WorldWideScience

Sample records for acute tritium releases

  1. 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...

  2. 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

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

    International Nuclear Information System (INIS)

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

  4. 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)

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

    International Nuclear Information System (INIS)

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

  6. Preliminary dimensioning of the IFMIF Tritium Release Test Module

    Energy Technology Data Exchange (ETDEWEB)

    Arbeiter, Frederik, E-mail: frederik.arbeiter@kit.edu; Abou-Sena, Ali; Chen, Yuming; Freund, Jana; Klix, Axel; Kondo, Keitaro; Vladimirov, Pavel

    2013-10-15

    Highlights: • The design of the IFMIF Tritium Release Test Module is explained. • Nuclear responses in the module and specimens are calculated. • Temperature fields during irradiation are calculated by 1D methods. • The tritium budget is calculated by 1D methods. -- Abstract: As part of the ongoing Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF), an experimental device suitable for the irradiation and online tritium release measurements of solid breeder ceramics and beryllium is investigated. This experimental device is called the Tritium Release Test Module (TRTM). In the preliminary design phase, the possible thermal conditions, the tritium diffusion budgets, and the mechanical loads have been studied by analytical calculations and numerical codes. The most important results concern the tritium production and nuclear heating induced in the structures, the temperature distribution in the specimen region and the structure, and the diffusion of tritium through the safety barriers.

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

    International Nuclear Information System (INIS)

    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.)

  8. Tritium releases and impact about EDF nuclear power stations

    International Nuclear Information System (INIS)

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

  9. 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.

  10. Tritium release from SS316 under vacuum condition

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

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

    International Nuclear Information System (INIS)

    DT+ ion irradiation with energy of 0.5 and 1.0 keV was performed on helium pre-irradiated tungsten and the amount of retained tritium and the long-term release of retained tritium in vacuum was investigated using an IP technique and BIXS. Tritium retention and long-term tritium release were significantly influenced by helium pre-irradiation. The amount of retained tritium increased until it reached 1 × 1017 He/cm2, and at 1 × 1018 He/cm2 it became smaller compared to 1 × 1017 He/cm2. The amount of retained tritium in tungsten without helium pre-irradiation largely decreased after several weeks preservation in vacuum, and the long-term release rate during vacuum preservation was retarded by helium pre-irradiation. The results indicate that the long-term tritium release and the helium irradiation effect on it should be taken into account for more precise estimation of tritium retention in the long-term use of tungsten in fusion devices

  12. Modeling unusual tritium release behavior from Li2O

    International Nuclear Information System (INIS)

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

  13. 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)

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

    International Nuclear Information System (INIS)

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

  15. An atmospheric tritium release database for model comparisons

    International Nuclear Information System (INIS)

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

  16. MODELING ATMOSPHERIC RELEASES OF TRITIUM FROM NUCLEAR INSTALLATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Okula, K

    2007-01-17

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

  17. 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)

  18. Release of tritium from fuel and collection for storage

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

  20. Computer simulation of tritium releases in inertial fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  1. 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.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-11-01

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

  6. 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.

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

    International Nuclear Information System (INIS)

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

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

    2013-07-01

    Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the main concept of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by company NGK, Japan. It is notable that beryllium pebbles of other types are commercially available at the market. Presented work is dedicated to a study of characteristics of microstructure, packaging density and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Company Materion, USA. (orig.).

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2004-08-01

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

  11. 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)

  12. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jacobsen, W.R.

    1976-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

  19. 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. PMID:23337314

  20. 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)

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

    International Nuclear Information System (INIS)

    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)

  2. 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

  3. 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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-01

    To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Walker, T. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, S. H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DelCul, Guillermo Daniel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Walker, T. B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); DelCul, Guillermo Daniel [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    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.

  12. WBN-1 Cycle 10 TPBAR Tritium Release, Deduced From Analysis of RCS Data TTP-1-3046-00, Rev 0

    Energy Technology Data Exchange (ETDEWEB)

    Shaver, Mark W.; Niehus, Mark T.; Love, Edward F.

    2012-02-19

    This document contains the calculation of the TPBAR tritium release from the Mark 9.2 design TPBARs irradiated in WBN cycle 10. The calculation utilizes the generalized cycle analysis methodology given in TTP-1-3045 Rev. 0.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  14. 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.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. 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)

  18. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, S; Raskob, W

    2007-08-15

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

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

    Science.gov (United States)

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

    2007-06-01

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

  1. 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.

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

    International Nuclear Information System (INIS)

    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

  3. Sources of tritium

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

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

  5. 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)

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

  6. Welsh tritium

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.)

  8. Spherical diffusion of tritium from a point of release in a uniform unsaturated soil. A deterministic model for tritium migration in an arid disposal site

    Energy Technology Data Exchange (ETDEWEB)

    Smiles, D.E.; Gardner, W.R.; Schulz, R.K. [Univ. of California, Berkeley, CA (United States). Dept. of Environmental Science, Policy and Management

    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{sup {minus}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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-06-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    International Nuclear Information System (INIS)

    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

  14. 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)

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

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

    Science.gov (United States)

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

    2014-05-01

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

  16. Analyses of Generation and Release of Tritium in Nuclear Power Plant%核电厂氚的产生和排放分析

    Institute of Scientific and Technical Information of China (English)

    黎辉; 梅其良; 付亚茹

    2015-01-01

    T ritium research including tritium generation in reactor core and in the primary coolant ,release pathways ,tritium chemical forms and release amount is a very impor‐tant part of environment assessment of nuclear power plant .Based on the international operation practice ,the primary coolant system ,auxiliary systems ,radwaste system and ventilation system were analysed ,and the tritium release pathways and chemical forms were investigated .The results indicate that the theoretic calculation results agree with the nuclear power plant operation data very well .The tritium contained in the primary coolant is mainly produced from the three‐fragment fission reaction ,boron activation in the burnable poison rods and boron ,lithium and deuterium activation w hen they pass through the core . The released tritium to the environment is mainly in the form of tritiated water and the percentage between the liquid and gaseous of release tritium mainly depends on the leakage rate from the primary coolant to the reactor building and auxiliary building .%研究核电厂中氚在堆芯和主冷却剂中的产生方式,以及进入环境的途径、形态和排放量,是核电厂辐射环境影响评价非常重要的内容之一。本文通过分析压水堆核电厂中的主冷却剂系统、辅助系统、三废系统和厂房通风系统的运行模式,结合国际上的运行经验参数,研究主冷却剂中的氚排放进入环境大气的途径和形态。研究结果表明:理论计算分析结果与电厂运行经验数据相吻合,氚主要通过燃料棒中的三元裂变,可燃毒物棒中硼的活化以及主冷却剂中硼、锂和氘流经堆芯时的活化产生,主要以液态氚水形式排放,影响气液两相分配份额的主要因素取决于主冷却剂向反应堆厂房和辅助厂房的泄漏率。

  17. Analysis of tritium release from LiAlO sub 2 in the TEQUILA experiment, using the MISTRAL code

    Energy Technology Data Exchange (ETDEWEB)

    Badawi, A.; Raffray, A.R.; Abdou, M.A. (Univ. of California, Dept. of Mechanical, Aerospace and Nuclear Engineering, Los Angeles, CA (United States))

    1991-12-01

    The tritium release behavior from LiAlO{sub 2} samples in the TEQUILA experiment was analyzed using the MISTRAL code. This was done in order to benchmark the code for analyzing the performance of a LiAlO{sub 2} blanket test section under ITER-like conditions. Material property data available from the experimental sample microstructure characterization and from the literature were used as input to the code. The microstructure characterization was quite thorough and included the pore size distribution which was used to estimate the pore diffusion coefficient. In the case of the bulk diffusion coefficient, since single crystal experimental measurements are not available, two different values from different experimental data were used. The strategy was to model four different transients for the same sample and to use the property data, in particular the diffusion coefficient, which will better reproduce all four transients. The transients studied were: Two temperature transients, in which the temperature changed by +50deg C and -50deg C and two hydrogen concentration transients in the purge, in which the concentration changes from 0.1% to 1% and from 1% to 0.1%. The results showed that the assumed bulk diffusion coefficient can change the output substantially. For each case, the effects of other parameters, such as the adsorption activation and pore diffusion coefficient, were also considered. The results are discussed in the paper. (orig.).

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

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, S R

    2006-09-27

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

  19. 锂铅合金释氚实验研究%Experimental Study of Tritium Release from Li_(17)Pb_(83) Alloy

    Institute of Scientific and Technical Information of China (English)

    谢波; 吴宜灿; 陈晓军; 翁葵平; 刘俊; 肖成建; FDS团队

    2011-01-01

    Lithium-lead alloy is considered to be one of the most prominent tritium breeding materials for the fusion reactor blanket because of its high breeding ratio,and low reactivity and possible use as coolant.An out-of-pile experiment of tritium release from Li17Pb83 alloy was performed after neutron irradiation on the base of mathematical model to describe tritium release behavior from an eutectic lithium-lead alloy.The results suggest that the dominant chemical form of the released tritium(99%) was the water-insoluble component(HT or T2).Tritium residence time decreased with increasing H2 pressure in carrier gas up to 1000 Pa,and above this concentration limit it became constant and not influenced by the plenum volume.The temperature dependence of the tritium release rate can be described by an Arrhenius law.Consequently,the present results on the kinetic parameters of tritium in molten Li17Pb83 alloy are considered to be different from the values in literature,but it is the same that the overall release process is governed by the diffusion of tritium atoms in the Li17Pb83 and by the heterogeneous reaction at the gas-eutectic interface of the tritium atom recombination at temperatures from 633 to 973 K.%由于锂铅合金因具有高增殖比、低活泼性和可能作为冷却剂的特点,被认为是最有潜力的能源堆包层氚增殖材料。在理论模型描述熔融锂铅合金氚释放行为的基础上,开展了中子辐照后Li17Pb83合金的离线氚释放实验。结果表明:释放氚的化学形式99%以上为难溶于水的成分(HT或T2);氚滞留时间随载气中氢分压的增加而减小,氢分压达到1000 Pa后变为常数,且与实体积无关;氚释放速率对温度的依赖性符合Arrhenius定律。以此为基础得到的氚在熔融锂铅中的动力学参数结果,虽与文献值有差异,但同样证明了在633—973 K的范围内,氚从液态锂铅到气相的整个释放过程中起决定作用的是氚在合金内

  20. 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)

  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. Tritium in the aquatic environment

    International Nuclear Information System (INIS)

    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

  3. Tritium in the aquatic environment

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Tritium oxidation and exchange: preliminary studies

    International Nuclear Information System (INIS)

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

  7. 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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

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

  10. 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)

    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

  11. Tritium in the environment. Knowledge synthesis

    International Nuclear Information System (INIS)

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

    OpenAIRE

    Alejandra Delgado; Erica H. Jaffé

    2011-01-01

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

  13. Transport of tritium in SS316 at moderate temperatures

    International Nuclear Information System (INIS)

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

  14. Tritium monitoring at the Sandia Tritium Research Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, S

    2004-06-30

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

  16. Tritium contamination and decontamination

    International Nuclear Information System (INIS)

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

  17. Recovery of tritium from a liquid lithium blanket

    Energy Technology Data Exchange (ETDEWEB)

    Talbot, J.B.

    1981-01-01

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

  18. 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.

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

    International Nuclear Information System (INIS)

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

  20. 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)

  1. Behaviour of tritium in the environment

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  3. Management of Tritium in European Spallation Source

    DEFF Research Database (Denmark)

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

    2015-01-01

    The European Spallation Source (ESS) will produce tritium via spallation and activation processes during operational activities. Within the location of ESS facility in Lund, Sweden site it is mandatory to demonstrate that the management strategy of the produced tritium ensures the compliance...... with the country regulation criteria. The aim of this paper is to give an overview of the different aspects of the tritium management in ESS facility. Besides the design parameter study of the helium coolant purification system of the target the consequences of the tritium releasing into the environment were also...

  4. Tritium handling in vacuum systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-10-01

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

  5. Universal tritium transmitter

    International Nuclear Information System (INIS)

    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

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

    DEFF Research Database (Denmark)

    Petersen, Lars Jelstrup; Pedersen, Juri Lindy; Skov, Per Stahl;

    2009-01-01

    BACKGROUND: Animal models have shown histamine to be released from the skin during the acute phase of a burn injury. The role of histamine during the early phase of thermal injuries in humans remains unclear. PURPOSE: The objectives of this trial were to study histamine release in human skin during.......6 +/- 1.8 nM vs. post-burn values of 14.8 +/- 1.8 nM, n = 8). CONCLUSIONS: Histamine is not released in human skin during the acute phase of a thermal injury....... the acute phase of a standardized thermal injury in healthy volunteers. METHODS: Histamine concentrations in human skin were measured by skin microdialysis technique. Microdialysis fibers were inserted into the dermis in the lower leg in male healthy volunteers. A standardized superficial thermal injury...

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

    International Nuclear Information System (INIS)

    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.)

  8. Interaction of energetic tritium with silicon carbide

    International Nuclear Information System (INIS)

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

  9. Magmatic tritium

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    NARCIS (Netherlands)

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

    2011-01-01

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

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

    NARCIS (Netherlands)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, S

    2007-08-15

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

  13. DOE handbook: Tritium handling and safe storage

    International Nuclear Information System (INIS)

    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

  14. Handling of tritium-bearing wastes

    International Nuclear Information System (INIS)

    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

  15. Tritium/ 3He dating of shallow groundwater

    Science.gov (United States)

    Schlosser, Peter; Stute, Martin; Dörr, Helmut; Sonntag, Christian; Münnich, Karl Otto

    1988-08-01

    Combined tritium/ 3He data from three multi-level sampling wells (DFG 1, DFG 4, DFG 7) located at Liedern/ Bocholt, West Germany, are presented and principles of the tritium/ 3He method in shallow groundwater studies are discussed. The 3He excess produced by radioactive decay of bomb tritium (released mainly between 1952 and 1963) is clearly reflected in the data. The tritiogenic 3He signal can be detected with a good resolution (signal/1σ error: ≈ 350). The confinement of the tritiogenic 3He is estimated to approximately 77-85% at site DFG 4. For the bomb tritium peak the deviation of the tritium/ 3He age from the age determined by identifying the groundwater layer recharged between 1962 and 1965 is about 3 years (15%). The deviation can be explained by diffusive 3He loss across the groundwater table and by flow dispersion.

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Strenge, D.L.; Peloquin, R.A.

    1981-04-01

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

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

    International Nuclear Information System (INIS)

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

  19. Behavior of tritium in heavy water reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

  2. 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.

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

    OpenAIRE

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

    2007-01-01

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

  4. Quick management of accidental tritium exposure cases.

    Science.gov (United States)

    Singh, Vishwanath P; Badiger, N M; Managanvi, S S; Bhat, H R

    2012-07-01

    Removal half-life (RHL) of tritium is one of the best means for optimising medical treatment, reduction of committed effective dose (CED) and quick/easy handling of a large group of workers for medical treatment reference. The removal of tritium from the body depends on age, temperature, relative humidity and daily rainfall; so tritium removal rate, its follow-up and proper data analysis and recording are the best techniques for management of accidental acute tritium exposed cases. The decision of referring for medical treatment or medical intervention (MI) would be based on workers' tritium RHL history taken from their bodies at the facilities. The workers with tritium intake up to 1 ALI shall not be considered for medical treatment as it is a derived limit of annual total effective dose. The short-term MI may be considered for tritium intake of 1-10 ALI; however, if the results show intake ≥100 ALI, extended strong medical/therapeutic intervention may be recommended based on the severity of exposure for maximum CED reduction requirements and annual total effective dose limit. The methodology is very useful for pressurized heavy water reactors (PHWRs) which are mainly operated by Canada and India and future fusion reactor technologies. Proper management will optimise the cases for medical treatment and enhance public acceptance of nuclear fission and fusion reactor technologies.

  5. The Tritium White Paper

    International Nuclear Information System (INIS)

    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

  6. Safe handling of tritium

    International Nuclear Information System (INIS)

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

  7. Tritium conference days

    International Nuclear Information System (INIS)

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

  8. Tritium contamination control

    International Nuclear Information System (INIS)

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

  9. Tritium pellet injector results

    International Nuclear Information System (INIS)

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

  10. Modeling tritium behavior in Li2ZrO3

    International Nuclear Information System (INIS)

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

  11. Modeling tritium behavior in Li2ZrO3

    International Nuclear Information System (INIS)

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

  12. Tritium glovebox stripper system seismic design evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

  13. Tritium Depth Profiles in 316 Stainless Steel

    Science.gov (United States)

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

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

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

    Directory of Open Access Journals (Sweden)

    Marc Afilalo

    2010-02-01

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

  15. Profile of extended-release oxycodone/acetaminophen for acute pain

    Directory of Open Access Journals (Sweden)

    Bekhit MH

    2015-10-01

    Full Text Available Mary Hanna Bekhit1–51David Geffen School of Medicine, 2Ronald Reagan UCLA Medical Center, 3UCLA Ambulatory Surgery Center, 4UCLA Wasserman Eye Institute, 5UCLA Martin Luther King Community Hospital, University of California Los Angeles, Los Angeles, CA, USA Abstract: This article provides a historical and pharmacological overview of a new opioid analgesic that boasts an extended-release (ER formulation designed to provide both immediate and prolonged analgesia for up to 12 hours in patients who are experiencing acute pain. This novel medication, ER oxycodone/acetaminophen, competes with current US Food and Drug Administration (FDA-approved opioid formulations available on the market in that it offers two benefits concurrently: a prolonged duration of action, and multimodal analgesia through a combination of an opioid (oxycodone with a nonopioid component. Current FDA-approved combination analgesics, such as Percocet (oxycodone/acetaminophen, are available solely in immediate-release (IR formulations. Keywords: opioid, analgesic, xartemis, acute postsurgical pain, substance abuse, acetaminophen, extended release 

  16. Impact of Serial Troponin Release on Outcomes in Patients With Acute Heart Failure Analysis From the PROTECT Pilot Study

    NARCIS (Netherlands)

    O'Connor, Christopher M.; Fiuzat, Mona; Lombardi, Carlo; Fujita, Kenji; Jia, Gang; Davison, Beth A.; Cleland, John; Bloomfield, Daniel; Dittrich, Howard C.; DeLucca, Paul; Givertz, Michael M.; Mansoor, George; Ponikowski, Piotr; Teerlink, John R.; Voors, Adriaan A.; Massie, Barry M.; Cotter, Gad; Metra, Marco

    2011-01-01

    Background-Cardiac troponin T (cTnT) elevation is common and is a predictor of outcomes in patients with acute heart failure (AHF). The degree and progression of cTnT release during hospitalization of patients with AHF is unclear. We evaluated the incidence of cTnT release during AHF hospitalization

  17. Acute release of tissue-type plasminogen activator (t-PA) from the endothelium ; regulatory mechanisms and therapeutic target

    NARCIS (Netherlands)

    Schrauwen, Y.; Vries, R.E.M. de; Kooistra, T.; Emeis, J.J.

    1994-01-01

    The acute release of t-PA was studied in vitro in human endothelial cells from different origin. It proved possible to enhance the amounts of t-PA which can be released by increasing t-PA synthesis, both in vitro, and in vivo in rats. These data suggest the feasibility to induce and (or) enhance acu

  18. Elemental tritium deposition and conversion in the terrestrial environment

    International Nuclear Information System (INIS)

    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. Confinement and Tritium Stripping Systems for APT Tritium Processing

    International Nuclear Information System (INIS)

    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

  20. Tritium in metals

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Investigation of tritium in groundwater at Pickering NGS

    International Nuclear Information System (INIS)

    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)

  3. Characterisation of redundant tritium light devices

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-12-01

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

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

    International Nuclear Information System (INIS)

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

  7. 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)

  8. 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

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

    International Nuclear Information System (INIS)

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

  10. Tritium Issues in Next Step Devices

    International Nuclear Information System (INIS)

    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

  11. Tritium liquid effluents from the Krsko NPP

    International Nuclear Information System (INIS)

    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)

  12. 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.)

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

    Directory of Open Access Journals (Sweden)

    Laura Musazzi

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  16. 2012 ACCOMPLISHMENTS - TRITIUM AGING STUDIES ON STAINLESS STEELS

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, M.

    2013-01-31

    This report summarizes the research and development accomplishments during FY12 for the tritium effects on materials program. The tritium effects on materials program is designed to measure the long-term effects of tritium and its radioactive decay product, helium-3, on the structural properties of forged stainless steels which are used as the materials of construction for tritium reservoirs. The FY12 R&D accomplishments include: (1) Fabricated and Thermally-Charged 150 Forged Stainless Steel Samples with Tritium for Future Aging Studies; (2) Developed an Experimental Plan for Measuring Cracking Thresholds of Tritium-Charged-and-Aged Steels in High Pressure Hydrogen Gas; (3) Calculated Sample Tritium Contents For Laboratory Inventory Requirements and Environmental Release Estimates; (4) Published report on “Cracking Thresholds and Fracture Toughness Properties of Tritium-Charged-and-Aged Stainless Steels”; and, (5) Published report on “The Effects of Hydrogen, Tritium, and Heat Treatment on the Deformation and Fracture Toughness Properties of Stainless Steels”. These accomplishments are highlighted here and references given to additional reports for more detailed information.

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

    DEFF Research Database (Denmark)

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

    2002-01-01

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

  18. Tritium Systems Test Assembly operator training program

    International Nuclear Information System (INIS)

    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

  19. Interactions of tritium and materials

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

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

  20. Tritium breeding in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    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. Tritium technology. A Canadian overview

    International Nuclear Information System (INIS)

    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)

  2. 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)

  3. A study and analysis on tritium radioactivity and environmental behavior in domestic NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Jun; Lee, Kyeong Jin; Yeom, Jeong Min; Shin, Dae Tewn [Chosun University, Gwangju (Korea, Republic of)

    2015-12-15

    Several analyses on tritium that is the largest release of gas or liquid radioactive waste from domestic PWR and PHWR NPPs were carried out, such as release comparison, directional frequency of wind and tritium behavior changes in environmental samples. First of all, analysis result showed that tritium released from PHWR was more than ten times as gas and double to three times as liquid in comparison to PWR in 2013. Independent release management in NPP units is needed to precisely control and analyze tritium, since there were 2 units of some NPPs having the same amount of release during analysis. In analysis on frequency of wind direction, average range showed 1.7 to 11.5% by 16-point compass. In case of analysis on sampling points by wind direction, Result showed most of the sampling points are right in places. However, There are some areas needed to examine. In analysis on tritium concentration changes in environmental samples, tritium concentration near NPPs was higher than one far away from NPPs. In case of environmental samples far from PWR, a trace of tritium occur. While, tritium concentration near NPPs was more than or equal to one further from PHWR. In conclusion, tritium occurs considerably in PHWR and is lower than standard in samples. but, it is still detected. Therefore, it is needed to strengthen control in system in NPPs and to consistently monitor tritium in environment.

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

    International Nuclear Information System (INIS)

    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)

  5. A review of tritium conversion reactions

    International Nuclear Information System (INIS)

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

  6. Modeling of tritium behavior in ceramic breeder materials

    International Nuclear Information System (INIS)

    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

  7. Tritium catalyzed deuterium tokamaks

    International Nuclear Information System (INIS)

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

  8. Tritium protective clothing

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

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

    2004-01-01

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

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

    Science.gov (United States)

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

    2016-01-01

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

  11. Tritium//sup 3/He dating of shallow groundwater

    Energy Technology Data Exchange (ETDEWEB)

    Schlosser, P.; Stute, M.; Doerr, H.; Sonntag, C.; Muennich, K.O.

    1988-08-01

    Combined tritium//sup 3/He data from three multi-level sampling wells (DFG 1, DFG 4, DFG 7) located at Liedern/Bocholt, West Germany, are presented and principles of the tritium//sup 3/He method in shallow groundwater studies are discussed. The /sup 3/He excess produced by radioactive decay of bomb tritium (released mainly between 1952 and 1963) is clearly reflected in the data. The tritiogenic /sup 3/He signal can be detected with a good resolution (signal/1sigma error: approx. = 350). The confinement of the tritiogenic /sup 3/He is estimated to approximately 77-85% at site DFG 4. For the bomb tritium peak the deviation of the tritium//sup 3/He age from the age determined by identifying the groundwater layer recharged between 1962 and 1965 is about 3 years (15%). The deviation can be explained by diffusive /sup 3/He loss across the groundwater table and by flow dispersion.

  12. Estimation of dose to man from environmental tritium

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  16. Tritium effects on germ cells and fertility

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Okamoto, Takanobu; Masuhara, Mitsuhiko; Ikuta, Komei

    2014-01-01

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

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

    Science.gov (United States)

    Okamoto, Takanobu; Masuhara, Mitsuhiko; Ikuta, Komei

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

  2. Hazards of exposure to tritium and tritium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, R.C.; Kornberg, H.A.

    1954-01-01

    Experimental data pertinent to the evaluation of hazards involved in the exposure of personnel to tritium and tritium oxide are reviewed. Conclusions are drawn and recommendations made with regard to the control of these hazards.

  3. Glovebox stripper system tritium capture efficiency-literature review

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-28

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

  4. Tritium analyses of COBRA-1A2 beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    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.

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

    International Nuclear Information System (INIS)

    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

  7. Tritium analysis at TFTR

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Nakamura, H.; Kawamura, Y.; Iwai, Y.; Isobe, K.; Yamada, M.; Kurata, R.; Edao, Y. [Tritium Technology Group, Japan Atomic Energy Agency, Tokai-mura (Japan); Suzuki, T.; Oyaizu, M.; Yamanishi, T. [Tritium Technology Group, Japan Atomic Energy Agency, Rokkasho-mura (Japan)

    2015-03-15

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

  9. Tritium - is it underestimated

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2007-11-01

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

  11. Environmental monitoring of molecular tritium

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Tritium breeding materials

    International Nuclear Information System (INIS)

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

  14. Tritium retention in TFTR

    International Nuclear Information System (INIS)

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

  15. Tritium Exchange in Biological Systems

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M. C.

    1998-01-13

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  20. Assessment of tritium in the Savannah River Site environment

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-02-01

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

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

    International Nuclear Information System (INIS)

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

  3. Breeding blanket development; Tritium release from breeder

    OpenAIRE

    土谷 邦彦; 河村 弘; 長尾 美春

    2006-01-01

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

  4. Hydroxyl radicals mediate acute pulmonary vasoconstriction and thromboxane release during protamine reversal of heparin in awake sheep

    Energy Technology Data Exchange (ETDEWEB)

    Nguyenduy, T.; Morel, D.R.; Collee, G.; Eberhard, M.; Melvin, C.; Robinson, D.R.; Repine, J.E.; Lowenstein, E.; Zapol, W.M.

    1986-03-01

    Neutralization of heparin by protamine sulfate activates the classical complement pathway and causes release of thromboxane leading to pulmonary hypertension (PH). Oxygen radicals are generated during granulocyte stimulation by complement fragments. The authors investigated the effect of dimethyl sulfoxide (DMSO), a non-specific hydroxyl radical scavenger, and dimethyl thiourea (DMTU), a specific enzymatic hydroxyl radical scavenger, on this acute reaction. In chronically instrumented sheep, IV injection of protamine (2 mg/kg) 5 min after heparin (200IU/kg) produced PH (246 +/- 16% of baseline at 1 min, anti x +/- SE), with leukopenia (to 37 +/- 8% of baseline at 2 min) associated with thromboxane B/sub 2/ (TxB/sub 2/) release (5.3 + 2.0 ng/ml). IV pretreatment with DMSO (lg/kg) had no effect on the response. DMTU, 0.5 g/kg, attenuated, and lg/kg completely abolished TxB/sub 2/ release and PH. Neither DMSO nor DMTU had any effect on the leukopenia. Arachidonic acid infusion (100 ..mu..g/kg/min x 5 min) released TxB/sub 2/ and produced PH despite pre-treatment with DMTU, demonstrating intact cyclooxygenase pathway. Thus, hydroxyl radicals appear to mediate TxB/sub 2/ release in classical pathway complement activation accompanying heparin neutralization by protamine.

  5. Tritium system for compact high field devices

    International Nuclear Information System (INIS)

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

  6. Tritium neutrino mass experiments

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Afilalo, Marc

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ricapito, I.; Calderoni, P.; Poitevin, Y. [Fusion for Energy, Barcelona (Spain); Aiello, A.; Utili, M. [ENEA, Camugnano (Italy); Demange, D. [Karlsruhe Institute of Technology - KIT, Karlsruhe (Germany)

    2015-03-15

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

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

    International Nuclear Information System (INIS)

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

  13. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    J. L. Pérez-Arellano

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-10-01

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

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

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

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

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

    Science.gov (United States)

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

    1984-01-01

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

  19. TRITIUM ACCOUNTANCY IN FUSION SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Klein, J. E.; Farmer, D. A.; Moore, M. L.; Tovo, L. L.; Poore, A. S.; Clark, E. A.; Harvel, C. D.

    2014-03-06

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

  20. Tritium accountancy in fusion systems

    Energy Technology Data Exchange (ETDEWEB)

    Klein, J.E.; Clark, E.A.; Harvel, C.D.; Farmer, D.A.; Tovo, L.L.; Poore, A.S. [Savannah River National Laboratory, Aiken, SC (United States); Moore, M.L. [Savannah River Nuclear Solutions, Aiken, SC (United States)

    2015-03-15

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

  1. Tritium uptake by SS316 and its decontamination

    Science.gov (United States)

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

    2004-08-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E. Jr.

    1990-01-01

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

  4. Tritium-surface interactions

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-08-15

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

  8. Trapping and depth profile of tritium in surface layers of metallic materials

    Energy Technology Data Exchange (ETDEWEB)

    Matsuyama, M., E-mail: masao@ctg.u-toyama.ac.jp [Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555 (Japan); Chen, Z. [The Southwestern Institute of Physics, Chengdu 610041, Sichuan (China); Nisimura, K. [National Institute for Fusion Science, Toki 509-5292 (Japan); Akamaru, S.; Torikai, Y.; Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555 (Japan); Ashikawa, N. [National Institute for Fusion Science, Toki 509-5292 (Japan); Oya, Y.; Okuno, K. [Radiochemistry Research Laboratory, Shizuoka University, Shizuoka 422-8529 (Japan); Hino, T. [Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo 060-8628 (Japan)

    2011-10-01

    Tritium amount retained in surface layers and release behavior from surface layers were examined using SS316L samples exposed to plasmas in the Large Helical Device and a commercial Cu-Be alloy plate. BIXS analyses and observation by SEM indicate that carbon and titanium deposited on the plasma-facing surface of the SS316L samples. Larger amount of tritium was trapped in the plasma-facing surface in comparison with the polished surface. Higher enrichment of tritium in surface layers was similarly found in the polished surface of both samples. The amount of surface tritium in both samples was almost same, while the bulk concentration of tritium in Cu-Be was much lower than that in SS316L. Tritium release from the SS316L and Cu-Be samples into water was examined by immersion experiments. Tritium elution was observed for both samples, but changes in the residual tritium amount in surface layers were different from each other.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-02-26

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

  11. Tritium waste control: July--September 1978

    International Nuclear Information System (INIS)

    The combined Electrolysis Catalytic Exchange system was modified to allow better control of experimental conditions and to prevent the overflow of water into the air detritation system. A program designed to regenerate the activity of the hydrophobic catalyst was also completed. Slight differences in the release rate of high specific activity tritiated liquid wastes from the drums are now beginning to appear. The three drums with the highest fractional permeation rate had the least amount of tritium when packaged. The fractional permeation rate of the two octane drums appears to have leveled off at about the same rate as the oil and water drums. Tests continued on samples of cement and cement-plaster mixtures which were injected with 386 Ci of tritiated water, cured, and then impregnated with catalyzed styrene monomer. After polymerization, the samples were put into uncontaminated water and the tritium concentration was monitored. No significant differences were noted except in two cases when the polyethylene bottle had been removed, which resulted in 35 to 80 times more tritium being released into the surrounding water. Full scale (cold) waste drum No. 5 was polymerized with excellent results. Pressure increase and gas composition were measured over (1) tritiated water without fixation, (2) polymer-impregnated concrete, and (3) nonpolymer concrete. Activities for all samples were 10 Ci/m3. Pressure buildup results are essentially the same for concrete made with tritiated distilled water and tritiated waste water. However, the pressure buildup rate is slightly higher for the polymer impregnated concrete than for the nonpolymer concrete. Mass analysis of the cover gas over tritiated water without fixation and over the polymer and nonpolymer concrete samples made with tritiated waste water show that hydrogen represents about 85% of the gas generated

  12. Reducing the tritium inventory in waste produced by fusion devices

    International Nuclear Information System (INIS)

    Highlights: • Fusion devices including ITER will generate tritiated waste, some of which will need to be detritiated before disposal. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Incineration is very attractive for VLLW and possibly SL-LILW soft housekeeping waste, since it offers higher tritium and waste volume reduction than the alternative thermal treatment technique. • For metallic waste, further R&D efforts should be made to optimize tritium release management and minimize the need for interim storage. - Abstract: The specific issues raised by tritiated waste resulting from fusion machines are described. Of the several categories of tritium contaminated waste produced during the entire lifespan of a fusion facility, i.e. operating phase and dismantling phase, only two categories are considered here: metal components and solid combustible waste, especially soft housekeeping materials. Some of these are expected to contain a high level of tritium, and may therefore need to be processed using a detritiation technique before disposal or interim storage. The reference solution for tritiated waste management in France is a 50-year temporary storage for tritium decay, with options for reducing the tritium content as alternatives or complement. An overview of the strategic issues related to tritium reduction techniques is proposed for each radiological category of waste for both metallic and soft housekeeping waste. For this latter category, several options of detritiation techniques by thermal treatment like heating up or incineration are described. A comparison has been made between these various technical options based on several criteria: environment, safety, technical feasibility and costs. For soft housekeeping waste, incineration is very attractive for VLLW and possibly SL-LILW. For metallic waste, further R&D efforts should be conducted

  13. Reducing the tritium inventory in waste produced by fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Pamela, J., E-mail: jerome.pamela@cea.fr [CEA, Agence ITER-France, F-13108 Saint-Paul-lez-Durance (France); Decanis, C. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Canas, D. [CEA, DEN/DADN, Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Liger, K.; Gaune, F. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2015-04-15

    Highlights: • Fusion devices including ITER will generate tritiated waste, some of which will need to be detritiated before disposal. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Incineration is very attractive for VLLW and possibly SL-LILW soft housekeeping waste, since it offers higher tritium and waste volume reduction than the alternative thermal treatment technique. • For metallic waste, further R&D efforts should be made to optimize tritium release management and minimize the need for interim storage. - Abstract: The specific issues raised by tritiated waste resulting from fusion machines are described. Of the several categories of tritium contaminated waste produced during the entire lifespan of a fusion facility, i.e. operating phase and dismantling phase, only two categories are considered here: metal components and solid combustible waste, especially soft housekeeping materials. Some of these are expected to contain a high level of tritium, and may therefore need to be processed using a detritiation technique before disposal or interim storage. The reference solution for tritiated waste management in France is a 50-year temporary storage for tritium decay, with options for reducing the tritium content as alternatives or complement. An overview of the strategic issues related to tritium reduction techniques is proposed for each radiological category of waste for both metallic and soft housekeeping waste. For this latter category, several options of detritiation techniques by thermal treatment like heating up or incineration are described. A comparison has been made between these various technical options based on several criteria: environment, safety, technical feasibility and costs. For soft housekeeping waste, incineration is very attractive for VLLW and possibly SL-LILW. For metallic waste, further R&D efforts should be conducted.

