS)

    Ritter, P.D.; Dolan, T.J.; Longhurst, G.R.

    1993-01-01

    Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a weak after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER)

  2. Tritium environmental transport studies at TFTR

    Science.gov (United States)

    Ritter, P. D.; Dolan, T. J.; Longhurst, G. R.

    1993-06-01

    Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a week after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER).

  3. Experiments on tritium behavior in beryllium, (2)

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Nakata, Hirokatsu; Sugai, Hiroyuki; Tanase, Masakazu.

    1990-02-01

    Beryllium has been used as the neutron reflector of material testing reactor and as the neutron multiplier for the fusion reactor lately. To study the tritium behavior in beryllium, we conducted the experiments, i.e., tritium release by recoil or diffusion by using the hot-pressed beryllium which had been produced both tritium and helium by neutron irradiation. From our experiments, we found that (1) amount of tritium production per one cycle irradiation (lasting 22 days) of JMTR is 10 mCi/g, (2) amount of tritium per surface area of hot-pressed beryllium released by recoil is 4 μCi/cm 2 , (3) diffusion coefficient of tritium in a temperature range of 800 ∼1180degC can be expressed with the following equation; D = 8.7 x 10 4 exp(-2.9x10 5 /R/T) cm 2 /s. (author)

  4. Tritium retention in S-65 beryllium after 100 eV plasma exposure

    Energy Technology Data Exchange (ETDEWEB)

    Causey, R.A. [Sandia National Labs., Livermore, CA (United States); Longhurst, G.R. [Idaho National Engineering Laboratories, Idaho Falls, 83415 (United States); Harbin, W. [Los Alamos National Laboratories, Los Alamos, NM 87545 (United States)

    1997-02-01

    The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 x 10{sup 17} (D+T)/cm{sup 2} s up to about 3 x 10{sup 18} (D+T)/cm{sup 2} s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H{sub 2} gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 x 10{sup 17} (D+T)/cm{sup 2} at 373 K to a low of 1 x 10{sup 16} (D+T)/cm{sup 2} at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C=0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters. (orig.).

  5. Tritium retention in S-65 beryllium after 100 eV plasma exposure

    Science.gov (United States)

    Causey, Rion A.; Longhurst, Glen R.; Harbin, Wally

    1997-02-01

    The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 × 10 17 ( D + T)/ cm2s up to about 3 × 10 18 ( D + T)/ cm2s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H 2 gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 × 10 17 ( D + T)/ cm2 at 373 K to a low of 1 × 10 16 ( D + T)/ cm2 at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C = 0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters.

  6. Tritium processing in JT-60U

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Masaki, Kei

    1997-01-01

    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 10 19 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/cm 3 for radiation work permit requirements. (author)

  7. EXOTIC: Development of ceramic tritium breeding materials

    International Nuclear Information System (INIS)

    Flipot, A.J.; Kennedy, P.; Conrad, R.

    1989-03-01

    As part of the joint European Programme on fusion blanket technology three laboratories, Northern Research Laboratories (NRL), Springfields in the UK, SCK/CEN-Mol in Belgium and ECN-Petten in conjunction with JRC-Petten in the Netherlands have worked closely together since 1983 on the development of ceramic breeder materials, the programme being codenamed EXOTIC. Lithium oxides, aluminates, silicates and zirconates have been produced, characterised and irradiated in the HFR-Petten in experiments EXOTIC-1, -2 and -3. EXOTIC-4 is in preparation. In this fourth annual progress report the work achieved in 1987 is reported. For EXOTIC-1 to -3 mainly post irradiation examinations have been carried out like: visual inspection, puncturing of closed capsules, tritium retention measurements and material characterisation. Moreover, tritium release experiments on small specimens have started. SCK/CEN performed a general study on lithium silicates, in particular on the thermal stability. Finally, the fabrication and the characterisation of the materials to be irradiated in experiment EXOTIC-4 are presented. The eight capsules of EXOTIC-4 will be loaed with samples of Li 2 SiO 3 , Li 2 O, Li 2 ZrO 3 , Li 6 Zr 2 O 7 and Li 8 ZrO 6 . The irradiation will last 4 reactor cycles or about 100, Full Power Day, FPD. The main objective is to determine the tritium residence time of the various lithium zirconates. 18 figs., 8 refs., 15 tabs

  8. Tritium retention in candidate next-step protection materials: engineering key issues and research requirements

    International Nuclear Information System (INIS)

    Federici, G.; Andrew, P.L.; Wu, C.H.

    1995-01-01

    Although a considerable volume of valuable data on the behaviour of tritium in beryllium and carbon-based armours exposed to hydrogenic fusion plasmas has been compiled over the past years both from operation of present-day tokamaks and from laboratory simulations, knowledge is far from complete and tritium inventory predictions for these materials remain highly uncertain. In this paper we elucidate the main mechanisms responsible for tritium trapping and release in next-step D-T tokamaks, as well as the applicability of some of the presently known data bases for design purposes. Owing to their strong anticipated implications on tritium uptake and release, attention is focused mainly on the interaction of tritium with neutron damage induced defects, on tritium codeposition with eroded carbon and on the effects of oxide and surface contaminants. Some preliminary quantitative estimates are presented based on most recent experimental findings and latest modelling developments as well. The influence of important working conditions such as target temperature, loading particle fluxes, erosion and redeposition rates, as well as material characteristics such as the type of morphology of the protection material (i.e. amorphous plasma-sprayed beryllium vs. solid forms), and design dependent parameters are discussed in this paper. Remaining issues which require additional effort are identified. (orig.)

  9. Tritium immobilisation

    International Nuclear Information System (INIS)

    Bridger, N.J.

    1982-01-01

    Tritium is immobilised for long term storage by absorption in a hydridable/tritidable material, such as zirconium. A gas permeable container is packed with the material in the form of sponge fragments, rods or tubes, and a gaseous mixture of hydrogen and tritium introduced into the container whilst the container is at a temperature of about 600 deg C or above. Thermal expansion of the material during reaction with the gaseous mixture compacts the material into a coherent body in the container relatively free from finely divided hydride/ tritide material. (author)

  10. Antibacterial activity of nitric oxide releasing silver nanoparticles

    Science.gov (United States)

    Seabra, Amedea B.; Manosalva, Nixson; de Araujo Lima, Bruna; Pelegrino, Milena T.; Brocchi, Marcelo; Rubilar, Olga; Duran, Nelson

    2017-06-01

    Silver nanoparticles (AgNPs) are well known potent antimicrobial agents. Similarly, the free radical nitric oxide (NO) has important antibacterial activity, and due to its instability, the combination of NO and nanomaterials has been applied in several biomedical applications. The aim of this work was to synthesize, characterize and evaluate the antibacterial activity of a new NO-releasing AgNPs. Herein, AgNPs were synthesized by the reduction of silver ions (Ag+) by catechin, a natural polyphenol and potent antioxidant agent, derived from green tea extract. Catechin acts as a reducing agent and as a capping molecule on the surface of AgNPs, minimizing particle agglomeration. The as-synthesized nanoparticles were characterized by different techniques. The results showed the formation of AgNPs with average hydrodynamic size of 44 nm, polydispersity index of 0.21, and zeta potential of -35.9 mV. X-ray diffraction and Fourier transform infrared spectroscopy revealed the presence of the AgNP core and cathecin as capping agent. The low molecular weight mercaptosuccinic acid (MSA), which contain free thiol group, was added on the surface of catechin-AgNPs, leading to the formation of MSA-catechin-AgNPs (the NO precursor nanoparticle). Free thiol groups of MSA-catechin-AgNPs were nitrosated leading to the formation of S-nitroso-mercaptosuccinic acid (S-nitroso-MSA), the NO donor. The amount of 342 ± 16 µmol of NO was released per gram of S-nitroso-MSA-catechin-AgNPs. The antibacterial activities of catechin-AgNPs, MSA-catechin-AgNPs, and S-nitroso-MSA-catechin-AgNPs were evaluated towards different resistant bacterial strains. The results demonstrated an enhanced antibacterial activity of the NO-releasing AgNP. For instance, the minimal inhibitory concentration values for Pseudomonas aeruginosa (ATCC 27853) incubated with AgNPs-catechin, AgNPs-catechin-MSA, and AgNPs-catechin-S-nitroso-MSA were found to be 62, 125 and 3 µg/mL, respectively. While in the case of

  11. Emergency response monitoring activities and environmental impact of the K-Reactor aqueous tritium release of December 1991

    International Nuclear Information System (INIS)

    Hamby, D.M.; Addis, R.P.; Beals, D.M.; Cadieux, J.R.; Carlton, W.H.; Dunn, D.L.; Hall, G.; Hayes, D.W.; Lorenz, R.; Kantelo, M.V.; Taylor, R.W.

    1992-01-01

    Approximately 150 gallons of tritiated water leaked from one of the K-Reactor heat exchangers between December 22 and December 25, 1991. Upon notification, the Environmental Technology Section (ETS) activated its emergency response team to provide predictions of river concentrations, transport times, and radiological effects to downstream water users. Additionally, within a few days of the release, ETS and the Environmental Monitoring Section (EMS) began a comprehensive program to collect and analyze surface water samples from SRS down to the Savannah River estuary. The TRAC mobile laboratory was deployed to the Beaufort-Jasper water treatment plant to provide initial analyses for downriver water samples. This document discusses the results of the sampling activities. Concentration levels are provided along with hypothetical maximum individual doses

  12. Tritium management for fusion reactors

    International Nuclear Information System (INIS)

    Rouyer, J.L.; Djerassi, H.

    1985-01-01

    To determine a waste management strategy, one has to identify first the wastes (quantities, activities, etc.), then to define options, and to compare these options by appropriate criteria and evaluations. Two European Associations are working together, i.e., Studsvik and CEA, on waste treatment and tritium problems. A contribution to fusion specific tritiated waste management strategy is presented. It is demonstrated that the best strategy is to retain tritium (outgas and recover, or immobilize it) so that residual tritium releases are kept to a minimum. For that, wastes are identified, actual regulations are described and judged inadequate without amendments for fusion problems. Appropriate criteria are defined. Options for treatment and disposal of tritiated wastes are proposed and evaluated. A tritium recovery solution is described

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

    International Nuclear Information System (INIS)

    Draghia, M.; Pasca, G.; Porcariu, F.

    2015-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  15. Magnetothermal release of payload from iron oxide/silica drug delivery agents

    Energy Technology Data Exchange (ETDEWEB)

    Luong, T.T., E-mail: thientai.luong@chem.kuleuven.be [KU Leuven, Department of Chemistry, Celestijnenlaan 200D, Heverlee 3001 (Belgium); Hanoi National University of Education, Faculty of Chemistry, Xuan Thuy 136, Cau Giay, Hanoi (Viet Nam); Knoppe, S.; Bloemen, M.; Brullot, W.; Strobbe, R. [KU Leuven, Department of Chemistry, Celestijnenlaan 200D, Heverlee 3001 (Belgium); Locquet, J.-P. [KU Leuven, Department of Physics, Celestijnenlaan 200D, Heverlee 3001 (Belgium); Verbiest, T. [KU Leuven, Department of Chemistry, Celestijnenlaan 200D, Heverlee 3001 (Belgium)

    2016-10-15

    The release of covalently bound Rhodamine B from iron oxide/mesoporous silica core/shell nanoparticles under magnetically induced heating was studied. The system acts as a model to study drug delivery and payload release under magnetothermal heating. - Graphical abstract: The release of covalently bound Rhodamine B from iron oxide/mesoporous silica core/shell nanoparticles under magnetically induced heating was studied. - Highlights: • Iron oxide/mesoporous-SiO{sub 2} core-shell NPs were synthesized. • The dye was covalently bound to SiO{sub 2} shells. • The release of dye under magnetothermal heating was studied. • The results are relevant for controlled drug release.

  16. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  17. Design options to minimize tritium inventories at Savannah River

    Energy Technology Data Exchange (ETDEWEB)

    Klein, J.E., E-mail: james.klein@srnl.doe.gov; Wilson, J.; Heroux, K.J.; Poore, A.S.; Babineau, D.W.

    2016-11-01

    Highlights: • La-Ni-Al alloys are used as tritium storage materials and retain He-3. • La-Ni-Al He-3 effects decrease useable process tritium inventory. • Use of Pd or depleted uranium beds decreases process tritium inventories. • Reduced inventory tritium facilities will lower public risk. - Abstract: Large quantities of tritium are stored and processed at the Savannah River Site (SRS) Tritium Facilities. In many design basis accidents (DBAs), it is assumed the entire tritium inventory of the in-process vessels are released from the facility and the site for inclusion in public radiological dose calculations. Pending changes in public dose calculation methodologies are driving the need for smaller in-process tritium inventories to be released during DBAs. Reducing the in-process tritium inventory will reduce the unmitigated source term for public dose calculations and will also reduce the production demand for a lower inventory process. This paper discusses process design options to reduce in-process tritium inventories. A Baseline process is defined to illustrate the impact of removing or replacing La-Ni-Al alloy tritium storage beds with palladium (Pd) or depleted uranium (DU) storage beds on facility in-process tritium inventories. Elimination of La-Ni-Al alloy tritium storage beds can reduce in-process tritium inventories by over 1.5 kg, but alternate process technologies may needed to replace some functions of the removed beds.

  18. Design options to minimize tritium inventories at Savannah River

    International Nuclear Information System (INIS)

    Klein, J.E.; Wilson, J.; Heroux, K.J.; Poore, A.S.; Babineau, D.W.

    2016-01-01

    Highlights: • La-Ni-Al alloys are used as tritium storage materials and retain He-3. • La-Ni-Al He-3 effects decrease useable process tritium inventory. • Use of Pd or depleted uranium beds decreases process tritium inventories. • Reduced inventory tritium facilities will lower public risk. - Abstract: Large quantities of tritium are stored and processed at the Savannah River Site (SRS) Tritium Facilities. In many design basis accidents (DBAs), it is assumed the entire tritium inventory of the in-process vessels are released from the facility and the site for inclusion in public radiological dose calculations. Pending changes in public dose calculation methodologies are driving the need for smaller in-process tritium inventories to be released during DBAs. Reducing the in-process tritium inventory will reduce the unmitigated source term for public dose calculations and will also reduce the production demand for a lower inventory process. This paper discusses process design options to reduce in-process tritium inventories. A Baseline process is defined to illustrate the impact of removing or replacing La-Ni-Al alloy tritium storage beds with palladium (Pd) or depleted uranium (DU) storage beds on facility in-process tritium inventories. Elimination of La-Ni-Al alloy tritium storage beds can reduce in-process tritium inventories by over 1.5 kg, but alternate process technologies may needed to replace some functions of the removed beds.

  19. Behaviour of tritium in the environment

    International Nuclear Information System (INIS)

    1979-01-01

    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

  20. Tritium migration studies at the Nevada Test Site

    International Nuclear Information System (INIS)

    Schulz, R.K.; Weaver, M.O.

    1993-05-01

    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

  1. Tritium behavior in the Caisson, a simulated fusion reactor room

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Yamada, Masayuki; Suzuki, Takumi; O'hira, Shigeru; Nakamura, Hirofumi; Shu, Weimin; Yamanishi, Toshihiko; Kawamura, Yoshinori; Isobe, Kanetsugu; Konishi, Satoshi; Nishi, Masataka

    2000-01-01

    In order to confirm tritium confinement ability in the deuterium-tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called 'Caisson', at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak-tight vessel of 12 m 3 , simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code

  2. Tritium Permeability of Incoloy 800H and Inconel 617

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Tritium Permeability of Incoloy 800H and Inconel 617

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

    Skinner, C.H.; Coad, J.P.; Federici, G.

    2003-01-01

    Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating

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

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

    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.)

