WorldWideScience

Sample records for tritium decay irradiation

  1. Study on radiation effect of poly (vinyl alcohol) films irradiated by tritium decay

    International Nuclear Information System (INIS)

    Li Hairong; Peng Shuming; Zhou Xiaosong; Yu Mingming; Xia Lidong; Chen Xiaohua; Liang Jianhua

    2014-01-01

    The radiation effect of poly(vinyl alcohol) films used as a kind of gas-barrier material for inertial confinement fusion (ICF) targets was studied under the different conditions of β-ray from tritium decay. The changes of physical and chemical properties of the irradiated material samples were analyzed by FTIR, XRD and AFM. The tritium-hydrogen isotopic exchange reaction of the irradiated samples mainly occurs at C-H bond and the IR absorption peak of C-T bond obviously increases with the irradiation dose. For strong hydrogen bonding interaction, the isotopic exchange reaction doesn't occur at O-H bond. The crystallinity degree and surface morphology of the irradiated samples were changed. The tensile properties of irradiated poly(vinyl alcohol) films were measured by universal material testing machine. The results show that the change trend of mechanical properties is in accordance with the microstructures of the irradiated samples. (authors)

  2. Tritium decay helium-3 effects in tungsten

    Directory of Open Access Journals (Sweden)

    M. Shimada

    2017-08-01

    Full Text Available Tritium (T implanted by plasmas diffuses into bulk material, especially rapidly at elevated temperatures, and becomes trapped in neutron radiation-induced defects in materials that act as trapping sites for the tritium. The trapped tritium atoms will decay to produce helium-3 (3He atoms at a half-life of 12.3 years. 3He has a large cross section for absorbing thermal neutrons, which after absorbing a neutron produces hydrogen (H and tritium ions with a combined kinetic energy of 0.76 MeV through the 3He(n,HT nuclear reaction. The purpose of this paper is to quantify the 3He produced in tungsten by tritium decay compared to the neutron-induced helium-4 (4He produced in tungsten. This is important given the fact that helium in materials not only creates microstructural damage in the bulk of the material but alters surface morphology of the material effecting plasma-surface interaction process (e.g. material evolution, erosion and tritium behavior of plasma-facing component materials. Effects of tritium decay 3He in tungsten are investigated here with a simple model that predicts quantity of 3He produced in a fusion DEMO FW based on a neutron energy spectrum found in literature. This study reveals that: (1 helium-3 concentration was equilibrated to ∼6% of initial/trapped tritium concentration, (2 tritium concentration remained approximately constant (94% of initial tritium concentration, and (3 displacement damage from 3He(n,HT nuclear reaction became >1 dpa/year in DEMO FW.

  3. Tritium Decay Helium-3 Effects in Tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Merrill, B. J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    A critical challenge for long-term operation of ITER and beyond to a Demonstration reactor (DEMO) and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to steady-state/transient heat fluxes and intense neutral/ion particle fluxes under the extreme fusion nuclear environment, while at the same time minimizing in-vessel tritium inventories and permeation fluxes into the PFC’s coolant. Tritium will diffuse in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [1,2]. Tritium decay into helium-3 may also play a major role in microstructural evolution (e.g. helium embrittlement) in tungsten due to relatively low helium-4 production (e.g. He/dpa ratio of 0.4-0.7 appm [3]) in tungsten. Tritium-decay helium-3 effect on tungsten is hardly understood, and its database is very limited. Two tungsten samples (99.99 at. % purity from A.L.M.T. Co., Japan) were exposed to high flux (ion flux of 1.0x1022 m-2s-1 and ion fluence of 1.0x1026 m-2) 0.5%T2/D2 plasma at two different temperatures (200, and 500°C) in Tritium Plasma Experiment (TPE) at Idaho National Laboratory. Tritium implanted samples were stored at ambient temperature in air for more than 3 years to investigate tritium decay helium-3 effect in tungsten. The tritium distributions on plasma-exposed was monitored by a tritium imaging plate technique during storage period [4]. Thermal desorption spectroscopy was performed with a ramp rate of 10°C/min up to 900°C to outgas residual deuterium and tritium but keep helium-3 in tungsten. These helium-3 implanted samples were exposed to deuterium plasma in TPE to investigate helium-3 effect on deuterium behavior in tungsten. The results show that tritium surface concentration in 200°C sample decreased to 30 %, but tritium surface concentration in 500°C sample did not alter over the 3 years storage period, indicating possible tritium

  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. Reduction of weighing errors caused by tritium decay heating

    International Nuclear Information System (INIS)

    Shaw, J.F.

    1978-01-01

    The deuterium-tritium source gas mixture for laser targets is formulated by weight. Experiments show that the maximum weighing error caused by tritium decay heating is 0.2% for a 104-cm 3 mix vessel. Air cooling the vessel reduces the weighing error by 90%

  6. Tritium and helium behavior in irradiated beryllium

    International Nuclear Information System (INIS)

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

    1990-11-01

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

  7. Irradiation of lithium aluminate and tritium extraction

    International Nuclear Information System (INIS)

    Roth, E.; Abassin, J.J.; Botter, F.; Briec, M.; Chenebault, P.; Masson, M.; Rasneur, B.; Roux, N.

    1984-12-01

    After preselection of the preparation procedures, following short irradiations, γ LiAl0 2 samples submitted to 2.10 19 fast neutrons cm -2 and 1.5 10 20 thermal neutrons cm -2 fluences experienced no apparent damage. Post-irradiation tritium extraction from samples irradiated to 2.10 17 neutrons/cm 2 in quartz ampoules produced mostly tritiated water. When in-pile experiments are performed the sample container material influences greatly the measured ratio of tritium gas to tritiated water - Stainless steel capsules yield more T 2 gas than quartz capsules probably because of a reduction process. Difficulties in interpretation arise from adsoption of tritiated water on the measuring lines. Both experiments showed that much faster extraction rates are obtained from small grain size samples than from large ones at the same open porosity. If diffusion in the grains controls the extraction rates, apparent D values vary from 10 -16 to 1.5 10 -15 cm 2 S -1 in the temperature range explored. Around 500 0 C small grain samples reached equilibrium tritium concentration of a few mCi in 4 hours. Such values are suitable for a blanket concept

  8. Tritium

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

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

  9. 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.)

  10. Progress report on the Los Alamos tritium beta decay experiment

    International Nuclear Information System (INIS)

    Wilkerson, J.F.; Bowles, T.J.; Knapp, D.A.; Robertson, R.G.H.; Wark, D.L.

    1988-01-01

    Measurements near the endpoint of the tritium beta-decay spectrum using a gaseous molecular tritium source yield an essentially model-independent upper limit of 27 eV on the /ovr ν//sub e/ mass at the 95% confidence level. Since demonstrating from this initial measurement the successful operation of a gaseous source based system, most of our effort has been concentrated towards the upgrade and optimization of the experimental apparatus. The emphasis of this work has been to eliminate or further reduce effects that generate systematic errors. Based on realistic projections from our initial measurement, an ultimate sensitivity to neutrino mass of 10 eV is expected. 12 refs., 1 fig

  11. Detection of helium-containing paramagnetic complex at decay of tritium incorporated in HTO and DTO

    International Nuclear Information System (INIS)

    Legasov, V.A.; Usatyj, A.F.; Ibragimov, R.A.; Myasoedov, N.F.