  14. Effect of thyrotropin-releasing hormone on cerebral free radical reactions following acute brain injury in rabbits

    Institute of Scientific and Technical Information of China (English)

    牛光明; 顾秀娟; 苏玉林; 万锋; 苏芳忠; 薛德麟

    2003-01-01

    Objective: To investigate the early effect of thyrotropin-releasing hormone (TRH) on cerebral free radical reactions after acute brain injury in rabbits.Methods: 30 healthy white rabbits were randomly divided into three groups: Group A (n=10), Group B (n=12) and Group C (n=8). The rabbits in Group A and Group B were injured by direct hit. At 0.5-4 hours after injury, the rabbits in Group A were injected with TRH (8 mg/kg body weight) through a vein and the rabbits in Group B were injected with normal saline of equal volume. The rabbits in Group C served as the normal control. Then all the rabbits were killed and brain tissues were obtained. The content of lipoperoxide (LPO), the activity of superoxide dismutase (SOD) and the water content of the brain tissues were measured.Results: The contents of LPO and water in brain tissues in Group A were lower and the activity of SOD was higher than those of Group B (P<0.05). After injury, intracranial pressure (ICP) rose rapidly and continuously with time passing by. When TRH was given to the animals in Group A, the rising speed of ICP slowed down significantly.Conclusions: TRH can decrease the cerebral free radical reactions and cerebral edema after acute brain injury in rats.

  15. Preliminary analysis of the safety and environmental impact of the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    The Tritium Systems Test Assembly (TSTA) is a facility dedicated to the development of technologies associated with the D-T fuel cycle of future fusion reactors while demonstrating that TSTA can be operated safely with no significant losses to the environment. During the initial design stage of TSTA, a safety analysis was performed which investigated the effects of major subsystem component failure, the meteorology and seismicity of the site and their possible effect on the facility, and accident scenarios which result in tritium releases. Major releases of tritium to the environment are considered highly improbable since they require a compound failure of primary and secondary containment, along with either a breach of the building or a failure of the Emergency Tritium Cleanup system. Accidental releases caused by natural phenomena (earthquake, tornado, etc.) are considered highly improbable

  16. Tritium Plasma Experiment Upgrade for Fusion Tritium and Nuclear Sciences

    Science.gov (United States)

    Shimada, Masashi; Taylor, Chase N.; Kolasinski, Robert D.; Buchenauer, Dean A.

    2015-11-01

    The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to directly study tritium retention and permeation in neutron-irradiated materials [M. Shimada et.al., Rev. Sci. Instru. 82 (2011) 083503 and and M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. Recently the TPE has undergone major upgrades in its electrical and control systems. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium. We discuss the electrical upgrade, enhanced operational safety, improved plasma performance, and development of tritium plasma-driven permeation and optical spectrometer system. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, Fusion Nuclear Science Facility (FNSF), and Demonstration reactor (DEMO). This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

    The results of tritium and helium gas release from lithium ceramics samples Li{sub 2}TiO{sub 3} irradiated at the WWR-K reactor (Almaty, Kazakhstan) and from beryllium samples irradiated at the BN-350 reactor (Aktau, Kazakhstan) and the IVG.1M reactor (Kurchatov, Kazakhstan) are presented. Experimentally obtained thermal desorption (TDS) spectra have shown that the dependence of tritium release from lithium ceramics has a complicated behavior and to a large extent depends on lithium ceramics type. Nevertheless, it was found that the total amount of tritium released from all types of lithium ceramics has the same order of magnitude, equal to about 10{sup 11} Bq/kg. It was found that in the temperature range from 523 K to 1373 K the process of tritium release from lithium ceramics involves volume diffusion and thermoactivated tritium release from the accumulation centers generated under irradiation. TDS of beryllium samples enables us to obtain characteristics of tritium and helium release during linear heating, to determine integrated quantities of generated helium and tritium, and to determine parameters of release processes.

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

    OpenAIRE

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

    1981-01-01

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

  19. Limitation of tritium outgassing from tritiated solid waste drums

    International Nuclear Information System (INIS)

    In the framework of the development of fusion thermonuclear reactors, tritiated solid waste is foreseen and will have to be managed. The management of tritiated waste implies limitations in terms of activity and tritium degassing. The degassing tritium can be under the form of tritiated hydrogen, tritiated water and, in some specific cases, negligible amount of tritiated volatile organic compound. Hence, considering the major forms of degassing tritium, CEA has developed a mixed-compound dedicated to tritium trapping in drums. Based on several experiments, the foreseen mixed compound is composed of MnO2, Ag2O, Pt and molecular sieve, the three first species having the ability to convert tritiated hydrogen into tritiated water and the last one acting as a trap for tritiated water. To assess the performance of the trapping mixture, experimental tests were performed at room temperature on tritiated dust composed of beryllium and carbon. It was shown that the metallic oxides mixture used for tritiated hydrogen conversion is efficient and that tritiated water adsorption was limited due to an inefficient regeneration of the molecular sieve prior to its use. Apart from this point, the tritium release from waste was reduced by a factor of 5.5, which can be improved up to 87 if the adsorption step is efficient

  20. Limitation of tritium outgassing from tritiated solid waste drums

    Energy Technology Data Exchange (ETDEWEB)

    Liger, K.; Trabuc, P.; Lefebvre, X.; Troulay, M.; Perrais, C. [CEA, Centre de Cadarache, DEN/DTN/STPA/LIPC, Saint-Paul-lez-Durance (France)

    2015-03-15

    In the framework of the development of fusion thermonuclear reactors, tritiated solid waste is foreseen and will have to be managed. The management of tritiated waste implies limitations in terms of activity and tritium degassing. The degassing tritium can be under the form of tritiated hydrogen, tritiated water and, in some specific cases, negligible amount of tritiated volatile organic compound. Hence, considering the major forms of degassing tritium, CEA has developed a mixed-compound dedicated to tritium trapping in drums. Based on several experiments, the foreseen mixed compound is composed of MnO{sub 2}, Ag{sub 2}O, Pt and molecular sieve, the three first species having the ability to convert tritiated hydrogen into tritiated water and the last one acting as a trap for tritiated water. To assess the performance of the trapping mixture, experimental tests were performed at room temperature on tritiated dust composed of beryllium and carbon. It was shown that the metallic oxides mixture used for tritiated hydrogen conversion is efficient and that tritiated water adsorption was limited due to an inefficient regeneration of the molecular sieve prior to its use. Apart from this point, the tritium release from waste was reduced by a factor of 5.5, which can be improved up to 87 if the adsorption step is efficient.

  1. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

    When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

  2. Tritium transport around nuclear faciliteis

    International Nuclear Information System (INIS)

    The transport and cycling of tritium around nuclear facilities is reviewed with special emphasis on studies at the Savannah River Laboratory, Aiken, South Carolina. These studies have shown that the rate of deposition from the atmosphere, the site of deposition, and the subsequent cycling are strongly influenced by the compound with which the tritium is associated. Tritiated hydrogen is largely deposited in the soil, while tritiated water is deposited in the greatest quantity in the vegetation. Tritiated hydrogen is converted in the soil to tritiated water that leaves the soil slowly, through drainage and transpiration. Tritiated water deposited directly to the vegetation leaves the vegetation more rapidly after exposure. Only a small part of the tritium entering the vegetation becomes bound in organic molecules. However, it appears that the existence of soil organic compounds with tritium concentrations greater than the equilibrium concentration in the associated water can be explained by direct metabolism of tritiated hydrogen in vegetation. (J.P.N.)

  3. Tritium transport around nuclear facilities

    International Nuclear Information System (INIS)

    The transport and cycling of tritium around nuclear facilities is reviewed with special emphasis on studies at the Savannah River Laboratory, Aiken, South Carolina. These studies have shown that the rate of deposition from the atmosphere, the site of deposition, and the subsequent cycling are strongly influenced by the compound with which the tritium is associated. Tritiated hydrogen is largely deposited in the soil, while tritiated water is deposited in the greatest quantity in the vegetation. Tritiated hydrogen is converted in the soil to tritiated water that leaves the soil slowly, through drainage and transpiration. Tritiated water deposited directly to the vegetation leaves the vegetation more rapidly after exposure. Only a small part of the tritium entering the vegetation becomes bound in organic molecules. However, it appears tht the existence of soil organic compounds with tritium concentrations greater than the equilibrium concentration in the associated water can be explained by direct metabolism of tritiated hydrogen in vegetation

  4. Evaluation of retention and disposal options for tritium in fuel reprocessing

    International Nuclear Information System (INIS)

    Five options were evaluated as means of retaining tritium released from light-water reactor or fast breeder reactor fuel during the head-end steps of a typical Purex reprocessing scheme. Cost estimates for these options were compared with a base case in which no retention of tritium within the facility was obtained. Costs were also estimated for a variety of disposal methods of the retained tritium. The disposal costs were combined with the retention costs to yield total costs (capital plus operating) for retention and disposal of tritium under the conditions envisioned. The above costs were converted to an annual basis and to a dollars per curie retained basis. This then was used to estimate the cost in dollars per man-rem saved by retaining the tritium. Only the options that used the least expensive disposal costs could approach the $1000/man-rem cost used as a guide by the Nuclear Regulatory Commission

  5. Quantification of exchangeable and non-exchangeable organically bound tritium (OBT) in vegetation.

    Science.gov (United States)

    Kim, S B; Korolevych, V

    2013-04-01

    The objective of this study is to quantify the relative amounts of exchangeable organically bound tritium (OBT) and non-exchangeable OBT in various vegetables. A garden plot at Perch Lake, where tritium levels are slightly elevated due to releases of tritium from a nearby nuclear waste management area and Chalk River Laboratories (CRL) operations, was used to cultivate a variety of vegetables. Five different kinds of vegetables (lettuce, cabbage, tomato, radish and beet) were studied. Exchangeable OBT behaves like tritium in tissue free water in living organisms and, based on past measurements, accounts for about 20% of the total tritium in dehydrated organic materials. In this study, the percentage of the exchangeable OBT was determined to range from 20% to 57% and was found to depend on the type of vegetables as well as the sequence of the plants exposure to HTO. PMID:23220540

  6. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellet. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  7. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) phase. The existing TFTR deuterium pellet injector (DPI) has been modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed to provide pellets ranging from 3.3 to 4.5 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single- stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller. A new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed, and the TPI was tested at ORNL with deuterium pellets. Results of the limited testing program at ORNL are described. The TPI is being installed on TFTR to support the D-D run period in 1992. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and secondary tritium containment systems and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  8. Correlation between balloon release pressure and no-reflow in patients with acute myocardial infarction undergoing direct percutaneous coronary intervention

    Institute of Scientific and Technical Information of China (English)

    Wang Yanfei; Yao Min; Liu Haibo; Yang Yuejin; Xie Junmin; Jia Xinwei; Pan Huanjun

    2014-01-01

    Background Balloon release pressure may increase the incidence of no reflow after direct percutaneous coronary intervention (PCI).This randomized controlled study was designed to analyze the correlation between balloon release pressure and no-reflow in patients with acute myocardial infarction (AMI) undergoing direct PCI.Methods There were 156 AMI patients who underwent PCI from January 1,2010 to December 31,2012,and were divided into two groups according to the stent inflation pressure:a conventional pressure group and a high pressure group.After PCI,angiography was conducted to assess the thrombolysis in myocardial infarction (TIMI) grade with related artery.Examinations were undertaken on all patients before and after the operation including cardiac enzymes,total cholesterol,low-density lipoprotein,blood glucose,homocysteine,β-thromboglobulin (β-TG),Hamilton depression scale (HAMD) and self-rating anxiety scale (SAS).After interventional therapy,the afore-mentioned parameters in both the conventional pressure group and high pressure group were again analyzed.Results The results showed that CK-MB,HAMD,SAS were significantly different (P <0.05) in all patients after PCI,especially the CK-MB in the high pressure group ((25.7±7.6) U/L vs.(76.7±11.8) U/L).CK-MB,HAMD,SAS,and β-TG were comparative before PCI but they were significantly changed (P <0.05) after intervention.No-reflow phenomenon occurred in 13 patients in the high pressure group,which was significantly higher than in the conventional pressure group (17.11% vs.6.25%,P<0.05).Conclusion In stent implantation,using a pressure less than 1823.4 kPa balloon to release pressure may be the better choice to reduce the occurrence of no-reflow following direct PCI.

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

    Science.gov (United States)

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

    1981-12-01

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

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

    Directory of Open Access Journals (Sweden)

    Zhang Weihui

    2010-05-01

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

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

    NARCIS (Netherlands)

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

    2007-01-01

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

  12. Trace tritium recovery from the residue of liquid Li17Pb83 alloy

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The liquid Li17Pb83 alloy is a prominent breeder material for use in a fusion reactor.In the design of an effective tritium extraction system for the liquid lithium lead bubbler of the test blanket module of such a reactor,finding ways to strictly limit the losses of tritium and to minimize radioactive risks is very important.For this purpose,the isotope exchange process has been investigated as a means of trace tritium recovery from a model of the residue from Li17Pb83 alloy.The results indicate that the isotope exchange process is an effective means of tritium recovery from the residue of Li17Pb83 alloy,and the optimum composition of the exchange carrier gas is He + 0.1% D2.The exchange temperature and number of exchange steps are the main factors influencing the efficiency of tritium recovery from the residue.Trace tritium recovery efficiency increases with increasing exchange temperature and number of times of exchange.Tritium recovery efficiency can approach 80% when the residue is treated six times at 823 K.A gas-liquid two-phase contact model to describe the proceeding of tritium release from the liquid Li17Pb83 alloy has been derived on the basis of this experiment.

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

    International Nuclear Information System (INIS)

    With the prospect of large fusion facilities coming on-line in the not-too-distant future, it is becoming increasingly important that an on-line tritium-monitoring system be developed which is capable of detecting small amounts of released tritium. Since tritium oxide is some 400 times as hazardous as elemental tritium, it is necessary to distinguish between the two in order to properly evaluate the hazard. Presently available on-line instrumentation has marginal sensitivity, is unable to distinguish between the two forms of tritium, and has poor discrimination against background gamma radiation and air activation products. The objective of our program is to develop a monitoring system with the capability of distinguishing between the two forms of tritium, detecting tritium with a sensitivity of a fraction of an MPC/sub a/ (1 MPC/sub a/ = 5. x 10-6 Ci/M3) for the oxide, and discriminating against gamma activity and airborne activity other than tritium

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-09-01

    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/sup 0/C. Beyond 600/sup 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.

  15. An Estimate of the History of Tritium Inventory in Wood Following Irrigation with Tritiated Water

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E.

    2001-06-15

    Some of the groundwater and surface water at the Savannah River Site (SRS) is contaminated with tritium as a legacy of nuclear materials production. An analysis of tritium remediation alternatives suggests that the most practical remediation alternative is to change in the path of tritium exposure to the public. Calculations based on many years of experience at the Savannah River Site indicate that a 40 percent reduction in dose can be achieved by releasing tritiated water to the atmosphere, as water vapor, as opposed to allowing it to flow off site in surface water streams.

  16. 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.

  17. 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.

  18. Overview of Recent Tritium Experiments in TPE

    Energy Technology Data Exchange (ETDEWEB)

    Masashi Shimada; T. Otsuka; R. J. Pawelko; P. Calderoni; J. P. Sharpe

    2010-10-01

    Tritium retention in plasma-facing components influences the design, operation, and lifetime of fusion devices such as ITER. Most of the retention studies were carried out with the use of either hydrogen or deuterium. Tritium Plasma Experiment is a unique linear plasma device that can handle radioactive fusion fuel of tritium, toxic material of beryllium, and neutron-irradiated material. A tritium depth profiling method up to mm range was developed using a tritium imaging plate and a diamond wire saw. A series of tritium experiments (T2/D2 ratio: 0.2 and 0.5 %) was performed to investigate tritium depth profiling in bulk tungsten, and the results shows that tritium is migrated into bulk tungsten up to mm range.

  19. Crediting Tritium Deposition in Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E. Jr.

    2001-06-20

    This paper describes the major aspects of tritium dispersion phenomenology, summarizes deposition attributes of the computer models used in the DOE Complex for tritium dispersion, and recommends an approach to account for deposition in accident analysis.

  20. Recommended radiological controls for tritium operations

    International Nuclear Information System (INIS)

    This informal report presents recommendations for an adequate radiological protection program for tritium operations. Topics include hazards analysis, facility design, personnel protection equipment, training, operational procedures, radiation monitoring, to include surface and airborne tritium contamination, and program management

  1. Toxicity and dosimetry of tritium

    International Nuclear Information System (INIS)

    Tritium doses to the general public are very low (currently about 0.2 μSv per year). Radiation doses from tritium to members of the public living in the vicinity of a CANDU power station are higher but rarely exceed 20 μSv per year or 1% of normal exposures to radiation from all natural sources, but doses to some radiation workers can approach ten mSv per year. The relative biological effectiveness (RBE) of tritium beta rays varies appreciably depending upon the biological endpoint. Observed RBE values at low doses and low dose-rates are usually about 2 to 3 when tritium beta rays are compared to 60Co gamma rays but are closer to 1 than to 2 when compared to 200 kVp X-rays. This conclusion is supported by microdosimetric considerations of the quality of tritium beta rays, 60Co gamma rays and X-rays. Since X-rays have traditionally been accepted as reference radiation by the International Commission on Radiological Protection, it seems reasonable that the quality factor (Q) assigned to tritium beta rays should be close to one. Recommended procedures in Canada for estimation of effective dose equivalents from exposures to HTO and HT assume that Q = 1 and that body water represents 67% of the mass of soft tissue; they take into account conversions of HTO to appear to be reasonable for radiation protection purposes when the source of exposure is HTO or HT, but will not be adequate for exposures to other tritiated compounds. (modified author abstract) (137 refs., 11 figs., 12 tabs.)

  2. Tritium accident containment within a large fusion enclosure: cost, benefit, and risk considerations

    International Nuclear Information System (INIS)

    Containment of a tritium accident within a large fusion device building will be difficult and costly. Complete containment is impossible, and with this fact in mind, the global dispersion and health effects of tritium are reviewed. Atmospheric tritium emissions lead to an estimated population dose to the Northern Hemisphere of 5.6 x 10-3 man-rem/Ci, which may also be interpreted as 1 cancer fatality per MCi. Updating the NRC $1000 per man-rem criterion to 1982 costs gives 9.5 $/y per Ci/y as the unit annual health benefit rate from averting tritium release at a continuous rate. Present worth considerations lead to an estimate of $100 per Ci/y for the maximum capital investment justified per expected curie per year of tritium release averted. A simplified enclosure model is used to explore the trade-off between processing capacity and recycle time with the health cost of residual tritium release included in the analysis

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

    International Nuclear Information System (INIS)

    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

  4. Development of a tritium transport analysis code for the LMFBR system

    Energy Technology Data Exchange (ETDEWEB)

    Iizawa, Katsuyuki; Torii, Tatsuo [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Tsuruga, Fukui (Japan)

    2001-03-01

    A tritium transport analysis code for the LMFBR system, TTT code, has been developed and validated using data from a power rising test conducted at Monju in 1995. The behavior of tritium during future long-term full power operation of Monju has been estimated. The TTT code was created from the tritium and hydrogen transport model devised by R. Kumar and ANL. Actual data from some plants has been used to improve the code. In this study, we used data from Monju to increase the accuracy of the calculated to measured ratio, the C/E ratio. As a result of the study, we were able to: 1. show that the calculated tritium concentration distribution and the change in the primary and secondary sodium, steam and water correlated sufficiently closely with the measured, C/E ratio of 1.1; 2. propose a transport model between sodium and the cover gas system taking into account the mechanisms affecting the partial pressure difference and the isotopic exchange of H and H3; 3. examine the considerable effect of the hydrogen source within the sodium cooling system of Monju on tritium behavior and clarify the characteristics at the initial stage of plant; 4. estimate the tritium transport and distribution for the long-term full power operation of Monju. The tritium release from the core will be 7,400 TBq during 30 years of operation. The primary and secondary cold trap will capture 99% of this and 1% or less will be released to the environment as gaseous radioactive waste from stack and its drainage water from SG; and 5. compare the best fitted tritium source rates from cores in Phenix and Monju and estimate the major release from Monju's helium bond closed type control rods. (author)

  5. Development of a tritium transport analysis code for the LMFBR system

    International Nuclear Information System (INIS)

    A tritium transport analysis code for the LMFBR system, TTT code, has been developed and validated using data from a power rising test conducted at Monju in 1995. The behavior of tritium during future long-term full power operation of Monju has been estimated. The TTT code was created from the tritium and hydrogen transport model devised by R. Kumar and ANL. Actual data from some plants has been used to improve the code. In this study, we used data from Monju to increase the accuracy of the calculated to measured ratio, the C/E ratio. As a result of the study, we were able to: 1. show that the calculated tritium concentration distribution and the change in the primary and secondary sodium, steam and water correlated sufficiently closely with the measured, C/E ratio of 1.1; 2. propose a transport model between sodium and the cover gas system taking into account the mechanisms affecting the partial pressure difference and the isotopic exchange of H and H3; 3. examine the considerable effect of the hydrogen source within the sodium cooling system of Monju on tritium behavior and clarify the characteristics at the initial stage of plant; 4. estimate the tritium transport and distribution for the long-term full power operation of Monju. The tritium release from the core will be 7,400 TBq during 30 years of operation. The primary and secondary cold trap will capture 99% of this and 1% or less will be released to the environment as gaseous radioactive waste from stack and its drainage water from SG; and 5. compare the best fitted tritium source rates from cores in Phenix and Monju and estimate the major release from Monju's helium bond closed type control rods. (author)

  6. Computer simulation of tritium removal facility design

    International Nuclear Information System (INIS)

    In this study, a computer simulation of tritium diffusion out of molten salt is performed using COMSOL Multiphysics. The purpose of the simulation is to investigate the efficiency of the permeation window type tritium removal facility, which is proposed for tritium control in FHRs. The result of the simulation suggests a large surface area is one of the key issues in the design of the tritium removal facility, and the simple tube bundle concept is insufficient to provide the surface area needed for an efficient tritium removal process. (author)

  7. Management of tritium at nuclear facilities

    International Nuclear Information System (INIS)

    This report presents extending summaries of the works of the participants to an IAEA co-ordinated research programme, ''Handling Tritium - bearing effluents and wastes''. The subjects covered include production of tritium in nuclear power plants (mainly heavy water and light water reactors), as well as at reprocessing plants; removal and enrichment of tritium at nuclear facilities; conditioning methods and characteristics of immobilized tritium of low and high concentration; some potential methods of storage and disposal of tritium. In addition to the conclusions of this three-years work, possible activities in the field are recommended

  8. Tritium hazard via the ingestion pathway

    International Nuclear Information System (INIS)

    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 that allows for 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 the 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. 10 refs., 1 fig., 1 tab

  9. Total tritium measurement in atmosphere

    International Nuclear Information System (INIS)

    Measurement of tritium in the atmosphere is of strong interest wherever this radionuclide is used. Therefore, a method is proposed for the joint measurement of burnable tritium, independently from its physico-chemical form, and of tritiated water. The method consists of transforming the tritiated molecules of the gases present in the air volume into tritiated water by burning them together with a known quantity of hydrogen. The water vapor is condensed and added to a liquid scintillator. The scintillator is also able to dissolve conventional filters so that the tritium attached to particulate and concentrated on these filters can be jointly measured, as will be discussed in a future report. The overall detection limit of the method is approximately 64 Bq m-3 for a combustion period of 10 min (which corresponds to sampling an air volume of 15 L) and a counting period of 10 min. This limit, much lower than the derived air concentrations in the most unfavorable cases, allows the application of the method for safety purposes. Moreover, the method can be integrated into a general procedure for the measurement of tritium in different chemical forms, to be applied in case of necessity

  10. Weapons engineering tritium facility overview

    Energy Technology Data Exchange (ETDEWEB)

    Najera, Larry [Los Alamos National Laboratory

    2011-01-20

    Materials provide an overview of the Weapons Engineering Tritium Facility (WETF) as introductory material for January 2011 visit to SRS. Purpose of the visit is to discuss Safety Basis, Conduct of Engineering, and Conduct of Operations. WETF general description and general GTS program capabilities are presented in an unclassified format.

  11. Tritium dynamics in soils and plants grown under three irrigation regimes at a tritium processing facility in Canada.

    Science.gov (United States)

    Mihok, S; Wilk, M; Lapp, A; St-Amant, N; Kwamena, N-O A; Clark, I D

    2016-03-01

    The dynamics of tritium released from nuclear facilities as tritiated water (HTO) have been studied extensively with results incorporated into regulatory assessment models. These models typically estimate organically bound tritium (OBT) for calculating public dose as OBT itself is rarely measured. Higher than expected OBT/HTO ratios in plants and soils are an emerging issue that is not well understood. To support the improvement of models, an experimental garden was set up in 2012 at a tritium processing facility in Pembroke, Ontario to characterize the circumstances under which high OBT/HTO ratios may arise. Soils and plants were sampled weekly to coincide with detailed air and stack monitoring. The design included a plot of native grass/soil, contrasted with sod and vegetables grown in barrels with commercial topsoil under natural rain and either low or high tritium irrigation water. Air monitoring indicated that the plume was present infrequently at concentrations of up to about 100 Bq/m(3) (the garden was not in a major wind sector). Mean air concentrations during the day on workdays (HTO 10.3 Bq/m(3), HT 5.8 Bq/m(3)) were higher than at other times (0.7-2.6 Bq/m(3)). Mean Tissue Free Water Tritium (TFWT) in plants and soils and OBT/HTO ratios were only very weakly or not at all correlated with releases on a weekly basis. TFWT was equal in soils and plants and in above and below ground parts of vegetables. OBT/HTO ratios in above ground parts of vegetables were above one when the main source of tritium was from high tritium irrigation water (1.5-1.8). Ratios were below one in below ground parts of vegetables when irrigated with high tritium water (0.4-0.6) and above one in vegetables rain-fed or irrigated with low tritium water (1.3-2.8). In contrast, OBT/HTO ratios were very high (9.0-13.5) when the source of tritium was mainly from the atmosphere. TFWT varied considerably through time as a result of SRBT's operations; OBT/HTO ratios showed no clear temporal

  12. Tritium inventory assessment for ITER using TRIMO

    Energy Technology Data Exchange (ETDEWEB)

    Cristescu, Ioana-R. [Forschungszentrum Karlsruhe GmbH, Tritiumlabor, Postfach 3640, D-76021 Karlsruhe (Germany)]. E-mail: ioana.cristescu@hvt.fzk.de; Cristescu, I. [Forschungszentrum Karlsruhe GmbH, Tritiumlabor, Postfach 3640, D-76021 Karlsruhe (Germany); Murdoch, D. [EFDA Close Support Unit, Boltzmannstrasse 2, D-85748 Garching bei Muenchen (Germany); Doerr, L. [Forschungszentrum Karlsruhe GmbH, Tritiumlabor, Postfach 3640, D-76021 Karlsruhe (Germany); Glugla, M. [Forschungszentrum Karlsruhe GmbH, Tritiumlabor, Postfach 3640, D-76021 Karlsruhe (Germany)

    2006-02-15

    Currently, the strategy for determination of ITER in-vessel tritium inventory envisages that at predetermined intervals, tritiated gases in all systems of fuel cycle will be transferred to the storage and delivery system (SDS) and tritium quantities measured by in-bed calorimetry. The isotope separation system (ISS) is the system used to separate hydrogen isotopes at the quality required to be stored in SDS, and is one of the systems with highest tritium inventory within the fuel cycle. Therefore, during tritium inventory procedure, ISS has to be 'milked down' of tritium, mainly as DT molecular species. Based on the dynamic modelling code TRIMO of the tritium content in the main sub-systems of ITER Fuel Cycle, the procedure for tritium extraction from ISS is presented and numerical examples given to assess the necessary time of transferring the tritium from ISS to SDS, and the residual amount of tritium in ISS after different milking scenarios. Consequently a fuel handling strategy during tritium inventory assessment in the ISS and SDS is described, with the constraint of mobilizable tritium inventory minimisation.

  13. Extraction of tritium from ceramic breeder material

    International Nuclear Information System (INIS)

    The first generation of fusion reactors will use deuterium and tritium as fuel since this reaction takes place at relatively low temperature. Since tritium is not available in nature, it must be produced in the fusion reactor blanket which surrounds the plasma zone. The lithium bearing compound is available in plenty in earths crust and by absorbing neutron, lithium produces tritium by the reactions 6Li (n, α) T and 7Li (n, n'α) T. Natural lithium consists of 93% 7Li and the remaining 7% as 6Li. Since the inelastic scattering of 7Li with fast neutrons produces one tritium and one neutron, more than one tritium atom can be produced per neutron. Hence by suitably designing the lithium blanket, more than one tritium atom per fusion reaction can be produced. In the absence of thermonuclear reactions, the (D,T) neutrons which are energetic 14-MeV neutrons, are produced in the accelerator based neutron generators. In order to ensure that sufficient amount of tritium would be produced in the future fusion reactor blankets, experiments are carried out to irradiate the lithium assembly using the available neutron source and measurements are done to estimate the tritium breeding. Also, it is required to extract the tritium produced in the lithium blanket. This work consists of tritium breeding measurement technique and a design of tritium extraction system. (author)

  14. Tritium issues in plasma wall interactions

    Science.gov (United States)

    Tanabe, Tetsuo

    2010-05-01

    Since tritium resources are very limited, not only for safety reason but also for tritium economy, tritium inventory in a reactor must be kept as small as possible. In the present tokamaks, however, hydrogen retention rate in their vacuum vessel is significantly large, i.e. more than 20% of fueled hydrogen is continuously piled up, which must not be allowed in a reactor. After the introduction of tritium as a hydrogen radioisotope, the paper summarizes present tritium issues in plasma wall interactions, in particular, fueling, retention and recovering, considering the handling of large amounts of tritium, i.e. confinement, leakage, contamination, permeation, regulations and tritium accountancy. Progress in overcoming such problems will be also presented.

  15. Fusion tritium program in the United States

    International Nuclear Information System (INIS)

    The fusion technology development program for tritium in the US is centered around the Tritium Systems Test Assembly (TSTA) at Los Alamos National Labortory. Objectives of this project are to develop and demonstrate the fuel cycle for processing the reactor exhaust gas (unburned deuterium and tritium plus impurities), and the necessary personnel and environemntal protection systems for the next generation of fusion devices. The TSTA is a full-scale system for an INTOR/ITER sized machine. That is, TSTA has the capacity to process tritium in a closed loop mode at the rate of 1 kg per day, requiring a tritium inventory of about 100 g. The TSTA program also interacts with all other tritium-related fusion technology programs in the US and all major programs abroad. This report is a summary of the results and interactions of the TSTA program since a previous summary was published and an overview of related tritium programs

  16. Evaluation of tritium diffusion through the Neutral Beam Injector calorimeter panel

    Energy Technology Data Exchange (ETDEWEB)

    Borgognoni, Fabio [ENEA, Dipartimento Fusione Tecnologie e Presidio Nucleare, C.R. ENEA Frascati, Via E. Fermi 45, Frascati (RM) I-00044 (Italy)], E-mail: fabio.borgognoni@frascati.enea.it; Moriani, Andrea [ENEA, Dipartimento Fusione Tecnologie e Presidio Nucleare, C.R. ENEA Frascati, Via E. Fermi 45, Frascati (RM) I-00044 (Italy); Sandri, Sandro [ENEA, Dipartimento Biotecnologie, Agroindustria e Protezione della Salute Istituto di Radioprotezione - C.R. ENEA Frascati, Via E. Fermi 45, Frascati (RM) I-00044 (Italy); Tosti, Silvano [ENEA, Dipartimento Fusione Tecnologie e Presidio Nucleare, C.R. ENEA Frascati, Via E. Fermi 45, Frascati (RM) I-00044 (Italy)

    2009-06-15

    The Neutral Beam Test Facility (NBTF) to be realized in Padoa will test the Neutral Beam Injection (NBI), one of the Heating and Current Drive Systems foreseen for ITER. The NBI is based on the acceleration of hydrogen or deuterium negative ions up to 1 MeV. This work has been aimed at assessing the tritium release from the NBTF in order to provide data for the safety analysis. In particular, the diffusion of the tritium through the neutral beam target material (the CuCrZr alloy calorimeter panels) has been assessed by using literature data of the diffusion coefficient. The tritium generated inside the calorimeter panels moves into both the vacuum and water side: the tritium diffusion flux has been evaluated during the beam-on (200 deg. C) and the beam-off (20 deg. C) phases of the NBTF experiments consisting of an interim campaign and a final test. The penetration depth of the tritium through the 2 mm thick CuCrZr alloy material has been also evaluated by using a Monte-Carlo code. As main result, the assessed diffusion flux of tritium during both the beam-on and the beam-off phases are modest. In fact, at the end of the interim campaign (100 days), about the 96% of the all generated tritium (626.5 MBq) exits the calorimeter while the residual tritium inventory (25 MBq) leaves the copper alloy with a diffusion time of about 1 month. At the end of the final test (14 days) about the 99% of the total generated tritium (1.023 x 10{sup 4} MBq) leaves the copper alloy and the remaining tritium inventory (152.2 MBq) is released by about 32 days. In both the interim campaign and the final test, more than the 99% of the total tritium is transferred into the vacuum side of the calorimeter panel while negligible tritium amounts enter the cooling water system thus showing a very low impact on the environ0010me.