  6. Universal tritium transmitter

    International Nuclear Information System (INIS)

    Cordaro, J. V.; Wood, M.

    2008-01-01

    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

  7. Radioecological behaviour of elementary tritium, especially dry deposition and its dependence on soil porosity

    International Nuclear Information System (INIS)

    Foerstel, H.

    1997-01-01

    The inventory of fusion reactors mainly consists of deuterium and tritium. The amount of tritium of each reactor is equal to the natural inventory of the earth's atmo- and hydrosphere. Elementary tritium (HT) itself is not dangerous to man, for it is hardly dissolved in water, that is neither taken up by human tissues nor metabolized anywhere in our body. In contrast to HT the tritiated water HTO quickly exchanges with any wet surface and with the humidity of air. After an accidental release into the atmosphere the main pathway of intake into the human body is as HTO via the lung; its surface is comparable to a soccer playground. HT released into air will be quickly oxidised within the upper centimetres of the soil when the plume touches the ground. Each soil tested by us until now had oxidized HT, that had shown hydrogenase activity. Neither the biological function nor the catalytic system wee identified yet. The hypothesis of a correlation between hydrogenase activity and soil nitrogen fixation could not be confirmed (nitrogen fixation shows a leakage of hydrogen): nitrogen fixing plants (nodules) do not oxidize HT. The presentation will summarize ten years of work in the laboratory and in the field. A concise picture of the radioecological behaviour of elementary tritium after an accidental release could be obtained. The work was partly done as cooperation within the frame of the EU or within the International Union of Radioecology

  8. Analysis of the organically bound tritium

    International Nuclear Information System (INIS)

    Baglan, N.; Alanic, G.

    2011-01-01

    In environmental samples, tritium is very often combined with the fraction of bulk water accumulated in the sample but also in the form of organically bound tritium. When the tritium is organically bound, 2 forms can coexist: the exchangeable fraction and the non-exchangeable fraction. The analysis of the different forms of tritium present in the sample is necessary to assess the sanitary hazards due to tritium. The total tritium is obtained from the analysis of the water released when the fresh sample is burnt while the organically bound tritium is obtained from the analysis of the water released when the dry extract of the sample is burnt. The measurement of the exchangeable fraction and the non-exchangeable fraction requires an additional stage of labile exchange. The exchangeable fraction is determined from the analysis of the water released during the labile exchange and the non-exchangeable fraction is determined from the water released during the combustion of the dry extract of the labile exchange

  9. Tritium Room Air Monitor Operating Experience Review

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader; B. J. Denny

    2008-09-01

    Monitoring the breathing air in tritium facility rooms for airborne tritium is a radiological safety requirement and a best practice for personnel safety. Besides audible alarms for room evacuation, these monitors often send signals for process shutdown, ventilation isolation, and cleanup system actuation to mitigate releases and prevent tritium spread to the environment. Therefore, these monitors are important not only to personnel safety but also to public safety and environmental protection. This paper presents an operating experience review of tritium monitor performance on demand during small (1 mCi to 1 Ci) operational releases, and intentional airborne inroom tritium release tests. The tritium tests provide monitor operation data to allow calculation of a statistical estimate for the reliability of monitors annunciating in actual tritium gas airborne release situations. The data show a failure to operate rate of 3.5E-06/monitor-hr with an upper bound of 4.7E-06, a failure to alarm on demand rate of 1.4E-02/demand with an upper bound of 4.4E-02, and a spurious alarm rate of 0.1 to 0.2/monitor-yr.

  10. Management of Tritium in European Spallation Source

    DEFF Research Database (Denmark)

    Ene, Daniela; Andersson, Kasper Grann; Jensen, Mikael

    2015-01-01

    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...... 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....

  11. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

    Tritium release from the candidate ceramic materials, Li 2 O, LiA10 2 , Li 2 SiO 3 , Li 4 SiO 4 and Li 2 ZrO 3 , is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

  12. Management of Tritium in ITER Waste

    International Nuclear Information System (INIS)

    Rosanvallon, S.; Benchikhoune, M.; Ciattaglia, S.; Uzan, J. Elbez; Na, B. C.; Taylor, N.; Gastaldi, O.

    2011-01-01

    ITER will use tritium as fuel. Procedures and processes are thus put in place in order to recover the tritium that is not used in the fusion reaction, including from waste and effluents. The tritium thus recovered can be re-injected into the fuel cycle. Moreover, tritium content and thus outgassing may be a safety concern, because of the potential for releases to the environment, both from the facility and from the final disposal (subjected to stringent acceptance criteria in the current waste final disposal). The aim of this paper is to present the measures considered to deal with the specific case of tritium in the liquid and solid waste that will arise from ITER operation and decommissioning. It concerns the processes that are considered from the waste production to its final disposal and in particular: the tritium removal stages (in-situ divertor baking at 350 C and tritium removal from solid waste and liquid and gaseous effluents), the removal of dust contamination (dust containing tritium produced by plasma-wall interaction and by the maintenance/ refurbishment processes) and the measures to enable safe processing and storage of the waste (wall-liner in the hot cell facility to limit concrete contamination and interim storage enabling tritium decay for waste that could not be directly accepted in the host-country final disposal facilities). (authors)

  13. Magmatic tritium

    International Nuclear Information System (INIS)

    Goff, F.; Aams, A.I.; McMurtry, G.M.; Shevenell, L.; Pettit, D.R.; Stimac, J.A.; Werner, C.

    1997-01-01

    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 ( 3 H) of deep origin ( 2 O). On the other hand, all volcanoes produce mixtures of meteoric and magmatic fluids that contain measurable 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

  14. Temporal sealing material of tritium-contaminated stainless steel

    International Nuclear Information System (INIS)

    Wen Wei; Dan Guiping; Zhang Dong; Qiu Yongmei; Zhang Li

    2010-01-01

    Tritium can be released from the exterior of tritium-contaminated stainless steel by slight stirring while decontaminating and disassembling. In order to avoid secondary tritium contamination to environment and operators, it is necessary to cover with an effective coating to tritium on the exterior of tritium-contaminated stainless steel and fill an effective substance to tritium inside. The results of tritium sealed experiments show that sealing efficiency of neutral silicone rubber is more than 85% for condition of static state and more than 99% for foam concrete condition of dynamic state. Neutral silicone rubber and foam concrete which have finer sealing efficiency can be used as temporal sealed material for the decontamination and disassembly of tritium-contaminated stainless steel. (authors)

  15. Tritium Removal from JET and TFTR Tiles by a Scanning Laser; TOPICAL

    International Nuclear Information System (INIS)

    C.H. Skinner; N. Bekris; J.P. Coad; C.A. Gentile; M. Glugla

    2002-01-01

    Fast and efficient tritium removal is needed for future D-T machines with carbon plasma-facing components. A novel method for tritium release has been demonstrated on co-deposited layers on tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) and from the Joint European Torus (JET). A scanning continuous wave neodymium laser beam was focused to=100 W/mm2 and scanned at high speed over the co-deposits, heating them to temperatures=2000 C for about 10 ms in either air or argon atmospheres. Fiber optic coupling between the laser and scanner was implemented. Up to 87% of the co-deposited tritium was thermally desorbed from the JET and TFTR samples. This technique appears to be a promising in-situ method for tritium removal in a next-step D-T device as it avoids oxidation, the associated de-conditioning of the plasma-facing surfaces, and the expense of processing large quantities of tritium oxide

  16. Tritium Removal from JET and TFTR Tiles by a Scanning Laser

    International Nuclear Information System (INIS)

    Skinner, C.H.; Bekris, N.; Coad, J.P.; Gentile, C.A.; Glugla, M.

    2002-01-01

    Fast and efficient tritium removal is needed for future D-T machines with carbon plasma-facing components. A novel method for tritium release has been demonstrated on co-deposited layers on tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) and from the Joint European Torus (JET). A scanning continuous wave neodymium laser beam was focused to =100 W/mm2 and scanned at high speed over the co-deposits, heating them to temperatures =2000 C for about 10 ms in either air or argon atmospheres. Fiber optic coupling between the laser and scanner was implemented. Up to 87% of the co-deposited tritium was thermally desorbed from the JET and TFTR samples. This technique appears to be a promising in-situ method for tritium removal in a next-step D-T device as it avoids oxidation, the associated de-conditioning of the plasma-facing surfaces, and the expense of processing large quantities of tritium oxide

  17. Tritium handling facilities at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Anderson, J.L.; Damiano, F.A.; Nasise, J.E.

    1975-01-01

    A new tritium facility, recently activated at the Los Alamos Scientific Laboratory, is described. The facility contains a large drybox, associated gas processing system, a facility for handling tritium gas at pressures to approximately 100 MPa, and an effluent treatment system which removes tritium from all effluents prior to their release to the atmosphere. The system and its various components are discussed in detail with special emphasis given to those aspects which significantly reduce personnel exposures and atmospheric releases. (auth)

  18. Distribution of tritium in a chronically contaminated lake

    International Nuclear Information System (INIS)

    Blaylock, B.G.; Frank, M.L.

    1978-01-01

    White Oak Lake located on the U.S. Department of Energy's Oak Ridge Reservation receives a continuous input of tritium from operating facilities and waste disposal operations at the Oak Ridge National Laboratory. The purpose of this paper was (1) to determine the distribution and concentration of tritium in an aquatic environment which has received releases of tritium significantly greater than expected releases from nuclear power plants, and (2) to determine the effect of fluctuating tritium concentrations in ambient water on the concentration of tritium in fish. Aquatic biota from White Oak Lake were analyzed for tissue water tritium and tissue bound tritium. Except for one plant species, the ratio of tissue water tritium to lake water tritium ranged from 0.80 to 1.02. The tissue water tritium in Gambusia affinis, the mosquito fish, followed closely the significant changes in tritium concentration in lake water. The turnover of tissue water tritium was very rapid; Gambusia from White Oak Lake eliminated 50% of their tissue water tritium in 14 min. The ratio of the specific activity of the tissue bound tritium to the specific activity of the lake water was greatest for the larger species of fish but never exceeded unity. The radiation dose to man from tritium which could be acquired through the aquatic food chain was relatively small when compared to other pathways. The whole body dose to a hypothetical individual taking in concentrations of tritium measured in White Oak Lake was 1.8 mrem/yr from eating fish and 10.0 mrem/yr from drinking water

  19. Low pressure tritium interaction with Inconel 625 and AISI 316 L stainless steel surfaces: an evaluation of the recombination and adsorption constants

    International Nuclear Information System (INIS)

    Perujo, A.; Douglas, K.; Serra, E.

    1996-01-01

    The surface constants for the recombination (σk 2 ) and adsorption (σk 1 ) of tritium in Inconel 625 and austenitic stainless steel AISI 316 L were determined from the measurement of tritium permeation through engineering components (bellows) typical of those used on large fusion devices which will operate with tritium. Experimental permeation measurements were performed over the temperature range 450-620 K and an interpretation of the data was attempted based on a surface-limited tritium release model. At the tritium partial pressure of 0.1 Pa present in a machine such as JET, the flow of tritium is strongly influenced by surface reactions. Furthermore, it is often assumed that oxide layers, acting as permeation barriers, are present on such components. However, for effectiveness, such barriers must be intact and this may not necessarily be the case for engineering components in which mechanical stresses can lead to oxide cracking. The recombination (σk 2 ) and adsorption (σk 1 ) constants of tritium were estimated for both stationary and continually flexing bellows. (orig.)

  20. Tritium loss in molten flibe systems

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Anderl, R.A. [Idaho National Eng. and Environ. Lab., Idaho Falls, ID (United States); Scott Willms, R. [Los Alamos National Lab., NM (United States)

    2000-04-01

    An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF.BeF{sub 2}, commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons

  1. Tritium loss in molten flibe systems

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Scott Willms, R.

    2000-01-01

    An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF.BeF 2 , commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons

  2. DOE handbook: Tritium handling and safe storage

    International Nuclear Information System (INIS)

    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

  3. DOE handbook: Tritium handling and safe storage

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. The organic tritium in the environment

    International Nuclear Information System (INIS)

    Kirchmann, R.

    1979-01-01

    Sources, organization process, and biological availability of organic tritium released in the environment, transfer of organic tritium in the environment from methane or soil to plants and from food to mammals, transfer of tritium in aquatic ecosystems, and dose to man resulting of the ingestion of tritiated food were reviewed and discussed. Some data about transfer of organic tritium in terrestrial and aquatic ecosystems reported by literatures were summarized and were supplied with recent data on biological accumulation of organic tritium in the food chain. It was stressed that more research must be done in future because data available were still insufficient. Last, some research programs in progress or planned were stated. (Tsunoda, M.)

  5. Handling of tritium-bearing wastes

    International Nuclear Information System (INIS)

    1981-01-01

    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

  6. Tritium in organic matter around Krsko Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kristof, Romana; Zorko, Benjamin; Kozar Logar, Jasmina; Kosenina, Suzana

    2017-01-01

    The aim of the research was to obtain first results of tritium in the organic matter of environmental samples in the vicinity of Krsko NPP. The emphasis was on the layout of suitable sampling network of crops and fruits in nearby agricultural area. Method for determination of tritium in organic matter in the form of Tissue Free Water Tritium (TFWT) and Organically Bound Tritium (OBT) has been implemented. Capabilities of the methods were tested on real environmental samples and its findings were compared to modeled activities of tritium from atmospheric releases and literature based results of TFWT and OBT. (author)

  7. Implantation measurements to determine tritium permeation in first wall structures

    International Nuclear Information System (INIS)

    Holland, D.F.; Causey, R.A.

    1983-01-01

    A principal safety concern for a D-T burning fusion reactor is release of tritium during routine operation. Tritium implantation into first wall structures, and subsequent permeation into coolants, is potentially an important source of tritium loss. This paper reports on an experiment in which an ion accelerator was used to implant deuterium atoms in a stainless steel disk to simulate tritium implantation in first wall structures. The permeation rate was measured under various operating conditions. These results were used in the TMAP computer code to determine potential tritium loss rates for fusion reactors

  8. Generation of gaseous tritium standards

    International Nuclear Information System (INIS)

    Hohorst, F.A.

    1994-09-01

    The determination of aqueous and non-aqueous tritium in gaseous samples is one type of determination often requested of radioanalytical laboratories. This determination can be made by introducing the sample as a gas into a sampling train containing two silica gel beds separated by.a catalytic oxidizer bed. The first bed traps tritiated water. The sample then passes into and through the oxidizer bed where non-aqueous tritium containing species are oxidized to water and other products of combustion. The second silica gel bed then traps the newly formed tritiated water. Subsequently, silica gel is removed to plastic bottles, deionized water is added, and the mixture is permitted to equilibrate. The tritium content of the equilibrium mixture is then determined by conventional liquid scintillation counting (LSC). For many years, the moisture content of inert, gaseous samples has been determined using monitors which quantitatively electrolyze the moisture present after that moisture has been absorbed by phosphorous pentoxide or other absorbents. The electrochemical reaction is quantitative and definitive, and the energy consumed during electrolysis forms the basis of the continuous display of the moisture present. This report discusses the experimental evaluation of such a monitor as the basis for a technique for conversion of small quantities of SRMs of tritiated water ( 3 HOH) into gaseous tritium standards ( 3 HH)

  9. Tritium accountancy

    International Nuclear Information System (INIS)

    Avenhaus, R.; Spannagel, G.

    1995-01-01

    Conventional accountancy means that for a given material balance area and a given interval of time the tritium balance is established so that at the end of that interval of time the book inventory is compared with the measured inventory. In this way, an optimal effectiveness of accountancy is achieved. However, there are still further objectives of accountancy, namely the timely detection of anomalies as well as the localization of anomalies in a major system. It can be shown that each of these objectives can be optimized only at the expense of the others. Recently, Near-Real-Time Accountancy procedures have been studied; their methodological background as well as their merits will be discussed. (orig.)