    1979-01-01

    EPR technique was used to study the paramagnetic centers appearing during long-term storage of frozen (77 K) non-deoxygenated light (H 2 O) and heavy (D 2 O) water containing 3.5% vol. highly active HTO (29 ci/cm 3 ), so that the resulting activity of the sample (0.1 ml) was about 100 mci. For comparison, the same samples containing no HTO but irradiated by Co-60 rays were studied under identical conditions. A schematic of the processes involved in tritium decay in DTO (or HTO) in the aqueous matrix frozen at low temperatures was suggested

  12. Research of CITP-II tritium production irradiation device design

    International Nuclear Information System (INIS)

    Zhang Zhihua; Deng Yongjun; Mi Xiangmiao; Li Rundong; Liu Zhiyong

    2012-01-01

    As the core component of CITP-II, the online tritium production irradiation device is the pivotal equipment in the research on tritium production and release of tritium breeders. The design of CITP-II online tritium production irradiation device creatively makes replacing the breeders online come true; as tritium production capacity, the self-shielding factor of device, and neutron flux were studied. The influence of different load models and load thicknesses of breeders to tritium production capacity was calculated. The hydrodynamics parameters of device in solid-gas phase were computed. Thermal parameters, such as the heat power of breeders, hotspot, temperature grads distributions, utmost temperature, uneven factors, were analyzed. Creatively designed nonlinear electric heater equalized breeders' even heat power. The influence laws of the components, pressure of gap gas and carrier gas to the balance temperature were got. And the key thermal parameters were ascertained. The key thermal parameters and the changing laws were got and provide the basis for structural optimization and safety analysis. They can also be referenced for the study of breeders' tritium production and release. (authors)

  13. Tritium extraction from neutron-irradiated lithium aluminate

    International Nuclear Information System (INIS)

    Garcia H, F.

    1995-01-01

    Lithium aluminate is being strongly considered as a breeder material because of its thermophysical, chemical and mechanical stability at high temperatures and its favorable irradiation behavior. Furthermore, it is compatible with other blanket and structural materials. In this work, the effects of calcination temperature during preparation, extraction temperature and sweep gas composition were observed. Lithium aluminate prepared by four different methods, was neutron irradiated for 30 minutes at a flux of 10 12 -10 13 n/cm 2 s in the TRIGA Mark III reactor at Salazar, Mexico; and the tritium extraction rate was measured. Calcination temperature do not affect the tritium extraction rate. However, using high calcination temperature, gamma lithium aluminate was formed. The tritium extraction at 600 Centigrade degrees was lower than at 800 Centigrade degrees and the tritium amount extracted by distillation of the solid sample was higher. The sweep gas composition showed that tritium extraction was less with Ar plus 0.5 % H 2 that with Ar plus 0.1 % H 2 . This result was contrary to expected, where the tritium extraction rate could be higher when hydrogen is added to the sweep gas. Probably this effect could be attributed to the gas purity. (Author)

  14. 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)

  15. Results of preliminary experiments on tritium decontamination by UV irradiation

    International Nuclear Information System (INIS)

    Oya, Yasuhisa; Shu, Wataru; O'hira, Shigeru; Hayashi, Takumi; Nishi, Masataka

    2000-03-01

    In the point of view of protection of workers from the radiation exposure and the limitation of the contamination with radioactive materials, it is important to decontaminate mobile tritium from plasma facing components of a nuclear fusion reactor at the beginning of their maintenance work. It is considered that the heating is the most effective method for decontamination. However, it is important to develop new decontamination method of adsorbed hydro-carbon based substances from the materials that cannot be heated or the inner pipe of double pipes. This report presents results of preliminary experiments performed for the development of the effective tritium decontamination technique pursuing under US/Japan collaborative program on technology for fusion-fuel processing (Annex IV). In the experiments, the effects of Ultra Violet (UV) irradiation on tritium removal from some kinds of materials, such as poly vinyl chloride -(CH 2 CHCl) n - film, polyethylene film and graphite samples coated by C 2 H 2 plasma were examined. As the result of UV irradiation, it was confined that hydrogen and carbon based compounds could be released from the specimen during UV irradiation. It is concluded that UV irradiation is one of the hopeful candidates for effective tritium decontamination. (author)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Vladimir Chakin

    2016-12-01

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

  18. Molecular effects in the neutrino mass determination from beta-decay of the tritium molecule

    International Nuclear Information System (INIS)

    Fackler, O.; Jeziorski, B.; Kolos, W.; Szalewicz, K.; Monkhorst, H.J.; Mugge, M.

    1986-03-01

    Molecular final state energies and transition probabilities have been computed for beta-decay of the tritium molecule. The results are of sufficient accuracy to make a determination of the electron neutrino rest mass with an error not exceeding a few tenths of an electron volt. Effects of approximate models of tritium beta-decay on the neutrino mass determination are discussed. 14 refs., 3 figs., 1 tab

  19. Experiment to study the β-decay of free atomic and molecular tritium

    International Nuclear Information System (INIS)

    Robertson, R.G.H.; Bowles, T.J.; Maley, M.; Browne, J.C.; Burritt, T.; Toevs, J.; Stelts, M.; Helfrick, J.; Knapp, D.; Ledebuhr, A.G.

    1982-01-01

    An apparatus is described which will allow the measurement of the β-decay of free tritium atoms and molecules for determining antineutrino mass. It consists of an RF dissociator, a long cylindrical decay region open at both ends, a guide field, and a magnetic spectrometer

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. Theoretical study of β-decay of a negative tritium ion

    International Nuclear Information System (INIS)

    Goryaev, F.F.; Sukhanov, L.P.

    1997-01-01

    A calculation of the probability of the main β-decay channel of a negative tritium ion is described as one of the processes that can affect the β-spectrum of a tritium source near the end point. The appropriate energy parameters have been calculated. This process should be taken into account in interpreting measured β-spectra near the end point in connection with determining the neutrino rest mass

  2. An assessment of anti-neutrino mass determination via electrostatic measurements of tritium beta-decay

    International Nuclear Information System (INIS)

    Le Bas, P.A.

    1984-01-01

    Data on the mass of the anti-neutrino determined via electrostatic measurements of tritium beta-decay are assessed. Relativistic calculations concerning the finite mass of the electron anti-neutrino and the recoil of the nucleus, are given for the theoretical end-point spectrum of tritium beta-decay. The specifications are given for an electrostatic Spherical Retarding Beta-Spectrometer, and an electrostatic Cylindrical Mirror Analyser, both used in the tritium beta-decay experiment. The electrostatic measurements lead to a value of less than 50 ev (90% C.L.) for the electron anti-neutrino mass. These results are discussed in terms of the resolution of the electrostatic equipment and the Monte Carlo simulations of the data collection. (UK)

  3. Status of the Los Alamos tritium beta decay experiment

    International Nuclear Information System (INIS)

    Robertson, R.G.H.; Bowles, T.J.; Wark, D.L.; Wilkerson, J.F.; Knapp, D.A.

    1989-01-01

    The Los Alamos tritium experiment employs a gaseous tritium source and a magnetic spectrometer to determine the mass of the electron antineutrino from the shape of the tritium beta spectrum. Since publication of the first result from this apparatus (m/sub nu/ < 27 eV at 95% confidence), work has concentrated on improving the data rates. A 96-element Si microstrip array detector has been installed to replace the single proportional counter at the spectrometer focus, resulting in greatly increased efficiency. Measurements of the 1s photoionization spectrum of Kr now obviate the need for reliance on the theoretical shakeup and shakeoff spectrum of Kr in determining the spectrometer resolution. 19 refs., 3 figs

  4. Determination of the neutrino mass from the beta decay of gaseous molecular tritium

    International Nuclear Information System (INIS)

    Decman, D.J.; Stoeffl, W.

    1992-06-01

    We set an upper limit of 8 eV for the mass of the electron antineutrino from studying the beta decay of tritium. We use a gaseous molecular tritium source, a high resolution magnetic spectrometer and a low background counting system to minimize the systematic errors encountered in these measurements. Our calibration data with radioactive 83m Kr enables us to measure our system response function and a good deal of atomic physics data. In addition to our end point results we have made the first measurement of the tritium beta decay spectrum below 200 keV. We find an excess of very low energy electrons which arise from molecular processes of the 3 He-T + ion

  5. Final results phase II of the Mainz neutrino mass search in tritium ß decay

    Czech Academy of Sciences Publication Activity Database

    Kraus, Ch.; Bornschein, B.; Bornschein, L.; Bonn, J.; Flatt, B.; Kovalík, Alojz; Ostrick, B.; Otten, E.W.; Schall, J.P.; Thümmler, Th.; Weinheimer, Ch.