  17. Tritium Burn-up Depth and Tritium Break-Even Time

    Institute of Scientific and Technical Information of China (English)

    LI Cheng-Yue; DENG Bai-Quan; HUANG Jin-Hua; YAN Jian-Cheng

    2006-01-01

    @@ Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder (FEB). It is found that the tritium burn-up depth is 317g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18kg that is more than theleast required amount of tritium storage to start up three of FEB-like fusion reactors.

  18. Development and Verification of Tritium Analyses Code for a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim

    2009-09-01

    A tritium permeation analyses code (TPAC) has been developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in the VHTR systems including integrated hydrogen production systems. A MATLAB SIMULINK software package was used for development of the code. The TPAC is based on the mass balance equations of tritium-containing species and a various form of hydrogen (i.e., HT, H2, HTO, HTSO4, and TI) coupled with a variety of tritium source, sink, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of HT and H2 through pipes, vessels, and heat exchangers were importantly considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including both high-temperature electrolysis and sulfur-iodine process. The TPAC has unlimited flexibility for the system configurations, and provides easy drag-and-drops for making models by adopting a graphical user interface. Verification of the code has been performed by comparisons with the analytical solutions and the experimental data based on the Peach Bottom reactor design. The preliminary results calculated with a former tritium analyses code, THYTAN which was developed in Japan and adopted by Japan Atomic Energy Agency were also compared with the TPAC solutions. This report contains descriptions of the basic tritium pathways, theory, simple user guide, verifications, sensitivity studies, sample cases, and code tutorials. Tritium behaviors in a very high temperature reactor/high temperature steam electrolysis system have been analyzed by the TPAC based on the reference indirect parallel configuration proposed by Oh et al. (2007). This analysis showed that only 0.4% of tritium released from the core is transferred to the product hydrogen

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

    International Nuclear Information System (INIS)

    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.)

  20. Preliminary tritium safety analysis on China DFLL-TBM for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Song Yong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)], E-mail: ysong@ipp.ac.cn; Huang Qunying [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Ni Muyi [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230027 (China); Wang Yongliang [College of Physical Science and Technology, Sichuan University, Chengdu, Sichuan, 610064 (China)

    2009-12-15

    The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R and D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.

  1. Preliminary tritium safety analysis on China DFLL-TBM for ITER

    International Nuclear Information System (INIS)

    The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R and D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.

  2. Determination of tritium in the metallic sodium - the BN-350 reactor coolant

    International Nuclear Information System (INIS)

    In the paper the results on tritium determination in the metallic sodium samples - BN-350 reactor coolant - are presented. Tritium activity measurement was carried out on liquid scintillation spectrometer - TriCarb-3100 ('Canberra'). For the spectrometer calibration the solutions prepared on the base NIS USA standards was used. One of principal difficulties in tritium determination in metallic sodium is sodium hydride extraction into solution. Interaction of metallic sodium with water leads to vigorous energy release, heating and explosion. A few methods of sodium dissolution were tested. The sodium dissolution in isopropyl alcohol with small water quantity addition is the most effective sodium dissolution method. Thr process is proceeding smoothly enough, and 1 gram sodium dissolution goes during 1-2 hours. Tritium determination limit estimated by the measurements results makes up 0.1 Bq/g

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

    International Nuclear Information System (INIS)

    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 1014 n/cm2. sec, the fast neutron flux; 1.0 x 1012 n/cm2. sec. Irradiation material is the four pellets of cylindrical Li2O 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 -- 7600C. Four cycles of irradiation tests were carried out from May to August in 1983, and the effective thermal neutron fluence and the burnup of 6Li were 5.9 x 1019nvt 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 6Li(n, α) 3H 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 -- 102μCi/cm3) in the helium gas stream were also investigated. (author)

  4. Handling of tritium contaminated effluents and wastes: Final Report

    International Nuclear Information System (INIS)

    This report deals with the work on: (1) applicability of cotton, woodpulp, sawdust and certian cellulosic derivatives for the removal of tritium from aqueous medium, (2) containment and fixation of tritiated water in nonleachable matrices. The absorption studies on cotton, woodpulp, sawdust, and cellulose acetates were carried out with a view to assess their potentialities as concentration media and also to choose a matrix which can concentrate tritium to the maximum extent possible. The experiments on water hyacinth plants were designed to see the applicability of concentrating tritium and also for providing a via medium for slow release of tritium into the atmosphere. The immobilisation of tritiated water in cement matrices was studied with combinations of portland cement and five filler materials namely sand, silica, vermiculite, portland cement aggregate and accoproof. If cement blocks come in contact with aqueous media as it may happen when the tritium bearing blocks are disposed to the ground, a considerable portion of the contained activity is likely to diffuse and leach out. In order to prevent this, it was proposed to try several coating materials as diffusion barriers over cement blocks. Screening of locally available coating materials was done using a diffusion cell. Shalismatic HD, Anticor and epoxy paint were found to be promising among the screened materials. Tritiated cement blocks with 29% vermiculite loading were coated with the above coating materials, and were subjected to leaching, both in sea water and distilled water. The cumulative leaching data for tritiated cement blocks over a period of 400 days show that Shalimastic HD, when used as a coating material, retards the leaching to the maximum extent. Further leaching studies were started on Shalimastic HD blocks in one ground water formulation, which is continued to this date. (author)

  5. Procurement of tritium for fusion reactor. 2. Transportation of large amounts of tritium for fusion reactors

    International Nuclear Information System (INIS)

    ITER will require kilograms of tritium to be transferred before and after the tritium experiment starts from tritium supplying facilities abroad and/or domestic. Currently, a Zr-Co type transfer container developed in JAERI with a capacity of 25 g tritium is available for international shipping; however, it does not seem enough large for tritium transfer for ITER. This article discusses the technical issues involving in developing a transfer container with a large tritium capacity and regulations governing radio isotope transport containers. (author)

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

    Directory of Open Access Journals (Sweden)

    E. J. Lewis

    1992-01-01

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

  7. The present status and recent applications of the accidental tritium assessment code UFOTRI

    Energy Technology Data Exchange (ETDEWEB)

    Raskob, W. [Forschungszentrum Karlsruhe GmbH, Institut fuer Neutronenphysik und Reaktortechnik, Karlsruhe (Germany)

    1999-03-01

    The computer program UFOTRI can be used for assessing the impact of accidental released tritium in the two chemical forms tritiated water vapour and tritium gas. By applying UFOTRI to potential European sites for ITER, it could be demonstrated that the main goal, the nonevacuation criteria, is fulfilled for the present release limits. Contributions in international studies together with the re-evaluation of experimental data showed that the plant sub-model as well as the soil sub-model are areas for further improvement. (author)

  8. Fabrication, properties, and tritium recovery from solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, C.E. (Argonne National Lab., IL (USA)); Kondo, T. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Roux, N. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)); Tanaka, S. (Tokyo Univ. (Japan)); Vollath, D. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  9. Radiological training for tritium facilities

    International Nuclear Information System (INIS)

    This program management guide describes a recommended implementation standard for core training as outlined in the DOE Radiological Control Manual (RCM). The standard is to assist those individuals, both within DOE and Managing and Operating contractors, identified as having responsibility for implementing the core training recommended by the RCM. This training may also be given to radiological workers using tritium to assist in meeting their job specific training requirements of 10 CFR 835

  10. DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Blanton, P.

    2013-10-10

    A new Bulk Tritium Shipping Package (BTSP) was designed by the Savannah River National Laboratory to be a replacement for a package that has been used to ship tritium in a variety of content configurations and forms since the early 1970s. The BTSP was certified by the National Nuclear Safety Administration in 2011 for shipments of up to 150 grams of Tritium. Thirty packages were procured and are being delivered to various DOE sites for operational use. This paper summarizes the design features of the BTSP, as well as associated engineered material improvements. Fabrication challenges encountered during production are discussed as well as fielding requirements. Current approved tritium content forms (gas and tritium hydrides), are reviewed, as well as, a new content, tritium contaminated water on molecular sieves. Issues associated with gas generation will also be discussed.

  11. HiPER Tritium factory elements

    Science.gov (United States)

    Guillaume, Didier

    2011-06-01

    HiPER will include a Tritium target factory. This presentation is an overview. We start from process ideas to go to first sketch passing through safety principles. We will follow the Tritium management process. We need first a gas factory producing the right gas mixture from hydrogen, Deuterium and Tritium storage. Then we could pass through the target factory. It is based on our LMJ single shot experiment and some new development like the injector. Then comes pellet burst and vapour recovery. The Tritium factory has to include the waste recovery, recycling process with gas purification before storage. At least, a nuclear plant is not a classical building. Tritium is also very special... All the design ideas have to be adapted. Many facilities are necessary, some with redundancy. We all have to well known these constraints. Tritium budget will be a major contributor for a material point of view as for a financial one.

  12. Tritium radioluminescent devices, Health and Safety Manual

    International Nuclear Information System (INIS)

    This document consolidates available information on the properties of tritium, including its environmental chemistry, its health physics, and safe practices in using tritium-activated RL lighting. It also summarizes relevant government regulations on RL lighting. Chapters are divided into a single-column part, which provides an overview of the topic for readers simply requiring guidance on the safety of tritium RL lighting, and a dual-column part for readers requiring more technical and detailed information

  13. Tritium radioluminescent devices, Health and Safety Manual

    Energy Technology Data Exchange (ETDEWEB)

    Traub, R.J.; Jensen, G.A.

    1995-06-01

    This document consolidates available information on the properties of tritium, including its environmental chemistry, its health physics, and safe practices in using tritium-activated RL lighting. It also summarizes relevant government regulations on RL lighting. Chapters are divided into a single-column part, which provides an overview of the topic for readers simply requiring guidance on the safety of tritium RL lighting, and a dual-column part for readers requiring more technical and detailed information.

  14. Overview of light sources powered by tritium

    International Nuclear Information System (INIS)

    Due to their long lifespan and stable intensity, light sources initiated by tritium instead of electricity or batteries are suitable for low level lighting applications. Therefore, tritium-based radioluminescent (RL) light sources are widely used in both military and civil applications. However, traditional tritium lights with the gas tube structure have several shortcomings: (1) the phosphors are opaque; (2) the glass tube is fragile and easily broken; and (3) the beta kinetic energy is attenuated due to the sorption by the gas; etc. As a result, further application of the tritium lights is limited. In this paper, the lighting mechanism and radiation safety of tritium-based RL light sources are briefly reviewed. Besides, the history and prospects of the development of tritium-based RL light source are discussed. Due to their long lifespan and stable intensity, light sources initiated by tritium instead of electricity or batteries are suitable for low level lighting applications. Therefore, tritium- based radioluminescent (RL) light sources are widely used in both military and civil applications. However, traditional tritium lights with the gas tube structure have several short- comings: (1) the phosphors are opaque; (2) the glass tube is fragile and easily broken; and (3) the beta kinetic energy is attenuated due to the sorption by the gas; etc. As a result, further application of the tritium lights is limited. In this paper, the lighting mechanism and radiation safety of tritium-based RL, light sources are briefly reviewed. Besides, the history and prospects of the development of tritium-based RL light source are discussed. (authors)

  15. Tritiated water retention on maize and beans after an acute contamination

    International Nuclear Information System (INIS)

    Although tritium is released in large quantities into environment by nuclear industries and peaceful radioisotope utilization, its behavior is not well known. The International Atomic Energy Agency is sponsoring an international study group to obtain more information about tritiated water (HTO) behavior in different ecological systems. This paper presents the studies made on corn and beans after an acute application of tritiated water during their early stages of growth on an experimental field. Sampling and radiochemical analytical methods of tritium and its behaviour on that plants during their growth cycle are outlined. It is shown that the tritiated water retention plot has at least two components, with effective half lifes of about 10 and 100 days for corn, and 8 and 40 days for beans. (author)

  16. Estimation of tritium and helium inventory in the tritium handling system in Korea

    International Nuclear Information System (INIS)

    In Korea, the Wolsong Tritium Removal Facility (WTRF) is under construction to reduce the amount of tritium present in the moderator and coolant of the CANDU type Wolsong nuclear power plants. Recently, a study on the tritium handling system for recovery of the tritium collected from the WTRF was started. Some tritium would enter the steel of the container walls and subsequently decay to helium. This helium can deteriorate the mechanical properties of the material of the tritium handling system. To evaluate the tritium and helium inventory in the stainless steel wall of this system, the time-dependent diffusion equation was developed, solved and the results are presented in this paper. These results were compared to previous work that evaluated the tritium inventory in the stainless steel wall of 50-L tritium containers. Tritium and helium concentration profiles and the corresponding inventories were evaluated with respect to the various parameters such as exposure time, temperature, and partial pressure. After 24 years, the helium inventory in the wall of the tritium handling system exceeds the tritium inventory. (authors)

  17. The trends of global tritium precipitations

    International Nuclear Information System (INIS)

    The trends of global tritium precipitation from 1953 to 1979 were estimated based on the tritium data published in seven volumes of Environmental Isotope Data by the International Atomic Energy Agency (IAEA). Tritium precipitation samples were collected from 342 stations in the world and tritium concentrations were measured by IAEA and 27 laboratories. Due to repeated atomospheric nuclear explosions, tritium precipitations showed maximum peak in 1963. After the agreement of the Partial Test Ban Treaty in 1964, they have gradually decreased until now showing seasonal variations. To obtain clear trends of tritium precipitations, seasonal and irregular factors were eliminated from the original time-series data using a code developed by the Japanese Economic Planning Agency. Results of analyses were as follows; a) Peak concentrations and precipitations of tritium were observed every year around the period of late spring to summer. b) The maximum annual tritium concentration and precipitation were observed in 1963 for northern hemisphere stations. c) A latitude effect was observed in the northern hemisphere. The maximum concentrations and precipitations were seen at the latitude of approximately 50 deg N. d) Continental stations always showed higher tritium concentrations and precipitations than comparable maritime stations. (author)

  18. Tritium stripping by a catalytic exchange stripper

    International Nuclear Information System (INIS)

    A catalytic exchange process for stripping elemental tritium from gas streams has been demonstrated. The process uses a catalyzed isotopic exchange reaction between tritium in the gas phase and protium or deuterium in the solid phase on alumina. The reaction is catalyzed by platinum deposited on the alumina. The process has been tested with both tritium and deuterium. Decontamination factors (ration of inlet and outlet tritium concentrations) as high as 1000 have been achieved, depending on inlet concentration. The test results and some demonstrated applications are presented

  19. Tritium immobilization and packaging using metal hydrides

    International Nuclear Information System (INIS)

    Tritium recovered from CANDU heavy water reactors will have to be packaged and stored in a safe manner. Tritium will be recovered in the elemental form, T2. Metal tritides are effective compounds in which to immobilize the tritium as a stable non-reactive solid with a high tritium capacity. The technology necessary to prepare hydrides of suitable metals, such as titanium and zirconium, have been developed and the properties of the prepared materials evaluated. Conceptual designs of packages for containing metal tritides suitable for transportation and long-term storage have been made and initial testing started. (author)

  20. Post-irradiation measurements of residual tritium and helium 4 quantities in lithium aluminate (γLiAlO2)

    International Nuclear Information System (INIS)

    The blanket of future fusion reactors will be used to recover heat and tritium. It is therefore essential to be able to extract tritium. It is possible to extract the totality of the residual tritium and helium contained in irradiated γLiAlO2 samples heated under vacuum or by means of a buyoant gas between 9300C and 10000C. The irradiation temperature and the nature of the barrel material play a dominant part in tritium recovery. The residual tritium content decreases when the irradiation temperature increases in the range 250-5200C. The reduction of the tritiated water released on the steel barrel minimizes the residual tritium content. This cannot be observed on a barrel made up of an Aluminium-Magnesium alloy. Beyond 5000C, surface desorption phenomena are dominant, the increase in the residual tritium content is proportional to the specific surface area of the samples. At 3000C, this content is higher in the coarse grained samples. The helium 4 retention is low and not so dependent on the irradiation temperature. The fraction of residual helium 4 in coarse grained samples is lower than in fine grained samples. At 9750C, the helium 4 release is controlled by diffusion phenomena, this is not the case at 6550C and 7500C

  1. Evidence for tritium production in the Earth's interior

    Institute of Scientific and Technical Information of China (English)

    JIANG SongSheng; HE Ming

    2008-01-01

    We have made a new investigation on the vertical profiles of tritium and helium isotopes in Lakes Van and Nemrut (Eastern Turkey), using experimental data from the reference by Kipfer et al. for study of long-term vertical mixing and deep water renewal in Lake Van. Lakes Van and Nemrut are crater lakes. Lake Nemrut is at the western border of Lake Van. The 3He and 4He are injected at the bottom of Lakes Van and Nemrut, and the both helium isotopes are confirmed from the mantle source. From 3H (tritium) data in Lakes Van and Nemrut, we have observed "3H anomaly" at the vertical profile of 3H concentrations in Lake Nemrut. The 3H concentration at the lake bottom is 10% higher than at the surface. The difference of 3H concentrations between surface and bottom is about 3.7±1.2 TU. This excess 3H should be injected from the lake bottom. An investigation on the origin of the injected tritium has been made. The results show the conventional origins are excluded, such as residence of precipitation tritium from nuclear testing in the early 1950s-1960s and tritium from known nuclear reactions. Based on the correlation of excess 3H with 3He and heat flow in Lake Nemrut, we infer that the 3He and 3H might be all from the mantle source, and produced by the supposed natural-nuclear-fusion, which might occur in an environment rich in water (H) and (U + Th) at high temperature and high pressure in the deep Earth. Detection of tritium in the Earth's interior is a key evidence for exploration of natural nuclear fusion in the deep Earth. Based on the published data, we have found that the excess 3He and 3H injected at the bottom of Lake Laacher (Germany) were also released from the mantle source. The present work will be helpful to the further study of mechanism of natural nuclear fusion in the Earth's interior.

  2. Use of passive sampling for atmospheric tritium monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Caldeira Ideias, P.; Pierrard, O.; Tournieux, D. [Institut de Radioprotection et de Surete Nucleaire - IRSN (France); Tenailleau, L. [Marine nationale (France)

    2014-07-01

    Tritium is one of the most important radionuclide in environmental radiological monitoring. In French civil and military nuclear facilities, the releases levels are between 100 to 100 000 higher than any other radionuclide (rare gas excluded). Moreover these levels will probably increase in the next decades. With an average energy of 6 keV, the beta particle from tritium radioactive decay is difficult to detect and quantify within the environmental levels. To monitor the tritium in the air, French actors (authorities, operator, and experts) commonly use atmospheric bubblers and water vapour condensers. This type of sampling approach is time-consuming and very costly. To simplify and complete these methods, the Institute for Radiological Protection and Nuclear Safety (IRSN), had developed an atmospheric tritium monitoring device based on passive sampling. The passive sampler developed consists in a small container designed with a patented specific geometry and filled with 13X molecular sieve. This system is based on free diffusion flow principle (Fick's law). The driving force is the partial pressure gradient existing between the environmental atmosphere and the passive sampler. The constancy of the sampling rate for different moisture conditions assures the representativeness of the proposed device. The desorption bench developed specifically allows the recovery of 99% of the water vapour sampled in the molecular sieve. More than 99% of the sampled tritium (HTO) activity is recovered in the range between 0 and 100 Bq.L{sup -1}. Above 100 Bq.L{sup -1} to 25 k Bq.L{sup -1} (max tested activity), it was verified that no more than 3% of the tritium remains in the molecular sieve.. Thus, the use of passive sampler provides: - a representative sampling method, - a good detection limit (0,01 Bq.m{sup -3}), - no electric power supply needs, - a wide range of sampling duration (1 day to 1 month), - a low-cost method for monitoring. Different performance tests were

  3. Sustained Liver Glucose Release in Response to Adrenaline Can Improve Hypoglycaemic Episodes in Rats under Food Restriction Subjected to Acute Exercise

    Directory of Open Access Journals (Sweden)

    Lucas K. R. Babata

    2014-01-01

    Full Text Available Background. As the liver is important for blood glucose regulation, this study aimed at relating liver glucose release stimulated by glucagon and adrenaline to in vivo episodes of hypoglycaemia. Methods. The blood glucose profile during an episode of insulin-induced hypoglycaemia in exercised and nonexercised male Wistar control (GC and food-restricted (GR, 50% rats and liver glucose release stimulated by glucagon and adrenaline were investigated. Results. In the GR, the hypoglycaemic episodes showed severe decreases in blood glucose, persistent hypoglycaemia, and less complete glycaemic recovery. An exercise session prior to the episode of hypoglycaemia raised the basal blood glucose, reduced the magnitude of the hypoglycaemia, and improved the recovery of blood glucose. In fed animals of both groups, liver glucose release was activated by glucagon and adrenaline. In fasted GR rats, liver glycogenolysis activated by glucagon was impaired, despite a significant basal glycogenolysis, while an adrenaline-stimulated liver glucose release was recorded. Conclusions. The lack of liver response to glucagon in the GR rats could be partially responsible for the more severe episodes of hypoglycaemia observed in vivo in nonexercised animals. The preserved liver response to adrenaline can partially account for the less severe hypoglycaemia in the food-restricted animals after acute exercise.

  4. Low-level measurements of tritium in water

    International Nuclear Information System (INIS)

    Using a liquid scintillation counter, an experimental procedure for measuring low-level activity concentrations of tritium in environmental water has been developed by our laboratory, using the electrolytic tritium enrichment. Additionally, some quality tests were applied in order to assure the goodness of the method. Well-known water samples collected in the Tagus River (West of Spain) and the Danube River (Bulgaria), both affected by nuclear plant releases, were analysed and results were compared to previous data. The analytical procedure was applied to drinking water samples from the public water supply of Seville and mineral waters from different springs in Spain in order to characterize its origin. Due to the very low levels of tritium in the analysed samples, some results were reported as lower than the minimum detectable activity concentration (MDA). However, the count rate of these measurements was over the background count rate of LS counter in all the cases. For that reason, an exhaustive discussion about the meaning of the MDA, using an experimental essay, was made in order to establish a rigorous criterion that leads to a reliable value in the case of low-level measurements

  5. Application of tritium behavior simulation code (TBEHAVIOR) to an actual-scale tritium handling room

    International Nuclear Information System (INIS)

    It is essential from the viewpoint of fusion safety to confine and remove tritium in a room since tritium handling room is placed as 'final barrier' of fusion plant to prevent the environmental discharge of tritium. At the Tritium Process Laboratory (TPL) of Japan Atomic Energy Agency (JAEA), the application of our original three-dimensional TBEHAVIOR code to the tritium behavior in a room of 3000 m3 was verified. The Renormalization Group Theory (RNG) model was selected as Low-Reynolds model for practical calculation time as well as to reasonable precision in evaluation of velocity from the engineering viewpoint. A series of evaluated results indicated that a flow adjacent to a wall surface plays an important role for tritium transport in a ventilated room. Evaluation of attenuating behavior is further important since the ventilation is normally stopped for the tritium confinement in the case of tritium leakage. We demonstrated that an attenuating behavior can also be evaluated well by the TBEHAVIOR code. Even an attenuating or stagnant flow of less than 10mm/s in a room mixed tritium concentration uniform promptly. The presence of apparatuses in a room did not generally affect tritium behavior. Although the effect of buoyancy was limited to the initial period after the leak, the spread of tritium was promoted by buoyancy. It led to the shortening of elapsed time until the concentration became uniform. (author)

  6. Development of CROPTRIT Model: The Dynamics of Tritium in Agricultural Crops

    Energy Technology Data Exchange (ETDEWEB)

    Galeriu, Dan; Melintescu, Anca [' Horia Hulubei' National Institute for Physics and Nuclear Engineering, Department of Environmental Physics and Life, 30 Reactorului St., POB MG-6, Bucharest-Magurele, RO-077125 (Romania); Lazar, Catalin [National Agricultural Research and Development Institute Fundulea, 915200 Fundulea, Calarasi County (Romania)

    2014-07-01

    Tritium has a complex behaviour once released into the environment. Tritium can be effectively incorporated into biological systems, including the human body, as organically bound tritium (OBT) with a larger residence time than tritiated water (HTO). In the last years robust models were developed for tritium dynamics in mammals (human included), birds and fish but all of them depend on the knowledge of intake for both terrestrial or aquatic food chain. The uncertainty of the present models for tritium in crops following an accidental atmospheric release, is very high and has impacts on the engineering actions for handling and decreasing the nuclear risk. The gaps in knowledge or the local variability of key parameters were recognised as source of uncertainty. Based on an interdisciplinary approach, CROPTRIT model was gradually developed in the last decade focusing on the detecting of the uncertainty sources. Crops of interest depends on each specific case but wheat and rice cover the majority of the practical needs for radiological risk modelling (the major food in Europe and Asia). An analysis of the processes involved in the Soil-Vegetation-Atmosphere Transfer (SVAT) of tritium was done in connection with the available experimental results. The agricultural research is focused on the improving of the yield and the crop growth models were developed in relation with the genotype, weather and management of fertilisation and water. For the radiological purposes, the interest lies in the pollutant concentration at harvest and the CROPTRIT model is focused on the influence of various processes contributing to variability and uncertainty of tritium (OBT and HTO) at harvest. The current results evidentiate the role of the stomatal conductance and difficulties at the day/night transitions, as well as the complex behaviour of the maintenance respiration. A review of the experimental results demonstrates the importance of OBT formation in night conditions and difficulties

  7. Detaching test of an irradiated mock-up containing with tritium from the core of JMTR

    International Nuclear Information System (INIS)

    The second in-situ irradiation experiment using a mock-up (ORIENT-II, JMTR capsule Number: 99M-54J) with a tritium breeder (Li2TiO3) pebble bed in the Japan Materials Testing Reactor (JMTR) was finished on Aug. 1, 2006. Correspondingly an investigation on the detaching procedure of the irradiated mock-up containing with tritium was carried out, followed by the actual detaching test of this mock-up. Firstly, tritium removal characteristics were studied for the irradiated mock-up, the sweep gas tube, the protective tube and the junction box, Out-of pile melting/enclosing tests of the sealing plug were also carried out for prevention of tritium leakage from sweep gas lines of Li2TiO3 pebble bed. From the results, tritium release amount were estimated during the detaching test of the real irradiated mock-up was estimated, and the melting/enclosing procedures of sealing plug were fixed. Then, the actual detaching test of the Li2TiO3 pebble bed was carried out. The tritium release to the area of detaching test was favorably suppressed, decreased, and the irradiated mock-up was safely detached from the core of JMTR as planned. This report describes the results of 1) tritium removal tests for the sweep gas line and the protective tube, 2) out-of pile melting/enclosing test of the sealing plug, 3) examination of the detaching procedure before the detaching test of the irradiated mock-up, and 4) the actual detaching test, as well as knowledge obtained from these tests and works. (author)

  8. Feasibility study of an experiment to measure the RBE of tritium for the induction of myeloid leukemia in animals

    International Nuclear Information System (INIS)

    A variety of RBE values ranging from about 1 to 3 for tritium have been measured by different investigators. The reason for this range in RBE can be attributed to differences in the biological endpoints measured, the reference radiation to which the effects of tritium were compared, and the tritium dosimetry of the particular study. Since the principal risk of low-level irradiation is the induction of cancers, it would be desirable to utilize this endpoint in tritium RBE experiments if these experiments are to be used to evaluate the quality factor for tritium. Furthermore, it would be desirable to use 200 kVp X-rays as the reference radiation since this radiation was suggested by ICRP as the standard reference to be used in the calculation of dose equivalents for purposes of radiation protection. Acute myeloid leukemia is one of the earliest recognized examples of radiogenic cancer in humans and this endpoint has also been the subject of animal studies. This report gives the results of a review of these animal studies to see if this endpoint is suitable for an experiment to measure the tritium RBE relative to 200 kVp X-rays. It was concluded that the male CBA/H mouse, would be a suitable species and an experiment involving 5000 animals in a four to five year study would be required to provide a useful estimate of the RBE for tritium. 72 refs

  9. Synthesis of tritium-labeled fosfomycin

    Energy Technology Data Exchange (ETDEWEB)

    Mertel, H.E.; Meriwether, H.T. (Merck Sharp and Dohme Research Labs., Rahway, NJ (USA))

    1982-03-01

    Tritium gas was used as a labeling agent for the preparation of (1,2-/sup 3/H)fosfomycin. Introduction of tritium into a precursor, the synthesis including resolution of the intermediate racemic 1,2-epoxypropylphosphonic acid, and preparation of both amine and calcium salts of the labeled antibiotic are described.

  10. Environmental tritium monitoring around Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    The environmental tritium monitoring in the sea near Tokai Reprocessing Plant has been performed since 1977, the year of having started the hot test operation of the plant. On the other hand, atmospheric tritium monitoring was started almost at the same time as a research program instead of a routine program. This paper is a review for tritium monitoring in the sea and in the air around the Tokai Reprocessing Plant. The plant is located in Tokai Village, Ibaraki Prefecture, on the Pacific coast. It is based on the Purex process, and the nominal capacity is 210 tons per year. Around the TRP, there are four uranium fabrication facilities, five research reactors, two power reactors and other research facilities. About 173,000 inhabitants are within 10 km range from the plant. The authorized discharge limit of tritium is 200 Ci per day and 51,100 Ci per year in the sea. That in the atmosphere is 50 Ci per day and about 15,000 Ci per year. The tritium from the TRP was discharged mainly into the sea. The sea water samples were distilled, and the tritium concentration was measured by liquid scintillation counting. During three years of the hot operation of TRP, discharged tritium was about 7,000 Ci into the sea and about 140 Ci into the atmosphere. The tritium level has been maintained, and its significant increase was not observed. (Kako, I.)

  11. Tritium waste disposal technology in the US

    International Nuclear Information System (INIS)

    Tritium waste disposal methods in the US range from disposal of low specific activity waste along with other low-level waste in shallow land burial facilities, to disposal of kilocurie amounts in specially designed triple containers in 65' deep augered holes located in an aird region of the US. Total estimated curies disposed of are 500,000 in commercial burial sites and 10 million curies in defense related sites. At three disposal sites in humid areas, tritium has migrated into the ground water, and at one arid site tritium vapor has been detected emerging from the soil above the disposal area. Leaching tests on tritium containing waste show that tritium in the form of HTO leaches readily from most waste forms, but that leaching rates of tritiated water into polymer impregnated concrete are reduced by as much as a factor of ten. Tests on improved tritium containment are ongoing. Disposal costs for tritium waste are 7 to 10 dollars per cubic foot for shallow land burial of low specific activity tritium waste, and 10 to 20 dollars per cubic foot for disposal of high specific activity waste. The cost of packaging the high specific activity waste is 150 to 300 dollars per cubic foot. 18 references

  12. Studies of the permeation and diffusion of tritium and hydrogen in TFTR

    International Nuclear Information System (INIS)

    This report documents the main features of studies conducted on the permeation and diffusion of tritium and hydrogen through the stainless steel sections comprising the vacuum vessel of TFTR. The overall conclusion of these studies is that tritium releases to the environment resulting from TFTR operations under normal conditions will be very small, less than one curie per year. A basis is described for predicting the magnitudes of the applicable transport properties for tritium-austenitic stainless steel systems as derived from a survey of the technical literature on tritium transport. The key characteristics of the TFTR vacuum vessel that are involved in the permeation and diffusion calculations are given. Information is given regarding the contemplated plasma scenarios and associated required gas injection quantities. Various issues involved in the bakeout of the vacuum vessel are discussed; focussing principally on the problems associated with in-situ bakeout and related means to reduce outgassing from the TFTR vessel and vacuum pumping system hardware. The anticipated tritium releases are studied considering the diffusion transients

  13. Transfer of tritium to foetuses and newborns from mother mice administered with tritiated water

    International Nuclear Information System (INIS)

    The kinetics of tritium release from the skin, liver and brain of pregnant and nursing mice or foetuses and sucking newborns were studied after single subcutaneous or intraperitoneal injections and the per oral uptake. When female mice were injected subcutaneously at various times during pregnancy, tritium in wet tissues at delivery was reduced lineally on a semi-log scale with prolongation of the exposure time in both pregnant females and foetuses. Tritium in dry tissues was not reduced for a short time before delivery. The estimated half-lives of tissues were longer in pregnant females than in non-pregnant ones. The half-lives of wet tissues were shorter in pregnant females than in foetuses but those of dry tissues were shorter in foetuses. The estimated doses in wet and dry tissues for 10 days after delivery were larger in foetuses than in pregnant females. In the oral uptake, the tritium in wet tissues of both pregnant females and foetuses increased with prolongation of the drinking period, reached a maximum after about 200 h and fell with the further prolongation of the period. The patterns of tritium release in nursing females injected intraperitoneally immediately after delivery were similar to those of normal females, and those of sucking newborns depended upon those of nursing females. The estimated doses in tissues at 300 h after delivery were larger in nursing females than in sucking newborns except for the amount in the skin. (author)

  14. TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2011-05-01

    This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05×10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.

  15. Analysis of a global database containing tritium in precipitation

    Energy Technology Data Exchange (ETDEWEB)

    Buckley, R. L. [Savannah River Site (SRS), Aiken, SC (United States); Rabun, R. L. [Savannah River Site (SRS), Aiken, SC (United States); Heath, M. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-02-17

    The International Atomic Energy Agency (IAEA) directed the collection of tritium in water samples from the mid-1950s to 2009. The Global Network of Isotopes in Precipitation (GNIP) data examined the airborne movement of isotope releases to the environment, with an objective of collecting spatial data on the isotope content of precipitation across the globe. The initial motivation was to monitor atmospheric thermonuclear test fallout through tritium, deuterium, and oxygen isotope concentrations, but after the 1970s the focus changed to being an observation network of stable hydrogen and oxygen isotope data for hydrologic studies. The GNIP database provides a wealth of tritium data collections over a long period of time. The work performed here primarily examined data features in the past 30 years (after much of the effects of above-ground nuclear testing in the late 1950s to early 1960s decayed away), revealing potentially unknown tritium sources. The available data at GNIP were reorganized to allow for evaluation of trends in the data both temporally and spatially. Several interesting cases were revealed, including relatively high measured concentrations in the Atlantic and Indian Oceans, Russia, Norway, as well as an increase in background concentration at a collector in South Korea after 2004. Recent data from stations in the southeastern United States nearest to the Savannah River Site do not indicate any high values. Meteorological impacts have not been considered in this study. Further research to assess the likely source location of interesting cases using transport simulations and/or literature searches is warranted.