  10. Risks of tritium and their mitigation

    International Nuclear Information System (INIS)

    Ichimasa, Y.; Shiba, H.; Ichimasa, M.; Chikuuti, M.; Akita, Y.

    1992-01-01

    In this study, the effects of an antibacterial drug, norfloxacin, and an antibiotic, clindamycin, on in vivo oxidation of tritium gas in rats were investigated. Wistar strain male rats were used. They were provided with a standard diet, water ad libitum, and maintained in glass metabolic cages of approximately 20 liters capacity. The air flow and temperature were controlled. To investigate the availability of norfloxacin and clindamycin on the inhibition effects of the oxidation of tritium gas, two types of the experiments were conducted one was that, before the exposure to tritium gas for 2 hours, norfloxacin or clindamycin was administrated to rats three times a day for 4 days, and the other was administration of a drug after tritium gas exposure. After the exposure to tritium gas, blood, the liver, urine and feces samples were collected from rats and the radioactivity of them was determined after combustion using a sample oxidizer. In the case of norfloxacin, tritium concentration in rat body decreased one fifth of that in non-treated rats. On the other hand, administration of clindamycin shortened the biological half-life of tritium in urine to three fifth of that of non-treated rats. (author)

  11. Tritium sorption on protective coatings for concrete

    International Nuclear Information System (INIS)

    Miller, J.M.; Senohrabek, J.A.; Allsop, P.A.

    1992-11-01

    Because of the high sorption level of tritium on unprotected concrete, a program to examine the effectiveness of various concrete coatings and sealants in reducing tritium sorption was undertaken, and various exposure conditions were examined. Coatings of epoxy, polyurethane, bituminous sealant, bituminous sealant covered with polyvinylidene chloride wrap, alkyd paint, and sodium silicate were investigated with tritium (HTO) vapor concentration, humidity and contact time being varied. An exposure to HT was also carried out, and the effect of humidity on the tritium desorption rate was investigated. The relative effectiveness of the coatings was in the order of bituminous sealant + wrap > bituminous sealant > solvent-based epoxy > 100%-solids epoxy > alkyd paint > sodium silicate. The commercially available coatings for concrete resulted in tritium sorption being reduced to less than 7% of unprotected concrete. This was improved to ∼0.1% with the use of the Saran wrap (polyvinylidene chloride). The amount of tritium sorbed was proportional to tritium concentration. The total tritium sorbed decreased with an increase in humidity. A saturation effect was observed with increasing exposure time for both the coated and unprotected samples. Under the test conditions, complete saturation was not achieved within the maximum 8-hour contact time, except for the solvent-based epoxy. The desorption rate increased with a higher-humidity air purge stream. HT desorbed more rapidly than HTO, but the amount sorbed was smaller. The experimental program showed that HTO sorption by concrete can be significantly reduced with the proper choice of coating. However, tritium sorption on concrete and proposed coatings will continue to be a concern until the effects of the various conditions that affect the adsorption and desorption of tritium are firmly established for both chronic and acute tritium release conditions. Material sorption characteristics must also be considered in

  12. On spray drying of oxidized corn starch cross-linked gelatin microcapsules for drug release

    International Nuclear Information System (INIS)

    Dang, Xugang; Yang, Mao; Shan, Zhihua; Mansouri, Shahnaz; May, Bee K; Chen, Xiaodong; Chen, Hui; Woo, Meng Wai

    2017-01-01

    Spray-dried gelatin/oxidized corn starch (G/OCS) microcapsules were produced for drug release application. The prepared microcapsules were characterized through a scanning electron microscope (SEM) picture and thermogravimetric analysis (TGA). The swelling characteristics of the G/OCS microcapsules and release properties of vitamin C were then investigated. The results from structural analysis indicated that the presence of miscibility and compatibility between oxidized corn starch and gelatin, and exhibits high thermal stability up to 326 °C. The swelling of G/OCS microcapsules increased with increasing pH and reduced with decreasing ionic strength, attributed to the cross-linking between gelatin and oxidized corn starch, ionization of functional groups. Vitamin C release characteristic revealed controlled release behavior in the first 3 h of contact with an aqueous medium. This release behavior was independent of the swelling behavior indicating the potential of the encapsulating matrix to produce controlled release across a spectrum of pH environment. - Highlights: • It's first time to prepare microencapsulation with gelatin and oxidized corn starch. • The microencapsulation material can be biodegradable completely. • The production technology of microcapsule is convenient. • This work explores the potential to use oxidized starch cross-linked gelatin. • The microencapsulation material can be used for drug release.

  13. On spray drying of oxidized corn starch cross-linked gelatin microcapsules for drug release

    Energy Technology Data Exchange (ETDEWEB)

    Dang, Xugang; Yang, Mao; Shan, Zhihua [National Engineering Laboratory for Clean Technology Leather Manufacture, Sichuan University, Chengdu, Sichuan 610065 (China); Mansouri, Shahnaz [Department of Chemical Engineering, Monash University, VIC 3800 (Australia); May, Bee K [School of Applied Science, RMIT University, 124 La Trobe St, Melbourne, VIC 3001 (Australia); Chen, Xiaodong [Department of Chemical Engineering, Monash University, VIC 3800 (Australia); School of Chemical and Environmental Engineering, College of Chemistry, Chemical Engineering and Material Science, Soochow University (China); Chen, Hui, E-mail: leather2088@sina.com [National Engineering Laboratory for Clean Technology Leather Manufacture, Sichuan University, Chengdu, Sichuan 610065 (China); Department of Chemical Engineering, Monash University, VIC 3800 (Australia); Woo, Meng Wai, E-mail: meng.woo@monash.edu [Department of Chemical Engineering, Monash University, VIC 3800 (Australia)

    2017-05-01

    Spray-dried gelatin/oxidized corn starch (G/OCS) microcapsules were produced for drug release application. The prepared microcapsules were characterized through a scanning electron microscope (SEM) picture and thermogravimetric analysis (TGA). The swelling characteristics of the G/OCS microcapsules and release properties of vitamin C were then investigated. The results from structural analysis indicated that the presence of miscibility and compatibility between oxidized corn starch and gelatin, and exhibits high thermal stability up to 326 °C. The swelling of G/OCS microcapsules increased with increasing pH and reduced with decreasing ionic strength, attributed to the cross-linking between gelatin and oxidized corn starch, ionization of functional groups. Vitamin C release characteristic revealed controlled release behavior in the first 3 h of contact with an aqueous medium. This release behavior was independent of the swelling behavior indicating the potential of the encapsulating matrix to produce controlled release across a spectrum of pH environment. - Highlights: • It's first time to prepare microencapsulation with gelatin and oxidized corn starch. • The microencapsulation material can be biodegradable completely. • The production technology of microcapsule is convenient. • This work explores the potential to use oxidized starch cross-linked gelatin. • The microencapsulation material can be used for drug release.

  14. Determination of low-level tritium concentrations in surface water and precipitation in the Czech Republic

    International Nuclear Information System (INIS)

    Maresova, Diana; Hanslik, Eduard; Sedlarova, Barbora; Juranova, Eva; Charles University, Prague

    2017-01-01

    Past tests of nuclear weapons in the atmosphere, nuclear energy facilities and tritium of natural origin are main sources of tritium in the environment. Thanks to its presence in environment and its favourable properties, tritium is used as a radiotracer. Since stopping of atmospheric nuclear tests, tritium in precipitation has been decreasing towards natural levels below 1 Bq l -1 and precise analyses of low level tritium activities are necessary. This paper focuses on tritium development at sites not influenced by any technogenic release of tritium in Elbe River basin (Bohemia) in the Czech Republic using liquid scintillation measurement with electrolytic enrichment. (author)

  15. Analysis of tritium migration and deposition in fusion-reactor systems

    International Nuclear Information System (INIS)

    Holland, D.F.; Merrill, B.J.

    1981-01-01

    EG and G Idaho, Inc., is developing a safety analysis code, TMAP (Tritium Migration Analysis Program), to determine tritium loss into the environment and tritium buildup in components, coolants, and walls during normal and accident conditions. TMAP determines the thermal response of structures, solves equations for hydrogen movement through surface and in bulk materials, and also includes equations for chemical reactions. TMAP calculations of tritium movement through metal barriers at low tritium pressure agree closely with experimental measurements. The code has been used to predict inventory buildup and loss to the coolant of tritium implanted in the first wall of a fusion device, and concentrations during cleanup of tritium released into an enclosure

  16. Tritium control: October 1982-March 1983

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Rogers, M.L.

    1983-01-01

    Surveys in gloveboxes indicated surface activity on stainless steel and its apparent dependence on time and atmospheric tritium levels. Surveys in fumehoods were completed to investigate the extent of surface contamination on surfaces of various materials. Gas generation rates caused by radiolysis of tritiated waste materials were determined for polymer and nonpolymer-impregnated tritiated concrete and fixated and nonfixated tritiated waste vacuum pump oil. In addition, the pressure change of hydrogen cover gas over tritiated water on cement-plaster was determined. The test program to measure and compare the release of tritium from tritiated concrete with and without styrene impregnation continued. Tritium permeation data from small test blocks are given. The drum study monitoring the release of tritium from actual burial packages continued. The maximum fractional release rate for the three types of high activity, tritiated liquid waste generated is 5.1 x 10 -5 , and the maximum total permeation is 179 mCi after 8.5 yr. These two values represent a 13% increase for the past 6 months. Tritium release from the polymer-impregnated, tritiated concrete (PITC) and from the control (non-PITC) remains very low. The Emergency Containment System (ECS), an automatically actuated system developed at Mound to remove tritium from room air, has been modified and upgraded to support new applications. The leakage rate in the ECS area has been lowered, a fast-start system installed for greater conversion efficiency at startup, and the alumina beds regenerated

  17. Anti-Biofilm Efficacy of Nitric Oxide-Releasing Silica Nanoparticles

    OpenAIRE

    Hetrick, Evan M.; Shin, Jae Ho; Paul, Heather S.; Schoenfisch, Mark H.

    2009-01-01

    The ability of nitric oxide (NO)-releasing silica nanoparticles to kill biofilm-based microbial cells is reported. Biofilms of Pseudomonas aeruginosa, Escherichia coli, Staphylococcus aureus, Staphylococcus epidermidis, and Candida albicans were formed in vitro and exposed to NO-releasing silica nanoparticles. Replicative viability experiments revealed that ≥ 99% of cells from each type of biofilm were killed via NO release, with the greatest efficacy (≥ 99.999% killing) against gram-negative...

  18. Improving tritium exposure reconstructions using accelerator mass spectrometry

    International Nuclear Information System (INIS)

    Love, A.H.; Hunt, J.R.; Vogel, J.S.; Knezovich, J.P.

    2004-01-01

    Direct measurement of tritium atoms by accelerator mass spectrometry (AMS) enables rapid low-activity tritium measurements from milligram-sized samples and permits greater ease of sample collection, faster throughput, and increased spatial and/or temporal resolution. Because existing methodologies for quantifying tritium have some significant limitations, the development of tritium AMS has allowed improvements in reconstructing tritium exposure concentrations from environmental measurements and provides an important additional tool in assessing the temporal and spatial distribution of chronic exposure. Tritium exposure reconstructions using AMS were 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. Although the current quantification limit of tritium AMS is not adequate to determine natural environmental variations in tritium concentrations, it is expected to be sufficient for studies assessing possible health effects from chronic environmental tritium exposure. (orig.)

  19. Improving tritium exposure reconstructions using accelerator mass spectrometry

    Science.gov (United States)

    Hunt, J. R.; Vogel, J. S.; Knezovich, J. P.

    2010-01-01

    Direct measurement of tritium atoms by accelerator mass spectrometry (AMS) enables rapid low-activity tritium measurements from milligram-sized samples and permits greater ease of sample collection, faster throughput, and increased spatial and/or temporal resolution. Because existing methodologies for quantifying tritium have some significant limitations, the development of tritium AMS has allowed improvements in reconstructing tritium exposure concentrations from environmental measurements and provides an important additional tool in assessing the temporal and spatial distribution of chronic exposure. Tritium exposure reconstructions using AMS were 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. Although the current quantification limit of tritium AMS is not adequate to determine natural environmental variations in tritium concentrations, it is expected to be sufficient for studies assessing possible health effects from chronic environmental tritium exposure. PMID:14735274

  20. Tritium in the Savannah River Site environment

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. Tritium in the Savannah River Site environment

    International Nuclear Information System (INIS)

    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

  2. The tritium and the controlled fusion reactors

    International Nuclear Information System (INIS)

    Leger, D.; Rouyer, J.L.

    1986-04-01

    It is shown how tritium is used how it is circulating in a fusion reactor. The great functions of tritium circuits are detailed: reprocessing of burnt gases, reprocessing of gases coming from neutral injectors, reprocessing from gaseous wastes, detritiation of cooling fluids. Current technologic developments are quoted. Then tritium confinement and containment, in normal or accidental situations, are displayed. Limitation devices of effluents and release for normal operating (noticeably the reprocessing systems of atmosphere) and safety and protection systems in case of accident are described [fr

  3. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    International Nuclear Information System (INIS)

    Gentile, C.A.; LaMarche, P.H.

    1995-01-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During '' time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems

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

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

    1993-01-01

    The Tritium Plasma Experiment was assembled at Sandia National Labs., Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Lab. 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 10 23 ions/m 2 .s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated 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. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures

  5. Tritium persistence pattern in some terrestrial plants-field investigations

    International Nuclear Information System (INIS)

    Soman, S.D.; Iyengar, T.S.; Krishnamoorthy, T.M.; Sadarangani, S.H.; Vaze, P.K.; Gogate, S.S.; Deo, J.V.

    1977-01-01

    The uptake and release pattern of tritium in certain trees in their natural conditions of growth were investigated by artificial simulation of active conditions by incorporating tritium in the system through stem or roots. These trees are grown in some of the nuclear sites wherein a number of nuclear facilities are located. The species studied include palms, casuarinas and banana trees. In most of the cases a single component corresponding to the tree compartment tritium was obtained. The second component of the tissue free water tritium and the tissue bound compartment of tritium were not easily resolvable due to tremendous variation caused by the environmental conditions such as rain, humidity etc. Repeated humps were observed in certain cases of root uptake studies due to the variation in the meteorological factors. In most of the cases the half residence times for tritium (Tsub(1/2)) (tissue free water tritium) were found to be below two days. (author)

  6. Effects of Muscle-Specific Oxidative Stress on Cytochrome c Release and Oxidation-Reduction Potential Properties.

    Science.gov (United States)

    Ke, Yiling; Mitacek, Rachel M; Abraham, Anupam; Mafi, Gretchen G; VanOverbeke, Deborah L; DeSilva, Udaya; Ramanathan, Ranjith

    2017-09-06

    Mitochondria play a significant role in beef color. However, the role of oxidative stress in cytochrome c release and mitochondrial degradation is not clear. The objective was to determine the effects of display time on cytochrome c content and oxidation-reduction potential (ORP) of beef longissimus lumborum (LL) and psoas major (PM) muscles. PM discolored by day 3 compared with LL. On day 0, mitochondrial content and mitochondrial oxygen consumption were greater in PM than LL. However, mitochondrial content and oxygen consumption were lower (P stress can affect cytochrome c release and ORP changes.

  7. Tritium and helium behavior in irradiated beryllium

    International Nuclear Information System (INIS)

    Billone, M.C.; Lin, C.C.; Baldwin, D.L.