    2005-01-01

    Roč. 40, - (2005), s. 447-468 ISSN 1434-6001 R&D Projects: GA MŠk 1P04LA213 Institutional research plan: CEZ:AV0Z10480505 Keywords : neutrino mass * tritium beta decay Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.659, year: 2005

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

  7. Tritium release from fast neutron irradiated boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1977-01-01

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

  8. Tritium

    International Nuclear Information System (INIS)

    Fiege, A.

    1992-07-01

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

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

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

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

  10. Theoretical investigation of the β decay of a negative tritium ion

    International Nuclear Information System (INIS)

    Goryaev, F.F.; Sukhanov, L.P.

    1997-01-01

    Calculation of probability of the main channel β decay of tritium negative ion as one of the possible process influencing the β spectrum form is carried out. The corresponding energy characteristics are estimated. The necessity of accounting for the investigated process by interpretation of the β spectrum experimental curves near the end point in connection with the problem of determining the neutrino rest mass, is pointed out

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  13. The effect of neutron irradiation on the trapping of tritium in carbon-based materials

    International Nuclear Information System (INIS)

    Kwast, H.; Werle, H.; Glugla, M.; Wu, C.H.; Federici, G.

    1993-11-01

    Carbon-based materials are considered for protection of plasma facing components in the next step fusion device. To investigate the effects of neutron damage on the tritium behaviour an experimental study on the tritium retention of various neutron irradiated graphites and carbon/carbon fibre composites was started. The irradiation dose of the specimens ranges from 10 -3 to 3.5 dpa.g and the irradiation temperature from 390 C to 1500 C. A comparison of tritium retention in pre- and post-irradiated carbon-based materials as a function of the sample temperature is reported in this paper and the results are discussed. The first results indicate that the retention of tritium is higher in irradiated graphite than in unirradiated graphite and depends largely on the density and microstructure. The retention is also influenced by the tritium-loading temperature. Graphite of type S 1611, irradiated at 400 C and 600 C up to a damage of 0.1 dpa.g, retained about two times more tritium than the unirradiated material. (orig.)

  14. Desorption of tritium and helium from high dose neutron irradiated beryllium

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Vlasov, V. V.; Kovalev, A. M.; Chakin, V. P.

    2007-08-01

    The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5-16) × 10 22 cm -2 at 70-100 °C and 380-420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250-1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312-445 °C and 500-740 °C, respectively. It is shown that most tritium (˜82%) from sample irradiated at 70-100 °C releases in temperature range of 312-700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380-420 °C, tritium release starts at a higher temperature ( Ts > Tann = 445 °C) and most of the tritium (˜99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at Tann > 600 °C and in helium release - at Tann > 750 °C.

  15. Initial decay process of radicals induced in irradiated food

    International Nuclear Information System (INIS)

    Kaimori, Yoshihiko; Sakamoto, Yuki; Nakamura, Hideo; Ukai, Mitsuko; Kikuchi, Masahiro; Shimoyama, Yuhei; Kobayashi, Yasuhiko; Kameya, Hiromi

    2011-01-01

    In order to determine radial decay behaviors of γ-irradiated food, we analyzed radicals in the food using ESR. We detected the ESR signal of specimens just several minutes after irradiation. The singlet signal intensity at g=2.0, originated from organic free radicals was increased as followed by the increasing radiation dose. Singlet signal intensity that increased by γ-irradiation was decreased with time. The phenomena of decay of the ESR singlet signal showed two phase that are rapid decay and slow decay. It was suggested that those two phase decay is due to at least the two radical species. Also we concluded that after three hours of radiation treatment long life radical as ESR signal intensity was detected in irradiated specimen; black pepper, green coffee bean and ginseng, showed the same decay phenomena. But the signal intensity of irradiated black pepper was three times larger than that of irradiated green coffee bean and irradiated ginseng. (author)

  16. Determination of effective cross sections and production rates for tritium in the irradiation experiment TRIDEX

    International Nuclear Information System (INIS)

    Weise, L.

    1986-04-01

    In the framework of the development of a fusion reactor blanket the irradiation experiment TRIDEX (Tritium Recovery Irradiation DIDO Experiment) takes place at the Juelich Research Reactor FRJ-2 (DIDO). For this equipment the required neutronic calculations have been performed. The aim was the determination of the neutron spectrum and several therefrom derived integral parameters for the irradiation positions in interest. From the calculated effective cross sections for the formation of Tritium resulting from irradiated Lithium samples on one hand and from measured neutron flux densities on the other hand, all needed quantities of the Tritium production could be determined. The calculation of the neutron spectrum has been performed in a two-dimensional x-y-geometry. The neutron flux densities have been gained by gamma-spectrometric measurement of the activities in irradiated activation samples. (orig.) [de

  17. Limit on the anti nu/sub e/ mass in free molecular tritium beta decay

    International Nuclear Information System (INIS)

    Bowles, T.J.; Wilkerson, J.F.; Browne, J.C.; Maley, M.P.; Robertson, R.G.H.; Knapp, D.A.; Helffrich, J.A.

    1986-01-01

    The question of a nonzero neutrino mass has received considerable attention since the claims of Lyubimov et al in 1980 were published which showed evidence for an electron antineutrino mass between 14 and 46 eV, with a best fit value of 35 eV. However, there are still considerable concerns about possible systematic problems in thier experiment. Many of these concerns revolve around the use of a tritiated valine source, in which the energy given up in final state excitations of the molecule following the veta decay of one of the tritium atoms is comparable to the size of the neutrino mass observed. The effect of these final state effects is difficult to calculate in a molecule as complex as valine. In addition, ionization energy loss and backscattering of the betas in traversing the solid source are appreciable and must be very accurately accounted for. These concerns have led us to carry out an experiment using free molecular tritium as the source material. The final state effects have been accurately calculated for the tritium molecule and the uncertainties in these calculations cannot generate a spurious neutrino mass greater than 1 eV. in addition, the energy loss in the source is small because the source consists of tritium only and there is no backscattering

  18. Tritium-doping enhancement of polystyrene by ultraviolet laser and hydrogen plasma irradiation for laser fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Iwasa, Yuki, E-mail: iwasa-y@ile.osaka-u.ac.jp [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Yamanoi, Kohei; Iwano, Keisuke; Empizo, Melvin John F.; Arikawa, Yasunobu; Fujioka, Shinsuke; Sarukura, Nobuhiko; Shiraga, Hiroyuki; Takagi, Masaru; Norimatsu, Takayoshi; Azechi, Hiroshi [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Noborio, Kazuyuki; Hara, Masanori; Matsuyama, Masao [Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama 930-8555 (Japan)

    2016-11-15

    Highlights: • Tritium-doped polystyrene films are fabricated by the Wilzbach method with UV laser and hydrogen plasma irradiation. • The 266-nm laser-irradiated, 355-nm laser-irradiated, and hydrogen plasma-irradiated polystyrene films exhibit higher PSL intensities and specific radioactivities than the non-irradiated sample. • Tritium doping by UV laser irradiation can be largely affected by the laser wavelength because of polystyrene’s absorption. • Hydrogen plasma irradiation results to a more uniform doping concentration even at low partial pressure and short irradiation time. • UV laser and plasma irradiations can be utilized to fabricate tritium-doped polystyrene shell targets for future laser fusion experiments. - Abstract: We investigate the tritium-doping enhancement of polystyrene by ultraviolet (UV) laser and hydrogen plasma irradiation. Tritium-doped polystyrene films are fabricated by the Wilzbach method with UV laser and hydrogen plasma. The 266-nm laser-irradiated, 355-nm laser-irradiated, and hydrogen plasma-irradiated polystyrene films exhibit higher PSL intensities and specific radioactivities than the non-irradiated sample. Tritium doping by UV laser irradiation can be largely affected by the laser wavelength because of polystyrene’s absorption. In addition, UV laser irradiation is more localized and concentrated at the spot of laser irradiation, while hydrogen plasma irradiation results to a more uniform doping concentration even at low partial pressure and short irradiation time. Both UV laser and plasma irradiations can nevertheless be utilized to fabricate tritium-doped polystyrene targets for future laser fusion experiments. With a high doping rate and efficiency, a 1% tritium-doped polystyrene shell target having 7.6 × 10{sup 11} Bq g{sup −1} specific radioactivity can be obtained at a short period of time thereby decreasing tritium consumption and safety management costs.