  16. A study on the safety evaluation of concentrated tritium storage

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. J.; Kim, K. K.; Lee, S. Y.; Lee, Y. E.; Hong, D. S.; Jung, H. Y.; Song, M. C.; Hwang, K. H.; Kim, S. I.; Yook, D. S.; Sheen, J. J. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2001-03-01

    In this study, hazards of hydrogen and the risk due to storage of tritium are reviewed. The safety related factors are suggested in terms of classification of hydrogen hazards and problems related to the tritium storage. The major design parameters of the vessel of foreign countries for the storage and transport of tritium are reviewed. By review of major safety parameters related to the tritium storage, the results of this study can be applied and helpful to the development and design of tritium storage vessel in Korea. Also, the results can be useful at design of the tritium treatment facility. The integrity of tritium storage vessel material was evaluated with considering the embrittlement of metal material in hydrogen environment. The tritium storage is one of the most important problems for the safety of tritium removal facility. The research for tritium storage could be divided into two parts, one is for the metal getter of tritium and the other is for the integrity of tritium storage vessel. Especially, the integrity of tritium storage vessel is up to the tritium embrittlement of vessel materials, for tritium vessel is mostly made of metal material. In this work, the evaluation of the tritium embrittlement for the tritium storage vessel material is performed with the equipment that is made for high temperature and high vacuum. 33 refs., 56 figs., 23 tabs. (Author)

  17. Percutaneous absorption of tritium-gas-contaminated pump oil

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A. [Radiation Biology and Health Physics Branch, Ontario (Canada)

    1995-08-01

    One of the radiological problems encountered in tritium handling facilities is the hazards associated with tritium`s ability to label and degrade organic materials. Experiments in which male hairless rats have been contaminated with tritium-gas-contaminated pump oil have demonstrated that tritium deposited on the skin provides an input of organically bound tritium and tritiated water in the body. The accumulation of organically bound tritium at the point of contact in the skin and in various tissues influenced tritium excretion in urine and feces. The retention of tritium in the body showed that tritium was mainly metabolized and assimilated as organically bound tritium. The distribution of tritiated water was rapid and uniform in the whole-body. Analyses of tritium excreted in animal urine and feces showed that a significant level of organically bound tritium was excreted shortly after exposure. The highest concentration of tritium activity was measured in the exposed area of the skin. An increased level of tritium accumulation in the liver and kidneys was seen. Dose calculations showed that the exposed skin had the highest dose, and the skin dose was primarily due to the retention of organically bound tritium at the point of contact. The interpretation of these data has indicated that the retention of short-term organically bound tritium in the skin may be a dominant factor for dosimetry purposes. 19 refs., 5 figs., 4 tabs.

  18. Investigation of the potential impacts from tritium soil contamination in the CP-5 yard.

    Energy Technology Data Exchange (ETDEWEB)

    Hysong, R. J.

    1998-12-21

    Based on a review of available data, significant contributions to low-level tritium soil contamination in the CP-5 yard have been made by airborne tritium fallout and rainout from the CP-5 ventilation system stack. Based on the distribution of tritium in the yard, it is also likely that leaks in secondary system piping which lead to the cooling towers were a significant contributor to tritium in CP-5 yard subsurface soil. Based on the foregoing analysis, low-level tritium contamination will not prohibit the release of the yard for unrestricted use in the future. Worst case dose estimates based on very conservative assumptions indicate that a 25 rmem annual effective dose equivalent limit will not be exceeded under the most restrictive residential-use family farm scenario. Given the impermeable nature of the glacial till under CP-5, low-level concentrations of tritium may be occasionally detected in the deep well (3300 12D), but the peak concentration will not approach the levels calculated by RESRAD; however, continued monitoring of the deep well is recommended. To ensure that all sources of potential tritium release have been removed from the CP-5 complex, removal of tritiated water from each rod-out hole and an evaluation of the physical integrity of the rod-out holes is recommended. This will also allow for an evaluation of tritium concentrations in shallow groundwater under CP-5 by sampling groundwater that is currently being forced into the drain tile system. Additional surface and subsurface soil sampling and analysis will be required to determine the final release status of soils around the Building 330 complex relative to elevated concentrations of CS-137, CO-60,Co-57, and Eu-152 identified during the 1993 IT Corporation characterization. The potential radiological impact from isolated elevations of the latter radionuclides is relatively low and can be evaluated as part of the final status survey of outdoor areas surrounding the Building 330 complex. In

  19. TRITIUM RESERVOIR STRUCTURAL PERFORMANCE PREDICTION

    Energy Technology Data Exchange (ETDEWEB)

    Lam, P.S.; Morgan, M.J

    2005-11-10

    The burst test is used to assess the material performance of tritium reservoirs in the surveillance program in which reservoirs have been in service for extended periods of time. A materials system model and finite element procedure were developed under a Savannah River Site Plant-Directed Research and Development (PDRD) program to predict the structural response under a full range of loading and aged material conditions of the reservoir. The results show that the predicted burst pressure and volume ductility are in good agreement with the actual burst test results for the unexposed units. The material tensile properties used in the calculations were obtained from a curved tensile specimen harvested from a companion reservoir by Electric Discharge Machining (EDM). In the absence of exposed and aged material tensile data, literature data were used for demonstrating the methodology in terms of the helium-3 concentration in the metal and the depth of penetration in the reservoir sidewall. It can be shown that the volume ductility decreases significantly with the presence of tritium and its decay product, helium-3, in the metal, as was observed in the laboratory-controlled burst tests. The model and analytical procedure provides a predictive tool for reservoir structural integrity under aging conditions. It is recommended that benchmark tests and analysis for aged materials be performed. The methodology can be augmented to predict performance for reservoir with flaws.

  20. PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Smith, P.; Sheetz, S.

    2013-09-30

    Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case of Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL

  1. Gp130-dependent release of acute phase proteins is linked to the activation of innate immune signaling pathways.

    Directory of Open Access Journals (Sweden)

    Maren Luchtefeld

    Full Text Available BACKGROUND: Elevated levels of acute phase proteins (APP are often found in patients with cardiovascular diseases. In a previous study, we demonstrated the importance of the IL-6-gp130 axis -as a key regulator of inflammatory acute phase signaling in hepatocytes-for the development of atherosclerosis. BACKGROUND/PRINCIPAL FINDINGS: Gp130-dependent gene expression was analyzed in a previously established hepatocyte-specific gp130 knockout mouse model. We performed whole transcriptome analysis in isolated hepatocytes to measure tissue specific responses after proinflammatory stimulus with IL-6 across different time points. Our analyses revealed an unexpected small gene cluster that requires IL-6 stimulus for early activation. Several of the genes in this cluster are involved in different cell defense mechanisms. Thus, stressors that trigger both general stress and inflammatory responses lead to activation of a stereotypic innate cellular defense response. Furthermore, we identified a potential biomarker Lipocalin (LCN 2 for the gp130 dependent early inflammatory response. CONCLUSIONS/SIGNIFICANCE: Our findings suggest a complex network of tightly linked genes involved in the early activation of different parts of the innate immune response including acute phase proteins, complement and coagulation cascade.

  2. Percutaneous absorption of tritium-gas-contaminated pump oil

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A

    1995-07-01

    One of the radiological problems encountered in tritium handling facilities is the hazards associated with tritium's ability to label and degrade organic materials. Experiments in which male hairless rats have been contaminated with tritium-gas-contaminated pump oil have demonstrated that tritium deposited on the skin provides an input of organically bound tritium and tritiated water in the body. The accumulation of organically bound tritium at the point of contact in the skin and in various tissues influenced tritium excretion in urine and feces. The retention of tritium in the body showed that tritium was mainly metabolized and assimilated as organically bound tritium. The distribution of tritiated water was rapid and uniform in the whole-body. Analyses of tritium excreted in animal urine and feces showed that a significant level of organically bound tritium was excreted shortly after exposure. The highest concentration of tritium activity was measured in the exposed area of the skin. An increased level of tritium accumulation in the liver and kidneys was seen. Dose calculations showed that the exposed skin had the highest dose, and the skin dose was primarily due to the retention of organically bound tritium at the point of contact. The interpretation of these data has indicated that the retention of short-term organically bound tritium in the skin may be a dominant factor for dosimetry purposes. (author)

  3. Risks of tritium and their mitigation

    International Nuclear Information System (INIS)

    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)

  4. Field analyses of tritium at environmental levels

    Energy Technology Data Exchange (ETDEWEB)

    Hofstetter, K.J.; Cable, P.R.; Beals, D.M

    1999-02-11

    An automated, remote system to analyze tritium in aqueous solutions at environmental levels has been tested and has demonstrated laboratory quality tritium analysis capability in near real time. The field deployable tritium analysis system (FDTAS) consists of a novel multi-port autosampler, an on-line water purification system, and a prototype stop-flow liquid scintillation counter (LSC) which can be remotely controlled for unmanned operation. Backgrounds of {approx}1.5 counts/min in the tritium channel are routinely measured with a tritium detection efficiency of {approx}25% for the custom 11 ml cell. A detection limit of <0.3 pCi/ml has been achieved for 100-min counts using a 50 : 50 mixture of sample and cocktail. To assess the long-term performance characteristics of the FDTAS, a composite sampler was installed on the Savannah River, downstream of the Savannah River Site, and collected repetitive 12-hour composite samples over a 14 day period. The samples were analyzed using the FDTAS and in the laboratory using a standard bench-top LSC. The results of the tritium analyses by the FDTAS and by the laboratory LSC were consistent for comparable counting times at the typical river tritium background levels ({approx}1 pCi/ml)

  5. Measurement of the tritium contamination of the biosphere

    International Nuclear Information System (INIS)

    Sources of natural and artificial tritium activities are discussed. Environmental tritium concentrations were determined either by a low background proportional counter after converting tritium to hydrogen, methane or ethane gas or by a liquid scintillation coincidence counter when tritium was found in the form of water. Tritium can be enriched electrolytically. The radioactivity of tritium in the Danube, in ground and rain water was determined before commissioning the Paks Nuclear Power Plant. Based on the analysis of tritium concentrations in wines and in annual rings of trees it is possible to detect local contaminations. (V.N.) 6 refs.; 16 figs

  6. Modelling the Environmental Transfer of Tritium and Carbon-14 to Biota and Man. Report of the Tritium and Carbon-14 Working Group of EMRAS Theme 1

    International Nuclear Information System (INIS)

    Hydrogen and carbon are biologically-regulated, essential elements that are highly mobile in the environment and the human body. As isotopes of these elements, tritium and 14C enter freely into water (in the case of tritium), plants, animals and humans. This complex behaviour means that there are substantial uncertainties in the predictions of models that calculate the transfer of tritium and 14C through the environment. The EMRAS Tritium/C14 Working Group (WG) was set up to establish the confidence that can be placed in the predictions of such models, to recommend improved modelling approaches, and to encourage experimental work leading to the development of data sets for model testing. The activities of the WG focused on the assessment of models for organically bound tritium (OBT) formation and translocation in plants and animals, the area where model uncertainties are largest. Environmental 14C models were also addressed because the dynamics of carbon and OBT are similar. The goals of the WG were achieved primarily through nine test scenarios in which model predictions were compared with observations obtained in laboratory or field studies. Seven of the scenarios involved tritium, covering terrestrial and aquatic ecosystems and steady-state and dynamic conditions. The remaining two scenarios concerned 14C, one addressing steady-state concentrations in plants and the other time-dependent concentrations in animals. The WG also considered one model intercomparison exercise involving the calculation of doses following a hypothetical, short-term release of tritium to the atmosphere in a farming area. Finally, the WG discussed the nature of OBT and proposed a definition to promote common understanding and usage within the international tritium community. The models used by the various participants varied in complexity from simple specific activity approaches to dynamic compartment models and process-oriented models, in which the various transfer processes were

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

    Energy Technology Data Exchange (ETDEWEB)

    Stamoulis, K.C., E-mail: kstamoul@cc.uoi.gr [Archaeometry Center, University of Ioannina, Ioannina 45110 (Greece); Karamanis, D. [Department of Environmental and Natural Resources Management, University of Ioannina, Agrinio 30100 (Greece); Ioannides, K.G. [Archaeometry Center, University of Ioannina, Ioannina 45110 (Greece); Nuclear Physics Laboratory, University of Ioannina, Ioannina 45110 (Greece)

    2011-03-15

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

  8. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

    A summary of the tritium system design studies for the engineering outline design of a fusion experimental breeder (FEB-E) is presented. This paper is divided into three sections. In first section, the geometry, loading features and tritium concentrations in liquid lithium of tritium breeding zones of blanket are described. The tritium flow chart corresponding to the tritium fuel cycle system has been constructed, and the inventories in ten subsystems are calculated using SWITRIM code in section 2. Results show that the necessary initial tritium storage to start up FEB-E with fusion power of 143 MW is about 319 g. In final section, the tritium leakage issues under different operation circumstances have been analyzed. It was found that the potential danger of tritium leakage could be resulted from the exhausted gas of the diverter system. It is important to elevate the tritium burnup fraction and reduce the tritium throughput. (authors)

  9. Chemical equilibrium studies of tritium--lithium and tritium--lithium alloy systems

    International Nuclear Information System (INIS)

    In deuterium-tritium fusion reactors currently under design, the production of tritium is accomplished by utilizing a lithium-bearing blanket. Lithium metal is presently the leading candidate for the blanket material, although molten Li2BeF4, solid Li--Al (50-50 at. percent) alloy and other lithium-containing materials are distinct possibilities. This paper summarizes progress of ongoing studies of the thermodynamics of some of these lithium containing systems. The individual solubilities of hydrogen, deuterium, and tritium in lithium as a function of temperature (700 to 10000C) and pressure are presented. Recent work with the solid alloy Li--Al (50-50 at. percent) has shown that the tritium solubility between 400 and 6000C is low. When the tritium pressure was between 0.14 and 0.52 torr, the Li--Al samples contained only 1 to 4 ppm tritium

  10. Electrolytic gettering of tritium from air

    International Nuclear Information System (INIS)

    We have removed 90% of 1 part-per-million tritium gas in air of 25% to 35% humidity by the dc electrical action of the solid proton electrolyte hydrogen uranyl phosphate (HUP). Gettering takes 5 to 24 hours for a 1 cm2 HUP disc at 2 to 4 V in a static, 1200 cc gas volume. Hydrogen gas may be used to flush captured tritium through the HUP. Liquid water leaches out the tritium but water vapor is ineffective. This technique promises an alternative to the conventional catalyst/zeolite method

  11. Design and trial fabrication of a dismantling apparatus for irradiation capsules of solid tritium breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, K. [Japan Atomic Energy Agency, Blanket Irradiation and Analysis Group, Fusion Research and Development Directorate, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: hayashi.kimio@jaea.go.jp; Nakagawa, T.; Onose, S.; Ishida, T.; Nakamichi, M. [Japan Atomic Energy Agency, Blanket Irradiation and Analysis Group, Fusion Research and Development Directorate, 4002 Narita-cho, Oarai-machi, Ibaraki-ken 311-1393 (Japan); Takatsu, H. [Fusion Energy and Development Directorate, Japan Atomic Energy Agency, 801-1 Mukouyama, Naka-shi, Ibaraki-ken 311-0193 (Japan); Nakamura, M.; Noguchi, T. [Kaken, Inc., 873-3 Shikada, Hokota-shi, Ibaraki-ken, 311-1416 (Japan)

    2009-04-30

    Irradiation experiments of solid breeder materials including Li{sub 2}TiO{sub 3} have been being carried out in preparation for a test blanket module (TBM) of the International Thermonuclear Experimental Reactor (ITER). The present paper deals with design and trial-fabrication works for developing a dismantling apparatus for the irradiation capsules. The dismantling process leads to release of tritium which is left in free volumes of the capsule or in the breeder specimens. In the design of the dismantling apparatus, the released tritium is recovered safely by a purge-gas system during the cutting of the irradiation capsule by a band saw, and then the tritium is consolidated into a radioactive waste. Furthermore, an inner-box enclosing the dismantling apparatus works as a countermeasure of possible release of tritium in accidental events. Good performance of a trial fabrication model of the dismantling apparatus has been demonstrated by preliminary cutting runs using some mockups simulating the irradiation capsules. Thus, the present design of the apparatus, together with the trial mock-up runs, will contribute to the design of the TBM structure and to the planning of the dismantling process of the TBM.

  12. Design and trial fabrication of a dismantling apparatus for irradiation capsules of solid tritium breeder materials

    International Nuclear Information System (INIS)

    Irradiation experiments of solid breeder materials including Li2TiO3 have been being carried out in preparation for a test blanket module (TBM) of the International Thermonuclear Experimental Reactor (ITER). The present paper deals with design and trial-fabrication works for developing a dismantling apparatus for the irradiation capsules. The dismantling process leads to release of tritium which is left in free volumes of the capsule or in the breeder specimens. In the design of the dismantling apparatus, the released tritium is recovered safely by a purge-gas system during the cutting of the irradiation capsule by a band saw, and then the tritium is consolidated into a radioactive waste. Furthermore, an inner-box enclosing the dismantling apparatus works as a countermeasure of possible release of tritium in accidental events. Good performance of a trial fabrication model of the dismantling apparatus has been demonstrated by preliminary cutting runs using some mockups simulating the irradiation capsules. Thus, the present design of the apparatus, together with the trial mock-up runs, will contribute to the design of the TBM structure and to the planning of the dismantling process of the TBM.

  13. Tritium Movement and Accumulation in the NGNP System Interface and Hydrogen Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hirofumi Ohashi; Steven R. Sherman

    2007-06-01

    Tritium movement and accumulation in a Next Generation Nuclear Plant with a hydrogen plant using a high temperature electrolysis process and a thermochemical water splitting sulfur iodine process are estimated by the numerical code THYTAN as a function of design, operational, and material parameters. Estimated tritium concentrations in the hydrogen product and in process chemicals in the hydrogen plant of the Next Generation Nuclear Plant using the high temperature electrolysis process are slightly higher than the drinking water limit defined by the U.S. Environmental Protection Agency and the limit in the effluent at the boundary of an unrestricted area of a nuclear plant as defined by the U.S. Nuclear Regulatory Commission. However, these concentrations can be reduced to within the limits through use of some designs and modified operations. Tritium concentrations in the Next Generation Nuclear Plant using the Sulfur-Iodine Process are significantly higher as calculated and are affected by parameters with large uncertainties (i.e., tritium permeability of the process heat exchanger, the hydrogen concentration in the heat transfer and process fluids, the equilibrium constant of the isotope exchange reaction between HT and H2SO4). These parameters, including tritium generation and the release rate in the reactor core, should be more accurately estimated in the near future to improve the calculations for the NGNP using the Sulfur-Iodine Process. Decreasing the tritium permeation through the heat exchanger between the primary and secondary circuits may be an an effective measure for decreasing tritium concentrations in the hydrogen product, the hydrogen plant, and the tertiary coolant.

  14. The Separation of Hydrogen Tritium and Tritium Hydride by Gas Chromatography

    International Nuclear Information System (INIS)

    Now that successful separation of hydrogen, deuterium and hydrogen deuteride has been achieved by gas chromatography, similar studies are being made dealing with mixtures of hydrogen, tritium and tritium hydride. Since tritium is used in tracer quantities the usual katharometer cannot be employed for its detection. This difficulty has been overcome by providing immediately following the katharometer a vibrating reed electrometer equipped with a high resistance leak which allows continuous monitoring of the activity of any tritium or tritium hydride emerging from the column by means of synchronized recorders. Separation of such mixtures has been tested with columns packed with palladium on silica, silica, alumina, and alumina coated with chromium oxide or ferric oxide. No effective separation was obtained with the palladium on silica column. Good separation was achieved with the plain silica column where hydrogen was employed as the carrier gas, but helium failed to elute the isotopes. Satisfactory results were obtained with the coated, partially deactivated alumina packing when helium or neon was the carrier gas, but the best separation was found with a column packing of uncoated activated alumina. Calibration with helium-tritium mixtures of known activity plus equilibrated hydrogen-tritium mixtures also of known activity allows quantitative estimation of tritium and tritium hydride. (author)

  15. Neutronic Comparison of Tritium-Breeding Performance of Candidate Tritium-Breeding Materials

    Institute of Scientific and Technical Information of China (English)

    郑善良; 吴宜灿

    2003-01-01

    Tritium self-sustainment, which will meet the fuel requirement of fusion reactor, isone of the key issues of fusion power development. The tritium breeding performances of varioustritium-breeding materials are compared based on a series of neutronics calculations using three-dimensional Monte Carlo neutron-photon transport code MCNP/4C with the IAEA FENDL-2data library. The effects of the dimensions of the tritium-breeding zone and the enrichment of 6Lion Tritium Breeding Ratio (TBR) are analyzed. The effects of Be as a neutron multiplier on TBRare also calculated.

  16. Overview of tritium processing development at the tritium systems test assembly

    International Nuclear Information System (INIS)

    The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been operating with tritium since June 1984. Presently there are some 50 g of tritium in the main processing loop. This 50 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation system and to integrate the fuel cleanup system into the main fuel processing loop. In January 1986 two major experiments were conducted. During these experiments the fuel cleanup system was integrated, through the transfer pumping system, with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems leaves only the main vacuum system to be integrated into the TSTA fuel processing loop. In September 1986 another major tritium experiment was performed in which the integrated loop was operated, the tritium inventory increased to 50 g and additional measurements on the performance of the distillation system were taken. In the period June 1984 through September 1986 the TSTA system has processed well over 108 Ci of tritium. Total tritium emissions to the environment over this period have been less than 15 Ci. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flows have been uncovered

  17. A method for assessing the annual dose to the most exposed individual from tritium and 14C reactor discharges to atmosphere

    International Nuclear Information System (INIS)

    A method is described for assessing the annual dose to the most exposed individual from routine releases of tritium and 14C to the atmosphere during normal reactor operations. A detailed assessment has been made of the resulting equilibrium contamination levels in a range of foodstuffs typical of an average UK diet and of the annual doses resulting from a chronic intake of tritium and 14C via inhalation, ingestion and, additionally, in the case of tritium, via skin absorption. Equilibrium annual doses from the global circulation of tritium and 14C have also been calculated. Upper limits to the effective annual dose-equivalents to the most exposed individual were found to be 0.6 rem.yr-1 and 100 rem.yr-1 per Ci.yr-1 release of tritium and 14C respectively, with the ingestion pathway contributing significantly to the overall exposure. The most exposed individual was found to be a Reference 10 year old child. The methods outlined for calculating the ingestion dose from tritium and 14C releases hav been incorporated into the more generally applicable code FOODDOSE. The code may be used to make more realistic dose calculations to the individuals based on site-specific surveys of variables such as local meteorology, local diet and local land use for agriculture, which may lead to doses smaller than the upper limit values quoted by factors of 20 and 200 for tritium and 14C respectively. (author)

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

    International Nuclear Information System (INIS)

    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

  19. Effects of acute treadmill running at different intensities on activities of serotonin and corticotropin-releasing factor neurons, and anxiety- and depressive-like behaviors in rats.

    Science.gov (United States)

    Otsuka, Tomomi; Nishii, Ayu; Amemiya, Seiichiro; Kubota, Natsuko; Nishijima, Takeshi; Kita, Ichiro

    2016-02-01

    Accumulating evidence suggests that physical exercise can reduce and prevent the incidence of stress-related psychiatric disorders, including depression and anxiety. Activation of serotonin (5-HT) neurons in the dorsal raphe nucleus (DRN) is implicated in antidepressant/anxiolytic properties. In addition, the incidence and symptoms of these disorders may involve dysregulation of the hypothalamic-pituitary-adrenal axis that is initiated by corticotropin-releasing factor (CRF) neurons in the hypothalamic paraventricular nucleus (PVN). Thus, it is possible that physical exercise produces its antidepressant/anxiolytic effects by affecting these neuronal activities. However, the effects of acute physical exercise at different intensities on these neuronal activation and behavioral changes are still unclear. Here, we examined the activities of 5-HT neurons in the DRN and CRF neurons in the PVN during 30 min of treadmill running at different speeds (high speed, 25 m/min; low speed, 15m/min; control, only sitting on the treadmill) in male Wistar rats, using c-Fos/5-HT or CRF immunohistochemistry. We also performed the elevated plus maze test and the forced swim test to assess anxiety- and depressive-like behaviors, respectively. Acute treadmill running at low speed, but not high speed, significantly increased c-Fos expression in 5-HT neurons in the DRN compared to the control, whereas high-speed running significantly enhanced c-Fos expression in CRF neurons in the PVN compared with the control and low-speed running. Furthermore, low-speed running resulted in decreased anxiety- and depressive-like behaviors compared with high-speed running. These results suggest that acute physical exercise with mild and low stress can efficiently induce optimal neuronal activation that is involved in the antidepressant/anxiolytic effects. PMID:26542811

  20. An acute bout of self-myofascial release increases range of motion without a subsequent decrease in muscle activation or force.

    Science.gov (United States)

    MacDonald, Graham Z; Penney, Michael D H; Mullaley, Michelle E; Cuconato, Amanda L; Drake, Corey D J; Behm, David G; Button, Duane C

    2013-03-01

    Foam rolling is thought to improve muscular function, performance, overuse, and joint range of motion (ROM); however, there is no empirical evidence demonstrating this. Thus, the objective of the study was to determine the effect of self-myofascial release (SMR) via foam roller application on knee extensor force and activation and knee joint ROM. Eleven healthy male (height 178.9 ± 3.5 cm, mass 86.3 ± 7.4 kg, age 22.3 ± 3.8 years) subjects who were physically active participated. Subjects' quadriceps maximum voluntary contraction force, evoked force and activation, and knee joint ROM were measured before, 2 minutes, and 10 minutes after 2 conditions: (a) 2, 1-minute trials of SMR of the quadriceps via a foam roller and (b) no SMR (Control). A 2-way analysis of variance (condition × time) with repeated measures was performed on all dependent variables recorded in the precondition and postcondition tests. There were no significant differences between conditions for any of the neuromuscular dependent variables. However, after foam rolling, subjects' ROM significantly (p < 0.001) increased by 10° and 8° at 2 and 10 minutes, respectively. There was a significant (p < 0.01) negative correlation between subjects' force and ROM before foam rolling, which no longer existed after foam rolling. In conclusion, an acute bout of SMR of the quadriceps was an effective treatment to acutely enhance knee joint ROM without a concomitant deficit in muscle performance.

  1. The diaplacental transfer and teratogenicity of tritium in organic compounds. Diaplazentare Verteilung und teratogene Wirkung von organisch gebundenem Tritium

    Energy Technology Data Exchange (ETDEWEB)

    Kistner, G.; Wiggenhauser, A.; Krestel, R.

    1987-10-01

    In the introductory section of this report the relevant literature on the environmental prevalence of tritium, its uptake and distribution in the organisms as well as on the calculation of tritium-related exposure risks is reviewed in brief. The literature study as well as recent workshops on tritium (as e.g. in Karlsruhe, 1986) emphasize the role of organically bound tritium in the assessment of tritium-related radiation exposure. (orig./MG).

  2. Synthesis of tritium-labelled natural prostaglandins

    International Nuclear Information System (INIS)

    The most suitable method for the preparation of tritium-labelled prostaglandins is the biosynthetic procedure. Polyunsaturated labelled fatty acids are converted into prostaglandins by a prostaglandin synthetase enzyme system produced from sheep seminal vesicule, and the crude product is purified using thin layer chromatography. Polyunsaturated fatty acids are prepared in a reaction series. Tritium is introduced at the very last step. A very little amount (10-20 mg) of tritium-labelled prostaglandin E2 can be converted into A2, B2 and F2 respectively, conversion and separation being carried out simultaneously on the same silica plate. After the separation on thin layer silica gel the obtained tritium-labelled prostaglandin (PC) was chemically and radiochemically pure, its activity was 3700 GBq/mmol (100 Ci/mmol) and it was suitable for RIA kits. (author)

  3. Methods of tritium recovery from molten lithium

    International Nuclear Information System (INIS)

    It is important to keep the tritium inventory in a blanket of a thermonuclear reactor at a low level both to eliminate possible hydriding of structural components and to reduce inventory cost. Removing the tritium from a lithium blanket by fractional distillation, flash vaporization, and fractional crystallization was investigated. No definitive data are available either on the vapor-liquid equilibrium between lithium and tritium at low T2 concentrations, or on the rate of formation and decomposition of lithium tritide. The final distinction between the recovery systems discussed in this report will depend on such data, but presently distillation appears to be the best alternate to the diffusion scheme proposed by A.P. Fraas. The capital cost of equipment necessary to remove tritium by distillation appears to be greater than 10 million dollars for a 5000 MW system, whereas the capital cost associated with the diffusion process has been estimated to be 4 million dollars

  4. Leaching of tritium from a cement composite

    International Nuclear Information System (INIS)

    Leaching of tritium from cement composites into an aqueous phase has been studied to evaluate the safety of incorporation of the tritiated liquid waste into cement. Leaching tests were performed by the method recommended by the International Atomic Energy Agency. The Leaching fraction was measured as functions of waste-cement ratio (Wa/C), temperature of leachant and curing time. The tritium leachability of cement in the long term test follows the order: alumina cement portland cement slag cement. The fraction of tritium leached increases with increasing Wa/C and temperature and decreasing curing period. A deionized water as a leachant gives a slightly higher leachability than synthetic sea water. The amount leached of tritium from a 200 l drum size specimen was estimated on the basis of the above results. (author)

  5. Origin, handling and storage of tritium

    International Nuclear Information System (INIS)

    The origin, handling and intermediate and/or final storage of tritium in the Federal Republic of Germany are described and evaluated. For this the following subjects - use, amounts of waste and emission, waste handling, transport, legal situation and points relevant to safety in respect to tritium handling, general valuation and future development - are completely presented. Presently and in future the waste volume will be determined by the fact that nearly the whole amount of tritium waste activity is contained in a small part of the overall waste volume. The rest is distributed to a relatively big waste volume accordingly showing low activity concentration levels. Future efforts are mainly necessary in respect to the handling of tritium waste. (orig.)

  6. Tritium proof-of-principle injector experiment

    International Nuclear Information System (INIS)

    The Tritium Proof-of-Principle (TPOP) pellet injector was designed and built by Oak Ridge National Laboratory (ORNL) to evaluate the production and acceleration of tritium pellets for fueling future fision reactors. The injector uses the pipe-gun concept to form pellets directly in a short liquid-helium-cooled section of the barrel. Pellets are accelerated by using high-pressure hydrogen supplied from a fast solenoid valve. A versatile, tritium-compatible gas-handling system provides all of the functions needed to operate the gun, including feed gas pressure control and flow control, plus helium separation and preparation of mixtures. These systems are contained in a glovebox for secondary containment of tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). 18 refs., 3 figs

  7. Technologies for immobilization and disposal of tritium

    International Nuclear Information System (INIS)

    This study was done within a program one of whose objectives was to know the state of the technology development for tritium separation in the moderator circuit at HWR and to define the possible technologies to be applied to the Argentine nuclear power plants. Within this framework the strategies adopted by each country and the available technologies for a safe disposal of tritium, not only in its gaseous state tritium but also as tritiated water were analyzed. It is considered that if the selected separation method is such that the tritium is in its gaseous state, the hydride formation for long periods of immobilization should be studied. whereas if it were triated water immobilization should be studied to choose the technology between cementation and drying agents, in both cases the final disposal site will have to be selected. (author). 8 refs

  8. Tritium effluent control project at Mound Laboratory

    International Nuclear Information System (INIS)

    Tritium control technology and philosophies that have been developed at the various weapons complex sites can be drawn upon for the design and operation of fusion research facilities and Controlled Thermonuclear Reactor fuel cycle and tritium confinement systems. Historically, tritium control has been based on high volume air flows and dilution of effluent systems. As a consequence of the ''as low as practical'' criterion, control philosophies were reevaluated and control efforts intensified. It is the results of these recent efforts that are most applicable to the CTR program. At Mound Laboratory the tritium control development efforts are centered around an advanced technology project. The philosophy and goals of this project and a nearly completed pilot scale effluent control laboratory are described. (auth)

  9. Biomedical tritium applications with AMS detection

    International Nuclear Information System (INIS)

    There are numerous applications for tritium (3 H) as a tracer isotope in biomedicine commonly combined with liquid scintillation counting method. The use of accelerator mass spectrometry (AMS), a rather new detection method will, enlarge and open new possibilities for tritium applications in biomedicine, especially when sample volumes are small. The tritium in the samples has to be transformed to solid form, which yields a high output of negative hydrogen ion current. The sample preparation is done in two steps: firstly extracting water from the biological sample and secondly, extracting hydrogen/tritium from the water and forming a chemically suitable compound for the AMS ion source. In this paper a chemical for the sample preparation is described. The results of the first measurements of tritiated water with known activity using the AMS detection technique will also be presented.(authors)

  10. Tritium measurements with a tandem accelerator

    Science.gov (United States)

    Middleton, R.; Klein, J.; Fink, D.

    1990-06-01

    Tritium concentrations ( 3H: 2H) of less than 10 -15 are readily measurable with almost any tandem accelerator and with an overall detection efficiency as high as 4.5%. The isobar, 3He, and other potential sources of interference (mainly 6Li, 2H and 1H) can all be removed by an absorber in front of the triton detector, so there is little need for analyzing elements other than the negative-and positive-ion magnets found on most tandems. The technique is particularly well suited for detecting tritium in deuterium absorbed in a metal and testing for cold fusion. We caution that tritium can occur in commercial deuterium and heavy water from sources other than cold fusion; one sample was observed to have a tritium-to-deuterium ratio of 10 -10.

  11. Permeability of protective coatings to tritium

    International Nuclear Information System (INIS)

    The permeability of four protective coatings to tritium gas and tritiated water was investigated. The coatings, including two epoxies, one vinyl and one urethane, were selected for their suitability in CANDU plant service in Ontario Hydro. Sorption rates of tritium gas into the coatings were considerably larger than for tritiated water, by as much as three to four orders of magnitude. However, as a result of the very large solubility of tritiated water in the coatings, the overall permeability to tritium gas and tritiated water is comparable. Marked differences were also evident among the four coatings, the vinyl exhibiting an abnormally high retention of free water because of a highly porous surface structure. It appears that epoxy coatings having a high pigment-to-binder ratio are most suited for coating concrete in tritium handling facilities

  12. Nuclear Overhauser effects in tritium NMR

    Energy Technology Data Exchange (ETDEWEB)

    Kaspersen, F.M.; Funke, C.W.; Sperling, E.M.G.; Wagenaars, G.N.