    1990-11-01

    Large quantities of Be (> 100 metric tons) are planned for use in the ITER blanket design to enhance tritium breeding and to act as a thermal barrier between coolant and breeder. Tritium retention/release and He-induced swelling are important issues in blanket design. The data base on tritium and helium behavior in Be is reviewed. New data on tritium retention/release and He bubble growth are presented for Be irradiated to 5 x 10 22 n(E > 1 MeV)/cm 2 at ∼75 degree C and postirradiation-annealed for 700 hours at 500 degree C. A model (diffusion/desorption) is proposed and tested against the data base to determine tritium diffusivity and the desorption rate constant. Similarly a model for He-induced swelling is developed and tested against the data base. The dependence of tritium retention and release on He content and impurities (e.g. BeO) is also explored. 11 refs., 6 figs

  8. STAR facility tritium accountancy

    International Nuclear Information System (INIS)

    Pawelko, R. J.; Sharpe, J. P.; Denny, B. J.

    2008-01-01

    The Safety and Tritium Applied Research (STAR) facility has been established to provide a laboratory infrastructure for the fusion community to study tritium science associated with the development of safe fusion energy and other technologies. STAR is a radiological facility with an administrative total tritium inventory limit of 1.5 g (14,429 Ci) [1]. Research studies with moderate tritium quantities and various radionuclides are performed in STAR. Successful operation of the STAR facility requires the ability to receive, inventory, store, dispense tritium to experiments, and to dispose of tritiated waste while accurately monitoring the tritium inventory in the facility. This paper describes tritium accountancy in the STAR facility. A primary accountancy instrument is the tritium Storage and Assay System (SAS): a system designed to receive, assay, store, and dispense tritium to experiments. Presented are the methods used to calibrate and operate the SAS. Accountancy processes utilizing the Tritium Cleanup System (TCS), and the Stack Tritium Monitoring System (STMS) are also discussed. Also presented are the equations used to quantify the amount of tritium being received into the facility, transferred to experiments, and removed from the facility. Finally, the STAR tritium accountability database is discussed. (authors)

  9. Fission gas release from oxide fuels at high burnups (AWBA development program)

    International Nuclear Information System (INIS)

    Dollins, C.C.

    1981-02-01

    The steady state gas release, swelling and densification model previously developed for oxide fuels has been modified to accommodate the slow transients in temperature, temperature gradient, fission rate and pressure that are encountered in normal reactor operation. The gas release predictions made by the model were then compared to gas release data on LMFBR-EBRII fuels obtained by Dutt and Baker and reported by Meyer, Beyer, and Voglewede. Good agreement between the model and the data was found. A comparison between the model and three other sets of gas release data is also shown, again with good agreement

  10. Dependency of irradiation damage density on tritium migration behaviors in Li2TiO3

    International Nuclear Information System (INIS)

    Kobayashi, Makoto; Toda, Kensuke; Oya, Yasuhisa; Okuno, Kenji

    2014-01-01

    Tritium migration behaviors in Li 2 TiO 3 with the increase of irradiation damage density were investigated by means of electron spin resonance and thermal desorption spectroscopy. The irradiation damages of F + -centers and O − -centers were formed by neutron irradiation, and their damage densities were increased with increasing neutron fluence. Tritium release temperature was clearly shifted toward higher temperature side with increasing neutron fluence, i.e. increasing damage density. The rate determining process for tritium release was also clearly changed depending on the damage density. Tritium release was mainly controlled by tritium diffusion process in crystalline grain of Li 2 TiO 3 at lower neutron fluence. The apparent tritium diffusivity was reduced as the damage density in Li 2 TiO 3 increased due to the introduction of tritium trapping/detrapping sites for diffusing tritium. Then, tritium trapping/detrapping processes began to control the overall tritium release with further damage introductions as the amount of tritium trapping sites increased enough to trap most of tritium in Li 2 TiO 3 . The effects of water vapor in purge gas on tritium release behaviors were also investigated. It was considered that hydrogen isotopes in purge gas would be dissociated and adsorbed on the surface of Li 2 TiO 3 . Then, hydrogen isotopes diffused inward Li 2 TiO 3 would occupy the tritium trapping sites before diffusing tritium reaches to these sites, promoting apparent tritium diffusion consequently. Kinetics analysis of tritium release for highly damaged Li 2 TiO 3 showed that the rate determining process of tritium release was the detrapping process of tritium formed as hydroxyl groups. The rate of tritium detrapping as hydroxyl groups was determined by the kinetic analysis, and was comparable to tritium release kinetics for Li 2 O, LiOH and Li 4 TiO 4 . The dangling oxygen atoms (O − -centers) formed by neutron irradiation would contribute strongly on the

  11. HYLIFE-II tritium management system

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Dolan, T.J.

    1993-06-01

    The tritium management system performs seven functions: (1) tritium gas removal from the blast chamber, (2) tritium removal from the Flibe, (3) tritium removal from helium sweep gas, (4) tritium removal from room air, (5) hydrogen isotope separation, (6) release of non-hazardous gases through the stack, (7) fixation and disposal of hazardous effluents. About 2 TBq/s (5 MCi/day) of tritium is bred in the Flibe (Li 2 BeF 4 ) molten salt coolant by neutron absorption. Tritium removal is accomplished by a two-stage vacuum disengager in each of three steam generator loops. Each stage consists of a spray of 0.4 mm diameter, hot Flibe droplets into a vacuum chamber 4 m in diameter and 7 m tall. As droplets fall downward into the vacuum, most of the tritium diffuses out and is pumped away. A fraction Φ∼10 -5 of the tritium remains in the Flibe as it leaves the second stage of the vacuum disengager, and about 24% of the remaining tritium penetrates through the steam generator tubes, per pass, so the net leakage into the steam system is about 4.7 MBq/s (11 Ci/day). The required Flibe pumping power for the vacuum disengager system is 6.6 MW. With Flibe primary coolant and a vacuum disengager, an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate vacuum disengager operation with Flibe. A secondary containment shell with helium sweep gas captures the tritium permeating out of the Flibe ducts, limiting leaks there to about 1 Ci/day. The tritium inventory in the reactor is about 190 g, residing mostly in the large Flibe recirculation duct walls. The total cost of the tritium management system is 92 M$, of which the vacuum disengagers cost = 56%, the blast chamber vacuum system = 15%, the cryogenic plant = 9%, the emergency air cleanup and waste treatment systems each = 6%, the protium removal system = 3%, and the fuel storage system and inert gas system each = 2%

  12. A new blanket tritium recovery experiment with intense DT neutron source at JAEA/FNS

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Kentaro, E-mail: ochiai.kentaro@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Edao, Yuki [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Kawamura, Yoshinori [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ohta, Masayuki; Kwon, Saerom; Konno, Chikara [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan)

    2016-11-01

    Highlights: • For detail investigation of the tritium recovery performance on the fusion reactor blanket, we have started a new blanket tritium recovery experiment with ionization chamber at JAEA/FNS. • A new improved container was provided for the appropriate tritium measurement by IC and also utilized for the enhancement of TPR in the new container. The TPR was calculated with a calculation code MCNP5 and some typical nuclear data libraries and then the radioactivity of the tritium recovery with LSC corresponded with that of calculation. • The tritium release curves by the IC outputs are similar to those by the LSC output. However, it was indicated that the quantitative measurement by IC needed further improvement for the tritium recovery. - Abstract: We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

  13. Nitric oxide synthesis and biological functions of nitric oxide released from ruthenium compounds

    Directory of Open Access Journals (Sweden)

    A.C. Pereira

    2011-09-01

    Full Text Available During three decades, an enormous number of studies have demonstrated the critical role of nitric oxide (NO as a second messenger engaged in the activation of many systems including vascular smooth muscle relaxation. The underlying cellular mechanisms involved in vasodilatation are essentially due to soluble guanylyl-cyclase (sGC modulation in the cytoplasm of vascular smooth cells. sGC activation culminates in cyclic GMP (cGMP production, which in turn leads to protein kinase G (PKG activation. NO binds to the sGC heme moiety, thereby activating this enzyme. Activation of the NO-sGC-cGMP-PKG pathway entails Ca2+ signaling reduction and vasodilatation. Endothelium dysfunction leads to decreased production or bioavailability of endogenous NO that could contribute to vascular diseases. Nitrosyl ruthenium complexes have been studied as a new class of NO donors with potential therapeutic use in order to supply the NO deficiency. In this context, this article shall provide a brief review of the effects exerted by the NO that is enzymatically produced via endothelial NO-synthase (eNOS activation and by the NO released from NO donor compounds in the vascular smooth muscle cells on both conduit and resistance arteries, as well as veins. In addition, the involvement of the nitrite molecule as an endogenous NO reservoir engaged in vasodilatation will be described.

  14. Safety analysis of tritium processing system based on PHA

    International Nuclear Information System (INIS)

    Fu Wanfa; Luo Deli; Tang Tao

    2012-01-01

    Safety analysis on primary confinement of tritium processing system for TBM was carried out with Preliminary Hazard Analysis. Firstly, the basic PHA process was given. Then the function and safe measures with multiple confinements about tritium system were described and analyzed briefly, dividing the two kinds of boundaries of tritium transferring through, that are multiple confinement systems division and fluid loops division. Analysis on tritium releasing is the key of PHA. Besides, PHA table about tritium releasing was put forward, the causes and harmful results being analyzed, and the safety measures were put forward also. On the basis of PHA, several kinds of typical accidents were supposed to be further analyzed. And 8 factors influencing the tritium safety were analyzed, laying the foundation of evaluating quantitatively the safety grade of various nuclear facilities. (authors)

  15. Tritium activities in Canada

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1995-01-01

    Canadian tritium activites comprise three major interests: utilites, light manufacturers, and fusion. There are 21 operating CANDU reactors in Canada; 19 with Ontario Hydro and one each with Hydro Quebec and New Brunswick Power. There are two light manufacturers, two primary tritium research facilities (at AECL Chalk River and Ontario Hydro Technologies), and a number of industry and universities involved in design, construction, and general support of the other tritium activities. The largest tritum program is in support of the CANDU reactors, which generate tritium in the heavy water as a by-product of normal operation. Currently, there are about 12 kg of tritium locked up in the heavy water coolant and moderator of these reactors. The fusion work is complementary to the light manufacturing, and is concerned with tritium handling for the ITER program. This included design, development and application of technologies related to Isotope Separation, tritium handling, (tritiated) gas separation, tritium-materials interaction, and plasma fueling

  16. Nitric oxide and non-quantal acetylcholine release

    Czech Academy of Sciences Publication Activity Database

    Vyskočil, František

    2003-01-01

    Roč. 7, č. 3 (2003), s. 241-243 ISSN 1211-7579. [Celostátní konference biologické psychiatrie /11./. Luhačovice, 11.06.2003-14.06.2003] R&D Projects: GA ČR GA305/02/1333 Institutional research plan: CEZ:AV0Z5011922; CEZ:MSM 113100003 Keywords : nitric oxide Subject RIV: ED - Physiology

  17. The Tritium White Paper

    International Nuclear Information System (INIS)

    2009-01-01

    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

  18. Nitric Oxide-Releasing Silica Nanoparticle-Doped Polyurethane Electrospun Fibers

    Science.gov (United States)

    Koh, Ahyeon; Carpenter, Alexis W.; Slomberg, Danielle L.; Schoenfisch, Mark H.

    2013-01-01

    Electrospun polyurethane fibers doped with nitric oxide (NO)-releasing silica particles are presented as novel macromolecular scaffolds with prolonged NO-release and high porosity. Fiber diameter (119–614 nm) and mechanical strength (1.7–34.5 MPa of modulus) were varied by altering polyurethane type and concentration, as well as the NO-releasing particle composition, size, and concentration. The resulting NO-releasing electrospun nanofibers exhibited ~83% porosity with flexible plastic or elastomeric behavior. The use of N-diazeniumdiolate- or S-nitrosothiol-modified particles yielded scaffolds exhibiting a wide range of NO release totals and durations (7.5 nmol mg−1–0.12 μmol mg−1 and 7 h to 2 weeks, respectively). The application of NO-releasing porous materials as coating for subcutaneous implants may improve tissue biocompatibility by mitigating the foreign body response and promoting cell integration. PMID:23915047

  19. TRIO-01 experiment: in-situ tritium-recovery results

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Billone, M.C.

    1983-08-01

    The TRIO-01 experiment is a test of in-situ tritium recovery from γ-LiAlO 2 with test conditions chosen to simulate those anticipated in fusion power reactors. A status report is presented which describes qualitatively the results observed during the irradiation phase of the experiment. Both the rate of tritium release and the chemical forms of tritium were measured using a helium sweep gas which flowed past the breeder material to a gas analysis system

  20. TRIO-01 experiment: in-situ tritium recovery results

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Billone, M.C.

    1983-10-01

    The TRIO-01 experiment is a test of in-situ tritium recovery from γ-LiAlO 2 with test conditions chosen to simulate those anticipated in fusion power reactors. A status report is presented which describes qualitatively the results observed during the irradiation phase of the experiment. Both the rate of tritium release and the chemical forms of tritium were measured using a helium sweep gas which flowed past the breeder material to a gas analysis system

  1. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

    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

  2. Tritium burning in inertial electrostatic confinement fusion facility

    Energy Technology Data Exchange (ETDEWEB)

    Ohnishi, Masami, E-mail: onishi@kansai-u.ac.jp [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Yamamoto, Yasushi; Osawa, Hodaka [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Hatano, Yuji; Torikai, Yuji [Hydrogen Isotope Science Center, University of Toyama, Gofuku, Toyama 930-8555 (Japan); Murata, Isao [Faculty of Engineering Environment and Energy Department, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Masuda, Kai [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hotta, Eiki [Interdisciplinary Graduate School of Science and Engineering, Tokyo Institute of Technology, 4259 Nagatsuda-cho, Midori-ku, Yokohama 226-8503 (Japan)

    2016-11-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10{sup 8}(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  3. Tritium burning in inertial electrostatic confinement fusion facility

    International Nuclear Information System (INIS)

    Ohnishi, Masami; Yamamoto, Yasushi; Osawa, Hodaka; Hatano, Yuji; Torikai, Yuji; Murata, Isao; Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki; Masuda, Kai; Hotta, Eiki

    2016-01-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10"8(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  4. Influence of corium oxidation on fission product release from molten pool

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V., E-mail: bechta@sbor.spb.s [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Krushinov, E.V.; Vitol, S.A.; Khabensky, V.B.; Kotova, S.Yu.; Sulatsky, A.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almyashev, V.I. [Grebenschikov Institute of Silicate Chemistry of the Russian Academy of Sciences (ISC RAS), St. Petersburg (Russian Federation); Ducros, G.; Journeau, C. [CEA, DEN, Cadarache, F-13108 St. Paul lez Durance (France); Bottomley, D. [Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Clement, B. [Institut de Radioprotection et Surete Nucleaire (IRSN), St. Paul lez Durance (France); Herranz, L. [CIEMAT, Madrid (Spain); Guentay, S. [PSI, Wuerenlingen (Switzerland); Trambauer, K. [GRS, Muenchen (Germany); Auvinen, A. [VTT, Espoo (Finland); Bezlepkin, V.V. [SPbAEP, St. Petersburg (Russian Federation)

    2010-05-15

    Qualitative and quantitative determination of the release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in a cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate, aerosol particle composition and size distribution. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA codes are made. Recommendations for further investigations are proposed following the major observations and discussions.

  5. Influence of corium oxidation on fission product release from molten pool

    International Nuclear Information System (INIS)

    Bechta, S.V.; Krushinov, E.V.; Vitol, S.A.