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

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

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

  20. Genotoxic and reprotoxic effects of tritium and external gamma irradiation on aquatic animals.

    Science.gov (United States)

    Adam-Guillermin, Christelle; Pereira, Sandrine; Della-Vedova, Claire; Hinton, Tom; Garnier-Laplace, Jacqueline

    2012-01-01

    Aquatic ecosystems are chronically exposed to natural radioactivity or to artificial radionuclides released by human activities (e.g., nuclear medicine and biology,nuclear industry, military applications). Should the nuclear industry expand in the future, radioactive environmental releases, under normal operating conditions or accidental ones, are expected to increase, which raises public concerns about possible consequences on the environment and human health. Radionuclide exposures may drive macromolecule alterations, and among macromolecules DNA is the major target for ionizing radiations. DNA damage, if not correctly repaired, may induce mutations, teratogenesis, and reproductive effects. As such, damage at the molecular level may have consequences at the population level. In this review, we present an overview of the literature dealing with the effects of radionuclides on DNA, development, and reproduction of aquatic organisms. The review focuses on the main radionuclides that are released by nuclear power plants under normal operating conditions, γ emitters and tritium. Additionally, we fitted nonlinear curves to the dose-response data provided in the reviewed publications and manuscripts, and thus obtained endpoints commonly associated with ecotoxicological studies, such as the EDR(10). These were then used as a common metric for comparing the values and data published in the literature.The effects of tritium on aquatic organisms were reviewed for dose rates that ranged from 29 nGy/day to 29 Gy/day. Although beta emission from tritium decay presents a rather special risk of damage to DNA, genotoxicity-induced by tritium has been scarcely studied. Most of the effects studied have related to reproduction and development. Species sensitivity and the form of tritium present are important factors that drive the ecotoxicity of tritium. We have concluded from this review that invertebrates are more sensitive to the effects of tritium than are vertebrates

  1. Preparation of Tritium from irradiated lithium compounds (study)

    International Nuclear Information System (INIS)

    1989-01-01

    The aim of the present study is the selection of a certain scheme for release, collection, measurement and analysis of Tritium as produced in Lithium compounds (Li 2 O, LiOH, Li 2 CO 3 LiF and some Lithium alloys) in accordance with the 6 Li(n, α)T reaction. Tritium technology is of vital concern to power reactor programmes as well as to fusion technology. Meanwhile the fields of activity include, tritium generation and mangement in fission and fusion reactors; enviromental studies, release modeling HT/HTO conversion and dose assessments, absorption/ deabsorption, monitoring and plant design; research and development, labeling, compatibility and physical chemical properties

  2. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  4. Present status of irradiation tests on tritium breeding blanket for fusion reactor

    International Nuclear Information System (INIS)

    Futamura, Yoshiaki; Sagawa, Hisashi; Shimakawa, Satoshi; Tsuchiya, Kunihiko; Kuroda, Toshimasa; Kawamura, Hiroshi.

    1994-01-01

    To develop a tritium breeding blanket for a fusion reactor, irradiation tests in fission reactors are indispensable for obtaining data on irradiation effects on materials, and neutronics/thermal characteristics and tritium production/recovery performance of the blanket. Various irradiation tests have been conducted in the world, especially to investigate tritium release characteristics from tritium breeding and neutron multiplier materials, and materials integrity under irradiation. In Japan, VOM experiments at JRR-2 for ceramic breeders and experiments at JMTR for ceramic breeders and beryllium as a neutron multiplier have been performed. Several universities have also investigated ceramic breeders. In the EC, the EXOTIC experiments at HFR in the Netherlands and the SIBELIUS, the LILA, the LISA and the MOZART experiments for ceramic breeders have carried out. In Canada, NRU has been used for the CRITIC experiments. The TRIO experiments at ORR(ORNL), experiments at RTNS-II, FUBR and ATR have been conducted in the USA. The last two are experiments with high neutron fluence aiming at investigating materials integrity under irradiation. The BEATRIX-I and -II experiments have proceeded under international collaboration of Japan, Canada, the EC and the USA. This report shows the present status of these irradiation tests following a review of the blanket design in the ITER CDA(Conceptual Design Activity). (author)

  5. Preparation of Tritium from irradiated lithium compounds (study)

    Energy Technology Data Exchange (ETDEWEB)

    1989-01-01

    The aim of the present study is the selection of a certain scheme for release, collection, measurement and analysis of Tritium as produced in Lithium compounds (Li[sub 2]O, LiOH, Li[sub 2]CO[sub 3] LiF and some Lithium alloys) in accordance with the [sup 6]Li(n, [alpha])T reaction. Tritium technology is of vital concern to power reactor programmes as well as to fusion technology. Meanwhile the fields of activity include, tritium generation and mangement in fission and fusion reactors; enviromental studies, release modeling HT/HTO conversion and dose assessments, absorption/ deabsorption, monitoring and plant design; research and development, labeling, compatibility and physical chemical properties.

  6. Tritium extraction mechanisms from lithium aluminates during in pile irradiation experiments

    International Nuclear Information System (INIS)

    Briec, M.; Roth, E.

    1987-04-01

    The principal aim was to determine ranges of parameters governing tritium release from γ lithium aluminates within which acceptable rates for their contemplated usage as tritium breeder material in a fusion reactor blanket could be obtained. in the first place values of every quantity involved should be known as well as possible. Reproducible results should be a criterium of validity of the selected parameters. It is shown from a description of a series of experiments that processes limiting tritium release rates are not the same in different temperature ranges. By varying the composition of purge gases used for tritium extraction, the level of irradiation fluxes, and by studying simultaneously samples of different textures, results were obtained and an assignment of the respective role of defect formation, texture, surface effect is attempted to interpret them

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

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

    International Nuclear Information System (INIS)

    Tanifuji, Takaaki; Yamaki, Daiju; Jitsukawa, Shiro

    2006-01-01

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

  9. Lethal and Mutagenic Effects of Tritium Decay Produced by Tritiated Compounds Incorporated into Bacteria and Bacteriophages

    Energy Technology Data Exchange (ETDEWEB)

    Person, S. [Pennsylvania State University, University Park, PA (United States)

    1968-06-15

    The work discussed below is predominantly from the author's laboratory. Some effects of tritium decay on bacteria and bacteriophages, using labelled compounds that are incorporated preferentially into DNA, RNA or protein, have been studied. A relatively high killing efficiency, the probability that a single decay produces loss of colony-forming ability in bacteria or loss of plaque-forming ability in bacteriophages, is observed for thymidine- methyl-{sup 3}H decay, but this is thought to be due to S-particle ionization damage. The most definitive experiments involved a comparison of killing efficiencies for decays originating as thymidine-methyl-{sup 3}H in phage DNA and as amino acid-{sup 3}H in phage protein with the calculated {beta}-particle path length through, the DNA in the two cases. Experimental and calculated values are essentially the same. Radiation-dose calculations to many different bacterial sub-volumes were determined for several different distributions of radioactivity. The high killing efficiency for thymidine-methyl- {sup 3}H decay can be explained by {beta}-particle ionizations, providing DNA is the sensitive target molecule and it is organized in a central volume of the cell (see also Ref.[24]). Although both the lethal and mutagenic effect of thymidine-methyl-{sup 3}H and amino acid-{sup 3}H decays can readily be explained on the basis of {beta}-particle ionizations, the observed large mutation frequency produced by uracil-5{sup 3}H decay cannot. The majority of mutations produced by uracil-5{sup 3}H decays are due to a specific chemical rearrangement of the parent molecule, since decays from uracil-6{sup 3}H are 6 to 7-fold less mutagenic. It is clear that the decays leading to mutations for cells labelled with uracil-5{sup 3}H originate as cytosine-5{sup 3}H in bacterial DNA. Recent work shows that the specific chemical rearrangement causes a C --> T(u) genetic coding change. (author)

  10. Search for a rest mass of the electron antineutrino in the beta-decay of tritium

    International Nuclear Information System (INIS)

    Daniel, H.; Hiddemann, K.H.; Schwentker, O.