    1987-02-01

    The accuracy of the quantification of the tritium distribution in labelled compounds may be reduced by differential nuclear Overhauser effects, especially for compounds in which the different tritiated positions differ in the number of protons surrounding them.

  13. Separation of tritium from aqueous effluents

    International Nuclear Information System (INIS)

    This report describes the further development of the so-called ELEX process, carried out from 1 July 1980 until 31 December 1982. The ELEX process is the combination of electrolysis with the catalytic tritium exchange between hydrogen and water in order to accumulate the tritium in the liquid phase. The experimental study of the catalytic tritium exchange between hydrogen and liquid water was continued and the overall exchange rate could be substantially increased. An alternative process based on bithermal exchange of tritium has been evaluated. In the 10 mol h-1 mini-pilot bench scale detritiation unit the ELEX process was successfully demonstrated by detritiating up to now more than 1m3 of water containing up to 100 mCi tritium per dm3, which is the feed concentration to be expected for application of the process in a reprocessing plant. A 280 mol h-1 pilot detritiation installation now being constructed is described. This installation will realize a volume reduction factor of 100 and a process decontamination factor of 100. The maximum total tritium inventory will be about 1000 Ci. The plant consists mainly of a 80 kW electrolyser and a 10 cm diameter exchange column and can be considered as the ultimate step before industrial application of the ELEX process

  14. Impact of Left Ventricular Hypertrophy on Troponin Release During Acute Myocardial Infarction: New Insights From a Comprehensive Translational Study

    Science.gov (United States)

    Fernández‐Jiménez, Rodrigo; Silva, Jacobo; Martínez‐Martínez, Sara; López‐Maderuelo, Mª Dolores; Nuno‐Ayala, Mario; García‐Ruiz, José Manuel; García‐Álvarez, Ana; Fernández‐Friera, Leticia; Pizarro, Tech Gonzalo; García‐Prieto, Jaime; Sanz‐Rosa, David; López‐Martin, Gonzalo; Fernández‐Ortiz, Antonio; Macaya, Carlos; Fuster, Valentin; Redondo, Juan Miguel; Ibanez, Borja

    2015-01-01

    Background Biomarkers are frequently used to estimate infarct size (IS) as an endpoint in experimental and clinical studies. Here, we prospectively studied the impact of left ventricular (LV) hypertrophy (LVH) on biomarker release in clinical and experimental myocardial infarction (MI). Methods and Results ST‐segment elevation myocardial infarction (STEMI) patients (n=140) were monitored for total creatine kinase (CK) and cardiac troponin I (cTnI) over 72 hours postinfarction and were examined by cardiac magnetic resonance (CMR) at 1 week and 6 months postinfarction. MI was generated in pigs with induced LVH (n=10) and in sham‐operated pigs (n=8), and serial total CK and cTnI measurements were performed and CMR scans conducted at 7 days postinfarction. Regression analysis was used to study the influence of LVH on total CK and cTnI release and IS estimated by CMR (gold standard). Receiver operating characteristic (ROC) curve analysis was performed to study the discriminatory capacity of the area under the curve (AUC) of cTnI and total CK in predicting LV dysfunction. Cardiomyocyte cTnI expression was quantified in myocardial sections from LVH and sham‐operated pigs. In both the clinical and experimental studies, LVH was associated with significantly higher peak and AUC of cTnI, but not with differences in total CK. ROC curves showed that the discriminatory capacity of AUC of cTnI to predict LV dysfunction was significantly worse for patients with LVH. LVH did not affect the capacity of total CK to estimate IS or LV dysfunction. Immunofluorescence analysis revealed significantly higher cTnI content in hypertrophic cardiomyocytes. Conclusions Peak and AUC of cTnI both significantly overestimate IS in the presence of LVH, owing to the higher troponin content per cardiomyocyte. In the setting of LVH, cTnI release during STEMI poorly predicts postinfarction LV dysfunction. LV mass should be taken into consideration when IS or LV function are estimated by troponin

  15. Appendix for blanket - University of Wisconsin: tritium issues

    International Nuclear Information System (INIS)

    The selection of the liquid metal alloys, Li17Pb83, as the tritium breeder with helium serving as the heat transfer fluid suggests two alternative techniques for the removal of tritium from the breeder. The low solubility of tritium in this liquid breeder requires only a simple vacuum degassing technique for tritium removal. Because of this high tritium partial pressure, tritium removal in the present design could potentially be achieved by either (a) slow circulation of the liquid LiPb alloy to an external degassing system, or (b) noncirculation of the liquid breeder so that the tritium permeates through the walls of the coolant tubes into the circulating helium for subsequent recovery. Both of these techniques were investigated with special attention given to the resultant tritium inventories in the liquid breeder and the helium system, and the potential for tritium permeation at the steam generator (SG)

  16. Parenteral versus early intrajejunal nutrition: Effect on pancreatitic natural course, entero-hormones release and its efficacy on dogs with acute pancreatitis

    Institute of Scientific and Technical Information of China (English)

    Huan-Long Qin; Zhen-Dong Su; Lei-Guang Hu; Zai-Xian Ding; Qing-Tian Lin

    2003-01-01

    -protein synthesis and release. EIN has no effect on the natural course of acute pancreatitis.

  17. Inhibitory Effect of Clopidogrel on Release of Soluble CD40 Ligand by ADP-activated Platelet in Patients With Non-ST-segment elevation Acute Coronary Syndromes

    Institute of Scientific and Technical Information of China (English)

    Wei Wei; Chufan Luo; Zhimin Du

    2008-01-01

    Objectives To investigate the inhibitory effect of clopidogrel on release of soluble CD40 ligand (sCD40L) by ADP-activated platelet in patients with non-ST-segment elevation acute coronary syndromes(NSTEACS).Methods Forty-two patients with NSTEACS were treated with clopidogrel for 6~8 days.In order to obtain platelet rich plasma (PRP) samples,the venous blood was drawn before and after treatment,respectively.The platelets were activated by adenosine diphosphate (ADP),thus releasing sCD4OL,sCD40L levels were determined by enzyme-linked immunosorbent assay (ELISA) at different time of the reaction.Results Plasma sCD40L concentration before treatment was (0.199±0.155 ) ng/mL,and (0.190±0.176) ng/mL after treatment (P>0.05).Before treatment the PRP sCD40L level at 20-minute of platelet activation was (4.34±2.51 )ng/mL,and decreased to (2.79±1.93 ) ng/mL after treatment (P<0.001).The corresponding level at 40-minute of platelet activation was (5.29±3.13 ) ng/mL before treatment and (2.87±1.59 ) ng/mL after treatment(P<0.001 ).Conclusions Short-term clopidogrel administration might inhibit the release of sCD40L by ADP-activated platelet in patients with NSTEACS,suggesting that,in addition to its antiplatelet potency,clopidogrel may still have an anti-inflammatory effect.

  18. Influence of the Fukushima Daiichi nuclear disaster on the tritium concentration in the precipitation of Kanazawa city

    International Nuclear Information System (INIS)

    The variation in tritium concentration in the precipitation of Kanazawa city was measured the day after the Fukushima Daiichi nuclear disaster, which occurred on 11 March, 2011. The most interesting result in the period from March to August 2011 is that a secondary peak of 4.6 Bq/L was observed on 22 March after the maximum peak of 15.0 Bq/L on 16 March. This fact suggests that a sudden release of a large amount of tritium from the Fukushima Daiichi nuclear power plant occurred between 21 March and 22 March, followed by the release from 11 March to 15 March. Another interesting result is that dramatic increases in the tritium concentration of 131.6 Bq/L and 99.9 Bq/L were observed on 30 May and 13 June in the variation patterns of the tritium concentration in the precipitation of Kanazawa city. These increases may have been caused by wind blowing down from the upper atmosphere during a storm. A large amount of tritium which had been released by hydrogen explosions from the Fukushima Daiichi nuclear reactors was considered to exist in the bulk air in the upper atmosphere. (author)

  19. Information for establishing bioassay measurements and evaluations of tritium exposure

    International Nuclear Information System (INIS)

    This report summarizes information and references used in developing regulatory guidance on programs for the bioassay of tritium as well as information useful in planning and conducting tritium bioassay programs and evaluating bioassay data. A review of literature on tritium radiobiology is included to provide a ready source of information useful for estimating internal doses of tritium and risks for the various tritium compounds and forms, including elemental (gaseous) tritium. Simplified and conservative dose conversion factors are derived and tabulated for easy reference in program planning, safety evaluations, and compliance determinations

  20. Mobile Tritium Removal Facility - an affordable option?

    International Nuclear Information System (INIS)

    Tritium removal facilities are only likely to be an issue when CANDU plants have matured and the increasing tritium levels in the water have become intolerable from a personnel health physics perspective. Even then some station owners claim that a Tritium removal facility is unnecessary, because improved health physics performance and practices is all that is required to protect against possible personnel exposure. To support this argument it is also true to say that the tritium accumulation does stabilize, and will reach a stage where the tritium content will no longer increase. However for station owners that support the view that they follow an ALARA principle in which only the lowest level achievable is acceptable, a tritium extraction plant when the plant is new or one built later when the plant is operating and in mid life, both have arguments to support the expense. For a CANDU reactor in mid-life, there are two options for siting the Tritium Removal Facility (TRF): Stationary Option which will require permanent structures for each station; and, Mobile Option which considers a complete TRF that can be moved from station to station. In most existing CANDU-6 stations, no provisions have been made to construct and operate a TRF. This would make the Stationary Option costly because space would have to be provided and newly added infrastructure would have to be installed. With appropriate seismic qualification and following the necessary codes and standards, a Mobile TRF unit could be more cost effective, particularly if there were a possibility to share the unit with other stations in like position. (author)

  1. Parametric analysis of LIBRETTO-4 and 5 in-pile tritium transport model on EcosimPro

    Energy Technology Data Exchange (ETDEWEB)

    Alcalde, Pablo Martínez, E-mail: pablomiguel.martinez@externos.ciemat.es [Universidad Nacional de Educación a Distancia (UNED), c/Juan del Rosal 12, 28040 Madrid (Spain); Moreno, Carlos; Ibarra, Ángel [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain)

    2014-10-15

    Highlights: • Introduction of a new tritium transport model of LIBRETTO-4 and 5 on EcosimPro{sup ®}. • Analysis of model input parameter and variable sensitivities and effects on tritium simulated fluxes. • Demonstrations of high tritium out-flux dependencies on lead-lithium parameters. • Rough fitting achievements proposed by Li17Pb solubility or recombination increase. - Abstract: A new model for LIBRETTO-4/1, 4/2 and 5 experiments have been developed on ECOSIMPro{sup ©} tool to simulate tritium in-pile breeding and transport into two separate purge gas channels with He + 0.1%H{sub 2}. Release from lead lithium eutectic plenum with coupled permeation through an austenitic steel wall on the first and single permeation through EUROFER-97 in the temperature ranges of 300–550 °C can be simulated tuning the transport parameters involved. A parametric study has been performed to reduce the degrees of freedom and to determine the error caused in the simulation due to the uncertainty in experimental input data. The information obtained is essential for the experimental benchmarking. The Tritium Permeation Percentage (TPP) is an output calculated parameter with low variations between 2 and 6% along the whole experimental time easy to compare (730 Full Power Days for LIBRETTO-4 and 520 for 5). Tritium transport parameter ranges verifying this output are defined herein.

  2. ITER SAFETY TASK NID-5D: Operational tritium loss and accident investigation for heat transport and water detritiation systems

    International Nuclear Information System (INIS)

    The task objectives are to: a) determine major pathways for tritium loss during normal operation of the cooling systems and water detritiation system, b) estimate operational losses and environmental tritium releases from the heat transport and water detritiation systems of ITER, and c) prepare a preliminary Failure Modes and Effects Analysis (FMEA) for the ITER Water Detritiation System. The analysis will be used to estimate chronic environmental tritium releases (airborne and waterborne) for the ITER Cooling Systems and Water Detritiation System. The assessment will form the basis for demonstrating the acceptability of ITER for siting in the Early Safety and Environmental Characterization Study (ESECS), to be issued in early 1995. (author). 7 refs., 10 tabs., 11 figs

  3. Mobility of Tritium in Engineered and Earth Materials at the NuMI Facility, Fermilab: Progress report for work performed between June 13 and September 30, 2006

    International Nuclear Information System (INIS)

    This report details the work done between June 13 and September 30, 2006 by Lawrence Berkeley National Laboratory (LBNL) scientists to assist Fermi National Accelerator Laboratory (Fermilab) staff in understanding tritium transport at the Neutrino at the Main Injector (NuMI) facility. As a byproduct of beamline operation, the facility produces (among other components) tritium in engineered materials and the surrounding rock formation. Once the tritium is generated, it may be contained at the source location, migrate to other regions within the facility, or be released to the environment

  4. Angiotensin II is related to the acute aortic dissection complicated with lung injury through mediating the release of MMP9 from macrophages

    Science.gov (United States)

    Wu, Zhiyong; Ruan, Yongle; Chang, Jinxing; Li, Bowen; Ren, Wei

    2016-01-01

    Background: Acute aortic dissection (AAD) patients usually show concurrent lung injury mainly featured by hyoxemia. To date, no effective treatment method has been established for the AAD complicated with acute lung injury (ALI). Matrix metalloproteinases (MMPs), especially MMP2 and MMP9, have been considered to be closely related to the onset of aortic disease including AAD. To investigate the roles of MMP in the pathogenesis of AAD complicated with ALI, we determined the expression of MMP2 and MMP9 in serum and lung tissues of AAD patients. In addition, a new rat model of AAD complicated with ALI was established to investigate the pathogenesis of such complicated conditions. Methods and results: Angiotensin II (Ang II) and MMP9 were up-regulated in the AAD complicated with ALI patients compared to those of the AAD without ALI patients, normal individuals and the patients with non-ruptured aneurysm. Besides, massive macrophages with MMP9 expression was noticed in the lung tissues in the AAD complicated with ALI patients. On this basis, AAD complicated with ALI rat model was established based on BAPN feeding and infusion of Ang II. Obvious lung injury was observed in the BAPN+Ang II group compared to that of the BAPN group, together with macrophage accumulation in lung tissues, as well as over-expression of MMP9 in lung tissues. After interference of MMP antagonist, a large number of macrophages were still accumulated in the lung tissues, but the lung injury was obviously attenuated. After the interference of AT1 receptor, the number of macrophages in the lung tissues was obviously decreased and the lung injury was obviously relieved. Conclusions: Ang II is closely related to the lung injury at the early stage of AAD through mediating the release of MMP9 in the macrophages in the lung tissues. PMID:27186269

  5. Investigation of ability of serum albumin to bind the tritium labeled drotaverine hydrochloride at virus hepatitis

    International Nuclear Information System (INIS)

    measurement of the radio-activity connected to fraction of proteins of blood serum. We investigated with the help of the developed radiochemical method the blood samples of 88 patients in the age of from 3 to 14 years with the diagnosis of acute virus hepatitis. From them with the acute virus hepatitis A there were 48 patients and with the acute virus hepatitis B - 40 patients. Etiological diagnostics were carried out by definition of specific markers of a virus hepatitis A and B. The control group consist of 10 practically health children of similar age. Results of research have shown that at the heavy form of a virus hepatitis B the binding ability of albumin in a stage of illness peak is reduced in comparison with group of the control practically at all investigated patients. At the moderate form of a acute virus hepatitis B the decrease of binding ability of albumin at 69 % of patients was observed. At acute virus hepatitis A the decrease of binding ability of albumin is expressed less, than at hepatitis B, and at moderate and heavy forms of disease achievement of control values is achieved in the period of convalescence. At moderate and heavy forms of acute virus hepatitis B on the background of traditional therapy in the period of early convalescence the binding ability of albumin did not reach the control values. The received results at whole testify the decrease of the binding capacity of serum albumin at virus hepatitis that allows to determine the optimal strategy of pharmaceutical loading on an organism and by that to optimize treatment of patients. Thus the developed radiochemical method with use of the tritium labeled drotaverine hydrochloride is the high informative method of definition of binding ability of serum albumin. It can be used for determination binding ability of serum albumin as universal criterion of estimation of weights of an intoxication and efficiency of ways and means of detoxication.

  6. Management of tritium contaminated wastes national strategies and practices at some European countries, USA and Canada

    International Nuclear Information System (INIS)

    The European Tritium Handling Experiment Laboratory (ETHEL) is the Commission of European Communities facility designed for handling multigram quantities of tritium for safety inherent R and D purposes. Tritium contamined wastes in gaseous, liquid and solid forms will be generated in ETHEL during the experiments as well as during the maintenance operations. All such wastes must be adequately managed under the safest operating conditions to minimize the releases of tritium to the environment and the consequent radiological risks to workers and general population. This safety requirement can be met by carefully defining strategies and practices to be applied for the safe management of these wastes. To this end an adequate background information must be collected which is the intent of this report. Through an exhaustive literature survey current strategies and practices applied in Europe, USA and Canada for managing tritiated wastes from specific tritium handling laboratories and plant have been assessed. For some countries, where only tritium bearing wastes simultaneously contaminated with nuclear fission products are generated, the attention has been focused on the strategies and practices currently applied for managing fission wastes. Operational criteria for waste collection, sorting, classification, conditioning and packaging as well as acceptance criteria for their storage or disposal have been identified. Waste storage or disposal options already applied in various countries or still being investigated in terms of safety have also been considered. Even if the radwaste management strategy is submitted to a nearly continuing process of review, some general comments resulting from the assessment of the present waste management scenario are presented. 60 refs., 16 figs., 13 tabs

  7. Evaluation of permeable and non-permeable tritium in normal condition in a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marta, V; Manuel, P J [Instituto de Fusion Nuclear (DENIM)/ETSII, Universidad Politecnica Madrid (UPM) (Spain); Sedano Luis, A [Ministerio de Educacion y Ciencia, Ciemat (Spain)], E-mail: marta@denim.upm.es

    2008-05-15

    The tritium cycle, technologies of process and control of the tritium in the plant will constitute a fraction of the environmental impact of the first generation of DT fusion reactors. The efforts of conceptual development of the tritium cycle are centered in the Internal Regenerator Cycle. The tritium could be recovered from a flow of He gas, or directly from solid breeder. The limits of transfers to the atmosphere are assumed {approx} 1 gr-T/a ({approx}20 Ci/a) (without species distinction). In the case of ITER, for example, we have global demands of control of 5 orders of magnitude have been demonstrated at experimental level. The transfer limits determine the key parameters in tritium Cycle (HT, HTO, as dominant, and T2, T2O as marginal). Presently, the transfer from the cycle to the environment is assumed through the exchange system of the power plant (primary to secondary). That transport is due to the permeation through HT, T2, or leakage to the coolant in the primary system. It is key the chemical optimization in the primary system, that needs to be reanalyzed in terms of radiological impact both for permeable, HT, T2, and non-permeable HTO, T2O. It is necessary considered the pathway of tritium from the reactor to the atmosphere, these processes are modelled adequately. Results of the assessments were early and chronic doses which have been evaluated for the Most Exposed Individual at particular distance bands from the release point. The impact evaluations will be performed with the computational tools (NORMTRI), besides national regulatory models, internationally accepted computer these code for dosimetric evaluations of tritiated effluents in operational conditions.

  8. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

    The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of ∼2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs

  9. Radioactive waste tank ventilation system incorporating tritium control

    Energy Technology Data Exchange (ETDEWEB)

    Rice, P.D. [ICF Kaiser Hanford Company, Richland, WA (United States)

    1997-08-01

    This paper describes the development of a ventilation system for radioactive waste tanks at the U.S. Department of Energy`s (DOE) Hanford Site in Richland, Washington. The unique design of the system is aimed at cost-effective control of tritiated water vapor. The system includes recirculation ventilation and cooling for each tank in the facility and a central exhaust air clean-up train that includes a low-temperature vapor condenser and high-efficiency mist eliminator (HEME). A one-seventh scale pilot plant was built and tested to verify predicted performance of the low-temperature tritium removal system. Tests were conducted to determine the effectiveness of the removal of condensable vapor and soluble and insoluble aerosols and to estimate the operating life of the mist eliminator. Definitive design of the ventilation system relied heavily on the test data. The unique design features of the ventilation system will result in far less release of tritium to the atmosphere than from conventional high-volume dilution systems and will greatly reduce operating costs. NESHAPs and TAPs NOC applications have been approved, and field construction is nearly complete. Start-up is scheduled for late 1996. 3 refs., 4 figs., 2 tabs.

  10. Tritium Specific Adsorption Simulation Utilizing the OSPREY Model

    Energy Technology Data Exchange (ETDEWEB)

    Veronica Rutledge; Lawrence Tavlarides; Ronghong Lin; Austin Ladshaw

    2013-09-01

    During the processing of used nuclear fuel, volatile radionuclides will be discharged to the atmosphere if no recovery processes are in place to limit their release. The volatile radionuclides of concern are 3H, 14C, 85Kr, and 129I. Methods are being developed, via adsorption and absorption unit operations, to capture these radionuclides. It is necessary to model these unit operations to aid in the evaluation of technologies and in the future development of an advanced used nuclear fuel processing plant. A collaboration between Fuel Cycle Research and Development Offgas Sigma Team member INL and a NEUP grant including ORNL, Syracuse University, and Georgia Institute of Technology has been formed to develop off gas models and support off gas research. This report is discusses the development of a tritium specific adsorption model. Using the OSPREY model and integrating it with a fundamental level isotherm model developed under and experimental data provided by the NEUP grant, the tritium specific adsorption model was developed.

  11. A compact tritium AMS system

    Energy Technology Data Exchange (ETDEWEB)

    Chiarappa, M L; Dingley, K H; Hamm, R W; Love, A H; Roberts, M L

    1999-09-23

    Tritium ({sup 3}H) is a radioisotope that is extensively utilized in biological and environmental research. For biological research, {sup 3}H is generally quantified by liquid scintillation counting requiring gram-sized samples and counting times of several hours. For environmental research, {sup 3}H is usually quantified by {sup 3}He in-growth which requires gram-sized samples and in-growth times of several months. In contrast, provisional studies at LLNL's Center for Accelerator Mass Spectrometry have demonstrated that Accelerator Mass Spectrometry (AMS) can be used to quantify {sup 3}H in milligram-sized biological samples with a 100 to 1000-fold improvement in detection limits when compared to scintillation counting. This increased sensitivity is expected to have great impact in the biological and environmental research community. However in order to make the {sup 3}H AMS technique more broadly accessible, smaller, simpler, and less expensive AMS instrumentation must be developed. To meet this need, a compact, relatively low cost prototype {sup 3}H AMS system has been designed and built based on a LLNL ion source/sample changer and an AccSys Technology, Inc. Radio Frequency Quadrupole (RFQ) linac. With the prototype system, {sup 3}/{sup 1}H ratios ranging from 1 x 10{sup -10} to 1 x 10{sup -13} have to be measured from milligram sized samples. With improvements in system operation and sample preparation methodology, the sensitivity limit of the system is expected to increase to approximately 1 x 10{sup -15}.

  12. Tritium in surface water of the Yenisei river Basin

    International Nuclear Information System (INIS)

    The paper reports an investigation of the tritium content in the surface waters of the Yenisei River basin near the Mining-and-Chemical Combine (MCC). In 2001-2003 the maximum tritium concentration in the Yenisei River did not exceed 4±1 Bq/L. It has been found that there are surface waters containing enhanced tritium, up to 168 Bq/L, as compared with the background values for the Yenisei River. There are two possible sources of tritium input. First, the last operating reactor of the MCC, which still uses the Yenisei water as coolant. Second, tritium may come from the deep aquifers at the Severny testing site. For the first time tritium has been found in two aquatic plant species of the Yenisei River with maximal tritium concentration 304 Bq/Kg wet weight. Concentration factors of tritium for aquatic plants are much higher than 1

  13. Advancement Of Tritium Powered Betavoltaic Battery Systems

    Energy Technology Data Exchange (ETDEWEB)

    Staack, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Gaillard, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hitchcock, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peters, B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Colon-Mercado, H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Teprovich, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coughlin, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Neikirk, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Fisher, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-14

    Due to their decades-long service life and reliable power output under extreme conditions, betavoltaic batteries offer distinct advantages over traditional chemical batteries, especially in applications where frequent battery replacement is hazardous, or cost prohibitive. Although many beta emitting isotopes exist, tritium is considered ideal in betavoltaic applications for several reasons: 1) it is a “pure” beta emitter, 2) the beta is not energetic enough to damage the semiconductor, 3) it has a moderately long half-life, and 4) it is readily available. Unfortunately, the widespread application of tritium powered betavoltaics is limited, in part, by their low power output. This research targets improving the power output of betavoltaics by increasing the flux of beta particles to the energy conversion device (the p-n junction) through the use of low Z nanostructured tritium trapping materials.

  14. Low-exposure tritium radiotoxicity in mammals

    Energy Technology Data Exchange (ETDEWEB)

    Dobson, R.L.

    1982-02-11

    Studies of tritium radiotoxicity involving chronic /sup 3/H0H exposures in mammals demonstrate in both mice and monkeys that biological effects can be measured following remarkably low levels of exposure - levels in the range of serious practical interest to radiation protection. These studies demonstrate also that deleterious effects of /sup 3/H beta radiation do not differ significantly from those of gamma radiation at high exposures. In contrast, however, at low exposures tritium is significantly more effective than gamma rays, rad for rad, by a factor approaching 3. This is important for hazard evaluation and radiation protection because knowledge concerning biological effects of chronic low-level radiation exposure has come mainly from gamma-ray data; and predictions based on gamma-ray data will underestimate tritium effects - especially at low exposures - unless the RBE is fully taken into account.

  15. Calibration for Radiation Protection Equipment for the Measuring Airborne Tritium

    Institute of Scientific and Technical Information of China (English)

    CHEN; Xi-lin; SHEN; En-wei; WEI; Ke-xin; WANG; Kong-zhao; LI; Hou-wen; GE; Jian-an; LV; Xiao-xia

    2012-01-01

    <正>Monitoring airborne tritium is an important routine work in heavy water reactor nuclear power stations and the units related with tritium. Nowadays direct measuring instruments like hand carrying tritium monitors are more often used in routine workshop environment check. Need for calibrating such monitors was suggested. A trial work about the calibration for radiation protection equipment for measuring airborne tritium was carried out with a domestic standard EJ/T 1077-1998 equivalent that of IEC 710.

  16. Global environmental transport models for tritium

    International Nuclear Information System (INIS)

    In this paper we discuss some of the obstacles to the construction of credible models of global tritium transport for use in dose assessments. We illustrate these difficulties by comparing model predictions of environmental tritium levels with measurements. Monitoring of tritium has shown that specific activities in precipitation over land are typically higher by a factor of three to four than those in precipitation over the oceans. Experience with modeling CO2 turnover in the oceans has led to the conclusion that two-box reservoir models of the ocean often give unsatisfactory representations of transient solutions. Failure to consider these factors in global models can lead to distorted estimates of collective dose and create difficulties in validation of the model against real data. We illustrate these problems with a seven-box model recommended by the National Council on Radiation Protection and Measurements in which we forced the atmospheric compartment to reproduce an exogenous function based on historic observations of HTO in precipitation at 500N. The fresh water response underestimates data from the Ottawa River by a factor of about five, and the ocean surface response overestimates tritium data from the surface waters of the Northern Pacific by nearly an order of magnitude. Revision of the model to include (1) separate over-land and over-ocean compartments of the atmosphere and (2) a box-diffusion model of the subsurface ocean brings the discrepant responses into good agreement with the environmental data. In a second exercise, we used a latitudinally disaggregated model and replaced a tropospheric compartment in the northern hemisphere by historic precipitation data. The model's response greatly underestimates the tritium specific activity in the southern hemisphere. These exercises lead us to doubt that a proper global transport model for tritium is available at present for collective dose assessment. 12 refs., 3 figs

  17. Comparison of different strategies for decommissioning a tritium laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Dylst, Kris, E-mail: Kris.Dylst@sckcen.be [SCK-CEN, Dismantling, Decontamination and Waste, Boeretang 200, 2400 Mol (Belgium); Slachmuylders, Frederik; Gilissen, Bart [SCK-CEN, Dismantling, Decontamination and Waste, Boeretang 200, 2400 Mol (Belgium)

    2013-10-15

    Highlights: ► Decontamination to below the free release limits is very labour intensive. ► Disposing of contaminated steel to a nuclear melting facility is cost effective. ► It can be advantageous to invest in decontamination of non-steel materials. -- Abstract: Between 2003 and 2009 two rooms that served as tritium laboratory at SCK• CEN and its ventilation system were decommissioned. Initially, the decommissioning strategy was to free release as much materials as possible. However, due to the imposed free release limit this was very labour intensive. Timing restrictions forced us to use a different strategy for the ventilation system. Most of the steel was disposed of to a nuclear melting facility. As a result there was a significant decrease in the required man labour. For the second laboratory room a similar strategy as for the ventilation was used: contaminated steel was disposed of to a nuclear melting facility and other materials that could not be easily decontaminated were disposed of as nuclear waste. At the expense of extra waste generation compared to the first laboratory the decommissioning was done using merely one third of the man hours. Comparison of the used strategies indicated opportunities for cost optimization. Even in absence of time constraints it is best to foresee a safe disposal of metals to a nuclear melting facility, whilst it is worth to invest in the labour intensive decontamination of the other materials to free release them.

  18. Status of R&D on Tritium Permeation Barrier Coatings for Tritium Breeding Blanket of Fusion Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The paper overviewed the recent progress in the application of several typical tritium permeation barrier (TPB) coatings and their corresponding fabrication technologies for tritium breeding blanket of fusion reactor. According to the design requirements of

  19. Tritium Aging Effects in Palladium on Kieselguhr

    Energy Technology Data Exchange (ETDEWEB)

    Shanahan, K.L. [Westinghouse Savannah River Company, AIKEN, SC (United States); Holder, J.S.; Wermer, J.R.

    1998-10-01

    50 weight % Pd on kieselguhr (Pd/k) is used in hydrogen isotope separation processes at the Savannah River Site. Long term aging studies on this material were undertaken in June, 1992. P-c-T data showing the aging effect of tritium loading for long periods will be presented and discussed covering from June, 1992 to March, 1997. Lowering of plateau pressures and increasing indications of in homogeneities have been observed in both tritium and deuterium absorption isotherms at 0 C, and desorption isotherms at 80 and 120 C.

  20. Tritium transport in lithium ceramics porous media

    Energy Technology Data Exchange (ETDEWEB)

    Tam, S.W.; Ambrose, V.

    1991-12-31

    A random network model has been utilized to analyze the problem of tritium percolation through porous Li ceramic breeders. Local transport in each pore channel is described by a set of convection-diffusion-reaction equations. Long range transport is described by a matrix technique. The heterogeneous structure of the porous medium is accounted for via Monte Carlo methods. The model was then applied to an analysis of the relative contribution of diffusion and convective flow to tritium transport in porous lithium ceramics. 15 refs., 4 figs.

  1. Tritium transport in lithium ceramics porous media

    Energy Technology Data Exchange (ETDEWEB)

    Tam, S.W.; Ambrose, V.

    1991-01-01

    A random network model has been utilized to analyze the problem of tritium percolation through porous Li ceramic breeders. Local transport in each pore channel is described by a set of convection-diffusion-reaction equations. Long range transport is described by a matrix technique. The heterogeneous structure of the porous medium is accounted for via Monte Carlo methods. The model was then applied to an analysis of the relative contribution of diffusion and convective flow to tritium transport in porous lithium ceramics. 15 refs., 4 figs.

  2. Atmospheric tritium. Progress report, July 1, 1975--March 31, 1976

    International Nuclear Information System (INIS)

    Progress is reported in the development of field equipment for sampling tritium in environmental samples. The performance of prototype tritiated hydrocarbon samples is discussed. Data are presented on the content of tritium in samples of rain water collected in Miami, Florida, Western Samoa, and Barbados during 1975, and tritium compounds in atmospheric samples collected at various world locations during 1975

  3. Tritium handling experience at Atomic Energy of Canada Limited

    Energy Technology Data Exchange (ETDEWEB)

    Suppiah, S.; McCrimmon, K.; Lalonde, S.; Ryland, D.; Boniface, H.; Muirhead, C.; Castillo, I. [Atomic Energy of Canad Limited - AECL, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-03-15

    Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritium powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.

  4. Procurement of tritium for fusion reactor. A design study of facility for production of fusion fuel tritium

    International Nuclear Information System (INIS)

    Tritium, a developmental fuel for use in fusion reactors, has been produced in fission research reactors in Japan by extraction from neutron-irradiated 6Li-targets. This paper describes the preliminary design of a large-scale production facility capable of producing 500 g of tritium annually. The present status of tritium production technology in Japan is also discussed. (author)

  5. Deuterium–tritium catalytic reaction in fast ignition: Optimum parameters approach

    Indian Academy of Sciences (India)

    B Khanbabaei; A Ghasemizad; S Khoshbinfar

    2014-09-01

    One of the main concerns about the currentworking on nuclear power reactors is the potential hazard of their radioactive waste. There is hope that this issue will be reduced in next generation nuclear fusion power reactors. Reactors will release nuclear energy through microexplosions that occur in a mixture of hydrogen isotopes of deuterium and tritium. However, there exist radiological hazards due to the accumulation of tritium in the blanket layer. A catalytic fusion reaction of DT mixture may stand between DD and an equimolar DT approach in which the fusion process continues with a small amount of tritium seed. In this paper, we investigate the possibility of DT reaction in the fast ignition (FI) scheme. The kinematic study of the main mechanism of the energy gain–loss term, which may disturb the ignition and burn process, was performed in FI and the optimum values of precompressed fuel and proton beam driver were derived. The recommended values of fuel parameters are: areal density $ρ R ≥ 5\\cdot$cm-2 and initial tritium fraction ≤ 0.025. For the proton beam, the corresponding optimum interval values are proton average energy $3≤ E_p ≤ 10$ MeV, pulse duration $5 ≤ t_p ≤ 15$ ps and power $5≤ W_p ≤ 12 × 10^{22}$ (keV$\\cdot$cm3$\\cdot$ps-1). It was proved that under the above conditions, a fast ignition DT reaction stays in the catalytic regime.