    2009-01-01

    Release of low-volatile fission products and core materials from molten oxidic corium was investigated in the EVAN project under the auspices of ISTC. The experiments carried out in cold crucible with induction heating and RASPLAV test facility are described. The results are discussed in terms of reactor application; in particular, pool configuration, melt oxidation kinetics, critical influence of melt surface temperature and oxidation index on the fission product release rate and aerosol particle composition. The relevance of measured high release of Sr from the molten pool for the reactor application is highlighted. Comparisons of the experimental data with those from the COLIMA CA-U3 test and the VERCORS tests, as well as with predictions from IVTANTHERMO and GEMINI/NUCLEA are set. (author)

  6. Savannah River Laboratory contribution to the Chalk River experimental HT release of June 1987

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.

    1987-01-01

    Seventeen years ago, a series of experiments were conducted at Bruyeres-Le-Chatel by French scientists interested in atmospheric diffusion. The scientists used tritium as a tracer in these tests. The results showed nearly complete oxidation of the gas by the time the released material had been carried 10 kilometers from the source. It became clear that if the results were general, the doses projected for fusion reactors would be greatly increased over what had been expected from earlier experience with HT gas. For this reason, the Next European Torus (NET) fusion program funded a series of experiments to look further at tritium oxidation in the environment. SRL participation in the tests provided an opportunity to make field comparisons of the SRL design tritium gas samplers and also provided vegetation that had been exposed under controlled conditions which could be used to study the incorporation of tritium into vegetation organics, the subject of ongoing SRL research. 25 refs., 4 figs., 2 tabs

  7. Tritium conference days

    International Nuclear Information System (INIS)

    Garnier-Laplace, J.; Lebaron-Jacobs, L.; Sene, M.; Devin, P.; Chretien, V.; Le Guen, B.; Guetat, Ph.; Baglan, N.; Ansoborlo, E.; Boyer, C.; Masson, M.; Bailly-Du-Bois, P.; Jenkinson, St.; Wakeford, R.; Saintigny, Y.; Romeo, P.H.; Thompson, P.; Leterq, D.; Chastagner, F.; Cortes, P.; Philippe, M.; Paquet, F.; Fournier, M.

    2009-01-01

    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/HTO air and OBT/HTO free (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.)

  8. Tritium and radon risks for humans

    International Nuclear Information System (INIS)

    Mauna, Traian; Mauna, Andriesica

    2008-01-01

    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)

  9. Tritium monitoring in environment at ICIT Tritium Separation Facility

    International Nuclear Information System (INIS)

    Varlam, Carmen; Stefanescu, I.; Vagner, Irina; Faurescu, I.; Toma, A.; Dulama, C.; Dobrin, R.

    2008-01-01

    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)

  10. Heat release from B4C oxidation in steam and air

    International Nuclear Information System (INIS)

    Belovsky, L.

    1996-01-01

    BWR and some PWR cores contain boron carbide (B 4 C) as neutron absorber. During a severe accident, the B 4 C can potentially react with steam under release of heat and hydrogen. Although models for B 4 C oxidation already exist in MELCOR and SCDAP/RELAP5, a development of a new model for another computer code seems to be difficult due to a missing comprehensive description of the current modelling methodology and scarce experimental data. The aim of this paper is to highlight the key points of the B 4 C oxidation using the existing available experimental data and to perform a simple heat balance analysis of the B 4 C/steam and B 4 C/air chemical reactions. The analysis of literature data shows that the B 4 C oxidation phenomenon is qualitatively well described below 1000 deg. C. However, no reliable data exist for the reaction kinetics especially above this temperature. It was found that the experimental results strongly depend on the experimental arrangement. The reaction heats, calculated in this study, indicate that the B 4 C oxidation is an exothermic reaction, releasing more heat in air than in steam. The formation of boric acids from the boron oxide increases the heat release from B 4 C by ∼ 10%, in the worst case. Although the total heat, released in a PWR core from the B 4 C oxidation, is probably much smaller than the heat released from the Zr/steam reaction, it is not excluded that the B 4 C oxidation can locally contribute to the damage of the control elements due to local overheating. Modelling of these phenomena is, however, very difficult due to the complex geometry of the liquefied control elements and due to absence of suitable data on the reaction kinetics. (author). 25 refs, 2 figs, 3 tabs

  11. Metal release behavior of surface oxidized stainless steels into flowing high temperature pure water

    International Nuclear Information System (INIS)

    Fujiwara, Kazuo; Tomari, Haruo; Nakayama, Takenori; Shimogori, Kazutoshi; Ishigure, Kenkichi; Matsuura, Chihiro; Fujita, Norihiko; Ono, Shoichi.

    1987-01-01

    In order to clarify the effect of oxidation treatment of Type 304 SS on the inhibition of metal release into high temperature pure water, metal release rate of individual alloying element into flowing deionized water containing 50 ppb dissolved oxygen was measured as the function of exposure time on representative specimens oxidized in air and steam. The behavior of metal release was also discussed in relation to the structure of surface films. Among the alloying elements the amount of Fe ion, Cr ion and Fe crud in high temperature pure water tended to saturate with the exposure time and that of Ni ion and Co ion tended to increase monotonously with the exposure time for all specimens tested. And the treatment of steam-oxidation was the most effective to decrease the metal release of alloying elements and the treatment by air-oxidation also decreased the metal release. These tendencies were confirmed to correlate well with the structure of the surface films as it was in the results in the static autoclave test. (author)

  12. Tritium Mitigation/Control for Advanced Reactor System

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Xiaodong; Christensen, Richard; Saving, John P

    2018-03-31

    A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent the residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: 1. To estimate tritium permeation behavior in FHRs; 2. To design a tritium removal system for FHRs; 3. To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; 4. To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities

  13. On spray drying of oxidized corn starch cross-linked gelatin microcapsules for drug release.

    Science.gov (United States)

    Dang, Xugang; Yang, Mao; Shan, Zhihua; Mansouri, Shahnaz; May, Bee K; Chen, Xiaodong; Chen, Hui; Woo, Meng Wai

    2017-05-01

    Spray-dried gelatin/oxidized corn starch (G/OCS) microcapsules were produced for drug release application. The prepared microcapsules were characterized through a scanning electron microscope (SEM) picture and thermogravimetric analysis (TGA). The swelling characteristics of the G/OCS microcapsules and release properties of vitamin C were then investigated. The results from structural analysis indicated that the presence of miscibility and compatibility between oxidized corn starch and gelatin, and exhibits high thermal stability up to 326°C. The swelling of G/OCS microcapsules increased with increasing pH and reduced with decreasing ionic strength, attributed to the cross-linking between gelatin and oxidized corn starch, ionization of functional groups. Vitamin C release characteristic revealed controlled release behavior in the first 3h of contact with an aqueous medium. This release behavior was independent of the swelling behavior indicating the potential of the encapsulating matrix to produce controlled release across a spectrum of pH environment. Copyright © 2016 Elsevier B.V. All rights reserved.

  14. Tritium pellet injector results

    International Nuclear Information System (INIS)

    Fisher, P.W.; Bauer, M.L.; Baylor, L.R.; Deleanu, L.E.; Fehling, D.T.; Milora, S.L.; Whitson, J.C.

    1988-01-01

    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 3 He 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

  15. Neuronal nitric oxide synthase-deficient mice have impaired Renin release but normal blood pressure

    DEFF Research Database (Denmark)

    Sällström, Johan; Carlström, Mattias; Jensen, Boye L

    2008-01-01

    BackgroundNitric oxide deficiency is involved in the development of hypertension, but the mechanisms are currently unclear. This study was conducted to further elucidate the role of neuronal nitric oxide synthase (nNOS) in blood pressure regulation and renin release in relation to different sodiu......-116; doi:10.1038/ajh.2007.16American Journal of Hypertension (2008) 21 111-116; doi:10.1038/ajh.2007.16....

  16. Release mechanisms of acetaminophen from polyethylene oxide/polyethylene glycol matrix tablets utilizing magnetic resonance imaging.

    Science.gov (United States)

    Tajiri, Tomokazu; Morita, Shigeaki; Sakamoto, Ryosaku; Suzuki, Masazumi; Yamanashi, Shigeyuki; Ozaki, Yukihiro; Kitamura, Satoshi

    2010-08-16

    Release mechanism of acetaminophen (AAP) from extended-release tablets of hydrogel polymer matrices containing polyethylene oxide (PEO) and polyethylene glycol (PEG) were achieved using flow-through cell with magnetic resonance imaging (MRI). The hydrogel forming abilities are observed characteristically and the layer thickness which is corresponding to the diffusion length of AAP has a good correlation with the drug release profiles. In addition, polymeric erosion contribution to AAP releasing from hydrogel matrix tablets was directly quantified using size-exclusion chromatography (SEC). The matrix erosion profile indicates that the PEG erosion kinetic depends primarily on the composition ratio of PEG to PEO. The present study has confirmed that the combination of in situ MRI and SEC should be well suited to investigate the drug release mechanisms of hydrogel matrix such as PEO/PEG. Copyright (c) 2010 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

    2010-01-01

    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)

  18. Elemental tritium deposition and conversion in the terrestrial environment

    International Nuclear Information System (INIS)

    Dunstall, T.G.; Ogram, G.L.; Spencer, F.S.

    1985-01-01

    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

  19. Experimental study and phenomenological modeling of the hydrolysis of tritiated sodium: influence of experimental conditions on the tritium distribution in the effluents

    International Nuclear Information System (INIS)

    Chassery, Aurelien

    2014-01-01

    Within the framework of the decommissioning of fast reactors, several processes are under investigation regarding sodium disposal. One of them rests on the implementation of the sodium-water reaction (SWR), in a controlled and progressive way, to remove residual sodium containing impurities such as sodium hydrides, sodium oxides and tritiated sodium hydrides. Such a hydrolysis releases some amount of energy and produces a liquid effluent, composed of a solution of soda, and a gaseous effluent, composed of hydrogen, steam and an inert gas. The tritium, originally into the sodium as a soluble (T - ) or precipitate form (NaT), will be distributed between the liquid and gaseous effluent, and according to two chemical forms, the tritium hydride HT and the tritiated water HTO. HTO being 10,000 times more radio-toxic than HT, a precise knowledge of the mechanisms governing the distribution of tritium is necessary in order to estimate the exhaust gas releases and design the process needed to treat the off-gas before its release into the environment. An experimental study has been carried out in order to determine precisely the phenomena involved in the hydrolysis. The influence of the experimental conditions on the tritium distribution has been tested. The results of this study leaded to a phenomenological description of the tritiated sodium hydrolysis that will help to predict the composition of the effluents, regarding tritium. (author) [fr

  20. Regulation of Ca2+ release from mitochondria by the oxidation-reduction state of pyridine nucleotides

    Science.gov (United States)

    Lehninger, Albert L.; Vercesi, Anibal; Bababunmi, Enitan A.

    1978-01-01

    Mitochondria from normal rat liver and heart, and also Ehrlich tumor cells, respiring on succinate as energy source in the presence of rotenone (to prevent net electron flow to oxygen from the endogenous pyridine nucleotides), rapidly take up Ca2+ and retain it so long as the pyridine nucleotides are kept in the reduced state. When acetoacetate is added to bring the pyridine nucleotides into a more oxidized state, Ca2+ is released to the medium. A subsequent addition of a reductant of the pyridine nucleotides such as β-hydroxybutyrate, glutamate, or isocitrate causes reuptake of the released Ca2+. Successive cycles of Ca2+ release and uptake can be induced by shifting the redox state of the pyridine nucleotides to more oxidized and more reduced states, respectively. Similar observations were made when succinate oxidation was replaced as energy source by ascorbate oxidation or by the hydrolysis of ATP. These and other observations form the basis of a hypothesis for feedback regulation of Ca2+-dependent substrate- or energy-mobilizing enzymatic reactions by the uptake or release of mitochondrial Ca2+, mediated by the cytosolic phosphate potential and the ATP-dependent reduction of mitochondrial pyridine nucleotides by reversal of electron transport. Images PMID:25436

  1. Preparation and characterization of oxidized starch polymer microgels for encapsulation and controlled release of functional ingredients

    NARCIS (Netherlands)

    Li, Y.; Vries, R. de; Slaghek, T.; Timmermans, J.; Cohen Stuart, M.A.; Norde, W.

    2009-01-01

    A novel biocompatible and biodegradable microgel system has been developed for controlled uptake and release of especially proteins. It contains TEMPO-oxidized potato starch polymers, which are chemically cross-linked by sodium trimetaphosphate (STMP). Physical chemical properties have been

  2. Medicinal chemistry and anti-inflammatory activity of nitric oxide-releasing NSAI drugs.

    Science.gov (United States)

    Koç And, Esra; Küçükgüzel, S Güniz

    2009-05-01

    Nitric Oxide, which acts as a non-specific cytotoxic mediator and a biological messenger on immunological competence, has been gaining significantly increasing importance. As an alternative to conventional NSAIDs having significant side effects, pharmacologically improved and therapeutically enhanced NO releasing non-steroidal anti-inflammatory drugs with less side effects are being planned to produce.

  3. Exhaustive Exercise-induced Oxidative Stress Alteration of Erythrocyte Oxygen Release Capacity.

    Science.gov (United States)

    Xiong, Yanlian; Xiong, Yanlei; Wang, Yueming; Zhao, Yajin; Li, Yaojin; Ren, Yang; Wang, Ruofeng; Zhao, Mingzi; Hao, Yitong; Liu, Haibei; Wang, Xiang

    2018-05-24

    The aim of the present study is to explore the effect of exhaustive running exercise (ERE) in the oxygen release capacity of rat erythrocytes. Rats were divided into sedentary control (C), moderate running exercise (MRE) and exhaustive running exercise groups. The thermodynamics and kinetics properties of the erythrocyte oxygen release process of different groups were tested. We also determined the degree of band-3 oxidative and phosphorylation, anion transport activity and carbonic anhydrase isoform II(CAII) activity. Biochemical studies suggested that exhaustive running significantly increased oxidative injury parameters in TBARS and methaemoglobin levels. Furthermore, exhaustive running significantly decreased anion transport activity and carbonic anhydrase isoform II(CAII) activity. Thermodynamic analysis indicated that erythrocytes oxygen release ability also significantly increased due to elevated 2,3-DPG level after exhaustive running. Kinetic analysis indicated that exhaustive running resulted in significantly decreased T50 value. We presented evidence that exhaustive running remarkably impacted thermodynamics and kinetics properties of RBCs oxygen release. In addition, changes in 2,3-DPG levels and band-3 oxidation and phosphorylation could be the driving force for exhaustive running induced alterations in erythrocytes oxygen release thermodynamics and kinetics properties.

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

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

    1993-01-01

    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 10 19 ions/cm 2 · 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

  5. Use of Tritium Accelerator Mass Spectrometry for Tree Ring Analysis

    Science.gov (United States)

    LOVE, ADAM H.; HUNT, JAMES R.; ROBERTS, MARK L.; SOUTHON, JOHN R.; CHIARAPPA - ZUCCA, MARINA L.; DINGLEY, KAREN H.

    2010-01-01

    Public concerns over the health effects associated with low-level and long-term exposure to tritium released from industrial point sources have generated the demand for better methods to evaluate historical tritium exposure levels for these communities. The cellulose of trees accurately reflects the tritium concentration in the source water and may contain the only historical record of tritium exposure. The tritium activity in the annual rings of a tree was measured using accelerator mass spectrometry to reconstruct historical annual averages of tritium exposure. Milligram-sized samples of the annual tree rings from a Tamarix located at the Nevada Test Site are used for validation of this methodology. The salt cedar was chosen since it had a single source of tritiated water that was well-characterized as it varied over time. The decay-corrected tritium activity of the water in which the salt cedar grew closely agrees with the organically bound tritium activity in its annual rings. This demonstrates that the milligram-sized samples used in tritium accelerator mass spectrometry are suited for reconstructing anthropogenic tritium levels in the environment. PMID:12144257

  6. Tritium contamination experience in an operational D-T fusion reactor

    International Nuclear Information System (INIS)

    Gentile, C.A.; Ascione, G.