    1986-01-01

    The authors performed a precision measurement of the end point region of the tritium β-ray spectrum. The data were recorded with an iron-free π√13/2 magnetic spectrometer. The source consisted of tritium atoms, absorbed in a thin layer of hafnium, which was evaporated on a silicon wafer. The instrumental resolution was set to 0.1% in Δrho/rho. The fractional solid angle was 1% of 4π. The authors identified the electrons with a proportional counter, filled with isobutane. The counter had two distinct features: First, the good energy resolution of the signal at the main counting wire over a wide energy range allowed us to discriminate between real and background events, and second, there was a ring of several secondary counting wires around the main wire within the same gas volume, which served as an anticoincidence counter. With this arrangement an additional reduction of the background could be obtained. The functions of the spectrometer were checked by using internal conversion lines from the decay of /sup 169/Yb

  11. Ion-molecular reactions initiated by β-decay of tritium in tritiated compounds

    International Nuclear Information System (INIS)

    Akulov, G.P.

    1976-01-01

    Ion-molecular reactions initiated by β-decay of tritium and tritiated hydrocarbons have been systematized. The review describes the theoretical and experimental foundation of the radiochemical method of this important type of chemical interactions investigation. The reactions of the molecular ions of HeT + with methane, ethane, propane, butane and also with cycloalkanes from C 3 to C 6 are discussed. The reactions under consideration have been united into two groups: reactions involving HeT + ions and those involving carbonic ions. From the experimental results available, the conclusions have been drawn about the reactivity of these intermediate formations, about the mechanism of their interaction with organic substances, and also about the perspectives of using the radiochemical method in studies of ion-molecular reactions. The experimental results published before May 1974 are examined. The bibliography includes 162 references

  12. Screening effect as an explanation of results of Simpson experiment on tritium β decay

    International Nuclear Information System (INIS)

    Drukarev, E.G.; Strikman, M.I.

    1989-01-01

    Results of the work carried out by Simpson who observed a wide maximum in electron spectrum as a result of tritium decay in the range 0.75< T<1.5 keV are discussed. The results are not confirmed by the subsequent experiments. It is shown in the work that after correct account of interaction in finite state the discrepancy between experimental data and theoretical predictions for T ≥ 1 keV disappears. It is also shown that results at T<1 keV depend greatly on the effects of interaction with medium. Gross evaluation of the effects eliminates the discrepancy in the range 0.75< T<1 keV

  13. Activation analysis of tritium breeder lithium lead irradiated by fusion neutrons in FDS-II

    International Nuclear Information System (INIS)

    Mingliang Chen

    2006-01-01

    R-and-D of fusion materials, especially their activation characteristics, is one of the key issues for fusion research in the world. Research on activation characteristics for low activation materials, such as reduced activation ferritic/martensitic steels, vanadium alloys and SiCf/SiC composites, is being done throughout the world to ensure the attractiveness of fusion power regarding safety and environmental aspects. However, there is less research on the activation characteristics of the other important fusion materials, such as tritium breeder etc.. Lithium lead (Li 17 Pb 83 ) is presently considered as a primary candidate tritium breeder for fusion power reactors because of its attractive characteristics. It can serve as a tritium breeder, neutron multiplier and coolant in the blanket at the same time. The radioactivity of Li 17 Pb 83 by D-T fusion neutrons in FDS-II has been calculated and analyzed. FDS-II is a concept design of fusion power reactor, which consists of fusion core with advanced plasma parameters extrapolated from the ITER (International Thermonuclear Experimental Reactor) and two candidates of liquid lithium breeder blankets (named SLL and DLL blankets). The neutron transport and activation calculation are carried out based on the one-dimensional model for FDS-II with the home-developed multi-functional code system VisualBUS and the multi-group data library HENDL1.0/MG and European Activation File EAF-99. The effects of irradiation time on the activation characteristics of Li 17 Pb 83 were analyzed and it concludes that the irradiation time has an important effect on the activation level of Li 17 Pb 83 . Furthermore, the results were compared with the activation levels of other tritium breeders, such as Li 4 SiO 4 , Li 2 TiO 3 , Li 2 O and Li etc., under the same irradiation conditions. The dominant nuclides to dose rate and activity of Li 17 Pb 83 were analyzed as well. Tritium generated by Li has a great contribution to the afterheat and

  14. Recent progress in the modelling of helium and tritium behaviour in irradiated beryllium pebbles

    International Nuclear Information System (INIS)

    Rabaglino, E.; Ronchi, C.; Cardella, A.

    2003-01-01

    One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release

  15. In-situ tritium recovery from Li2O irradiated in fast neutron flux - Beatrix-II temperature change specimen

    International Nuclear Information System (INIS)

    Slagle, O.D.; Hollenberg, G.W.; Kurasawa, T.; Verrall, R.A.

    1992-01-01

    The Beatrix-II irradiation experiment is an in-situ tritium release experiment to evaluate the stability and tritium release characteristics of Li 2 O under fast neutron irradiation to extended burnups. A thin annular ring specimen capable of temperature changes was irradiated in Phase I of the experiment to a lithium burnup of 5%. The primary emphasis of the test plan was to determine the effect and interrelationship of gas composition and temperature on the tritium inventory with increasing temperature and a series of specific temperature changes were carried out at intervals throughout the experiment to characterize the effect of burnup. Decreasing the amount of hydrogen in the sweep gas resulted in an increase in the tritium inventory in the Li 2 O specimen. The tritium recovery during startup and shutdown was observed to be strongly influenced by the composition of the sweep gas

  16. 2017 Status report-Tritium aging studies on stainless steel: Effect of hydrogen, tritium and decay helium on the fracture-toughness properties of stem, cup and block forgings

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Michael J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The materials of construction of tritium reservoirs are forged stainless steels. During service, the structural properties of the stainless steel change over time because of the diffusion of tritium into the reservoir wall and its radioactive decay to helium-3. This aging effect can cause cracks to initiate and grow which could result in a tritium leak or delayed failure of a tritium reservoir. Numerous factors affect the tendency for crack formation and propagation and are being investigated in this program. The goal of the research is to provide relevant fracture mechanics data that can be used by the design agencies in their assessments of tritium reservoir structural integrity. In this status report, new experimental results are presented on the effects of tritium and decay helium on the cracking properties of specimens taken from actual tritium reservoir forgings instead of the experimental forgings of past programs. The properties measured are more representative of actual reservoir properties because the microstructure of the specimens tested are more like that of the actual tritium reservoirs. The program was designed to measure the effects of material variables on tritium compatibility and includes two stainless steels (Type 304L and 316L stainless steel), multiple yield strengths (360-500 MPa), and multiple forging shapes (Stem, Cup, and Block).