  6. Tritium contamination studies involving test materials and jet remote handling tools

    International Nuclear Information System (INIS)

    To determine the potential contamination of remote cutting and welding tools to be used in the JET torus after the introduction of tritium, experiments were performed using these tools on INCONEL pipe specimens which had been exposed to elemental tritium (HT) at a concentration of 4.6 x 1010 Bq/m3. A maximum tritium release of ∼15,600 Bq was measured during welding, resulting in the tool's surface contamination of 0.5 Bq/cm2. A second series of tests was performed in order to determine the degree of surface contamination of various materials when exposed to HTO as a function of the exposure time and the relative efficacy of different decontamination techniques. Stainless steel, aluminium alloy and PVC rigid were exposed to HTO (liquid) at a concentration 4.4 x 1010 Bq/1 for 1, 24, 120 hours and decontaminated. The decontamination techniques used included; leaching in water, baking at 100 degree C, hot air stream, weathering. The maximum levels of tritium surface contamination measured during the test were ∼12 Bq/cm2 for stainless steel, ∼ Bq/cm2 for aluminium alloy and ∼1,700 Bq/cm2 for PVC. A decontamination factor of about 80% as measured by smears was achieved using hot air stream at 125 degree C on stainless steel and aluminium alloy and baking PVC at 100 degree C. 6 figs., 2 tabs

  7. Development of tritium transport package for ITER SDS supply

    International Nuclear Information System (INIS)

    ITER is the next generation fusion machine with the fuel of deuterium and tritium. The transport of large amounts of tritium is an important issue from viewpoints of fuel supply and safety. For the shipment of tritium to the ITER site, a transport container needs to be developed and licensed as type B(U) package. It is an ITER requirement to transport tritium as metal tritide, which has been considered to be the safest way for tritium transport. There are not many available licensed packages on the market today. Examples are the WSRC Hydride Transport Vessel (HTV), which can be loaded with up to 18 g tritium in uranium tritide powder and JAERI Type B(U) package with capacity up to 25 g tritium in ZrCo tritide material. JAERI (now JAEA) has proposed a 250 g capacity tritium transport package for future fusion reactors. The design would utilize ZrCo to form the metal tritide to store the tritium. This new package would have a volume of only approximately 50% more than that of the 25 g capacity package and would be capable of repeated use. The tritium will be transported from tritium production sites, mainly the CANDU type reactor sites to ITER tritium plant building. According to the tritium supply plan derived from the operation and experiment plan of ITER, it is necessary to develop a large capacity tritium transport package which is licensed for international transportation. In 2009, Korea Atomic Energy Research Institute (KAERI) was commissioned the work of developing the tritium transport package from ITER Organization and the first stage of the development has been just finished. The interfaces of the package with related equipment/facilities were identified and the basic design and preliminary safety analyses were successfully performed. This paper describes the design requirements, basic design and the structural and thermal evaluation results of the developed package under the hypothetical accident conditions

  8. [Mechanism of tritium persistence in porous media like clay minerals].

    Science.gov (United States)

    Wu, Dong-Jie; Wang, Jin-Sheng; Teng, Yan-Guo; Zhang, Ke-Ni

    2011-03-01

    To investigate the mechanisms of tritium persistence in clay minerals, three types of clay soils (montmorillonite, kaolinite and illite) and tritiated water were used in this study to conduct the tritium sorption tests and the other related tests. Firstly, the ingredients, metal elements and heat properties of clay minerals were studied with some instrumental analysis methods, such as ICP and TG. Secondly, with a specially designed fractionation and condensation experiment, the adsorbed water, the interlayer water and the structural water in the clay minerals separated from the tritium sorption tests were fractionated for investigating the tritium distributions in the different types of adsorptive waters. Thirdly, the location and configuration of tritium adsorbed into the structure of clay minerals were studied with infrared spectrometry (IR) tests. And finally, the forces and mechanisms for driving tritium into the clay minerals were analyzed on the basis of the isotope effect of tritium and the above tests. Following conclusions have been reached: (1) The main reason for tritium persistence in clay minerals is the entrance of tritium into the adsorbed water, the interlayer water and the structural water in clay minerals. The percentage of tritium distributed in these three types of adsorptive water are in the range of 13.65% - 38.71%, 0.32% - 5.96%, 1.28% - 4.37% of the total tritium used in the corresponding test, respectively. The percentages are different for different types of clay minerals. (2) Tritium adsorbed onto clay minerals are existed in the forms of the tritiated hydroxyl radical (OT) and the tritiated water molecule (HTO). Tritium mainly exists in tritiated water molecule for adsorbed water and interlayer water, and in tritiated hydroxyl radical for structural water. (3) The forces and effects driving tritium into the clay minerals may include molecular dispersion, electric charge sorption, isotope exchange and tritium isotope effect.

  9. Elements of thought on the health risk associated to tritium

    International Nuclear Information System (INIS)

    This report addresses and analyses the health problematic set by tritium and assesses the robustness of the radiation protection system with respect to this radionuclide by highlighting the lack of scientific knowledge on biological effects, and researches to be promoted. After a presentation of epidemiologic and dosimetric approaches of the radiological risk assessment, the authors discuss results and knowledge gained by epidemiologic studies on the risk associated to tritium for mankind, and discuss the knowledge on biological effects of tritium and on the relative biological effectiveness of tritium. The report finally discusses the possibility of reconsidering the radiation weighting factor in the case of tritium

  10. Development of nuclear micro-battery with solid tritium source

    International Nuclear Information System (INIS)

    A micro-battery powered by tritium is being developed to utilize tritium produced from the Wolsong Tritium Removal Facility. The 3D p-n junction device has been designed and fabricated for energy conversion. Titanium tritide is adopted to increase tritium density and safety. Sub micron films or nano-powders of titanium tritide is applied on silicon semiconductor device to reduce the self absorption of beta rays. Until now protium has been used instead of tritium for safety. Hydrogen was absorbed up to atomic ratio of ∼1.3 and ∼1.7 in titanium powders and films, respectively.

  11. Comparative study of the tritium distribution in metals

    Energy Technology Data Exchange (ETDEWEB)

    Perevezentsev, A.N. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)], E-mail: alexander.perevezentsev@iter.org; Bell, A.C. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Rivkis, L.A.; Filin, V.M.; Gushin, V.V.; Belyakov, M.I.; Bulkin, V.I.; Kravchenko, I.M.; Ionessian, I.A. [All-Russia Institute of Inorganic Materials, 123060, P.O. Box 369, VNIINM, Moscow (Russian Federation); Torikai, Y.; Matsuyama, M.; Watanabe, K. [Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555 (Japan); Markin, A.I. [State Scientific Center TRINITI, 142190 Moscow Region (Russian Federation)

    2008-01-31

    Coupons of stainless steel, Inconel, beryllium, copper and aluminium bronze were exposed to tritium in hydrogen gas mixtures over a wide range of parameters: temperature up to 770 K, pressure from 1 x 10{sup -4} MPa to 0.05 MPa, tritium concentration from 1 at.% to 98 at.%. The tritium concentration on the surface and distribution through the metals were measured using radiography, radioluminography, {beta}-ray induced X-ray spectroscopy and acid etching methods. The effect of metal processing, such as forging, polishing and heat treatment on the tritium distribution was studied along with parameters relating to the exposure of the metal to tritium.

  12. Quantitative determination of tritium in metals and oxides

    International Nuclear Information System (INIS)

    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

  13. Organically bound tritium, OBT: Its true constitution

    International Nuclear Information System (INIS)

    Full text: Tritium, which is analytically determined to be non-exchangeable bound in tissue solids, is assumed to be bound to carbon. Furthermore, it follows that the biochemical passways by photosynthesis or enzymatic transfer reactions are retarded by the kinetic isotope effect leading to discrimination of tritium in biomolecules. In contrast, the logistic growth analysis of plants discloses a larger intrinsic growth rate of OBT than of OBH, resulting in tritium accumulation in biomolecules. Exchange experiments providing fractionation factors of 1.4 and 2 confirm this accumulation. In summary a larger part of the so called OBT is not carbon bound but consists of tritium positioned in hydrogen bridges of biopolymers which have been occupied during formation of the molecules and which became later inaccessible for exchange (so called buried hydrogens). Furthermore, there are experimental results indicating even rapid exchange during the in vivo state but inhibited in the in vitro state, which is commonly given in bio samples prepared for analysis. (author)

  14. Assessment of Tritium in Production Workers

    International Nuclear Information System (INIS)

    The tritium bioassay programme at the Savannah River Plant is geared for rapid urinalysis of large numbers of samples. More than 300 000. urine samples have been analysed in the past ten years. A liquid scintillation counting procedure currently used for analysis of urine samples is described. Untreated samples containing as little as 1μc of tritium per litre can be assayed in one minute. The detection limit for distilled urine is 5 x 10-4 μc of tritium per litre. Automation of equipment, optimum scintillation mixture and sample volumes, selection of reagents and counting containers, and elimination of interfering radionuclides are discussed: Empirical studies of biological half-life are summarized. In 310, cases where the initial-tritium conr centrations in urine ranged from 20 to 118 μc/l the average biological half-life was 9.5 d. The half-life varied inversely fwith ambient temperature and'age of employees. (author)

  15. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 1013Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 1010Bq (0.5 Ci) per day per ton of fuel

  16. Stereo and regioselectivity in ''Activated'' tritium reactions

    International Nuclear Information System (INIS)

    To investigate the stereo and positional selectivity of the microwave discharge activation (MDA) method, the tritium labeling of several amino acids was undertaken. The labeling of L-valine and the diastereomeric pair L-isoleucine and L-alloisoleucine showed less than statistical labeling at the α-amino C-H position mostly with retention of configuration. Labeling predominated at the single β C-H tertiary (methyne) position. The labeling of L-valine and L-proline with and without positive charge on the α-amino group resulted in large increases in specific activity (greater than 10-fold) when positive charge was removed by labeling them as their sodium carboxylate salts. Tritium NMR of L-proline labeled both as its zwitterion and sodium salt showed also large differences in the tritium distribution within the molecule. The distribution preferences in each of the charge states are suggestive of labeling by an electrophilic like tritium species(s). 16 refs., 5 tabs

  17. Small system for tritium accelerator mass spectrometry

    Science.gov (United States)

    Roberts, Mark L.; Davis, Jay C.

    1993-01-01

    Apparatus for ionizing and accelerating a sample containing isotopes of hydrogen and detecting the ratios of hydrogen isotopes contained in the sample is disclosed. An ion source generates a substantially linear ion beam including ions of tritium from the sample. A radio-frequency quadrupole accelerator is directly coupled to and axially aligned with the source at an angle of substantially zero degrees. The accelerator accelerates species of the sample having different mass to different energy levels along the same axis as the ion beam. A spectrometer is used to detect the concentration of tritium ions in the sample. In one form of the invention, an energy loss spectrometer is used which includes a foil to block the passage of hydrogen, deuterium and .sup.3 He ions, and a surface barrier or scintillation detector to detect the concentration of tritium ions. In another form of the invention, a combined momentum/energy loss spectrometer is used which includes a magnet to separate the ion beams, with Faraday cups to measure the hydrogen and deuterium and a surface barrier or scintillation detector for the tritium ions.

  18. Standardization of Tritium Water by TDCR Method

    Institute of Scientific and Technical Information of China (English)

    吴永乐; 梁珺成; 柳加成; 熊文俊; 姚顺和; 郭晓清; 陈细林; 杨元第; 袁大庆

    2012-01-01

    The triple-to-double coincidence ratio (TDCR) method of liquid scintillation count- ing is an absolute measurement method of radioactivity. The formulation of the TDCR method and the established TDCR liquid scintillation counter are presented in this paper. The NIST standard reference material (SRM) of tritium water was measured to verify the performance of the TDCR liquid scintillation counter.

  19. Tritium in the Savannah River environment addendum to WSRC-RP--90-424-1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E. Jr.; Carlton, W.H.

    1992-05-28

    This document is an addendum to Tritium in the Savannah River Site Environment,'' WSRC-RP90-424- 1, released in May of 1991. The purpose of this report is to update the information found in WSRC-RP-90-424-1 for the four year period 1987--1990. Some data has also been included from 1991. The report includes summaries of atmospheric and aqueous monitoring of tritium and estimates of the dose to the population surrounding the Savannah River Site.

  20. Tritium in the Savannah River environment addendum to WSRC-RP--90-424-1, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E. Jr.; Carlton, W.H.

    1992-05-28

    This document is an addendum to ``Tritium in the Savannah River Site Environment,`` WSRC-RP90-424- 1, released in May of 1991. The purpose of this report is to update the information found in WSRC-RP-90-424-1 for the four year period 1987--1990. Some data has also been included from 1991. The report includes summaries of atmospheric and aqueous monitoring of tritium and estimates of the dose to the population surrounding the Savannah River Site.

  1. Investigation of tritium in groundwater at Site 300

    Energy Technology Data Exchange (ETDEWEB)

    Buddemeier, R.W.

    1985-12-30

    In 1984, landfill monitoring wells at Site 300, a Lawrence Livermore National Laboratory (LLNL) explosive test site, revealed the presence of groundwater contaminated with tritium. These tritium levels were in excess of the State of California drinking water standard. A major investigation was initiated that included a search of records concerning tritium use, disposal, and previous analyses, and a survey of tritium levels in soil, vegetation, and water in contaminated and potentially contaminated areas. Over 50 boreholes were drilled for this investigation to characterize the local hydrogeology and tritium distributions, and a network of soil moisture and groundwater monitoring points was installed. This report presents the work completed through the end of September 1985: the records search; records for drilling completed as part of this study; characterization of the geology, hydrology, and tritium distributions in the contaminated area; and an initial assessment of the probable tritium sources, pathways, and migration rates. 19 refs.

  2. Global cycling of tritium and iodine-129

    International Nuclear Information System (INIS)

    Dynamic linear compartmnt models are used widely to describe global cycling of environmental tritium and 129I. Important tests of these models by comparison of predictions with environmental data from anthropogenic sources are discussed. A tritium model, based on the global hydrologic cycle that reproduces time-series data from atmospheric nuclear weapons testing 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 atmosperic tritium in both hemispheres is presented. The model includes: hemispheric stratosphere compartments; disaggregation of the troposphere and ocean surface waters into eight latitude zones; consideration of the different concentrations of water in air over land and the ocean in calculating the specific activity of atmospheric tritium; and use of a box-diffusion model for transport in the ocean. An important prediction of a global model for 129I, which we developed previously from data on cycling of naturally occurring stable iodine, is that the mean residence time in the first 1 m of surface soil is about 4000 y. However, a recent analysis of measured soil profiles of 129I near the Savannah River Plant, based on a linear compartment model for downward transport through soil, suggested that the mean residence time in the first 0.3 m is only about 40 y. A diffusion model is used to describe the measured soil profiles, and the resulting diffusion coefficient is shown to correspond to mean residence times in the first 0.3 m and 1 m of soil of about 80 and 900 y, respectively. The value for the first 1 m can be reconciled with the prediction of the global model

  3. The Reconcentration of Tritium by Distillation

    International Nuclear Information System (INIS)

    A method of tritium reconcentration by the total reflux distillation of water under reduced pressure using random-packed columns was investigated. For the maximum removal of tritium from a one liter reservoir, operating periods of several weeks were required. For this a fully automatic fractionating system incorporating an apparatus for taking samples automatically under reduced pressure was developed to enable the distillation to proceed with the minimum of manual adjustment. To reduce the possibility of flooding at the base of the column due to gravity settling of the packing over long periods of time, a novel design feature was incorporated at the junction of the column and the reflux meter. The performance of several commercially available column packings was investigated in an aqueous environment. Details of the packing pre-treatment to inhibit maldistribution in a liquid of high surface tension are given and enrichment factors calculated. A low H.E.T.P. (height equivalent to a theoretical plate) of about 0.8 in has been achieved with pre-treated phosphor-bronze gauze rings in an aqueous environment. With a reservoir-to-boiler volume ratio of the order of 7 : 1, a maximum of 98% of the tritium in the reservoir was removed in 28 d continuous distillation with a throughput of 100 ml/h. This indicated a tritium reconcentration factor of 6.3. By increasing the throughput to 140 ml/h, 92% of the tritium was extracted in 11 d. The reproducibility of the reconcentration factor with time was, however, shown to vary, and the reasons for this are discussed in the paper. (author)

  4. Fluorine 18 in tritium generator ceramic materials

    International Nuclear Information System (INIS)

    At present time, the ceramic materials generators of tritium are very interesting mainly by the necessity of to found an adequate product for its application as fusion reactor shielding. The important element that must contain the ceramic material is the lithium and especially the isotope with mass=6. The tritium in these materials is generated by neutron irradiation, however, when the ceramic material contains oxygen, then is generated too fluorine 18 by the action of energetic atoms of tritium in recoil on the 16 O, as it is showed in the next reactions: 1) 6 Li (n, α) 3 H ; 2) 16 O(3 H, n) 18 F . In the present work was studied the LiAlO2 and the Li2O. The first was prepared in the laboratory and the second was used such as it is commercially expended. In particular the interest of this work is to study the chemical behavior of fluorine-18, since if it would be mixed with tritium it could be contaminate the fusion reactor fuel. The ceramic materials were irradiated with neutrons and also the chemical form of fluorine-18 produced was studied. It was determined the amount of fluorine-18 liberated by the irradiated materials when they were submitted to extraction with helium currents and argon-hydrogen mixtures and also it was investigated the possibility about the fluorine-18 was volatilized then it was mixed so with the tritium. Finally it was founded that the liberated amount of fluorine-18 depends widely of the experimental conditions, such as the temperature and the hydrogen amount in the mixture of dragging gas. (Author)

  5. Transport of tritium contamination to the atmosphere in an arid environment

    Science.gov (United States)

    Garcia, C.A.; Andraski, B.J.; Stonestrom, D.A.; Cooper, C.A.; Johnson, M.J.; Michel, R.L.; Wheatcraft, S.W.

    2009-01-01

    Soil-plant-atmosphere interactions strongly infl uence water movement in desert unsaturated zones, but litile is known about how such interactions aff ect atmospheric release of subsurface water-borne contaminants. This 2-yr study, performed at the U.S. Geological Survey's Amargosa Desert Research Site in southern Nevada, quantifi ed the magnitude and spatiotemporal variability of tritium (3H) transport from the shallow unsaturated zone to the atmosphere adjacent to a low-level radioactive waste (LLRW) facility. Tritium fl uxes were calculated as the product of 3H concentrations in water vapor and respective evaporation and transpiration water-vapor fl uxes. Quarterly measured 3H concentrations in soil water vapor and in leaf water of the dominant creosote-bush [Larrea tridentata (DC.) Coville] were spatially extrapolated and temporally interpolated to develop daily maps of contamination across the 0.76-km2 study area. Maximum plant and root-zone soil concentrations (4200 and 8700 Bq L-1, respectively) were measured 25 m from the LLRW facility boundary. Continuous evaporation was estimated using a Priestley-Taylor model and transpiration was computed as the diff erence between measured eddy-covariance evapotranspiration and estimated evaporation. The mean evaporation/transpiration ratio was 3:1. Tritium released from the study area ranged from 0.12 to 12 ??g d-1 and totaled 1.5 mg (8.2 ?? 1010 Bq) over 2 yr. Tritium fl ux variability was driven spatially by proximity to 3H source areas and temporally by changes in 3H concentrations and in the partitioning between evaporation and transpiration. Evapotranspiration removed and limited penetration of precipitation beneath native vegetation and fostered upward movement and release of 3H from below the root zone. ?? Soil Science Society of America.

  6. Plant-Based Plume-Scale Monitoring Reveals the Extents and Pathways of Tritium Transport

    Science.gov (United States)

    Andraski, B. J.; Michel, R. L.; Halford, K. J.; Stonestrom, D. A.; Abraham, J. D.

    2005-05-01

    Cost-effective methods are needed to detect contamination near radioactive-waste and other contaminated sites. Such methods should be capable of providing an early warning of contaminant releases and be accurate and robust enough for monitoring the long-term performance of waste-isolation facilities and remediation measures. Plant-based methods were developed adjacent to a closed low-level radioactive waste (LLRW) facility in the Amargosa Desert, Nevada. Objectives were to (i) characterize and map the spatial variability of plant-water tritium, (ii) develop empirical relations to predict subsurface tritium contamination from plant-water concentrations, and (iii) gain insight into transport pathways and processes. Tritium was selected because it is a common radionuclide disposed at radioactive waste sites and it is a good tracer of water movement. Solar-distillation and solid-phase-extraction were used to collect and prepare plant (creosote bush, Larrea tridentata) foliage water for direct-scintillation counting. The maximum plant-water tritium concentration was 4,890 Bq/L; background values averaged 2.5 Bq/L. Geostatistical analysis showed that plant concentrations were spatially correlated to a distance of 380 m. Simple-contour and kriged maps of plant concentrations identified "hot spots" that were verified by soil-water-vapor measurements. Empirical linear relations between plant water and soil-water-vapor concentrations measured at the 0.5- and 1.5-m sampling depths were used to map the spatial distributions of root-zone and sub-root-zone tritium, respectively. Results showed that tritium migration away from the waste source primarily occurs in the gas phase with preferential transport through a dry, gravelly layer beneath the root zone, from which it moves upward and is subsequently released to the surface environment. Shallow and deep geologic units controlling preferential transport through the unsaturated zone were mapped by direct-current electrical

  7. Tritium concentrations in natural waters in Japan before use of a large quantity of tritium on its fusion program

    International Nuclear Information System (INIS)

    To clarify environmental tritium levels in Japan before use of a large quantity of tritium on its fusion program, the authors analyzed the tritium concentrations in various water samples, such as rain, river, lake, coastal sea and deep sea waters in Japan. The tritium concentrations in rain water were high at higher latitude. The definite differences of the tritium concentrations due to the weather conditions or seasons were not observed. The average tritium concentration in river water was 51.5 pCi/l in 1982 and that in lake water was 63.5 pCi/l in 1983. The vertical profiles of the tritium concentrations in the representative lakes were almost homogeneous except surface water. The average tritium concentrations in coastal seawater were about 20 pCi/l in both 1982 and 1983. The tendency of the increased tritium level with latitude as reported in literature was not observed by these experiments. Tritium levels in natural water in small isolated islands were lower than those at other places. In the Japan Sea, it was recognized that tritium was distributed down to around 2000 m in depth. This means that the more active vertical mixing of water masses than that in the Pacific Ocean is taking place. (author)

  8. Isotope exchange reactions on ceramic breeder materials and their effect on tritium inventory

    Energy Technology Data Exchange (ETDEWEB)

    Nishikawa, M.; Baba, A. [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Kawamura, Y.; Nishi, M.

    1998-03-01

    Though lithium ceramic materials such as Li{sub 2}O, LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 2}TiO{sub 3} and Li{sub 4}SiO{sub 4} are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials has not been fully understood. The isotope exchange reaction rate between hydrogen isotopes in the purge gas and tritium on the surface of breeding materials have not been quantified yet, although helium gas with hydrogen or deuterium is planned to be used as the blanket purge gas in the recent blanket designs. The mass transfer coefficient representing the isotope exchange reaction between H{sub 2} and D{sub 2}O or that between D{sub 2} and H{sub 2}O in the ceramic breeding materials bed is experimentally obtained in this study. Effects of isotope exchange reactions on the tritium inventory in the bleeding blanket is discussed based on data obtained in this study where effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions are considered. The way to estimate the tritium inventory in a Li{sub 2}ZrO{sub 3} blanket used in this study shows a good agreement with data obtained in such in-situ experiments as MOZART, EXOTIC-5, 6 and TRINE experiments. (author)

  9. Tritium Transport at the Rulison Site, a Nuclear-stimulated Low-permeability Natural Gas Reservoir

    Energy Technology Data Exchange (ETDEWEB)

    C. Cooper; M. Ye; J. Chapman

    2008-04-01

    The U.S. Department of Energy (DOE) and its predecessor agencies conducted a program in the 1960s and 1970s that evaluated technology for the nuclear stimulation of low-permeability natural gas reservoirs. The second project in the program, Project Rulison, was located in west-central Colorado. A 40-kiltoton nuclear device was detonated 2,568 m below the land surface in the Williams Fork Formation on September 10, 1969. The natural gas reservoirs in the Williams Fork Formation occur in low permeability, fractured sandstone lenses interbedded with shale. Radionuclides derived from residual fuel products, nuclear reactions, and activation products were generated as a result of the detonation. Most of the radionuclides are contained in a cooled, solidified melt glass phase created from vaporized and melted rock that re-condensed after the test. Of the mobile gas-phase radionuclides released, tritium ({sup 3}H or T) migration is of most concern. The other gas-phase radionuclides ({sup 85}Kr, {sup 14}C) were largely removed during production testing in 1969 and 1970 and are no longer present in appreciable amounts. Substantial tritium remained because it is part of the water molecule, which is present in both the gas and liquid (aqueous) phases. The objectives of this work are to calculate the nature and extent of tritium contamination in the subsurface from the Rulison test from the time of the test to present day (2007), and to evaluate tritium migration under natural-gas production conditions to a hypothetical gas production well in the most vulnerable location outside the DOE drilling restriction. The natural-gas production scenario involves a hypothetical production well located 258 m horizontally away from the detonation point, outside the edge of the current drilling exclusion area. The production interval in the hypothetical well is at the same elevation as the nuclear chimney created by the detonation, in order to evaluate the location most vulnerable to

  10. Investigation and design of the dismantling process for irradiation capsules containing tritium. 1. Conceptual investigation and basic design

    International Nuclear Information System (INIS)

    In-pile functional tests of tritium breeding blankets for fusion reactors have been planned by Japan Atomic Energy Agency (JAEA), using a test blanket module (TBM) which will be loaded in ITER. In preparation for the in-pile functional tests, JAEA has been being performed irradiation experiments of solid breeder materials including Li2TiO3, which is the first candidate of tritium breeder materials for the blanket of the demonstration reactor (DEMO) in a water-cooled solid-breeder design concept in Japan. The present report describes conceptual investigation and basic design of the dismantling process for irradiation capsules which were used in irradiation experiments by the Japan Materials Testing Reactor (JMTR) of JAEA. An irradiation capsule to be dismantled is comprised of a cylindrical outer-container (65mm in outer diameter) and an inner-container which is loaded with Li2TiO3 pebbles. In the present design, the irradiation capsule is cut by a band saw; the released tritium is recovered safely by a purge-gas system, and is consolidated into a radioactive waste form. Furthermore, an inner-box enclosing the dismantling apparatus has been designed as a safety countermeasure of possible tritium release from the dismantling apparatus in accidental events. The adoption of the inner-box has brought a prospect to be able to utilize an existing hot cell (β γ cell) equipped with usual wall material permeable to tritium, without extensive refurbishing of the cell. Thus, the present study has indicated the feasibility of the dismantling process for the irradiated JMTR capsules containing tritium. The results of the present investigation and design will contribute to the design of the TBM structure and to the planning of the dismantling process of the TBM. (author)

  11. DECOMMISSIONING THE HIGH PRESSURE TRITIUM LABORATORY AT LOS ALAMOS NATIONAL LABORATORY

    International Nuclear Information System (INIS)

    In May 0f 2000, the Cerro Grande wild land fire burned approximately 48,000 acres in and around Los Alamos. In addition to the many buildings that were destroyed in the town site, many structures were also damaged and destroyed within the 43 square miles that comprise the Los Alamos National Laboratory (LANL). A special Act of Congress provided funding to remove Laboratory structures that were damaged by the fire, or that could be threatened by subsequent catastrophic wild land fires. The High Pressure Tritium Laboratory (HPTL) is located at Technical Area (TA) 33, building 86 in the far southeast corner of the Laboratory property. It is immediately adjacent to Bandelier National Park. Because it was threatened by both the Cerro Grande fire in 2000, and the 16,000- acre Dome fire in 1996, the former tritium processing facility was placed on the list of facilities scheduled for Decontamination and Decommissioning under the Cerro Grande Rehabilitation Project. The work was performed through the Facilities and Waste Operations (FWO) Division and is integrated with other Laboratory D and D efforts. The primary demolition contractor was Clauss Construction of San Diego, California. Earth Tech Global Environmental Services of San Antonio, Texas was sub-contracted to Clauss Construction, and provided radiological decontamination support to the project. Although the forty-seven year old facility had been in a state of safe-shutdown since operations ceased in 1990, a significant amount of tritium remained in the rooms where process systems were located. Tritium was the only radiological contaminant associated with this facility. Since no specific regulatory standards have been set for the release of volumetrically contaminated materials, concentration guidelines were derived in order to meet other established regulatory criteria. A tritium removal system was developed for this project with the goal of reducing the volume of tritium concentrated in the concrete of the

  12. Impact of low-level radiation with special reference to tritium in environment

    International Nuclear Information System (INIS)

    Radiation is invisible, but exists in various types, in the form of particles and/or energy bundles. The effects of low-level radiation seem very abstract since these can not be perceived by our sensory organs. The increase in natural background radiation from various inadvertent sources like tritium has the prospect of altering the entire scenario of billions of years' slow and steady biogenetic evolution. Mankind, by developing atomic technologies, is unleashing forces which it does understand but not beyond experimental findings. There is no wise sorcerer who can undo the damage we are causing. Tritium is a radioactive form of hydrogen that is produced in the reactor core. The released tritium replaces hydrogen in water. Tritium in water when gets ingested, causes continuos internal low-level beta radiation exposure over a long period. Proposed presentation will focus on the possible long term damage caused by its low-level exposure is dependent on the length of duration living tissue spends in the radiation field, not on the relative radiation field strength. As internal radiation pulses never stop, impact is continuous by the ambient radiation atmosphere. There is no chance to heal at the molecular level, except small chances of DNA repair since the organically bound tritium has greater severe influence with the slow turnover. Though the situation needs not be alarming with tritium, the studies on radiation damage on various parameters have given evidence of two compartments of radiation damage; the reparable or potentially lethal and the irreparable or lethal. With emerging new reports on the stochastic effects, those for which the probability, rather than the severity of an effect from tritium occurring as a function of dose also can not be ruled out. Biotoxicity of tritium in the form of induction of cancer, hereditary effects, teratogenesis and life shortening really needs an exhaustive investigation and warrants careful evaluation. However, a positive

  13. Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

    International Nuclear Information System (INIS)

    Highlights: •The ITER TBM Program is one of the ITER missions. •We model a TBM port cell with CFD to optimize the design choices. •The heat and tritium releases management in TBM port cells has been optimized. •It is possible to reduce the T-concentration below one DAC in TBM port cells. •The TBM port cells can have human access within 12 h after shutdown. -- Abstract: Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown

  14. Analysis and speciation of the tritium in environmental matrices

    International Nuclear Information System (INIS)

    This study deals with environmental monitoring. The main aims are (i) the optimisation of the analytical procedure for the tritium in organic form determination, and (ii) the identification of the tritium bearing molecules which are responsible for its transfer from the environment to man. The study was divided into three stages. First an analytical method was developed to determine hydrogen content of several samples, which is a key element to calculate accurate organically bound tritium activities. Secondly, the impact of the organically bound tritium fractions separation (labile exchange) for the determination of the representative fraction of the level of environmental tritium activity was then evaluated. For that, the amount of solubilised sample was estimated. Finally, the speciation of tritium in environmental samples was investigated. Several molecules classes and organic compounds dissolved in the labile exchanges solvent were identified. The results show that the distribution of tritium in organisms depends on both properties of the chemical bond in which it is involved and chemical properties of tritium bearing molecules. The identified compounds belong to the molecules classes such as carbohydrates or amino acids, constitutive of living organisms. It would now be of interest to study the tritium distribution in an environmental sample to target molecules of interest and study the impact of tritium from the environment to man. (author)

  15. Best practices in management of heavy water and tritium

    International Nuclear Information System (INIS)

    The heavy water inventory of a typical HWR constitutes about 12% of the capital cost of the HWR. The typical tritium production in a single unit HWR is about 2 x 106 Ci/y.1 Heavy water and tritium control are important aspects of HWR operation, and this involves people, procedures, equipment and heavy water and tritium separation systems. Station personnel are trained to understand the importance of heavy water management and the economics and environmental impact of tritiated heavy water losses. The tritium and heavy water losses from a HWR are both airborne and waterborne in nature. Tritium is of particular concern in the HWR industry given the nature of heavy water reactors to build up high levels of tritium over time. Recent increased interest from regulators and the public has led more HWR utilities to pay increasing attention to occupational safety and environmental emissions of tritium at their power stations. As competing reactor technologies improve, a simple and economic means for tritium removal from heavy water in HWRs is essential for the long- term attractiveness of HWR technology. Tritium safety, occupational and environmental issues are of central importance in HWR licensing and operation. Building upon GE's extensive operational experience in tritium management in HWR reactors and its own tritium handling facility, GE2 has developed a large-scale diffusion-based isotope separation process as an alternative to conventional cryogenic distillation. Having a tritium inventory an order of magnitude lower than conventional cryogenic distillation, this process is very attractive for heavy water detritiation, applicable to single and multi-unit HWR and research reactors. Additionally, the new process has significant benefits to an operating HWR utility such as reducing environmental emissions and significantly lowering reactor vault tritium MPC(a) levels to a point where station capacity factors can be improved by shorter outages - representing best

  16. Tritium in some typical ecosystems

    International Nuclear Information System (INIS)

    The environmental significance of 3H releases prompted an IAEA-sponsored coordinated research programme on various aspects. Data were collected to help health physicists, radioecologists, radiobiologists and environmentalists to predict the behaviour of 3H in the major terrestrial ecosystems of the world. A common methodology was used to carry out a variety of projects in widely varying biomes, from tropical to arctic regions: in Belgium, on terrestrial food chains, with deposition of tritiated water (HTO) on crops and pasture, and incorporation of 3H into proteins, nucleic acids, etc.; in Finland, plots of pasture and forest were labelled by HTO, and plant uptake were studied; in France, 3H-content in water, in relation to different parts of vines, orange and olive trees in a Mediterranean climate; in the Federal Republic of Germany, contamination due to 3H-releases; in India, mean 3H-residence time in some tropical trees; in Mexico, 3H-persistence as free-water 3H and tissue-bound 3H in crops; in the Netherlands, 3H-metabolism in ruminants; in the Philippines, residence time in soil and in various commonly edible crops, and excretion time; in Thailand, half residence time in soil and local vegetation; in the USA, the effects of HTO vapour and liquid exposure in a wide range of climatic conditions, including organic fixation and concentration factors. An extensive bibliography is attached, and also annexes of laboratories and project titles; plant species, exposure and residence times; comparable lists for animals studied; scientific and common names of the species, and a glossary

  17. Oxidation of zirconium alloys in 2.5 kPa water vapor for tritium readiness.

    Energy Technology Data Exchange (ETDEWEB)

    Mills, Bernice E.