    1994-01-01

    During December 1993, the Tokamak Fusion Test Reactor (TFTR) injected a mixture of deuterium and tritium in the TFTR vacuum vessel for the purpose of creating D-T plasmas. The tritium used in these D-T plasmas was stored, delivered and processed in the TFTR tritium facility that includes the tritium vault, waste handling area, clean-up area, and gas holding tank room. During this time period, several components in the tritium process system were found to have tritium leaks which led to tritium deposition on process skids, components and floor area. Radiological surveys of surfaces contaminated with tritium oxide indicate a decrease in surface contamination in time (on the order of 12 to 36 hours) as the result of room ventilation. In instances where the facility HVAC system was maintained in the purge mode, a dramatic decrease in surface contamination was observed. Areas contaminated with tritium oxide (> 16.6 Bq/100 cm 2 ) were found to be clean ( 2 ) after several hours of continuous purging by the facility HVAC system. In instances where relative humidity was not decreased, the tritium surface contamination was found to be attenuated. During the months of December 1993, January and February 1994 tritium leaking components were either replaced, redesigned or repaired. During this time period, data were collected in the form of contamination surveys, real time tritium monitor output, and HVAC configuration indicating the correlation of purge ventilation leading to a decrease in tritium oxide surface contamination

  7. Biokinetic aspects of tissue-bound tritium in algae

    International Nuclear Information System (INIS)

    Strack, S.; Kistner, G.

    1978-01-01

    For the estimate of the radiation exposure of man and for the calculation of the risk of artificial tritium from nuclear power plants, organic tissue-bound tritium is of decisive importance. In model experiments, a tritium incorporation of 61 to 71% was found from tritiated water (HTO) into organic matter of planctonic algae under reproducible conditions and this was related to the theoretical value. In further experiments the tritium release from these high tritiated algae was of interest. Kept in darkness in tritium-free, non-sterile river water, so that autolytic processes and bacterial decomposition could occur, the concentration of HTO was measured over a period of three weeks. A relatively long half-life of tissue-bound tritium was found under various temperature conditions. Therefore it must be considered that a significant retention of tritium in biological matter has to be taken into account in a natural ecosystem. In streams into which the cooling water of a nuclear reactor is released all conditions are found already for a long turnover and cycling of artificial tritium in living organisms as well as the conditions for a favourable transport of tritium by food chains to man. (Auth.)

  8. Tritium in groundwater investigation at the Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    DeWilde, J.; Yu, L.; Wootton, R.; Belanger, D.; Hansen, K.; McGurk, E.; Teare, A.

    2001-01-01

    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 radionuclides, 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 identity 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)

  9. Investigation of tritium in groundwater at Pickering NGS

    International Nuclear Information System (INIS)

    DeWilde, J.; Yu, L.; Belanger, D.; Wootton, R.; Hansen, K.; McGurk, E.; Teare, A.

    2001-01-01

    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)

  10. TRIF - an intermediate approach to environmental tritium modelling

    International Nuclear Information System (INIS)

    Higgins, N.A.

    1997-01-01

    The movement of tritium through the environment, from an initial atmospheric release to selected end points in the food chain, involves a series of closely coupled and complex processes which are, consequently, difficult to model. TRIF (tritium transfer into food) provides a semi-empirical approach to this transport problem, which can be adjusted to bridge the gap between simple steady state approximations and a fully coupled model of tritium dispersion and migration (Higgins et al., 1996). TRIF provides a time-dependent description of the behaviour of tritium in the form of tritium gas (HT) and tritiated water (HTO) as it enters and moves through the food chain into pasture, crops and animals. This includes a representation of the production and movement of organically bound tritium (OBT). (Author)

  11. Apparent enrichment of organically bound tritium in rivers explained by the heritage of our past.

    Science.gov (United States)

    Eyrolle-Boyer, Frédérique; Boyer, Patrick; Claval, David; Charmasson, Sabine; Cossonnet, Catherine

    2014-10-01

    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

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

    International Nuclear Information System (INIS)

    Barbey, Pierre

    2009-06-01

    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

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

    International Nuclear Information System (INIS)

    Finn, P.A.; Reedy, G.T.; Homa, M.I.; Clemmer, R.G.; Pappas, G.; Slawecki, M.A.; Graczyk, D.G.; Bowers, D.L.; Clemmer, E.D.

    1983-01-01

    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. Chemical forms and discharge ratios to stack and sea of tritium from Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Mikami, Satoshi; Akiyama, Kiyomitsu; Miyabe, Kenjiro

    2002-03-01

    Chemical forms and discharge ratios to stack and sea of tritium form Tokai Reprocessing Plant of Japan Nuclear Cycle Development Institute (JNC) were investigated by analyzing monitoring data. It was ascertained that approximately 70-80% of tritium discharged from the main stack was tritiated water vapor (HTO) and approximately 20-30% was tritiated hydrogen (HT) as a result of analyzing the data taken from reprocessing campaign's in 1994, 1995, 1996, 1997, 2000 and 2001, and also that the amount of tritium released from the stack was less than 1% of tritium inventory in spent fuel and the amount of tritium released into sea was approximately 20-40% of inventory. (author)

  15. Confinement and Tritium Stripping Systems for APT Tritium Processing

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Heung, L.K.

    1997-10-20

    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.

  16. Confinement and Tritium Stripping Systems for APT Tritium Processing

    International Nuclear Information System (INIS)

    Hsu, R.H.; Heung, L.K.

    1997-01-01

    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

  17. Environmental tritium in trees

    International Nuclear Information System (INIS)

    Brown, R.M.

    1979-01-01

    The distribution of environmental tritium in the free water and organically bound hydrogen of trees growing in the vicinity of the Chalk River Nuclear Laboratories (CRNL) has been studied. The regional dispersal of HTO in the atmosphere has been observed by surveying the tritium content of leaf moisture. Measurement of the distribution of organically bound tritium in the wood of tree ring sequences has given information on past concentrations of HTO taken up by trees growing in the CRNL Liquid Waste Disposal Area. For samples at background environmental levels, cellulose separation and analysis was done. The pattern of bomb tritium in precipitation of 1955-68 was observed to be preserved in the organically bound tritium of a tree ring sequence. Reactor tritium was discernible in a tree growing at a distance of 10 km from CRNL. These techniques provide convenient means of monitoring dispersal of HTO from nuclear facilities. (author)

  18. Tritium monitoring techniques

    International Nuclear Information System (INIS)

    DeVore, J.R.; Buckner, M.A.

    1996-05-01

    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

  19. FDMH - The tritium model in RODOS

    International Nuclear Information System (INIS)

    Galeriu, D.; Mateescu, G.; Melintescu, A.; Turcanu, C.; Raskob, W.

    2000-01-01

    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

  20. Modeling of tritium behavior in Li2O

    International Nuclear Information System (INIS)

    Billone, M.C.; Attaya, H.; Kopasz, J.P.

    1992-08-01

    The TIARA and DISPL2 codes are being developed at Argonne National Laboratory to predict tritium retention and release from lithium ceramics under steady-state and transient conditions, respectively. Tritium retention and release are important design and safety issues for tritium-breeding blankets of fusion reactors. Emphasis has been placed on tritium behavior in Li 2 O because of the selection of this ceramic as a first option for the ITER driver blanket and because of the relatively good material properties data base for Li 2 O. Models and correlations for diffusion, surface desorption/adsorption, and solubility/precipitation of tritium in Li 2 0 have been developed based on well-controlled laboratory data from as-fabricated and irradiated samples. With the models and correlations, the codes are validated to the results of in-reactor purge flow tests. The results of validation of TIARA to tritium retention data from VOM-15H, EXOTIC-2, and CRITIC-1 are presented, along with predictions of tritium retention in BEATRIX-II. For DISPL2, results are presented for tritium release predictions vs. data for MOZART, CRITIC-1, and BEATRIX-II. Recommendations are made for improving both the data base and the modeling to allow extrapolation with reasonable uncertainty levels to fusion reactor design conditions

  1. Effect of etorphine on the spontaneous and field stimulation-mediated release of norepinephrine and total tritium from perfused guinea pig hearts

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Isolated guinea pig hearts were prelabeled with 3 H-norepinephrine ( 3 H-NE) and perfused with modified Krebs-bicarbonate solution at 37 0 C. Spontaneous release of total 3 H and field stimulation-mediated (supramax. V., 1 Hz, 2 msec duration for a total of 60 pulses) overflow of NE and 3 H-NE were measured in the absence or presence of etorphine. Etorphine (0.1 - 100 μM) was added to the perfusion fluid 15 min. before the stimulation. To study the effect of etorphine on spontaneous release of total 3 H, etorphine was added cumulatively without stimulation. Etorphine (1.0 - 100 μM) caused a significant decrease in the stimulation-mediated overflow of NE and the inhibition was dose-related. The overflow of NE was 5.1 +/- 0.3 ng in the absence and 4.0 +/- 0.2 ng in the presence of etorphine (1.0 μM). Low concentrations of etorphine (0.1 - 1.0 μM) had no effect on the spontaneous release of total 3 H while 10 μM and 100 μM caused a 3 and 6-fold increase respectively. The results show that etorphine inhibited neuronal release of NE at a dose which had no effect on spontaneous release. It is suggested that opiate receptors might be involved in the prejunctional modulation of the release of NE in the guinea pig heart

  2. Develop of omni-tritium sample preparation device

    International Nuclear Information System (INIS)

    Tian Junhua; Zheng Min; Zhang Dong

    2008-06-01

    The content of total tritium analysis is required in order to know the tritium contaminated degree of biological samples accurately. But the conversion and collection of organic tritium are difficult. A device to treat total tritium samples was developed. Plant samples were treated by combustion and catalysis. After expelling the free HTO in the samples when heated in abundant oxygen, the samples were ignited. Combustion gas passed the catalysts at 800 degree C and its oxidation was catalyzed, and then the combined tritium in tissues was converted into HTO. HTO was collected by water-cooling tube and condenser. For other samples, HTO was treated and collected by high temperature (The highest temperature is 1000 degree C)-catalysis-double condensation method. This device had solved the problem that organic tritium is difficult to gather. (authors)

  3. A model for global cycling of tritium

    International Nuclear Information System (INIS)

    Killough, G.G.; Kocher, D.C.

    1988-01-01

    Dynamic compartment models are widely used to describe global cycling of radionuclides for purposes of dose estimation. In this paper the authors present a new global tritium model that reproduces environmental time-series data on concentrations in precipitation, ocean surface waters, and surface fresh waters in the northern hemisphere, concentrations of atmospheric tritium in the southern hemisphere, and the latitude dependence of tritium in both hemispheres. Names TRICYCLE (for TRItium CYCLE) the model is based on the global hydrologic cycle and includes hemispheric stratospheric compartments, disaggregation of the troposphere and ocean surface waters into eight latitude zones, consideration of the different concentrations of atmospheric tritium over land and over the ocean, and a diffusive model for transport in the ocean. TRICYCLE reproduces the environmental data if it is assumed that about 50% of the tritium from atmospheric weapons testing was injected directly into the northern stratosphere as HTO. The model's latitudinal disaggregation permits taking into account the distribution of population. For a uniformly distributed release of HTO into the worldwide troposphere, TRICYCLE predicts a collective dose commitment to the world population that exceeds the NCRP model's corresponding prediction by a factor of three

  4. A model for global cycling of tritium

    International Nuclear Information System (INIS)

    Killough, G.G.; Kocher, D.C.

    1988-01-01

    Dynamic compartment models are widely used to describe global cycling of radionuclides for purposes of dose estimation. In this paper, we present a new global tritium model that reproduces environmental time-series data on concentrations in precipitation, ocean surface waters, and surface fresh waters in the northern hemisphere, concentrations of atmospheric tritium in the soutehrn hemisphere, and the latitude dependence of tritium in both hemispheres. Named TRICYCLE for Tritium CYCLE, the model is based on the global hydrologic cycle and includes hemisphereic stratospheric compartments, disaggregation of the troposphere and ocean surface waters into eight latitudezones, consideration of the different concentrations of atmospheric tritium over land and over the ocean, and a diffusive model for transport in the ocean. TRICYCLE reproduces the environmental data if we assume that about 50% of the tritium from atmospheric weapons testing was injected directly into the northern stratosphere as HTO. The models latitudinal disaggregation permits taking into account the distribution of population. For a unfiormaly distributed release of HTO into the worldwide troposphere, TRICYCLE predicts a collective dose commitment to the world population that exceeds the corresponding prediction by the NCRP model by about a factor of 3. 11 refs., 5 figs., 1 tab

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

    International Nuclear Information System (INIS)

    Galeriu, D.; Melintescu, A.; Crout, N. M. J.; Bersford, N. A.; Peterson, S. R.; Hess, M. van

    2001-01-01

    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

  6. Tritium in metals

    International Nuclear Information System (INIS)

    Schober, T.

    1990-01-01

    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 3 He in the samples. (orig.)

  7. Tritium sources; Izvori tricijuma

    Energy Technology Data Exchange (ETDEWEB)

    Glodic, S [Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Boreli, F [Elektrotehnicki fakultet, Belgrade (Yugoslavia)

    1993-07-01

    Tritium is the only radioactive isotope of hydrogen. It directly follows the metabolism of water and it can be bound into genetic material, so it is very important to control levels of contamination. In order to define the state of contamination it is necessary to establish 'zero level', i.e. actual global inventory. The importance of tritium contamination monitoring increases with the development of fusion power installations. Different sources of tritium are analyzed and summarized in this paper. (author)

  8. High-pressure tritium

    International Nuclear Information System (INIS)

    Coffin, D.O.

    1976-01-01

    Some solutions to problems of compressing and containing tritium gas to 200 MPa at 700 0 K are discussed. The principal emphasis is on commercial compressors and high-pressure equipment that can be easily modified by the researcher for safe use with tritium. Experience with metal bellows and diaphragm compressors has been favorable. Selection of materials, fittings, and gauges for high-pressure tritium work is also reviewed briefly

  9. Tritium Research Laboratory safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    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)

  10. Tritium Research Laboratory safety analysis report

    International Nuclear Information System (INIS)

    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

  11. On the fission gas release from oxide fuels during normal grain growth

    International Nuclear Information System (INIS)

    Paraschiv, M.C.; Paraschiv, A.; Glodeanu, F.

    1997-01-01

    A mathematical formalism for calculating the fission gas release from oxide fuels considering an arbitrary distribution of fuel grain size with only zero boundary condition for gas diffusion at the grain boundary is proposed. It has also been proved that it becomes unnecessary to consider the grain volume distribution function for fission products diffusion when the grain boundary gas resolution is considered, if thermodynamic forces on grain boundaries are only time dependent. In order to highlight the effect of the normal grain growth on fission gas release from oxide fuels Hillert's and Lifshitz and Slyozov's theories have been selected. The last one was used to give an adequate treatment of normal grain growth for the diffusion-controlled grain boundary movement in oxide fuels. It has been shown that during the fuel irradiation, the asymptotic form of the grain volume distribution functions given by Hillert and Lifshitz and Slyozov models can be maintained but the grain growth rate constant becomes time dependent itself. Experimental results have been used to correlate the two theoretical models of normal grain growth to the fission gas release from oxide fuels. (orig.)

  12. A model for the release of low-volatility fission products in oxidizing conditions

    International Nuclear Information System (INIS)

    Cox, D.S.; Hunt, C.E.L.; Liu, Z.; Keller, N.A.; Barrand, R.D.; O'Connor, R.F.