  17. Segmentation strategies for the irradiated and tritium contaminated PPPL TFTR

    International Nuclear Information System (INIS)

    Walton, G.R.; Spampinato, P.T.

    1995-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory is scheduled to complete its final experiments in the Fall of 1995. As a result, the TFTR will be activated and tritium contaminated. After the experiments are complete, the TFTR will undergo Shutdown and Removal (S and R). The space vacated by the TFTR will be used for a new test reactor, the Tokamak Physics Experiment (TPX). Remote methods may be required to remove components and to segment the Vacuum Vessel. The TFTR has been studied to determine alternatives for the segmentation of the Vacuum Vessel from the inside (In-Vessel). The methodology to determine suitable strategies to segment the Vacuum Vessel from In-Vessel included several areas of concentration. These areas were segmentation locations, cutting/removal technologies, pros and cons, and cutting/removal technology delivery systems. The segmentation locations for easiest implementation and minimal steps in cutting and removal have been identified. Each of these will also achieve the baseline for packaging and shipment. The methods for cutting and removal of components were determined. In addition, the delivery systems were conceptualized

  18. Segmentation strategies for the irradiated and tritium contaminated PPPL TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Walton, G.R. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Litka, T.J. [Advanced Consulting Group, Inc., Chicago, IL (United States); Spampinato, P.T. [RHD Consultants, Inc., Princeton, NJ (United States)

    1995-02-09

    The Tokamak Fusion Test Reactor (TFTR) at Princeton Plasma Physics Laboratory is scheduled to complete its final experiments in the Fall of 1995. As a result, the TFTR will be activated and tritium contaminated. After the experiments are complete, the TFTR will undergo Shutdown and Removal (S and R). The space vacated by the TFTR will be used for a new test reactor, the Tokamak Physics Experiment (TPX). Remote methods may be required to remove components and to segment the Vacuum Vessel. The TFTR has been studied to determine alternatives for the segmentation of the Vacuum Vessel from the inside (In-Vessel). The methodology to determine suitable strategies to segment the Vacuum Vessel from In-Vessel included several areas of concentration. These areas were segmentation locations, cutting/removal technologies, pros and cons, and cutting/removal technology delivery systems. The segmentation locations for easiest implementation and minimal steps in cutting and removal have been identified. Each of these will also achieve the baseline for packaging and shipment. The methods for cutting and removal of components were determined. In addition, the delivery systems were conceptualized.

  19. Atomic effects in tritium beta-decay. II. Muon to electron conversion in atoms

    International Nuclear Information System (INIS)

    Wampler, K.D.

    1989-01-01

    I. The final-state, atomic effects in the low energy end of the tritium beta decay spectrum are studied in detail. The author treats the instantaneous, two-electron repulsion in the final state, effectively to all orders in perturbation theory, by solving the eigenvalue problem with a discretized and truncated form of the Hamiltonian. He finds that these effects fail to explain the distortion in the spectrum observed by Simpson (Phys. Rev. Lett. 54, 649 (1985)). Simpson attributed this distortion to the admixture of a heavy mass antineutrino in the outgoing electron antineutrino state. In fact, the final-state Coulomb effects enhance the distortion. This calculation clears up some of the ambiguities of other theoretical analyses based on considerations of screening functions and perturbation theory. II. He presents a phenomenological study of separate lepton number violating muon to electron conversion in atoms. Previous work on this process has concentrated on elastic transitions where the nucleus characteristics have the gate on the substrate and the source-drain contacts on the top of the sample. The first use as an FET dielectric is reported of hydrogenated amorphous silicon-carbon (prepared from silane and propane mixture), photo-oxidised by UV lamp or laser. These FETs have similar characteristics to those with silicon nitride gate insulator but without the difficulties of preparing good insulator/semiconductor interfaces. Using the same materials attempts have been made to produce charge coupled devices

  20. LIBRETTO-3: Performance of tritium permeation barriers under irradiation at the HFR Petten

    International Nuclear Information System (INIS)

    Conrad, R.; Fuetterer, M.A.; Giancarli, L.; May, R.; Perujo, A.; Sample, T.

    1994-01-01

    The LIBRETTO-3 irradiation was performed at the HFR Petten during 77 full power days in three cycles to compare the efficiency of three different tritium permeation barriers in presence of Pb-17Li to uncoated AISI 316L steel. For this purpose four steel capsules (arnothing o =10 mm, arnothing i =8 mm, l=300 mm) were filled with 28 g Pb-17Li. The coatings included CVD TiC (outside), CVD TiC+Al 2 O 3 (inside), and pack cementation aluminisation (inside). The generated tritium was partly extracted by bubbling, partly it permeated through the capsules. Permeated and extracted tritium were measured as a function of temperature (280-450 C), H 2 doping (0-1 vol%) and purge gas flow rate. The driving partial pressure in the coated capsules were from an extraction model calibrated by the uncoated capsule for which tritium partial pressure could be calculated. In LIBRETTO-3 conditons, the best barrier was pack cementation aluminisation. The first interpretation of the experiment could, however, not confirm permeation reduction factors of 100 or more expected from this barrier. ((orig.))

  1. Comparison of the tritium residence times of various ceramic breeder materials irradiated in EXOTIC experiments 4 and 5

    International Nuclear Information System (INIS)

    Kwast, H.; Elen, J.D.; Conrad, R.; Casadio, S.; Werle, H.; Verstappen, G.

    1990-09-01

    Tritium residence times have been determined for various ceramic tritium breeding materials from in-situ release measurements. The irradiations, codenamed EXOTIC (EXtraction Of Tritium In Ceramics), were carried out in the High Flux Reactor (HFR) Petten. During the irradiation more than 450 transients were performed and the corresponding tritium release measured. Materials supplied by SCK/CEN (Li 2 ZrO 3 ), CEA (Li 2 ZrO 3 and LiAlO 2 ), ENEA (LiAlO 2 ), KfK (Li 4 SiO 4 ), NRL (Li 6 Zr 2 O 7 ) and ECN (Li 8 ZrO 6 ) were irradiated in EXOTIC-5 to compare the tritium residence times obtained under equal conditions. Apart from differences in density, grain size, pore size and OPV it appeared that the tritium residence times of the lithium zirconates (pellets) were shorter than those of the Li 4 SiO 4 pebbles. The tritium residence times of the Li 4 SiO 4 pebbles were shorter than those of the LiAlO 2 pellets. (author). 7 refs.; 5 figs.; 3 tabs

  2. Tritium trick

    Science.gov (United States)

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

    1971-01-01

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

  3. 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.)

  4. Advanced electrolytic cascade process for tritium recovery from irradiated heavy water moderator (Preprint No. PD-15)

    International Nuclear Information System (INIS)

    Ragunathan, P.; Mitra, S.K.; Jain, D.K.; Nayar, M.G.; Ramani, M.P.S.

    1989-04-01

    The paper briefly describes a design study of an electrolytic cascade process plant for enrichment and recovery of tritium from irradiated heavy water moderators from Rajasthan Atomic Power Station Reactors. In direct multistage electrolysis process, tritiated heavy water from the reactor units is fed to the electrolytic cell modules arranged in the form of a cascade where it is enriched and decomposed into O 2 gas stream and D 2 /DT gas stream. The direct electrolysis of tritiated heavy water allows tritium to be concentrated in the aqueous phase. Several stages are used to achieve the necessary enrichment. The cascade plant incorporates the advanced electrolyser technology developed in Bhabha Atomic Research Centre (Bombay) using porous nickel electrodes, capable o f high current density operation at reduced energy consumption for electrolysis. (author). 3 tabs

  5. Irradiation of carabao mangoes for decay control

    International Nuclear Information System (INIS)

    Alabastro, E.F.; Pineda, A.S.; Pangan, A.C.; Valle, M.J. del.