    2007-11-01

    A more reactive liner material is needed for use as liner and cruciform material in tritium producing burnable absorber rods (TPBAR) in commercial light water nuclear reactors (CLWR). The function of these components is to convert any water that is released from the Li-6 enriched lithium aluminate breeder material to oxide and hydrogen that can be gettered, thus minimizing the permeation of tritium into the reactor coolant. Fourteen zirconium alloys were exposed to 2.5 kPa water vapor in a helium stream at 300 C over a period of up to 35 days. Experimental alloys with aluminum, yttrium, vanadium, titanium, and scandium, some of which also included ternaries with nickel, were included along with a high nitrogen impurity alloy and the commercial alloy Zircaloy-2. They displayed a reactivity range of almost 500, with Zircaloy-2 being the least reactive.

  18. IN-SITU TRITIUM BETA DETECTOR

    International Nuclear Information System (INIS)

    The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye. Based on the results achieved, it is premature to initiate Phase 2 and commit to a prototype

  19. Tritium uptake kinetics in crayfish (Orconectes immunis)

    International Nuclear Information System (INIS)

    Uptake of tritiated water (HTO) by Orconectes immunis was investigated under laboratory conditions. Tritium uptake in the tissue-free water fraction (TFWT) was described using an exponential model. When steady-state was reached, the ratio of TFWT to HTO was approximately 0.9. Uptake of tritium in the organically-bound fraction (OBT) proceeded slowly, and had not reached steady-state after 117 days of culture. Although steady-state was never reached, the maximum observed ration of OBT to TFWT in whole animals was approximately 0.6. However, this ratio exceeded unity in the exoskeleton. Specific activity ratios of OBT between crayfish and lettuce (food source) were less than or at unity for various test conditions

  20. Tritium labelling of two new analgesic drugs

    International Nuclear Information System (INIS)

    The labelling with tritium of two arylpropionic esters was studied. The synthesis between 3H-Ibuprofen and the two unlabelled alcoholic moieties (Cl-Alkanol and CF3-Alkanol) was performed. Assuming that we got ready the acidic moiety, 3H-Ibuprofen, in our Laboratory, we attempted to label with tritium the alcoholic moiety and then go on to its esterification. Prior to labelling, thermic stability of 2-(4-(3-chlorophenyl)-1-piperazinyl) ethanol (Cl-Alkanol) was studied. As result of this study we had to change the labelling method, so that the Cl-Alkanol was unstable at 700C. Purification was accomplished through thin layer chromatography (TLC) and high performance liquid chromatography (HPLC). Concentration, purity and specific activities of the two labelled compounds were determined by ultraviolet, HPLC and liquid scintillation techniques. (author)

  1. IN-SITU TRITIUM BETA DETECTOR

    Energy Technology Data Exchange (ETDEWEB)

    J.W. Berthold; L.A. Jeffers

    1998-04-15

    The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye. Based on the results achieved, it is premature to initiate Phase 2 and commit to a prototype

  2. Low-exposure tritium radiotoxicity in mammals

    International Nuclear Information System (INIS)

    A special feature of tritium radiation is the very low energy of the beta particles (5.7 keV average, 18 keV maximum). This low energy results in very short particle ranges in tissue, ranges that are less than cell dimensions. Another result of the low energy is that the ionization density along beta-ray tracks (even though the tracks are short) is significantly greater than that associated with secondary electrons from gamma rays. The studies of tritium radiotoxicity reviewed, involving chronic 3HOH exposures in mammals, demonstrate in both mice and monkeys that biological effects can be measured following remarkably low levels of exposure --- levels in the range of serious practical interest to radiation protection. (Namekawa, K.)

  3. Experimental setup for the determination of exchangeable hydrogen in environmental samples using deuterium and tritium

    Energy Technology Data Exchange (ETDEWEB)

    Pastor, L.; Siclet, F. [EDF R et D (France); Peron, O.; Gegout, C.; Montavon, G.; Landesman, C. [Laboratoire SUBATECH, IN2P3/CNRS, EMN, Universite de Nantes (France); Fourre, E.; Jean-Baptiste, P. [LSCE, UMR 8112 CEA-CNRS-UVSQ/IPSL (France)

    2014-07-01

    Tritium ({sup 3}H or T) is a radioactive isotope of the element hydrogen with a half-life of 12.32 yrs. It is naturally produced in the upper atmosphere, but also by the nuclear industry. It is used in many fields like medical research and watch making. It is thus released in the environment on gaseous and liquid form by these facilities and is currently the major released radionuclide in liquid effluent from French nuclear power plants (in HTO form). Current studies dealing with the fate and behavior of tritium in the environment focus mainly on its organic form, i.e. the organically bound tritium (OBT). It is indeed more resilient in the environment than the tritiated water (HTO) as it is part of the organic matter cycle. There is nevertheless a distinction to be made between the exchangeable and the non-exchangeable fraction of OBT. When hydrogen is linked to nitrogen, sulfur or oxygen, it is considered to be exchangeable with the H contained in the surrounding solution or in the atmospheric water phase. Thus, its residence time within the molecule will be reduced and closely linked to the surrounding parameters. When hydrogen is linked to carbon, it is assumed that the link is more stable and thus the residence time in the molecule will be enhanced. It is thus important to know the fraction of exchangeable OBT when addressing the residence time of tritium in the environment. The present study aims at assessing this fraction in different environmental matrixes using deuterium and/or tritium. Compared to several others studies on exchangeable hydrogen where experiments were conducted at high temperature and/or high pressure, this study follows a different approach with experiments conducted at ambient temperature and atmospheric pressure (natural conditions) with a controlled hygrometric value within the system. The system itself consists in a glove box modified to fulfill the requirements for an efficient control on the experimental parameters (temperature

  4. Atmospheric tritium sampling at the NTS

    International Nuclear Information System (INIS)

    A modification of the method for the simultaneous collection of gaseous tritium and tritiated water vapor in air is under investigation. It is believed that the auxiliary hydrogen stream is unnecessary if a small volume of distilled water is added at the point of collection of water generated by the Pt-H2-O2 reaction. To test this hypothesis, two samplers were set up to sample the same air stream. Results are encouraging

  5. Decommissioning a tritium glove-box facility

    International Nuclear Information System (INIS)

    A large glove-box facility for handling reactive metal tritides was decommissioned. Major sections of the glove box were decontaminated and disassembled for reuse at another tritium facility. To achieve the desired results, decontamnation required repeated washing, first with organic liquids, then with water and detergents. Worker protection was provided by simple ventilation combined with careful monitoring of the work areas and employees. Several innovative techniques are described

  6. Determination of tritium in wine yeast samples

    International Nuclear Information System (INIS)

    Analytical procedures were developed to determine tritium in wine and wine yeast samples. The content of organic compounds affecting the LSC measurement is reduced by fractioning distillation for wine samples and azeotropic distillation/fractional distillation for wine yeast samples. Finally, the water samples were normally distilled with K MO4. The established procedures were successfully applied for wine and wine samples from Murfatlar harvests of the years 1995 and 1996. (authors)

  7. Tritium Behaviour in the Fusion Reactor Materials

    OpenAIRE

    Pajuste, Elīna

    2012-01-01

    ABSTRACT Doctoral thesis is devoted to the development of future energy source nuclear fusion. The objective of this research is to evaluate fusion reactor material suitability regarding their behaviour and tritium retention in the fusion reactor relevant conditions. Methods and technique developed in the UL Institute of Chemical Physics Laboratory of Radiation Chemistry of Solid State has been used in this study. Synergetic facilitating effect of accelerated electrons and high magnetic fi...

  8. Decommissioning a tritium glove-box facility

    Energy Technology Data Exchange (ETDEWEB)

    Folkers, C.L.; Homann, S.G.; Nicolosi, A.S.; Hanel, S.L.; King, W.C.

    1979-08-08

    A large glove-box facility for handling reactive metal tritides was decommissioned. Major sections of the glove box were decontaminated and disassembled for reuse at another tritium facility. To achieve the desired results, decontamnation required repeated washing, first with organic liquids, then with water and detergents. Worker protection was provided by simple ventilation combined with careful monitoring of the work areas and employees. Several innovative techniques are described.

  9. Continuously tritium monitoring of the pipe of liquid effluents at the Cea Cadarache; Controle en continu du tritium de la conduite des effluents liquides du CEA Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    Pira, Y

    2004-07-01

    This report is oriented toward the radiation protection of environment that is an essential component of radiation protection. It is necessary to detect any solid, liquid or gaseous abnormal release and to find its origin. The present study bears on a detection instrument in continuously to find tritium in liquid effluents of the Cea Cadarache. After having study the functioning principle of this device, an evaluation of its performances has been realised ( back noise, yield, detection limit) and to a checking in real conditions of utilization. (N.C.)

  10. Tritium labeling for bio-med research

    International Nuclear Information System (INIS)

    A very large fraction of what we know about biochemical pathways in the living cell has resulted from the use of radioactively-labeled tracer compounds; the use of tritium-labeled compounds has been particularly important. As research in biochemistry and biology has progressed the need has arisen to label compounds of higher specific activity and of increasing molecular complexity - for example, oligo-nucleotides, polypeptides, hormones, enzymes. Our laboratory has gradually developed special facilities for handling tritium at the kilocurie level. These facilities have already proven extremely valuable in producing labeled compounds that are not available from commercial sources. The principal ways employed for compound labeling are: (1) microwave discharge labeling, (2) catalytic tritio-hydrogenation, (3) catalytic exchange with T2O, and (4) replacement of halogen atoms by T. Studies have also been carried out on tritiation by the replacement of halogen atoms with T atoms. These results indicate that carrier-free tritium-labeled products, including biomacromolecules, can be produced in this way

  11. Development of tritium analysis system TAS 1.0

    International Nuclear Information System (INIS)

    Tritium is one of the fuels used in fusion reactors. Design and analysis on the tritium system are one of key research for fusion reactor study. Based on the research for the some concepts of Chinese liquid metal LiPb blanket fusion reactor, a Tritium Analysis System (TAS1.0) for fusion reactor had been developed by using Software Engineering method and Object-Oriented technology for tritium self-sustaining analysis, tritium management and tritium safety analysis. In addition, TAS 1.0 can also support the design of blanket and fuel circulation system. A series of tests and applications had shown the maturity and effectiveness of' the system. This paper gives a brief overview of the design of the system, main technical features and the related tests. (authors)

  12. Removal and recovery of tritium from light and heavy water

    International Nuclear Information System (INIS)

    A method and apparatus for removing tritium from light water are described, comprising contacting tritiated feed water in a catalyst column in countercurrent flow with hydrogen gas originating from an electrolysis cell so as to enrich this feed water with tritium from the electrolytic hydrogen gas and passing the tritium enriched water to an electrolysis cell wherein the electrolytic hydrogen gas is generated and then fed upwards through the catalyst column or recovered as product. The tritium content of the hydrogen gas leaving the top of the enricher catalyst column is further reduced in a stripper column containing catalyst which transfers the tritium to a countercurrent flow of liquid water. Anodic oxygen and water vapour from the anode compartment may be fed to a drier and condensed electrolyte recycled with a slip stream or recovered as a further tritium product stream. A similar method involving heavy water is also described. (author)

  13. Early experience with the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    The Tritium Systems Test Assembly (TSTA) project at Los Alamos is charged with developing and demonstrating the tritium technology required to fuel a deuterium-tritium burning fusion reactor and to develop and evaluate the personnel and environmental safety systems associated with the tritium facility. The TSTA project completed the construction phase in late 1982 and is currently in the component checkout and early experimental phase. Tritium introduction is scheduled for mid-summer 1983. Several major systems have been operated and tested with hydrogen and deuterium. These include the vacuum pump, the isotope separation system and the emergency tritium cleanup system. The results of the early experiments are summarized and the experimental programs for other systems are presented

  14. 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)

    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)

  15. Operation of the TSTA (Tritium Systems Test Assembly) with 100 gram tritium

    International Nuclear Information System (INIS)

    In March of 1988 full operation of the 4-column isotope separation system (ISS) was realized in runs that approximated the design load of tritium. Previous operations had been fraught with operating difficulties principally due to external systems. This report will examine the recent highly successful 6-day period of operation. During this time the system was cooled from room temperature, loaded with hydrogen isotopes including 109 grams of tritium, integrated with the transfer pumping, impurity injection, and impurity removal systems, as well as the remote computer control system. At the end of the operation 12 grams of tritium having a measured purity of 99.987% (remainder deuterium) were offloaded from the system. Observed profiles in the columns in general agree with computer models. A Height Equivalent to a Theoretical Plate (HETP) of 5.0 cm is confirmed. 3 refs., 5 figs

  16. Tritium control and activation in the Pulse*Star reactor

    International Nuclear Information System (INIS)

    Pulse*Star is an inertial fusion reactor that uses LiPb coolant in a pool type geometry. LiPb does not release great quantities of chemical energy in a fire, and the pool geometry reduces the difficulty of safely transporting the extremely dense fluid. The compact geometry and good neutronics qualities of LiPb lead to a thermal-to-fusion energy ratio of 1.26, a tritium breeding ratio of 1.22, and a net electric power density 29 times higher than in a fission reactor containment building. The afterheat of the coolant and steel is low enough that emergency cooling systems will be either simple or not required. The gamma dose rate of the bell jar or screen is high enough to require remote maintenance of these components. The steam generators and pumps are on the borderline between limited hands-on and remote maintenance. With additional design attention, limited hands-on maintenance could be feasible for these components. The biological hazard potential indicates that only 10-7 to 10-6 of the reactor central region can be vaporized and released; these are values typical of other fusion reactor designs

  17. Study and Application of Foreign Gaseous Tritium Light Sources

    Institute of Scientific and Technical Information of China (English)

    DENG; Bei; LI; Si-jie; ZHANG; Li-feng; SUN; Yu-hua; HAN; Shi-quan

    2013-01-01

    Light is given out by the phosphor material which is excited by theβrays from tritium,as is the way tritium light sources work.For tritium light sources,there is no need for maintenance and additional power,and it is not affected by temperature,humidity,altitude and use technology,which makes it widely used in some special areas of national economy,just like the sight lighting of varieties of instrument panel,

  18. Recent results on tritium technology in JAEA under BA program

    International Nuclear Information System (INIS)

    Highlights: • The multi-purpose RI facility has been constructed at Rokkasho site in DEMO R and D building until 2011. • The material of the column of the micro gas chromatograph has been studied to develop a real time analysis tool for the hydrogen isotope composition in gas phase. • A set of basic data on the interaction between materials and tritium has been measured by various methods. • As a study for the tritium durability, the endurance of ion exchange membrane has been tested by using high concentration tritium water. -- Abstract: The multi-purpose RI facility has been constructed at Rokkasho site in DEMO R and D building until 2011. The facility is the first and quite unique facility in Japan, where tritium, beta and gamma RI species, and beryllium (Be) can simultaneously be used. The amounts of tritium used and stored are 3.7 TBq per glove box and 7.4 TBq, respectively. Some tritium water samples of 38 GBq has been stored at the equipment on March 2012. The material of the column of the micro gas chromatograph has been studied to develop a real time analysis tool for the hydrogen isotope composition in gas phase. The calorimeter has also been studied as a possible tritium measurement method in solid waste. A set of basic data on the interaction between materials and tritium has been measured by various methods. The behavior of tritium in Fe and W has been studied as a typical subject. As a study for the tritium durability, the endurance of ion exchange membrane has been tested by using high concentration tritium water. The curves of strength vs. dose for the Nafion membranes in tritium water were well consistent with those by gamma rays and electron beams irradiations

  19. Five years of tritium handling experience at the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory is a facility designed to develop and demonstrate, in full scale, technologies necessary for safe and efficient operation of tritium systems required for tokamak fusion reactors. TSTA currently consists of systems for evacuating reactor exhaust gas with compound cryopumps; for removing impurities from plasma exhaust gas and recovering the chemically-combined tritium; for separating the isotopes of hydrogen; for transfer pumping; or storage of hydrogen isotopes; for gas analysis; and for assuring safety by the necessary control, monitoring, and tritium removal from effluent streams. TSTA also has several small scale experiments to develop and test new equipment and processes necessary for fusion reactors. In this paper, data on component reliability, failure types and rates, and waste quantities are presented. TSTA has developed a Quality Assurance program for preparing and controlling the documentation of the procedures required for the design, purchase, and operation of the tritium systems. Operational experience under normal, abnormal, and emergency conditions is presented. One unique aspect of operations at TSTA is that the design personnel for the TSTA systems are also part of the operating personnel. This has allowed for the relatively smooth transition from design to operations. TSTA has been operated initially as a research facility. As the system is better defined, operations are proceeding toward production modes. The DOE requirements for the operation of a tritium facility like TSTA include personnel training, emergency preparedness, radiation protection, safety analysis, and preoperational appraisals. The integration of these requirements into TSTA operations is discussed. 4 refs., 3 figs., 3 tabs

  20. The synthesis of tritium-labelled cyclic hydrocarbons by using tritium recoil nuclei

    International Nuclear Information System (INIS)

    The authors discuss the results of investigating the interaction of tritium recoil atoms produced by the reaction Li6 (n, α)T with cyclohexane, cyclohexene, cyclohexadiene, methyl cyclohexane, cyclohexanol, cyclohexylammine and benzene. Mixtures of these compounds with lithium carbonate were neutron-irradiated. From 1 g of lithium, 4 mc/h of tritium was obtained with a 4 x 1012 n/cm2 s neutron flux. The total yield of the products depends on the amount of tritium yielded by the crystals, and, so, on the irradiation conditions. The yield from the separate components is determined by analysis. The irradiation products were analysed by vacuum distillation, using carriers and gas-liquid chromatography. The results obtained show that 20-40% of the tritium yielded by the lithium carbonate crystals is embedded in the parent molecule of the irradiated compound. When, for instance, cyclohexene is irradiated together with 22% of the labelled parent-compound, 16% cyclohexane, 4% methyl cyclopentane and small amounts of other products are obtained. The specific activity of cyclohexane and methyl cyclopentane separated on a chromatographic column may be high, and the only dilution is with products of radiolysis. When other compounds are irradiated, there is a good yield only from the irradiated parent-compound, and a small yield from other products. For purposes of preparation, cyclohexane and methyl cyclopentane are best obtained by irradiating cyclohexane; other cyclic hydrocarbons can be obtained by irradiating the compounds directly with lithium salts. The paper describes a preparation column for separating tritium-labelled cyclohexane, cyclohexene and methyl cyclopentane from irradiated cyclohexene and for separating the products yielded by the reaction of tritium recoil atoms with other cyclic hydrocarbons. (author)

  1. Effect of tritium (tritium water) on prenatal and postnatal development of rats

    Energy Technology Data Exchange (ETDEWEB)

    Bajrakova, A.; Baev, I.; Yagova, A. (Meditsinska Akademiya, Sofia (Bulgaria). Nauchen Inst. po Rentgenologiya i Radiobiologiya)

    1983-01-01

    Female rats were injected intraperitoneally on the first day after their fecundation with 3,7 kBq/g b.w. tritium water - activity which under these conditions does not increase prenatal death rate. The postnatal development of the born alive was traced in respect to the lethality rate and growth rate (mean bodily weight in dynamics up to the 60-th day p.p.) and compared with that of the offsprings from the control group. It was shown that the used activity tritium water during the initial stages of embryonic development does not result in deviations from the norm.

  2. Effect of tritium (tritium water) on prenatal and postnatal development of rats

    International Nuclear Information System (INIS)

    Female rats were injected intraperitoneally on the first day after their fecundation with 3,7 kBq/g b.w. tritium water - activity which under these conditions does not increase prenatal death rate. The postnatal development of the born alive was traced in respect to the lethality rate and growth rate (mean bodily weight in dynamics up to the 60-th day p.p.) and compared with that of the offsprings from the control group. It was shown that the used activity tritium water during the initial stages of embryonic development does not result in deviations from the norm. (authors)

  3. Overview of tritium systems for the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    The Compact Ignition Tokamak (CIT) is being designed at several laboratories to produce and study fully ignited plasma discharges. The tritium systems which will be needed for CIT include fueling systems and radiation monitoring and safety systems. Design of the tritium systems is the responsibility of the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. Major new tritium systems for CIT include a pellet injector, an air detritiation system and a glovebox atmosphere detritiation system. The pellet injector is being developed at Oak Ridge National Laboratory. 7 refs., 2 figs

  4. Concentration of tritium in precipitation and river water

    International Nuclear Information System (INIS)

    The concentration of tritium in precipitation and river water has been measured sice 1973 in Aichi, Japan. The tritium in water samples was enriched by electrolysis, and measured by liquid scintillation counting. The concentration of tritium in precipitation decreased from 27 TU in 1973 to 17 TU in 1979, and showed seasonal variation. During this period, there was a rise of concentration because of Chinese nuclear detonation. The concentration of tritium in river water gradually decreased from 44 TU in 1973 to 24 TU in 1979, and the seasonal variation was not observed. Based on the observed values, the relation among precipitation, river water and ground water was analyzed. (J.P.N.)

  5. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  6. Tritium Formation and Mitigation in High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-10-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  7. TRITIUM BARRIER MATERIALS AND SEPARATION SYSTEMS FOR THE NGNP

    Energy Technology Data Exchange (ETDEWEB)

    Sherman, S; Thad Adams, T

    2008-07-17

    Contamination of downstream hydrogen production plants or other users of high-temperature heat is a concern of the Next Generation Nuclear Plant (NGNP) Project. Due to the high operating temperatures of the NGNP (850-900 C outlet temperature), tritium produced in the nuclear reactor can permeate through heat exchangers to reach the hydrogen production plant, where it can become incorporated into process chemicals or the hydrogen product. The concentration limit for tritium in the hydrogen product has not been established, but it is expected that any future limit on tritium concentration will be no higher than the air and water effluent limits established by the NRC and the EPA. A literature survey of tritium permeation barriers, capture systems, and mitigation measures is presented and technologies are identified that may reduce the movement of tritium to the downstream plant. Among tritium permeation barriers, oxide layers produced in-situ may provide the most suitable barriers, though it may be possible to use aluminized surfaces also. For tritium capture systems, the use of getters is recommended, and high-temperature hydride forming materials such as Ti, Zr, and Y are suggested. Tritium may also be converted to HTO in order to capture it on molecular sieves or getter materials. Counter-flow of hydrogen may reduce the flux of tritium through heat exchangers. Recommendations for research and development work are provided.

  8. Occurrence of organically bound tritium in the Mohelno lake system

    International Nuclear Information System (INIS)

    This study was focused on the 3Hactivitylevels in the unique 'tritium valley' around the Mohelno reservoir, which receives outlet cooling waters from the Dukovany nuclear power plant. Tritium activity levels above the background reference value were found in water from the reservoir and from the effluent part of the Jihlava water, in air moisture and in plant tissues tissue free water tritium(TFWT), and nonexchangeable organically bound tritium (NE-OBT). These zones were discernible that had noticeably different TFWT and NE-OBT values: (1) littoral zones, (2) slopes above the reservoir, (3) plateaus above the reservoir/river. (author)

  9. Comparison and Evaluation of Various Tritium Decontamination Techniques and Processes

    International Nuclear Information System (INIS)

    In support of fusion energy development, various techniques and processes have been developed over the past two decades for the removal and decontamination of tritium from a variety of items, surfaces, and components. Tritium decontamination, by chemical, physical, mechanical, or a combination of these methods, is driven by two underlying motivational forces. The first of these motivational forces is safety. Safety is paramount to the established culture associated with fusion energy. The second of these motivational forces is cost. In all aspects, less tritium contamination equals lower operational and disposal costs. This paper will discuss and evaluate the various processes employed for tritium removal and decontamination

  10. Analysis of residual tritium in an LP 50 product container

    Energy Technology Data Exchange (ETDEWEB)

    Wermer, J.R.

    1996-06-04

    The analysis was done by sampling coupons cut from the side of the vessel. Tests were performed to analyze the amount of residual tritium in the container wall, as well as the amount of tritium removed through exposure to moist air. Based on this data, the PC contained 62 curies of residual tritium. Air exposure and leaching of the coupons in aqua regia accounted for 27 curies. Recommendations are given for final processing of these containers in order to reduce the final tritium content.

  11. Treatment of tritiated exhaust gases at the Tritium Laboratory Karlsruhe

    Energy Technology Data Exchange (ETDEWEB)

    Hutter, E.; Besserer, U. [Kernforschungszentrum Karlsruhe GmbH (Germany); Jacqmin, G. [NUKEM GmbH, Industreistr, Alzenau (Germany)

    1995-02-01

    The Tritium Laboratory Karlsruhe (TLK) accomplished commissioning; tritium involving activities will start this year. The laboratory is destined mainly to investigating processing of fusion reactor fuel and to developing analytic devices for determination of tritium and tritiated species in view of control and accountancy requirements. The area for experimental work in the laboratory is about 800 m{sup 2}. The tritium infrastructure including systems for tritium storage, transfer within the laboratory and processing by cleanup and isotope separation methods has been installed on an additional 400 m{sup 2} area. All tritium processing systems (=primary systems), either of the tritium infrastructure or of the experiments, are enclosed in secondary containments which consist of gloveboxes, each of them connected to the central depressurization system, a part integrated in the central detritiation system. The atmosphere of each glovebox is cleaned in a closed cycle by local detritiation units controlled by two tritium monitors. Additionally, the TLK is equipped with a central detritiation system in which all gases discharged from the primary systems and the secondary systems are processed. All detritiation units consist of a catalyst for oxidizing gaseous tritium or tritiated hydrocarbons to water, a heat exchanger for cooling the catalyst reactor exhaust gas to room temperature, and a molecular sieve bed for adsorbing the water. Experiments with tracer amounts of tritium have shown that decontamination factors >3000 can be achieved with the TLK detritiation units. The central detritiation system was carefully tested and adjusted under normal and abnormal operation conditions. Test results and the behavior of the tritium barrier preventing tritiated exhaust gases from escaping into the atmosphere will be reported.

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

    International Nuclear Information System (INIS)

    Although a considerable volume of valuable data on the behavior of tritium in beryllium and carbon-based armors 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 being complete and tritium inventory predictions for these materials remain highly uncertain. In this paper the authors 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 database for design purposes. Due to their strong anticipated implications on the design, attention is focused mainly on codeposition and neutron damage effects. Some preliminary quantitative estimates are presented based on most recent experimental findings and latest modeling 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

  13. Evaluation of acute cardiovascular effects of immediate-release methylphenidate in children and adolescents with attention-deficit hyperactivity disorder

    Directory of Open Access Journals (Sweden)

    Lamberti M

    2015-05-01

    Full Text Available Marco Lamberti,1,2 Domenico Italiano,2 Laura Guerriero,1 Gessica D’Amico,3 Rosamaria Siracusano,1,4 Massimo Ingrassia,5 Eva Germanò,1 Maria Pia Calabrò,3 Edoardo Spina,2 Antonella Gagliano1 1Division of Child Neurology and Psychiatry, Department of Pediatrics, 2Department of Clinical and Experimental Medicine, 3Division of Pediatric Cardiology, Department of Pediatrics, University of Messina, Messina, Italy; 4Institution of Clinical Physiology, CNR, Pisa, 5Division of Psychology, Department of Humanities and Social Sciences, University of Messina, Messina, Italy Abstract: Attention-deficit hyperactivity disorder is a frequent condition in children and often extends into adulthood. Use of immediate-release methylphenidate (MPH has raised concerns about potential cardiovascular adverse effects within a few hours after administration. This study was carried out to investigate acute effects of MPH on electrocardiogram (ECG in a pediatric population. A total of 54 consecutive patients with attention-deficit hyperactivity disorder (51 males and 3 females; mean age =12.14±2.6 years, range 6–19 years, receiving a new prescription of MPH, underwent a standard ECG 2 hours before and after the administration of MPH 10 mg per os. Basal and posttreatment ECG parameters, including mean QT (QT interval when corrected for heart rate [QTc], QTc dispersion (QTd interval duration, T-peak to T-end (TpTe intervals, and TpTe/QT ratio were compared. Significant modifications of both QTc and QTd values were not found after drug administration. QTd fluctuated slightly from 25.7±9.3 milliseconds to 25.1±8.4 milliseconds; QTc varied from 407.6±12.4 milliseconds to 409.8±12.7 milliseconds. A significant variation in blood pressure (systolic blood pressure 105.4±10.3 vs 109.6±11.5; P<0.05; diastolic blood pressure 59.2±7.1 vs 63.1±7.9; P<0.05 was observed, but all the data were within normal range. Heart rate moved from 80.5±15.5 bpm to 87.7±18.8

  14. Synthesis of tritium labelled 24-epibrassinolide

    Energy Technology Data Exchange (ETDEWEB)

    Kolbe, A.; Marquardt, V.; Adam, G. (Inst. of Plant Biochemistry Halle, Halle/Saale (Germany))

    1992-10-01

    Deuterium and tritium 5,7,7-tris-labelled 24-epibrassinolide were prepared by base catalyzed exchange reaction using 24-epicastasterone tetraacetate 1 or bis-isopropylidenedioxy-24-epicastasterone 8 and labelled water. Baeyer-Villiger oxidation of the obtained labelled 6-ketones 2 and 3 with CF[sub 3]CO[sub 3]H gave after alkaline deacetylation of the resulting 4 and 5 the desired tris-labelled 24-epibrassinolides 6 and 7, respectively, or starting from 9 under simultaneous oxidation and deprotection in one step the same final products. (author).

  15. High accuracy tritium measurement for the verification of the tritium production rate calculations with MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Rovni, István, E-mail: rovni@reak.bme.hu [Budapest University of Technology and Economics (BME), Institute of Nuclear Techniques, 1111. Budapest, Műegyetem rkp 3-9 (Hungary); Szieberth, Máté [Budapest University of Technology and Economics (BME), Institute of Nuclear Techniques, 1111. Budapest, Műegyetem rkp 3-9 (Hungary); Palcsu, László; Major, Zoltán [Hertelendi Laboratory of Environmental Studies, 4026 Debrecen, Bem tér 18/C (Hungary); Fehér, Sándor [Budapest University of Technology and Economics (BME), Institute of Nuclear Techniques, 1111. Budapest, Műegyetem rkp 3-9 (Hungary)

    2013-06-21

    This paper presents high accuracy tritium production rate measurement results compared with calculations using the MCNPX Monte Carlo particle transport code. The experimental results are regarded as reference values for a new passive technique based on the secondary charged particle activation method developed for measuring the tritium production rate in the test blanket modules of the ITER Tokamak. The {sup 16}O(t,n){sup 18}F reaction, which is one of the possible tritium monitor reactions, was also extensively investigated, and the experimentally determined reaction rates were compared with simulations. Li{sub 2}CO{sub 3} solution was filled and sealed into quartz ampoules which were irradiated in the Training Reactor of the Budapest University of Technology and Economics. The amount of {sup 18}F was determined using γ-spectroscopy. Then the precise tritium measurements were carried out in the Hertelendi Laboratory of Environmental Studies using the {sup 3}H–{sup 3}He ingrowth method, where the {sup 3}He produced during the storage time is measured by a static noble gas mass spectrometer (VG-5400). The HT/HTO ratio in the irradiated aqueous solutions was found to be 0.1323±0.0034. Based on the comparison of the measurements and the simulations it was pointed out that the model calculations underestimate the reaction rate of both the {sup 6}Li(n,t)α and the {sup 16}O(t,n){sup 18}F reactions by 5–10% and 15%, respectively. -- Highlights: ► Tritium measurements for verifying the {sup 6}Li6(n, t)α reaction rate calculated by MCNPX. ► The HT/HTO ratio was determined in the neutron irradiated aqueous solution of Li{sub 2}CO{sub 3}. ► The reaction rate of {sup 16}O(t,n){sup 18}F was measured in thermal neutron spectrum.

  16. Development and Verification of Behavior of Tritium Analytic Code (BOTANIC)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Min Young; Kim, Eung Soo [Seoul National University, Seoul (Korea, Republic of)

    2014-10-15

    VHTR, one of the Generation IV reactor concepts, has a relatively high operation temperature and is usually suggested as a heat source for many industrial processes, including hydrogen production process. Thus, it is vital to trace tritium behavior in the VHTR system and the potential permeation rate to the industrial process. In other words, tritium is a crucial issue in terms of safety in the fission reactor system. Therefore, it is necessary to understand the behavior of tritium and the development of the tool to enable this is vital.. In this study, a Behavior of Tritium Analytic Code (BOTANIC) an analytic tool which is capable of analyzing tritium behavior is developed using a chemical process code called gPROMS. BOTANIC was then further verified using the analytic solutions and benchmark codes such as Tritium Permeation Analysis Code (TPAC) and COMSOL. In this study, the Behavior of Tritium Analytic Code, BOTANIC, has been developed using a chemical process code called gPROMS. The code has several distinctive features including non-diluted assumption, flexible applications and adoption of distributed permeation model. Due to these features, BOTANIC has the capability to analyze a wide range of tritium level systems and has a higher accuracy as it has the capacity to solve distributed models. BOTANIC was successfully developed and verified using analytical solution and the benchmark code calculation result. The results showed very good agreement with the analytical solutions and the calculation results of TPAC and COMSOL. Future work will be focused on the total system verification.

  17. Influence of neutron irradiation on the tritium retention in beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Rolli, R.; Ruebel, S.; Werle, H. [Forschungszentrum Karlsruhe, Inst. fuer Neutronenphysik und Reaktortechnik, Karlsruhe (Germany); Wu, C.H.

    1998-01-01

    Carbon-based materials and beryllium are the candidates for protective layers on the components of fusion reactors facing plasma. In contact with D-T plasma, these materials absorb tritium, and it is anticipated that tritium retention increases with the neutron damage due to neutron-induced traps. Because of the poor data base for beryllium, the work was concentrated on it. Tritium was loaded into the samples from stagnant T{sub 2}/H{sub 2} atmosphere, and afterwards, the quantity of the loaded tritium was determined by purged thermal annealing. The specification of the samples is shown. The samples were analyzed by SEM before and after irradiation. The loading and the annealing equipments are contained in two different glove boxes with N{sub 2} inert atmosphere. The methods of loading and annealing are explained. The separation of neutron-produced and loaded tritium and the determination of loaded tritium in irradiated samples are reported. Also the determination of loaded tritium in unirradiated samples is reported. It is evident that irradiated samples contained much more loaded tritium than unirradiated samples. The main results of this investigation are summarized in the table. (K.I.)