    1991-07-01

    A thermodynamic and kinetic model has been developed for calculating low-volatility fission-product releases from UO 2 at high temperatures in oxidizing conditions. Volatilization of the UO 2 matrix is assumed to be the rate controlling process. Oxidation kinetics of the UO 2 are modelled by either interfacial rate control, gas phase oxidant transport control, or solid-state diffusion of oxygen. The vapour pressure of UO 3 in equilibrium with the oxidizing fuel is calculated from thermodynamic data, and volatilization rates are determined using a model for forced convective mass transport. Low-volatility fission-product releases are calculated from the volume of vapourized fuel. Model calculations are conservative compared to experimental data for Zr, La, Ce and Nb fission-product releases from irradiated UO 2 exposed to air at 1973-2350 K. The implications of this conservatism are discussed in terms of possible rate control by processes other than convective mass transport of UO 3 . Coefficients for effective surface area (based on experimental data) and for heterogeneous rate controlling reaction kinetics are introduced to facilitate agreement between calculations and the experimental data.

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

    International Nuclear Information System (INIS)

    Kobayashi, K.; Miya, N.; Ikeda, Y.; Torikai, Y.; Saito, M.; Alimov, V.

    2015-01-01

    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 2 (92.8 %) - T 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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Problems bound to the tritium in materials for the nuclear - some illustrations; Problematiques liees au tritium dans les materiaux dans le domaine nucleaire - quelques illustrations

    Energy Technology Data Exchange (ETDEWEB)

    Gastaldi, O. [CEA Cadarache (DTN/STPA/LPC), 13 - Saint-Paul-lez-Durance (France)

    2007-07-01

    The tritium control takes more and more importance in the nuclear industry because of the release more and more limited, in the environment. After a presentation on the tritium sources in the environment, the author presents the different ways of its production. Then for each reactor channel, the main problems are presented (fission and fusion). The last part deals with the behavior of the tritium in materials: the tritium inventory control in a fusion system, the tritium management after the reactor exploitation. (A.L.B.)

  16. Incorporation of zinc oxide to dispersions of biopolymers and release of the metallic ion in vitro

    International Nuclear Information System (INIS)

    Barreto, Marina S.R.; Ferreira, Willian H.; Andrade, Cristina T.

    2015-01-01

    Zinc oxide (ZnO) nanoparticles, obtained from a commercial product, were dispersed in different biopolymers, to be added to piglet feeds. The resulting products, prepared with sodium alginate (SA), chitosan (CH) and low methoxyl pectin (LMP) were characterized by X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FTIR) and scanning electron microscopy (SEM). The release of Zn"2"+ was investigated under simulated conditions of the gastrointestinal tract of piglets, and analyzed by atomic absorption spectroscopy (AA). The results showed that the structural factors, which have influence on the biopolymer/ZnO interactions, govern the behavior of Zn"2"+ release. (author)

  17. Tritium waste control: April-September 1982

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Rogers, M.L.

    1983-01-01

    The pilot-scale, water feed cleanup system was used to successfully remove organic and inorganic impurities from Effluent Removal System (ERS) water. Tests with activated carbon traps removed organic impurities to as low as 2.5 ppM total carbon. Traps containing Amberlite resins for removing organic impurities were not successful and actually contaminated the water with higher levels (>2000 ppM) of organics. Gas generation rates caused by radiolysis of tritiated waste materials were determined for polymer and nonpolymer-impregnated tritiated concrete and fixated and nonfixated tritiated waste vacuum pump oil. In addition, the pressure change of hydrogen cover gas over tritiated water on cement-plaster was determined. The test program to measure and compare the release of tritium from tritiated concrete with and without styrene impregnation continued. Tritium permeation data from small test blocks are given. The drum study monitoring the release of tritium from actual burial packages continued. The maximum fractional release rate for the three types of high activity, tritiated liquid waste generated is 2.97 x 10 -5 , and the maximum total permeation is 158 mCi after 8 yr. These two values represent a 13% increase for the past 6 months. Tritium release from the polymer-impregnated, tritiated concrete (PITC) and from the control (non-PITC) remains very low

  18. Tritium liquid effluents from the Krsko NPP

    International Nuclear Information System (INIS)

    Savli, S.; Krizman, M.; Nemec, T.; Cindro, M.; Stritar, A.; Vokal Nemec, B.; Janzekovic, H.

    2007-01-01

    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)

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

    International Nuclear Information System (INIS)

    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

  20. Preparation of Tritium from irradiated lithium compounds (study)

    International Nuclear Information System (INIS)

    1989-01-01

    The aim of the present study is the selection of a certain scheme for release, collection, measurement and analysis of Tritium as produced in Lithium compounds (Li 2 O, LiOH, Li 2 CO 3 LiF and some Lithium alloys) in accordance with the 6 Li(n, α)T reaction. Tritium technology is of vital concern to power reactor programmes as well as to fusion technology. Meanwhile the fields of activity include, tritium generation and mangement in fission and fusion reactors; enviromental studies, release modeling HT/HTO conversion and dose assessments, absorption/ deabsorption, monitoring and plant design; research and development, labeling, compatibility and physical chemical properties

  1. Tritium in plants; Le tritium dans la matiere organique des vegetaux

    Energy Technology Data Exchange (ETDEWEB)

    Vichot, L.; Losset, Y. [CEA Valduc, 21 - Is-sur-Tille (France)

    2009-07-01

    The presence of tritium in the environment stems from its natural production by cosmic rays, from the fallout of the nuclear weapon tests between 1953 and 1964, and locally from nuclear industry activities. A part of the tritiated water contained in the foliage of plants is turned into organically bound tritium (OBT) by photosynthesis. The tritium of OBT, that is not exchangeable and then piles up in the plant, can be used as a marker of the past. It has been shown that the quantity of OBT contained in the age-rings of an oak that grew near the CEA center of Valduc was directly correlated with the tritium releases of the center. (A.C.)

  2. Increased norepinephrine release from dog pulmonary artery caused by nitrous oxide

    International Nuclear Information System (INIS)

    Rorie, D.K.; Tyce, G.M.; Sill, J.C.

    1986-01-01

    The effects of nitrous oxide on the release and metabolism of norepinephrine (NE) at neuroeffector junctions in dog pulmonary artery were examined. Helical strips of artery were incubated in Krebs-Ringer solution containing L-( 3 H)NE and mounted for superfusion. The arterial strips were studied in the presence of 95% oxygen-5% carbon dioxide, 70% nitrogen-30% oxygen, or 70% nitrous oxide-30% oxygen. During the 60 min of each experiment, five samples of superfusion fluid were collected for analysis and the effluxes of ( 3 H)NE and its radiolabeled metabolites were measured before and during electrical stimulation and during recovery from stimulation. ( 3 H)Norepinephrine was separated from its metabolites in the superfusate and in extracts of artery by column chromatography and quantitated by liquid scintillation spectrometry. Nitrous oxide significantly increased the fractional loss of total radioactivity and the amount of NE in the superfusate both during resting conditions and during stimulation. Nitrous oxide had no effect on the proportions of radioactivity among metabolites of NE in the superfusate or on the profile of NE metabolites remaining in the tissue after experimentation. These findings are consistent with increased NE release as a direct effect of nitrous oxide on nerve endings

  3. Caffeinated nitric oxide-releasing lozenge improves cycling time trial performance.

    Science.gov (United States)

    Lee, J; Kim, H T; Solares, G J; Kim, K; Ding, Z; Ivy, J L

    2015-02-01

    Boosting nitric oxide production during exercise by various means has been found to improve exercise performance. We investigated the effects of a nitric oxide releasing lozenge with added caffeine (70 mg) on oxygen consumption during steady-state exercise and cycling time trial performance using a double-blinded randomized, crossover experimental design. 15 moderately trained cyclists (7 females and 8 males) were randomly assigned to ingest the caffeinated nitric oxide lozenge or placebo 5 min before exercise. Oxygen consumption and blood lactate were assessed at rest and at 50%, 65% and 75% maximal oxygen consumption. Exercise performance was assessed by time to complete a simulated 20.15 km cycling time-trial course. No significant treatment effects for oxygen consumption or blood lactate at rest or during steady-state exercise were observed. However, time-trial performance was improved by 2.1% (p<0.01) when participants consumed the nitric oxide lozenge (2,424±69 s) compared to placebo (2,476±78 s) and without a significant difference in rating of perceived exertion. These results suggest that acute supplementation with a caffeinated nitric oxide releasing lozenge may be a practical and effective means of improving aerobic exercise performance. © Georg Thieme Verlag KG Stuttgart · New York.

  4. Tritium control in fusion reactor materials: A model for Tritium Extracting System

    Energy Technology Data Exchange (ETDEWEB)

    Zucchetti, Massimo [DENERG, Politecnico di Torino (Italy); Utili, Marco, E-mail: marco.utili@enea.it [ENEA UTIS – C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino (Italy); Ying, Alice [University of California Los Angeles (UCLA), Los Angeles, CA (United States); Franza, Fabrizio [Karlsruhe Institute of Technology, Karlsruhe (Germany); Abdou, Mohamed [University of California Los Angeles (UCLA), Los Angeles, CA (United States)

    2015-10-15

    Highlights: • A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a Molecular sieve as adsorbent material. • A computational model has been setup and tested in this paper. • The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. • It turns out the capability to model the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT). - Abstract: In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT).

  5. Tritium control in fusion reactor materials: A model for Tritium Extracting System

    International Nuclear Information System (INIS)

    Zucchetti, Massimo; Utili, Marco; Nicolotti, Iuri; Ying, Alice; Franza, Fabrizio; Abdou, Mohamed

    2015-01-01

    Highlights: • A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a Molecular sieve as adsorbent material. • A computational model has been setup and tested in this paper. • The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. • It turns out the capability to model the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT). - Abstract: In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT).

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

    International Nuclear Information System (INIS)

    Oestlund, H.G.; Mason, A.S.

    1985-04-01

    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 T 2 , 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

  7. Radionuclide Basics: Tritium

    Science.gov (United States)

    Tritium is a hydrogen atom that has two neutrons in the nucleus and one proton. It is radioactive and behaves like other forms of hydrogen in the environment. Tritium is produced naturally in the upper atmosphere and as a byproduct of nuclear fission.

  8. Tritium waste control: April--September 1977

    International Nuclear Information System (INIS)

    1978-01-01

    A pilot scale system was used in an initial experiment to investigate the combined-electrolysis-catalytic-exchange process (CECE) for the detritiation of water. Data taken during the experiment indicate the process does indeed strip tritium from gaseous hydrogen at the top and concentrate it in water at the bottom of the catalyst-filled column. A high activity tritiated water electrolysis system was designed and built using a solid polymer electrolyte (SPE) cell. The system was successfully operated at currents up to 50 A using deionized tap water. Triplicate samples of cement, cement-plaster (1:1 ratio by weight), and cement-plaster (1:1 ratio by volume) were injected with 386 Ci of tritium. Preliminary results indicate Type III Portland cement retains the tritium slightly better than the cement-plaster mixtures. The tritium release study of actual waste burial packages is continuing. The fractional release is 1 x 10 -5 on a 4-y old package, only 4 x 10 -7 on 3-y old packages, and 1 x 10 -9 on a 1-y old package. Pressure increase and gas composition determinations were repeated for octane (activity = 1000 Ci/liter) with and without tritium fixation using argon as the initial overgas. Pressure buildup measurements for octane without fixation were repeated a third time using hydrogen gas. The rate of pressure increase did not change significantly from previously determined values. Four elemental tritium samples were released into a laboratory to determine the efficiency of the Emergency Containment System. The ventilation system was modified during the fourth experiment to minimize leakage

  9. In-vessel tritium

    International Nuclear Information System (INIS)

    Ueda, Yoshio; Ohya, Kaoru; Ashikawa, Naoko; Ito, Atsushi M.; Kato, Daiji; Kawamura, Gakushi; Takayama, Arimichi; Tomita, Yukihiro; Nakamura, Hiroaki; Ono, Tadayoshi; Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Nakano, Tomohide; Nakamura, Makoto; Hoshino, Kazuo; Kenmotsu, Takahiro; Wada, Motoi; Saito, Seiki; Takagi, Ikuji; Tanaka, Yasunori; Tanabe, Tetsuo; Yoshida, Masafumi; Toma, Mitsunori; Hatayama, Akiyoshi; Homma, Yuki; Tolstikhina, Inga Yu.

    2012-01-01

    The in-vessel tritium research is closely related to the plasma-materials interaction. It deals with the edge-plasma-wall interaction, the wall erosion, transport and re-deposition of neutral particles and the effect of neutral particles on the fuel recycling. Since the in-vessel tritium shows a complex nonlinear behavior, there remain many unsolved problems. So far, behaviors of in-vessel tritium have been investigated by two groups A01 and A02. The A01 group performed experiments on accumulation and recovery of tritium in thermonuclear fusion reactors and the A02 group studied theory and simulation on the in-vessel tritium behavior. In the present article, outcomes of the research are reviewed. (author)

  10. Protection against tritium radiations

    International Nuclear Information System (INIS)

    Bal, Georges

    1964-05-01

    This report presents the main characteristics of tritium, describes how it is produced as a natural or as an artificial radio-element. It outlines the hazards related to this material, presents how materials and tools are contaminated and decontaminated. It addresses the issue of permissible maximum limits: factors of assessment of the risk induced by tritium, maximum permissible activity in body water, maximum permissible concentrations in the atmosphere. It describes the measurement of tritium activity: generalities, measurement of gas activity and of tritiated water steam, tritium-induced ionisation in an ionisation chamber, measurement systems using ionisation chambers, discontinuous detection of tritium-containing water in the air, detection of surface contamination [fr

  11. Circulating nitric oxide products do not solely reflect nitric oxide release in cirrhosis and portal hypertension

    DEFF Research Database (Denmark)

    Afzelius, Pia; Bazeghi, Nassim; Bie, Peter

    2011-01-01

    Patients with cirrhosis often develop a systemic vasodilatation and a hyperdynamic circulation with activation of vasoconstrictor systems such as the renin-angiotensin-aldosterone system (RAAS), and vasopressin. Increased nitric oxide (NO) synthesis has been implicated in the development of this ...

  12. Tritium Systems Test Assembly operator training program

    International Nuclear Information System (INIS)

    Kerstiens, F.L.

    1985-01-01

    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

  13. Nitric oxide donors enhance the frequency dependence of dopamine release in nucleus accumbens.

    Science.gov (United States)

    Hartung, Henrike; Threlfell, Sarah; Cragg, Stephanie J

    2011-08-01

    Dopamine (DA) neurotransmission in the nucleus accumbens (NAc) is critically involved in normal as well as maladaptive motivated behaviors including drug addiction. Whether the striatal neuromodulator nitric oxide (NO) influences DA release in NAc is unknown. We investigated whether exogenous NO modulates DA transmission in NAc core and how this interaction varies depending on the frequency of presynaptic activation. We detected DA with cyclic voltammetry at carbon-fiber microelectrodes in mouse NAc in slices following stimuli spanning a full range of DA neuron firing frequencies (1-100 Hz). NO donors 3-morpholinosydnonimine hydrochloride (SIN-1) or z-1-[N-(3-ammoniopropyl)-N-(n-propyl)amino]diazen-1-ium-1,2-diolate (PAPA/NONOate) enhanced DA release with increasing stimulus frequency. This NO-mediated enhancement of frequency sensitivity of DA release was not prevented by inhibition of soluble guanylyl cyclase (sGC), DA transporters, or large conductance Ca(2+)-activated K(+) channels, and did not require glutamatergic or GABAergic input. However, experiments to identify whether frequency-dependent NO effects were mediated via changes in powerful acetylcholine-DA interactions revealed multiple components to NO modulation of DA release. In the presence of a nicotinic receptor antagonist (dihydro-β-erythroidine), NO donors increased DA release in a frequency-independent manner. These data suggest that NO in the NAc can modulate DA release through multiple GC-independent neuronal mechanisms whose net outcome varies depending on the activity in DA neurons and accumbal cholinergic interneurons. In the presence of accumbal acetylcholine, NO promotes the sensitivity of DA release to presynaptic activation, but with reduced acetylcholine input, NO will promote DA release in an activity-independent manner through a direct action on dopaminergic terminals.