    Control of postharvest diseases of carabao mangoes by gamma radiation has been the subject of several local studies, notably that of Herrera and Valencia (1964) and Pablo et al (1971) which showed dose levels of 16.6 Kr. and 60 Kr. respectively, to be effective in extending the market life. The wide range of resorted effective dose levels for mangoes prompted this investigation which aimed to establish the pre-dominant spoilage-causing micro-organisms and the effectivity of irradiation in controlling or delaying the growth of these micro-organisms. (author)

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

    International Nuclear Information System (INIS)

    Bertone, P.C.

    1986-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  8. High resolution studies by Secondary Ion Mass Spectrometry of the spatial distribution of tritium in neutron irradiated beryllium pebbles

    International Nuclear Information System (INIS)

    Rabaglino, E.; Tamborini, G.; Hiernaut, J.-P.; Betti, M.

    2006-01-01

    A key issue of beryllium as a neutron multiplier in the blanket of future fusion reactors is tritium retention. Models are under development in order to predict tritium release kinetics in the typical operating conditions of the material in the blanket: the absence of experimental data in this range imposes an extrapolation of the models, therefore a detailed characterization and understanding of microscopic diffusion phenomena related to macroscopic tritium release is necessary. It has been recently shown, that the availability of evidence on such phenomena at a scale of 1 micron down to tens of nanometers enables a relevant progress in the effectiveness of model validation: therefore the need for applying and developing advanced analytical techniques based on mass spectrometry at this scale. A study of tritium spatial distribution in neutron irradiated beryllium pebbles (2 mm diameter, 480 appm 4 He, 7 appm 3 H) by means of Secondary Ion Mass Spectrometry (SIMS) is presented. Samples in different conditions (non-irradiated, at end of irradiation and at different temperatures during thermal ramp annealing) are examined by an oxygen ion primary beam with a spatial resolution of 1 micron along a diameter. The sample preparation is optimized in order to enable a quantitative comparison among the different conditions. Under an oxygen ion beam tritium is detected in the irradiated samples in a molecular form (3H 2 ), with a continuous distribution inside the grains, which suggests the presence of small clusters in agreement with TEM analyses, and in the form of peaks at grain boundaries, corresponding to large grain boundary bubbles. The evolving of molecular tritium distribution measured by SIMS during a typical thermal ramp release experiment shows precisely tritium diffusion from the centre of the grain to grain boundaries as the temperature increases: at the same time the remaining intragranular tritium inventory, given by the integral of the distribution

  9. Tritium extraction from neutron-irradiated lithium aluminate.; Extraccion del tritio generado por irradiacion neutronica de aluminato de litio.

    Energy Technology Data Exchange (ETDEWEB)

    Garcia H, F

    1995-10-01

    Lithium aluminate is being strongly considered as a breeder material because of its thermophysical, chemical and mechanical stability at high temperatures and its favorable irradiation behavior. Furthermore, it is compatible with other blanket and structural materials. In this work, the effects of calcination temperature during preparation, extraction temperature and sweep gas composition were observed. Lithium aluminate prepared by four different methods, was neutron irradiated for 30 minutes at a flux of 10{sup 12} -10{sup 13} n/cm{sup 2} s in the TRIGA Mark III reactor at Salazar, Mexico; and the tritium extraction rate was measured. Calcination temperature do not affect the tritium extraction rate. However, using high calcination temperature, gamma lithium aluminate was formed. The tritium extraction at 600 Centigrade degrees was lower than at 800 Centigrade degrees and the tritium amount extracted by distillation of the solid sample was higher. The sweep gas composition showed that tritium extraction was less with Ar plus 0.5 % H{sub 2} that with Ar plus 0.1 % H{sub 2}. This result was contrary to expected, where the tritium extraction rate could be higher when hydrogen is added to the sweep gas. Probably this effect could be attributed to the gas purity. (Author).

  10. In situ tritium from Li2O and Li2ZrO3 irradiated in a fast neutron flux: BEATRIX-II, Phase 1 and 2

    International Nuclear Information System (INIS)

    Slagle, O.D.; Hollenberg, G.W.; Kurasawa, T.; Verrall, R.A.

    1994-06-01

    BEATRIX-II was an in situ tritium recovery experiment. This in situ tritium recovery experiment provided data on the performance of Li 2 O and Li 2 ZrO 3 under irradiation conditions covering a range of sweep gas compositions and temperatures. The experiment consisted of two separate irradiation cycles which in turn included two vented tritium recovery canisters each. Phase 1 operated for 300 Effective Full Power Days (EFPD) while Phase 2 operated for 203 EFPD of irradiation. The tritium recovery behavior of Li 2 O and Li 2 ZrO 3 was characterized using temperature transients, sweep gas composition changes and reactor power changes to effect changes in the tritium inventory. The high neutron flux level in FFTF resulted in high tritium generation rates which combined with a responsive tritium measurement system allowed detailed observations on the tritium recovery behavior of Li 2 O and Li 2 ZrO 3 . During the course of the experiment a number of observations were made which did not appear consistent with a simplified view of the tritium recovery behavior of these materials. These observations included small negative tritium recovery peaks preceding the typical primary positive peaks; and, for specific temperature ranges, changes in tritium inventory which where opposite to the characteristic expected changes

  11. In-situ tritium recovery from Li2O irradiated in fast neutron flux: BEATRIX-II initial results

    International Nuclear Information System (INIS)

    Kurasawa, T.; Slagle, O.D.; Hollenberg, G.W.; Verrall, R.A.

    1990-10-01

    The BEATRIX-II experiment in FFTF is an in-situ tritium recovery experiment to evaluate the tritium release characteristics of Li 2 O and its stability under fast neutron irradiation to extended burnups. This experiment includes two specimens: a thin annular specimen capable of temperature transients and a larger temperature gradient specimen. During the first 85 days of the operating cycle of the reactor, the tritium recovery rate of a temperature transient capsule was examined as a function of temperature, gas flow rate, gas composition and burnup. Temperature changes in the range from 525 to 625 degree C resulted in decreasing tritium inventory with increasing temperature. Lower gas flow rates resulted in slightly lower tritium release rates while gas composition changes affected the tritium release rate significantly, more than either flow rate or temperature changes. Three different sweep gases were used: He with 0.1% H 2 , He with 0.01% H 2 , and pure He. Decreasing the amount of hydrogen in the sweep gas decreased the steady-state release rate by as much as a factor of two. A temperature gradient capsule is more prototypic of the conditions expected in a fusion blanket and was designed to provide data that can be used in evaluating the operational parameters of a solid breeder in a blanket environment. The operation of this canister during the first 85 EFPD cycle suggests that Li 2 O is a viable solid breeder material. 9 refs., 5 figs., 3 tabs

  12. Progress report on the accelerator production of tritium materials irradiation program

    International Nuclear Information System (INIS)

    Maloy, S.A.; Sommer, W.F.; Brown, R.D.; Roberts, J.E.

    1997-01-01

    The Accelerator Production of Tritium (APT) project is developing an accelerator and a spoliation neutron source capable of producing tritium through neutron capture on He-3. A high atomic weight target is used to produce neutrons that are then multiplied and moderated in a blanket prior to capture. Materials used in the target and blanket region of an APT facility will be subjected to several different and mixed particle radiation environments; high energy protons (1-2 GeV), protons in the 20 MeV range, high energy neutrons, and low energy neutrons, depending on position in the target and blanket. Flux levels exceed 10 14 /cm 2 s in some areas. The APT project is sponsoring an irradiation damage effects program that will generate the first data-base for materials exposed to high energy particles typical of spallation neutron sources. The program includes a number of candidate materials in small specimen and model component form and uses the Los Alamos Spallation Radiation Effects Facility (LASREF) at the 800 MeV, Los Alamos Neutron Science Center (LANSCE) accelerator

  13. 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)

  14. Tritium dosimetry and standardization

    International Nuclear Information System (INIS)

    Balonov, M.I.