  18. Study of tritium labeling of complex mixture of polychlorinated biphenyls

    International Nuclear Information System (INIS)

    The method for tritium labeling of technical mixture (commercial mark - SOVOL, USSR) of polychlorinated biphenyls (PCBs) was developed. The influence of procedure of labeling by thermally activated tritium on the nativity of polychlorinated biphenyls (PCBs) was studied. The method of labeling by thermally activated tritium has some factors, which are able to destroy organic compounds - photodegradation, thermo-degradation and degradation caused by reaction of substitution of organic compounds hydrogen atoms by activated tritium atoms. To develop a method of labeling of every organic compound by thermally activated tritium it is necessary to determine and optimize the conditions of labeling. In our case procedure of labeling is complicated because of technical mixture of PCBs consist from more than 20 isomers, chlorinated with different degree. We studied the dependence of appearance of products of degradation PCBs from duration of labeling procedure. It was found that some part of PCBs and product of its degradation were evaporated under vacuum and were collected on the glass flask cooled by liquid nitrogen. It was found that correlation between labeled of PCBs and products of degradation did not changed with increasing time of labeling, and radiochemical yield of tritium labeled of PCBs was stable - about 15-20 %. The optimum regime of labeling was selected. It was found that purification of labeled PCBs by TLC on silica gel with hexane allows obtaining tritium-labeled PCBs purified from by-products. Thus, TLC purification seems inexpensive, fast and suitable for purification of tritium-labeled PCBs

  19. A Gas Target with a Tritium Gas Handling System

    International Nuclear Information System (INIS)

    A detailed description is given of a simple tritium gas target and its tritium gas filling system, and how to put it into operation. By using the T (p,n) He reaction the gas target has been employed for production of monoenergetic fast neutrons of well defined energy and high intensity. The target has been operated successfully for a long time

  20. Tritium Content of Rainwater from the Eastern Mediterranean Area

    International Nuclear Information System (INIS)

    About 50 samples of rainwater collected during the years 1958-1960 in Israel and neighbouring countries were assayed for their tritium content by gas counting following electrolytic enrichment. The samples included single showers collected at two stations in Israel and one in Cyprus, as well as composite samples accumulated during each rainy season at a number of sites in Israel, Cyprus, Turkey and Greece. In addition samples, of cistern water, representative of rain from the 1956/57 and 1957/58 rainy seasons, were also analyzed. From the air circulation pattern and the timing of thermonuclear tests relative to the local rainy season it is inferred that little direct tropospheric transport of tritium from the test sites into the area occurs. The measured tritium levels hence are due to tritium leakage from high altitudes into lower air layers. Mean atmospheric residence times are estimated for stratospheric tritium from different sources. There is evidence that tritium is only slowly mixed throughout the stratosphere. Systematic differences between the tritium levels at various sites are explained in terms of the different rain producing situations. It is shown that the extent of mixing of maritime and continental air masses are of paramount importance for the resulting tritium content of rain and that the dimen- sions of the Mediterranean sea are small compared to the scale of meteorological phenomena involved. (author)

  1. Final programmatic environmental impact statement for tritium supply and recycling

    International Nuclear Information System (INIS)

    Tritium, a radioactive gas used in all of the Nation's nuclear weapons, has a short half-life and must be replaced periodically in order for the weapon to operate as designed. Currently, there is no capability to produce the required amounts of tritium within the Nuclear Weapons Complex. The PEIS for Tritium Supply and Recycling evaluates the alternatives for the siting, construction, and operation of tritium supply and recycling facilities at each of five candidate sites: the Idaho National Engineering Laboratory, the Nevada Test Site, the Oak Ridge Reservation, the Pantex Plant, and the Savannah River Site. Alternatives for new tritium supply and recycling facilities consist of four different tritium supply technologies: Heavy Water Reactor, Modular High Temperature Gas-Cooled Reactor, Advanced Light Water Reactor, and Accelerator Production of Tritium. The PEIS also evaluates the impacts of the DOE purchase of an existing operating or partially completed commercial light water reactor or the DOE purchase of irradiation services contracted from commercial power reactors. Additionally, the PEIS includes an analysis of multipurpose reactors that would produce tritium, dispose of plutonium, and produce electricity. Evaluation of impacts on land resources, site infrastructure, air quality and acoustics, water resources, geology and soils, biotic resources, cultural and paleontological resources, socioeconomics, radiological and hazardous chemical impacts during normal operation and accidents to workers and the public, waste management, and intersite transport are included in the assessment

  2. Tritium measurement technique using ''in-bed'' calorimetry

    International Nuclear Information System (INIS)

    One of the new technologies that has been introduced to the Savannah River Site (SRS) is the production scale use of metal hydride technology to store, pump, and compress hydrogen isotopes. For tritium stored in metal hydride storage beds, a unique relationship does not exist between the amount of tritium in the bed and the pressure-volume-temperature properties of the hydride material. Determining the amount of tritium in a hydride bed after desorbing the contents of the bed to a tank and performing pressure, volume, temperature, and composition (PVTC) measurements is not practical due to long desorption/absorption times and the inability to remove tritium ''heels'' from the metal hydride materials under normal processing conditions. To eliminate the need to remove tritium from hydride storage beds for measurement purposes, and ''in-bed'' tritium calorimetric measurement technique has been developed. The steady-state temperature rise of a gas stream flowing through a jacketed metal hydride storage bed is measured and correlated with power input to electric heaters used to simulate the radiolytic power generated by the decay of tritium to 3He. Temperature rise results for prototype metal hydride storage beds and the effects of using different gases in the bed are shown. Linear regression results shows that for 95% confidence intervals, temperature rise measurements can be obtained in 14 hours and have an accuracy of ±1.6% of a tritium filled hydride storage bed

  3. Risks involved in tritium compounds handling in the laboratory

    International Nuclear Information System (INIS)

    More and more laboratories are using tritium and its compounds of varied activities and in very different conditions. Whatever the importance of handled activity may be, we come up against complex radioprotection problems specific to tritium compounds. This paper is an attempt to give a general idea of the main difficulties encountered and the method used to overcome them

  4. Effectiveness Monitoring Report, MWMF Tritium Phytoremediation Interim Measures.

    Energy Technology Data Exchange (ETDEWEB)

    Hitchcock, Dan; Blake, John, I.

    2003-02-10

    This report describes and presents the results of monitoring activities during irrigation operations for the calendar year 2001 of the MWMF Interim Measures Tritium Phytoremediation Project. The purpose of this effectiveness monitoring report is to provide the information on instrument performance, analysis of CY2001 measurements, and critical relationships needed to manage irrigation operations, estimate efficiency and validate the water and tritium balance model.

  5. Fast Tritium Separation From the Low Level Radioactive Liquid Waste

    Institute of Scientific and Technical Information of China (English)

    LIANG; Xiao-hu; YANG; Su-liang; YANG; Lei; YANG; Jin-ling

    2012-01-01

    <正>Due to the needed of high efficiency monitoring and controlling of the waste water generated from the spent fuel reprocessing process, analyzing work need to be done quickly. Tritium is an important nuclide in the liquid waste and its content must be determined. But the existing tritium analysis method

  6. Key processes and input parameters for environmental tritium models

    International Nuclear Information System (INIS)

    The primary objective of the work reported here is to define key processes and input parameters for mathematical models of environmental tritium behaviour adequate for use in safety analysis and licensing of fusion devices like NET and associated tritium handling facilities. (author). 45 refs., 3 figs

  7. Confinement and heating of a deuterium-tritium plasma

    International Nuclear Information System (INIS)

    The Tokamak Fusion Test Reactor (TFTR) has performed initial high-power experiments with the plasma fueled by deuterium and tritium to nominally equal densities. Compared to pure deuterium plasmas, the energy stored in the electron and ions increased by ∼20%. These increases indicate improvements in confinement associated with the use of tritium and possibly heating of electrons by α-particles

  8. Design and construction of the Tritium Systems Test Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, J.L.

    1980-01-01

    The objective of TSTA is to develop those aspects of tritium technology related to the fuel cycle for fusion power reactors and to develop the environmental and personnel safety systems required for such a tritium facility. The TSTA schedule calls for construction to be completed and the facility to be operational by the end of FY-1981. The project is now somewhat more than halfway through the design-construction phase and is currently on schedule for the 1981 operational milestone. In this paper the current status of the major subsystems will be discussed. The subsystems to be discussed include the: Vacuum Facility; Fuel Cleanup; Isotope Separation; Transfer Pump; Emergency Tritium Cleanup; Tritium Waste Treatment; Tritium Monitoring; Secondary Containment; and, the Master Data Acquisition and Control System.

  9. The tritium systems test assembly: Overview and recent results

    International Nuclear Information System (INIS)

    The fusion technology development program for tritium in the US is centered around the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The TSTA is a full-scale system of reactor exhaust gas reprocessing for an ITER-sized machine. That is, TSTA has the capacity to process tritium in a closed loop mode at the rate of 1 kg per day, requiring a tritium inventory of about 100 g. The TSTA program also interacts with all other tritium-related fusion technology programs in the US and all major programs abroad. This report summarizes the current status, results and interactions of the TSTA. Special emphasis is given to operations in May/June using large compound cryopumps that completed the fuel loop integration of all TSTA subsystems for the first time. 6 refs., 2 figs

  10. An approach to the proposal of an annual limit of intake for the general public (ALIgp) for organically bound tritium

    International Nuclear Information System (INIS)

    Values for ALIs for the general public for adults, newborn, 1 and 10 year old children are proposed. These are based on a simple two component model in which all the ingested tritium is assumed to be absorbed and 50% enters carbon-hydrogen bonds from which it is released with the biological half-time of carbon; the remaining 50% is assumed to equilibrate rapidly with the total body water pool. (author)

  11. Detection of tritium sorption on four soil materials

    International Nuclear Information System (INIS)

    In order to measure groundwater age and design nuclear waste disposal sites, it is important to understand the sorption behavior of tritium on soils. In this study, batch tests were carried out using four soils from China: silty clays from An County and Jiangyou County in Sichuan Province, both of which could be considered candidate sites for Very Low Level Waste disposal; silty sand from Beijing; and loess from Yuci County in Shanxi Province, a typical Chinese loess region. The experimental results indicated that in these soil media, the distribution coefficient of tritium is slightly influenced by adsorption time, water/solid ratio, initial tritium specific activity, pH, and the content of humic and fulvic acids. The average distribution coefficient from all of these influencing factors was about 0.1-0.2 mL/g for the four types of soil samples. This relatively modest sorption of tritium in soils needs to be considered in fate and transport studies of tritium in the environment. - Research highlights: → In this study, batch sorption tests validate the adsorption of tritium on all of the four tested soil samples collected in China, and the distribution coefficient is found to be non-zero and less than 0.4 mL/g. The experimental results indicated that in these soil media, the distribution coefficient of tritium is slightly influenced by adsorption time, water/solid ratio, initial tritium specific activity, pH, and the content of humic and fulvic acids. This relatively modest sorption of tritium in soils needs to be considered in fate and transport studies of tritium in the environment.

  12. Variations in environmental tritium doses due to meteorological data averaging and uncertainties in pathway model parameters

    Energy Technology Data Exchange (ETDEWEB)

    Kock, A.

    1996-05-01

    The objectives of this research are: (1) to calculate and compare off site doses from atmospheric tritium releases at the Savannah River Site using monthly versus 5 year meteorological data and annual source terms, including additional seasonal and site specific parameters not included in present annual assessments; and (2) to calculate the range of the above dose estimates based on distributions in model parameters given by uncertainty estimates found in the literature. Consideration will be given to the sensitivity of parameters given in former studies.

  13. Variations in environmental tritium doses due to meteorological data averaging and uncertainties in pathway model parameters

    International Nuclear Information System (INIS)

    The objectives of this research are: (1) to calculate and compare off site doses from atmospheric tritium releases at the Savannah River Site using monthly versus 5 year meteorological data and annual source terms, including additional seasonal and site specific parameters not included in present annual assessments; and (2) to calculate the range of the above dose estimates based on distributions in model parameters given by uncertainty estimates found in the literature. Consideration will be given to the sensitivity of parameters given in former studies

  14. Tritium method oil consumption and its relation to oil film thicknesses in a production diesel engine

    OpenAIRE

    Hartman, Richard M.

    1990-01-01

    CIVINS Approved for public release ; distribution is unlimited Oil consumption was measured in a modern production diesel engine using tritium as a radiotracer. The measurements were made primarily at two speeds and one load using first a single-grade lubricant and then a multi-grade lubricant. These values were then compared to oil flow rates up/down the liner which were based on film thickness traces of a sister engine under the same loads and speeds. The traces were obtained using th...

  15. Organically bound tritium analysis in environmental samples

    Energy Technology Data Exchange (ETDEWEB)

    Baglan, N. [CEA/DAM/DIF, Arpajon (France); Kim, S.B. [AECL, Chalk River Laboratories, Chalk River, ON (Canada); Cossonnet, C. [IRSN/PRP-ENV/STEME/LMRE, Orsay (France); Croudace, I.W.; Warwick, P.E. [GAU-Radioanalytical, University of Southampton, Southampton (United Kingdom); Fournier, M. [IRSN/DG/DMQ, Fontenay-aux-Roses (France); Galeriu, D. [IFIN-HH, Horia-Hulubei, Inst. Phys. and Nucl. Eng., Bucharest (Romania); Momoshima, N. [Kyushu University, Radioisotope Ctr., Fukuoka (Japan); Ansoborlo, E. [CEA/DEN/DRCP/CETAMA, Bagnols-sur-Ceze (France)

    2015-03-15

    Organically bound tritium (OBT) has become of increased interest within the last decade, with a focus on its behaviour and also its analysis, which are important to assess tritium distribution in the environment. In contrast, there are no certified reference materials and no standard analytical method through the international organization related to OBT. In order to resolve this issue, an OBT international working group was created in May 2012. Over 20 labs from around the world participated and submitted their results for the first intercomparison exercise results on potato (Sep 2013). The samples, specially-prepared potatoes, were provided in March 2013 to each participant. Technical information and results from this first exercise are discussed here for all the labs which have realised the five replicates necessary to allow a reliable statistical treatment. The results are encouraging as the increased number of participating labs did not degrade the observed dispersion of the results for a similar activity level. Therefore, the results do not seem to depend on the analytical procedure used. From this work an optimised procedure can start to be developed to deal with OBT analysis and will guide subsequent planned OBT trials by the international group.

  16. Derivation of dose conversion factors for tritium

    Energy Technology Data Exchange (ETDEWEB)

    Killough, G. G.

    1982-03-01

    For a given intake mode (ingestion, inhalation, absorption through the skin), a dose conversion factor (DCF) is the committed dose equivalent to a specified organ of an individual per unit intake of a radionuclide. One also may consider the effective dose commitment per unit intake, which is a weighted average of organ-specific DCFs, with weights proportional to risks associated with stochastic radiation-induced fatal health effects, as defined by Publication 26 of the International Commission on Radiological Protection (ICRP). This report derives and tabulates organ-specific dose conversion factors and the effective dose commitment per unit intake of tritium. These factors are based on a steady-state model of hydrogen in the tissues of ICRP's Reference Man (ICRP Publication 23) and equilibrium of specific activities between body water and other tissues. The results differ by 27 to 33% from the estimate on which ICRP Publication 30 recommendations are based. The report also examines a dynamic model of tritium retention in body water, mineral bone, and two compartments representing organically-bound hydrogen. This model is compared with data from human subjects who were observed for extended periods. The manner of combining the dose conversion factors with measured or model-predicted levels of contamination in man's exposure media (air, drinking water, soil moisture) to estimate dose rate to an individual is briefly discussed.

  17. Current status of tritium calorimetry at TLK

    International Nuclear Information System (INIS)

    Inside a tritium facility, calorimetry is an important analytical method as it is the only reference method for accountancy (it is based on the measurement of the heat generated by the radioactive decay). Presently, at Tritium Laboratory Karlsruhe (TLK), 4 calorimeters are in operation, one of isothermal type and three of inertial guidance control type (IGC). The volume of the calorimeters varies between 0.5 and 20.6 liters. About two years ago we started an extensive work to improve our calorimeters with regard to reliability and precision. We were forced to upgrade 3 of our 4 calorimeters due to the outdated interfaces and software. This work involved creating new LabView programs driving the devices, re-tuning control loops and replacing obsolete hardware components. In this paper we give a review on the current performance of our calorimeters, comparing it to recently available devices from the market and in the literature. We also show some ideas for a next generation calorimeter based on experiences with our IGC calorimeters and other devices reported in the literature. (authors)

  18. Preliminary Experimental Results for Tritium Accountancy Measurement

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Do Yeon; Chung, Hong Suk; Chung, Dong You; Koo, Dae Seo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The SDS (storage and delivery system) is one of the major components of ITER fuel cycle. The main function of the SDS is to store the hydrogen isotopes and deliver them to the fuel injection system. The tritium inventory of the bed is determined from the decay heat of the tritium without removing the inventory from bed. The decay heat is measured by the in-bed calorimetry. He through the ZrCo bed and measuring the resultant temperature increase of the He flow. Korea has been various test results for the experimental ZrCo beds. Based on this result, we propose concept of tray type ZrCo bed. ZrCo was reacted with the hydrogen ingressed through SUS filter(120mesh) placed in the tray. The heating coils and the helium loop for the inbed calorimetry are installed bottom of the tray. In this paper, we performed thermo analysis on the in-bed calorimetry performance of the bed. Using the software, LABVIEW, the time-dependent temperature distribution of the bed, the temperature difference ({Delta} T) between the inlet and outlet of the flow through the helium loop

  19. Current status of tritium calorimetry at TLK

    Energy Technology Data Exchange (ETDEWEB)

    Buekki-Deme, A.; Alecu, C.G.; Kloppe, B.; Bornschein, B. [Institute of Technical Physics, Tritium Laboratory Karsruhe - TLK, Karlsruhe Institute of Technology - KIT, Karlsruhe (Germany)

    2015-03-15

    Inside a tritium facility, calorimetry is an important analytical method as it is the only reference method for accountancy (it is based on the measurement of the heat generated by the radioactive decay). Presently, at Tritium Laboratory Karlsruhe (TLK), 4 calorimeters are in operation, one of isothermal type and three of inertial guidance control type (IGC). The volume of the calorimeters varies between 0.5 and 20.6 liters. About two years ago we started an extensive work to improve our calorimeters with regard to reliability and precision. We were forced to upgrade 3 of our 4 calorimeters due to the outdated interfaces and software. This work involved creating new LabView programs driving the devices, re-tuning control loops and replacing obsolete hardware components. In this paper we give a review on the current performance of our calorimeters, comparing it to recently available devices from the market and in the literature. We also show some ideas for a next generation calorimeter based on experiences with our IGC calorimeters and other devices reported in the literature. (authors)

  20. Gaseous Tritium Light Sources in armament and watches industries; Tritium-Gas-Lichtquellen in der Ruestungs- und Uhrenindustrie

    Energy Technology Data Exchange (ETDEWEB)

    Amme, Marcus; Siegenthaler, Roger [mb-microtec ag, Niederwangen (Switzerland)

    2015-07-01

    The industrial application of Tritium gas enclosed in glass tubes is a modern way illuminating instruments and items wherever instant and independent readability is prerequisite. The GTLS (Gaseous Tritium Light Sources) technology follows the principle of radiation-induced luminescence and supersedes the luminous radioactive paints and their hazards such as particles erasure or heavy isotope use. Enclosure of tritium in glass is a demanding micro technology process and work needs to be performed in controlled areas due to handling of open sources. The storage and transport of the Tritium is done via licensed B(U)-containers coming from heavy water reactor sites, and disposal of radioactive Tritium wastes has to be compliant with national and international regulations for transport and waste management.

  1. The tritium monitoring requirements of fusion and the status of research

    International Nuclear Information System (INIS)

    This report is a summary of an investigation into the tritium monitoring requirements of tritium laboratories, D-T burning ignition experiments, and fusion reactors. There is also a summary of the status of research into tritium monitoring and a survey of commercially available tritium monitors

  2. 1997 evaluation of tritium removal and mitigation technologies for Hanford Site wastewaters

    International Nuclear Information System (INIS)

    This report contains results of a biennial assessment of tritium separation technology and tritium nitration techniques for control of tritium bearing wastewaters at the Hanford Site. Tritium in wastewaters at Hanford have resulted from plutonium production, fuel reprocessing, and waste handling operations since 1944. this assessment was conducted in response to the Hanford Federal Facility Agreement and Consent Order

  3. Tritium removal from inert gases using ST 198 alloy

    International Nuclear Information System (INIS)

    Tritium handling invariably requires multiple containment to ensure personnel safety and to control emissions to the environment. For this measure to be effective, tritiated environments require periodic or continuous tritium removal. Currently, this is primarily achieved by the conventional method of catalytic oxidation of tritium bearing compounds followed by adsorption of tritiated water on molecular sieves. An experimental program was initiated to examine the potential of a metal getter detritiation process. The tritium removal characteristics of a zirconium-iron alloy, Zr2Fe, which is essentially chemically inert in nitrogen at operating temperatures of approximately 300-400 degrees C, were studied. Over 50 tritium removal tests were reproducibly conducted and the following results emerged: the zirconium alloy getter, operating at a mean temperature of 360 degrees C, is found to effectively remove tritium from nitrogen and noble gases to levels of a few μCi/m3 for initial concentrations of approximately 1300 μCi/m3 or less. In the cases of higher initial tritium concentrations, the removal is not complete, probably a consequence of getter history. The performance of the alloy is not highly sensitive to the presence of impurities. The getter effectively absorbs the impurities CO, O2 and NH3, while the impurity methane is not removed noticeably

  4. Design of a tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    The TFTR tritium pellet injector (TPI) is designed to provide a tritium pellet fueling capability with pellet speeds in the 1- to 3 km/s-range for the TFTR D-T phase. The existing TFTR deuterium pellet injector is being modified at Oak Ridge National Laboratory to provide a fourshot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns a two -stage light gas gun driver. The pipe gun concept has been qualified for tritium operation by the tritium proof-of-principle injector experiments conducted on the Tritium Systems Test Assembly at Los Alamos National Laboratory. In these experiments, tritium and D-T pellets were accelerated to speeds near 1.5 km/s. The TPI is being designed for pellet sizes in the range from 3.43 to 4.0 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation will be controlled by a programmable logic controller. 7 refs., 4 figs

  5. Design of a tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    This paper reports on the TFTR tritium pellet injector (TPI) designed to provide a tritium pellet fueling capability with pellet speeds in the 1-to 3 km/s-range for the TFTR D-T phase. The existing TFTR deuterium pellet injector (DPI) is being modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The pipe gun concept has been qualified for tritium operation by the tritium proof-of-principle (TPOP) injector experiments conducted on the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. In these experiments, tritium and D-T pellets were accelerated to speeds near 1.5 km/s. The TPI is being designed for pellet sizes in the range from 3.43 to 4.0 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation will be controlled by a programmable logic controller (PLC)

  6. Natural tritium determination in groundwater on Mt. Etna (Sicily, Italy)

    International Nuclear Information System (INIS)

    Tritium is a naturally occurring radionuclide, due to interactions of cosmic-rays with the upper layers of the atmosphere; but its presence in the environment is mainly due to residual fallout from nuclear weapons atmosphere tests, carried out from 1952 till 1980. Tritium reaches the Earth's surface mainly in the form of precipitation, becoming part of the hydrological cycle, then the interest of tritium content analysis in drinking water is both for dosimetry and health-risk and for using tritium as a natural tracer in the groundwater circulation system. This paper presents results from a survey carried out in the Mt. Etna area (east and west flanks) and in the southern side of Nebrodi in Sicily (Italy), in order to determine tritium activity concentrations in water samples by using liquid scintillation counter. The investigated areas show quite low tritium concentrations, much below the Italian limit of 100 Bq L-1 for drinking water and even comparable with the minimum detectable activity value. The effective dose due to tritium for public drinking water consumption was also evaluated. (author)

  7. In situ measurement of tritium permeation through stainless steel

    Science.gov (United States)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2013-06-01

    The TMIST-2 irradiation experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory to evaluate tritium permeation through Type 316 stainless steel (316 SS). The interior of a 316 SS seamless tube specimen was exposed to a 4He carrier gas mixed with a specified quantity of tritium (T2) to yield partial pressures of 0.1, 5, and 50 Pa at 292 °C and 330 °C. In situ tritium permeation measurements were made by passing a He-Ne sweep gas over the outer surface of the specimen to carry the permeated tritium to a bubbler column for liquid scintillation counting. Results from in situ permeation measurements were compared with predictions based on an ex-reactor permeation correlation in the literature. In situ permeation data were also used to derive an in-reactor permeation correlation as a function of temperature and pressure over the ranges considered in this study. In addition, the triton recoil contribution to tritium permeation, which results from the transmutation of 3He to T, was also evaluated by introducing a 4He carrier gas mixed with 3He at a partial pressure of 1013 Pa at 330 °C. Less than 3% of the tritium resulting from 3He transmutation contributed to tritium permeation.

  8. Characterization of tritium exposures by measuring tritiated metabolites in urine

    International Nuclear Information System (INIS)

    A high-performance liquid chromatography (HPLC)-based method was developed to look for the presence of characteristic urinary metabolites associated with different tritium-exposure situations. Non-volatile metabolites in urine were isolated by evaporating an aliquot of urine samples, at room temperature under nitrogen, from animals percutaneously exposed to tritiated thymidine, tritiated formaldehyde, tritium-gas-contaminated metal surfaces and tritiated pump oil. A total of 40 fractions were collected at 1 min intervals with a flow rate of 1 ml x min-1, and their tritium activities were measured. The activity profile of tritium showed that the ratios of non-volatile tritiated metabolites in fraction I (0-20 min) to fraction II (20-40 min) were noticeably different among the animals exposed to tritiated thymidine (77.2±4.5), tritiated formaldehyde (40.9±3.3), tritium-gas-contaminated metal surfaces (16.5±2.5), or tritiated pump oil (8.7±0.4) 24 h post-exposure. Our results suggest that, if the nature of a tritium exposure is unknown, comparison of the ratio of fraction I to fraction II in non-volatile tritiated metabolites may be useful in characterizing the source and the nature of tritium exposure. (author)

  9. Apparatus for monitoring tritium in tritium contaminating environments using a modified Kanne chamber

    Science.gov (United States)

    Anderson, David F.

    1984-01-01

    A conventional Kanne tritium monitor has been redesigned to reduce its sensitivity to such contaminants as tritiated water vapor and tritiated oil. The high voltage electrode has been replaced by a wire cylinder and the collector electrode has been reduced in diameter. The area sensitive to contamination has thereby been reduced by about a factor of forty while the overall apparatus sensitivity and operation has not been affected. The design allows for in situ decontamination of the chambers, if necessary.

  10. Apparatus for monitoring tritium in tritium-contaminating environments using a modified Kanne chamber

    Science.gov (United States)

    Anderson, D.F.

    1981-01-27

    A conventional Kanne tritium monitor has been redesigned to reduce its sensitivity to such contaminants as tritiated water vapor and tritiated oil. The high voltage electrode has been replaced by a wire cylinder and the collector electrode has been reduced in diameter. The area sensitive to contamination has thereby been reduced by about a factor of forty while the overall apparatus sensitivity and operation has not been affected. The design allows for in situ decontamination of the chambers, if necessary.

  11. The lichens, tritium and carbon 14 integrators; Les lichens, integrateurs de tritium et de carbone 14

    Energy Technology Data Exchange (ETDEWEB)

    Daillant, O

    2007-07-01

    The present report concerns a research for the tritium and for the carbon 14 in lichens in a spirit of bio-indication: the first results appear in Daillant and al (2004 ) and additional results were presented to the congress B.I.O.M.A.P. in Slovenia, organized collectively by the institute Josef Stefan from Ljubljana and the international atomic energy agency from Vienna (Daillant and al 2003). (N.C.)

  12. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  13. Final report of the tritium issues working group. Vol. 1

    International Nuclear Information System (INIS)

    Early in 1985 the proposed sale of the isotope 'tritium' by Ontario Hydro became a public issue. A number of community groups claimed in public forum that tritium recovered from Ontario Hydro's nuclear reactors would be sold or diverted to American thermonuclear (fusion) weapons. Their position was based on the following presumptions: that tritium was a major component in American nuclear weapons, that the United States has a supply problem with or shortage of this material, and that Ontario Hydro would directly or indirectly support the American nuclear weapons program: a) by providing tritium directly to the U.S. Department of Energy for use in nuclear weapons, or b) by supplying tritium to certain buyers - either traditional commercial facilities or the developing fusion research agencies associated with the Department of Energy, thus allowing or making possible the diversion of this isotope to nuclear weapons purposes, or c) by answering the needs of the commercial market, at present supplied from production reactors dedicated to supplying U.S. military requirements, indirectly allowing the U.S. government to concentrate its efforts on the production of tritium for nuclear weapons. When members of what has become known as the 'Tritium Issues Working Group' were first approached by Dr. T.S. Drolet in mid-April 1985, we were asked if we would agree to participate in a study to assess whether Canadian tritium, which is to be produced only for commercial and research purposes, could be inadvertantly utilized, either directly or indirectly, in the American nuclear weapons program. Our discussion of these issues is covered in Volume 1 of this report and is supplemented by appropriate Appendices in Volume 2. We could find absolutely nothing of a factual nature to justify the hypothesis that Canadian tritium would find its way into the American weapons program

  14. The enteral nutrition containing slow release starch improves the outcome of severe acute pancreatitis patients%含缓释淀粉的肠内营养对重症急性胰腺炎病人预后的影响

    Institute of Scientific and Technical Information of China (English)

    董朝晖

    2011-01-01

    目的:探讨含缓释淀粉的肠内营养对重症急性胰腺炎(SAP)病人的影响.方法:选择接受EN治疗的SAP病人180例,随机分为研究组和对照组,每组各90例,分别给予同等热量的EN支持,研究组用含缓释淀粉的瑞代,对照组用瑞素,观察EN支持前和支持后第10天病人空腹血糖(FBG)、餐后2 h血糖(2h BG)、糖化血红蛋白(HbAIc)的变化.并比较两组病人感染率、并发症发生率、住院时间和病死率的差异.结果:在营养支持后第10天,研究组病人FBG、2h BG、HbAlc水平均明显低于对照组(P0.05).结论:含缓释淀粉的EN应用于SAP病人显著优于标准配方的EN制剂.%Objective: To evaluate the impacts of enteral nutrition contained with slow release starch in severe acute pancreatitis (SAP) patients. Methods: 180 cases of SAP patients were randomly divided into slow release starch group ( Fresubin diabetes , SSPC, n = 90) and control group ( Fresubin ,SSPC, n =90). The fasting blood glucose (FBG), 2-hour postprandial blood glucose (2h BG) and glycosylated hemoglobin (HbAlc) were examined at base line and 10 days after nutrition support . The infectious rates, complication rates , mortality and hospital stay were compared between the two groups. Results: FBG,2h BG and HbAlc were significantly lower in slow release starch group than control group (P <0.05 )10 days after nutrition treatment. The infectious rate and complication rate were also significantly lower compared with control group (P < 0.05 ) . The mortality rate was not significantly different between the two groups ( P > 0. 05 ) , and the hospital stay in slow release starch group was shorter than control group (P < 0. 05 ). Conclusion: Enteral nutrition contained with slow release starch is the preferred enteral nutrition in severe acute pancreatitis (SAP) patients.

  15. Tritium Formation and Mitigation in High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  16. Accumulation of tritium in beryllium material under neutron irradiation

    International Nuclear Information System (INIS)

    In the present work the programming code is created on the basis of which the accumulation kinetics of tritium and isotope of He4 in the Be9 sample is analyzed depending on the time. The program is written in C++ programming language and for the calculations Monte Carlo method was applied. This program scoped on the calculation of concentration of helium and tritium in beryllium samples depending on the spectrum of the neutron flux in different experimental reactors such as JMTR, JOYO and IPEN/MB. The processes of accumulation of helium and tritium for each neutron energy spectrum of these reactors were analyzed. (author)

  17. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  18. Synthesis of some useful tritium labelled auxins

    Energy Technology Data Exchange (ETDEWEB)

    Buchman, O.; Pri-Bar, I.; Shimoni, M.; Azran, J. (Israel Atomic Energy Commission, Beersheba (Israel). Nuclear Research Center-Negev)

    1992-06-01

    The synthesis of six useful auxins labelled with tritium is described. The following compounds were prepared: 3-indoleacetic acid-5-[sup 3]H (28.9 Ci-1.07 TBq/mmol), 3-indolebutyric acid-5-[sup 3]H (7.3 Ci-270 GBq/mmol), 1-naphthylacetic acid-4-[sup 3]H (27.6 Ci-1.02 TBq/mmol), 2,4-dichloropheno-xyacetic acid-5-[sup 3]H (18.5 Ci-685 GBq/mmol), 2(2,4-dichlorophenoxy-5-[sup 3]H) -propionic acid (20.7 Ci-766 GBq/mmol), 2(2,4-dichlorophenoxy)-propionic acid-3-[sup 3]H (0.39 Ci-14.4 GMq/mmol), and 4-chlorophenoxyacetic acid-2-[sup 3]H (13.3 Ci-492 GBq/mmol). (author).

  19. Tritium permeation model for plasma facing components

    International Nuclear Information System (INIS)

    This report documents the development of a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. The model is developed for solution using commercial spread-sheet software such as Lotus 123. Comparison calculations are provided with the verified and validated TMAP4 transient code with good agreement. Results of calculations for the ITER CDA diverter are also included

  20. Decontamination tests on tritium-contaminated materials

    International Nuclear Information System (INIS)

    These tests are designed to try out various processes liable to be applied to the decontamination of a material contaminated with tritium. The samples are thin stainless- steel slabs contaminated in the laboratory with elements extracted from industrial installations. The measurement of the initial and residual activities is carried out using an open-window BERTHOLD counter. The best results are obtained by passing a current of pre-heated (300 deg. C) air containing water vapour. This process makes it possible to reach a decontamination factor of 99.5 per cent in 4 hours. In a vacuum, the operation has to be prolonged to 100 hours in order to obtain a decontamination factor of 99.2 per cent. Wet-chemical or electrolytic treatments are efficient but their use is limited by the inherent corrosion risks. A study of the reappearance of the contamination has made it possible to observe that this phenomenon occurs whatever the process used. (authors)