  14. Prediction of tritium behavior in rice plant after a short-term exposure of HTO

    International Nuclear Information System (INIS)

    Yook, Dae Sik; Lee, Kun Jai; Choi, Heui Joo; Lee, Chang Min

    2001-01-01

    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. Because four CANDU reactors are in operation in Korea, relatively large amounts of tritium are released into the environment and the dose by these tritium in the rice plant must be estimated. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out experimental studies to obtain domestic data on various parameters related to the direct tritium contamination of plant. But the analysis of the tritium behavior in the rice plant has been insufficient. In this study, the behavior of the 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, transient tritium concentrations in soil and leaves were predicted. If the effect of tritium concentration in the soil is taken into account, the tritium concentration in leaves can be described by a double exponential model, however if the tritium concentration in the soil is disregarded, the tritium concentration in leaves can be described by a single exponential term like other relevant models e.g. UFOTRI or STAR-H3 model. The results can be used to predict the tritium concentration in the rice plant near the plant site and to estimate the ingestion dose after the release of tritium to the environment

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

    International Nuclear Information System (INIS)

    Iyengar, T.S.; Sadarangani, S.H.; Vaze, P.K.; Soman, S.D.

    1977-01-01

    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)

  16. Tritium waste management on the La Hague AREVA NC site: associated impact and monitoring

    International Nuclear Information System (INIS)

    Devin, P.; Deguette, H.

    2009-01-01

    The authors propose an analysis of tritium behaviour in the nuclear fuel processed in the AREVA NC plant in La Hague, of its presence in the plant and in its wastes, and of the impact of these wastes and the tritium monitoring in the environment. First, they present the AREVA NC plant and evoke the legal context concerning the waste management. They report and discuss the analysis of the presence and behaviour of tritium in irradiated fuel, of its behaviour during spent fuel processing, the evolution of tritium releases (legal limitations, evolutions since 1992), of measurement of activity in effluents, and discuss a study of possible reductions of tritium releases by La Hague plants (mainly in sea waters). They also report the computational assessment of the dosimetric impact of tritium on neighbouring population. They describe how the presence of tritium in the environment is monitored within the annual radioactivity monitoring programme

  17. Direct measurement of tritium in urine by liquid scintillation method

    International Nuclear Information System (INIS)

    Zhang Caihong; Wen Qinghua; Chen Kefei; Li Huaixin

    1999-01-01

    The author introduces the method for direct measurement of tritium concentration in urine using liquid scintillation. Effects of sampling containers, store patterns and storage time are studied. Meanwhile, results of two methods are compared with direct measurement method and oxidation distillation method. The results shows that direct measurement method is a economic and simple method, which can meet the need of determination of urine tritium for NPP workers. There is no significant difference compared with the data obtained by oxidation distillation method

  18. ANCLI's conclusions and recommendations made after the ANCLI colloquium 'Tritium, discrete, but present everywhere'

    International Nuclear Information System (INIS)

    Sene, M.

    2009-01-01

    The authors briefly state the conclusions of the colloquium about the presence of tritium in the environment, its sanitary impact, the re-examination of a management based on release, the need to reduce tritium production. The recommendations are also indicated: to continue researches on organically bound tritium, not to allow any release increase as long as effects are not better known. The role of the ANCLI is outlined

  19. Production sources and management of tritium generated by nuclear facilities. Works reporting from the reflection group 'tritium defense-in-depth'

    International Nuclear Information System (INIS)

    2001-01-01

    This report synthesizes current knowledge of tritium, its sources of tritium and its behavior in reactors and nuclear facilities. In addition, liquid and gaseous tritiated releases and the processes of separation and possible trapping of this radionuclide are displayed. This report also provides elements of reflection on the management of tritium produced today by these facilities, both in France and abroad. It presents insights on the relevance of current management modes of tritiated liquid and solid waste and also on the associated measures to reduce releases of tritium into the environment. (author)

  20. Tritium breeding in fusion reactors

    International Nuclear Information System (INIS)

    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

  1. Modeling of tritium behavior in ceramic breeder materials

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Tam, S.W.; Johnson, C.E.

    1988-11-01

    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

  2. Dependence of CuO particle size and diameter of reaction tubing on tritium recovery for tritium safety operation

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Cui, E-mail: cdxohc10000@163.com [Shizuoka University, 836 Ohya, Suruga-ku Shizuoka 422-8529 (Japan); Uemura, Yuki; Yuyama, Kenta; Fujita, Hiroe; Sakurada, Shodai; Azuma, Keisuke [Shizuoka University, 836 Ohya, Suruga-ku Shizuoka 422-8529 (Japan); Taguchi, Akira; Hara, Masanori; Hatano, Yuji [University of Toyama, 3190 Gofuku, Toyama 939-8555 (Japan); Chikada, Takumi; Oya, Yasuhisa [Shizuoka University, 836 Ohya, Suruga-ku Shizuoka 422-8529 (Japan)

    2016-12-15

    Highlights: • Influence of CuO particle size and diameter of reaction tubing on the tritium recovery was evaluated. • Reaction rate constant of tritium with CuO particle has been calculated by the combination of experimental results and a simulation code. • Dependence of reaction tubing length on tritium conversion ratio has been explored. - Abstract: Usage of CuO and water bubbler is one of the conventional and convenient methods for tritium recovery. In present work, influence of CuO particle size and diameter of reaction tubing on the tritium recovery was evaluated. Reaction rate constant of tritium with CuO particle has been calculated by the combination of experimental results and a simulation code. Then, these results were applied for exploring the dependence of reaction tubing length on tritium conversion ratio. The results showed that the surface area of CuO has a great influence on the oxidation rate constant. The frequency factor of the reaction would be approximately doubled by reducing the CuO particle size from 1.0 mm to 0.2 mm. Cross section of reaction tubing mainly affected on the duration of tritium at the temperature below 600 K. Reaction tubing with length of 1 m at temperature of 600 K would be suitable for keeping the tritium conversion ratio above 99.9%. The length of reaction tubing can be reduced by using the smaller CuO particle or increasing the CuO temperature.

  3. Tritium trapping states induced by lithium-depletion in Li{sub 2}TiO{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Makoto, E-mail: kobayashi.makoto@LHD.nifs.ac.jp [National Institute for Fusion Science (Japan); Oya, Yasuhisa; Okuno, Kenji [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan)

    2017-04-15

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

  4. Tritium permeation through iron

    International Nuclear Information System (INIS)

    Hagi, Hideki; Hayashi, Yasunori

    1989-01-01

    An experimental method for measuring diffusion coefficients and permeation rates of tritium in metals around room temperature has been established, and their values in iron have been obtained by using the method. Permeation rates of tritium and hydrogen through iron were measured by the electrochemical method in which a tritiated aqueous solution was used as a cathodic electrolyte. Tritium and hydrogen were introduced from one side of a membrane specimen by cathodic polarization, while at the other side of the specimen the permeating tritium and hydrogen were extracted by potentiostatical ionization. The amount of permeated hydrogen was obtained by integrating the anodic current, and that of tritium was determined by measuring the radioactivity of the electrolyte sampled from the extraction side. Diffusion coefficients of tritium (D T ) and hydrogen (D H ) were determined from the time lag of tritium and hydrogen permeation. D T =9x10 -10 m 2 /s and D H =4x10 -9 m 2 /s at 286 K for annealed iron specimens. These values of D T and D H were compared with the previous data of the diffusion coefficients of hydrogen and deuterium, and the isotope effect in diffusion was discussed. (orig.)

  5. Tritium permeation barriers for fusion technology

    International Nuclear Information System (INIS)

    Perujo, A.; Forcey, K.

    1994-01-01

    An important issue concerning the safety, feasibility and fueling (i.e., tritium breeding ratio and recovery from the breeding blanket) of a fusion reactor is the possible tritium leakages through the structural materials and in particular through those operating at high temperatures. The control of tritium permeation could be a critical factor in determining the viability of a future fusion power reactor. The formation of tritium permeation barriers to prevent the loss of tritium to the coolant by diffusion though the structural material seems to be the most practical method to minimize such losses. Many authors have discussed the formation of permeation barriers to reduce the leakage of hydrogen isotopes through proposed first wall and structural materials. In general, there are two routes for the formation of such a barrier, namely: the growth of oxide layers (e.g., Cr 2 O 3 , Al 2 O 3 , etc.) or the application of surface coatings. Non-metals are the most promising materials from the point of view of the formation of permeation barriers. Oxides such as Al 2 O 3 or Cr 2 O 3 or carbides such as SiC or TiC have been proposed. Amongst the metals only tungsten or gold are sufficiently less permeable than steel to warrant investigation as candidate materials for permeation barriers. It is of course possible to grow oxide layers on steel directly by heating in the atmosphere or under a variety of conditions (first route above). The direct oxidizing is normally done in an environment of open-quotes wet hydrogenclose quotes to promote the growth of chromia on, for example, nickel steels or ternary oxides on 316L to prevent corrosion. The application of surface layers (second route above), offers a greater range of materials for the formation of permeation barriers. In addition to reducing permeation, such layers should be adhesive, resistant to attack by corrosive breeder materials and should not crack during thermal cycling

  6. Tritium technology. A Canadian overview

    Energy Technology Data Exchange (ETDEWEB)

    Hemmings, R.L. [Canatom NPM (Canada)

    2002-10-01

    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)

  7. Tritium technology. A Canadian overview

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    2002-01-01

    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. Metabolism and dosimetry of tritium

    International Nuclear Information System (INIS)

    Hill, R.L.; Johnson, J.R.

    1993-01-01

    This document was prepared as a review of the current knowledge of tritium metabolism and dosimetry. The physical, chemical, and metabolic characteristics of various forms of tritium are presented as they pertain to performing dose assessments for occupational workers and for the general public. For occupational workers, the forms of tritium discussed include tritiated water, elemental tritium gas, skin absorption from elemental tritium gas-contaminated surfaces, organically bound tritium in pump oils, solvents and other organic compounds, metal tritides, and radioluminous paints. For the general public, age-dependent tritium metabolism is reviewed, as well as tritiated water, elemental tritium gas, organically bound tritium, organically bound tritium in food-stuffs, and tritiated methane. 106 refs

  9. Improvement in energy release properties of boron-based propellant by oxidant coating

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Daolun; Liu, Jianzhong, E-mail: jzliu@zju.edu.cn; Chen, Binghong; Zhou, Junhu; Cen, Kefa

    2016-08-20

    Highlights: • NH{sub 4}ClO{sub 4}, KNO{sub 3}, KClO{sub 4} and HMX coated B were used to prepare propellant samples. • FTIR, XRD and SEM were used for the microstructure analysis of the prepared B. • Thermal oxidation and combustion characteristics of the propellants were studied. • HMX coating was the most beneficial to the energy release of the samples. - Abstract: The energy release properties of a propellant can be improved by coating boron (B) particles with oxidants. In the study, B was coated with four different oxidants, namely, NH{sub 4}ClO{sub 4}, KNO{sub 3}, LiClO{sub 4}, and cyclotetramethylenetetranitramine (HMX), and the corresponding propellant samples were prepared. First, the structural and morphological analyses of the pretreated B were carried out. Then, the thermal analysis and laser ignition experiments of the propellant samples were carried out. Coating with NH{sub 4}ClO{sub 4} showed a better performance than mechanical mixing with the same component. Coating with KNO{sub 3} efficiently improved the ignition characteristics of the samples. Coating with LiClO{sub 4} was the most beneficial in reducing the degree of difficulty of B oxidation. Coating with HMX was the most beneficial in the heat release of the samples. The KNO{sub 3}-coated sample had a very high combustion intensity in the beginning, but then it rapidly became weak. Large amounts of sparks were ejected during the combustion of the LiClO{sub 4}-coated sample. The HMX-coated sample had the longest self-sustaining combustion time (4332 ms) and the highest average combustion temperature (1163.92 °C).

  10. Tritium and helium-3 in metals

    International Nuclear Information System (INIS)

    Lasser, R.

    1989-01-01

    The book surveys recent results on the behaviour of tritium and its decay product helium-3 metals. In contrast to many earlier books which discuss the properties of the stable hydrogen isotopes without mentioning tritium, this book reviews mainly the results on tritium in metals. Due to the difficulties in preparing metal tritide samples, very important quantities such as diffusivity, superconductivity, solubility, etc. have only been determined very recently. The book not only presents the measured tritium data, but also the isotopic dependency of the different physical properties by comparing H, D and T results. A chapter is devoted to helium-3 in metals. Aspects such as helium release, generation of helium bubbles, swelling, and change of the lattice parameter upon aging are discussed. The book provides the reader with up-to-date information and deep insight into the behaviour of H, D, T and He-3 in metals. Further important topics such a tritium production, its risks, handling and discharge to the environment are also addressed

  11. Metal ion release from metallothioneins: proteolysis as an alternative to oxidation.

    Science.gov (United States)

    Peroza, Estevão A; dos Santos Cabral, Augusto; Wan, Xiaoqiong; Freisinger, Eva

    2013-09-01

    Metallothioneins (MTs) are among others involved in the cellular regulation of essential Zn(II) and Cu(I) ions. However, the high binding affinity of these proteins requires additional factors to promote metal ion release under physiological conditions. The mechanisms and efficiencies of these processes leave many open questions. We report here a comprehensive analysis of the Zn(II)-release properties of various MTs with special focus on members of the four main subfamilies of plant MTs. Zn(II) competition experiments with the metal ion chelator 4-(2-pyridylazo)resorcinol (PAR) in the presence of the cellular redox pair glutathione (GSH)/glutathione disulfide (GSSG) show that plant MTs from the subfamilies MT1, MT2, and MT3 are remarkably more affected by oxidative stress than those from the Ec subfamily and the well-characterized human MT2 form. In addition, we evaluated proteolytic digestion with trypsin and proteinase K as an alternative mechanism for selective promotion of metal ion release from MTs. Also here the observed percentage of liberated metal ions depends strongly on the MT form evaluated. Closer evaluation of the data additionally allowed deducing the thermodynamic and kinetic properties of the Zn(II) release processes. The Cu(I)-form of chickpea MT2 was used to exemplify that both oxidation and proteolysis are also effective ways to increase the transfer of copper ions to other molecules. Zn(II) release experiments with the individual metal-binding domains of Ec-1 from wheat grain reveal distinct differences from the full-length protein. This triggers the question about the roles of the long cysteine-free peptide stretches typical for plant MTs.

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

    International Nuclear Information System (INIS)

    Till, J.E.; Etnier, E.L.; Meyer, H.R.

    1980-01-01

    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

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

    International Nuclear Information System (INIS)

    Finn, P.A.; Clemmer, R.G.; Misra, B.

    1981-10-01

    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

  14. Implantation measurements to determine tritium permeation in first-wall structures

    International Nuclear Information System (INIS)

    Holland, D.F.; Causey, R.A.; Sattler, M.L.

    1983-01-01

    A principal safety concern for a D-T burning fusion reactor is release of tritium during routine operation. Tritium implantation into first-wall structures, and subsequent permeation into coolants, is potentially an important source of tritium loss. This paper reports on an experiment in which an ion accelerator was used to implant deuterium