    1983-01-01

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

  15. Measurement of tritium production in 6LiD irradiated with neutrons from a critical system

    International Nuclear Information System (INIS)

    Duan Shaojie

    1998-03-01

    The tritium production rate and its distribution, in a 6 LiD semisphere on a critical assembly neutron source are measured with a 6 Li sandwich gold-silicon surface barrier detector. Then tritium production rate and the average tritium production length of the neutrons in the whole 6 LiD sphere are derived from approximate sphere symmetry

  16. Improved limit on the electron anti neutrino rest mass from tritium β decay

    International Nuclear Information System (INIS)

    Weinheimer, C.; Przyrembel, M.; Backe, H.; Barth, H.; Bonn, J.; Degen, B.; Edling, T.; Fischer, H.; Fleischmann, L.; Grooss, J.U.; Haid, R.; Hermanni, A.; Kube, G.; Loeken, T.; Molz, A.; Osipowicz, A.; Otten, E.W.; Picard, A.; Schrader, M.; Steininger, M.

    1993-01-01

    This paper presents the results of a re-measurement of the endpoint region of the beta-spectrum of tritium by an electrostatic spectrometer with magnetic guiding field. It enables the search for a rest mass of the electron anti neutrino with improved precision. The result is: m ν 2 =(-39±34 stat ±15 syst ) (eV/c 2 ) 2 , from which an upper limit of: m ν 2 may be derived. The experiment yields the atomic mass difference: m(T)-m( 3 He)=(18591±3) eV/c 2 . (TEC). 31 refs., 4 figs., 3 tabs

  17. Tritium autoradiography

    International Nuclear Information System (INIS)

    Caskey, G.R. Jr.

    1981-01-01

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

  18. Rapid heating tensile tests of high-energy-rate-forged 316L stainless steel containing internal helium from radioactive decay of absorbed tritium

    International Nuclear Information System (INIS)

    Mosley, W.C.

    1990-01-01

    316L stainless steel is a candidate material for construction of equipment that will be exposed to tritium. This austenitic stainless steel is frequently used in the high-energy-rate-forged (HERF) metallurgical condition to take advantage of increased strength produced by cold work introduced by this process. Proper design of tritium-handling equipment will require an understanding of how helium-3, the product of radioactive decay of tritium, affects mechanical properties. This report describes results of elevated-temperature tensile testing of HERF 316L stainless steel specimens containing helium concentrations of 171 (calculated) atomic parts per million (appm). Results are compared with those reported previously for specimens containing 0 and 94 (measured) appm helium

  19. LIBRETTO-3: modelling tritium extraction/permeation and evaluation of permeation barriers under irradiation

    International Nuclear Information System (INIS)

    Sedano, L.A.; Fuetterer, M.A.; Viola, R.; Dies, X.

    1996-01-01

    Permeation barriers are required in order to limit the size and cost of the detritiation plants for future fusion reactor blankets of the water-cooled Pb-17Li type. The LIBRETTO irradiations were performed to evaluate the efficiency of permeation barriers under high flux reactor (HFR) conditions. Tritium extraction and permeation characteristics from Pb-17Li under variable temperatures 553-723 K, H 2 doping (0-1 vol%) and purge gas flow rates 20-100 scc/min were tested in LIBRETTO-3. An external TiC coating, an internal (TiC+Al 2 O 3 ), both produced by chemical vapour deposition (CVD), and an internal Al 2 O 3 produced by pack cementation (PC) on AISI 316L steel were tested as permeation barriers. The release mechanisms, experimental uncertainties and method for permeation barriers qualification are presented. As a result permeation reduction factors (PRF) at 0.1 dpa of 17 and 34 were obtained for the CVD-Al 2 O 3 at 498 K and for the PC-Al 2 O 3 at 508 K, respectively. These values were confirmed by a residence time analysis and are higher than in a preliminary analysis. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-04-01

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

  1. Method and device for secure, high-density tritium bonded with carbon

    Science.gov (United States)

    Wertsching, Alan Kevin; Trantor, Troy Joseph; Ebner, Matthias Anthony; Norby, Brad Curtis

    2016-04-05

    A method and device for producing secure, high-density tritium bonded with carbon. A substrate comprising carbon is provided. A precursor is intercalated between carbon in the substrate. The precursor intercalated in the substrate is irradiated until at least a portion of the precursor, preferably a majority of the precursor, is transmutated into tritium and bonds with carbon of the substrate forming bonded tritium. The resulting bonded tritium, tritium bonded with carbon, produces electrons via beta decay. The substrate is preferably a substrate from the list of substrates consisting of highly-ordered pyrolytic graphite, carbon fibers, carbon nanotunes, buckministerfullerenes, and combinations thereof. The precursor is preferably boron-10, more preferably lithium-6. Preferably, thermal neutrons are used to irradiate the precursor. The resulting bonded tritium is preferably used to generate electricity either directly or indirectly.

  2. Tritium monitoring in the GCFR sweep gas fuel irradiation capsule BG-10

    International Nuclear Information System (INIS)

    Gat, U.; Pruitt, M.E.; Longest, A.W.; Epstein, B.D.

    1980-01-01

    The release of tritium and its transport pathways were studied in a vented, pressure-equalized fuel rod which simulated a fuel rod in a Gas-Cooled Fast Reactor (GCFR). The purpose was to determine the fraction of total tritium production transported via the various pathways and to determine its chemical form (tritiated hydrogen or water). It was concluded that the fuel rod and its effluent venting lines retained low concentrations of HT (or T 2 ) and any HTO (or T 2 O) present. However, the addition of 1% hydrogen to the helium carrier gas quantitatively eluted the tritium from the charcoal trap integral to the fuel rod and from the effluent lines. The chemical composition of the tritium arriving at the monitoring system could be determined by means of converters which convert HT to HTO and vice versa. Ht was the dominant species in the samples measured

  3. Damage to E. coli cells induced by tritium decay: secondary lethality under nongrowth conditions

    International Nuclear Information System (INIS)

    Koukalova, B.; Kuhrova, V.

    1980-01-01

    Cells containing incorporated 3 H-thymidine are damaged by its decay. It was found with E.coli TAU-bar cells that a small part of the damage is lethal whereas most of it is reparable and only potentially lethal. If cells are subjected to nongrowth conditions, the potentially lethal damage changes to lethal damage. This process is called secondary lethality (SL). The extent of SL and some changes in DNA under three different modes of growth inhibition were determined. It was found that: (i) SL is maximal under conditions of amino acid starvation (-AA), the viable count decreasing by two orders of magnitude. (ii) SL is 4 times lower in the presence of chloramphenicol (-AA+CLP) and 6.5 times lower under +AA+CLP conditions. Changes in the sedimentation rate of DNA determined in alkaline sucrose gradient correlate with the differences in SL: under -AA conditions the sedimentation rate of DNA decreases whereas in the presence of CLP no decrease occurs. The results suggest that certain enzymatic processes take place under -AA conditions which lead to irreparable changes in DNA. (author)

  4. Damage to E. coli cells induced by tritium decay: secondary lethality under nongrowth conditions

    Energy Technology Data Exchange (ETDEWEB)

    Koukalova, B; Kuhrova, V [Ceskoslovenska Akademie Ved, Brno. Biofysikalni Ustav

    1980-05-01

    Cells containing incorporated /sup 3/H-thymidine are damaged by its decay. It was found with E.coli TAU-bar cells that a small part of the damage is lethal whereas most of it is reparable and only potentially lethal. If cells are subjected to nongrowth conditions, the potentially lethal damage changes to lethal damage. This process is called secondary lethality (SL). The extent of SL and some changes in DNA under three different modes of growth inhibition were determined. It was found that: (i) SL is maximal under conditions of amino acid starvation (-AA), the viable count decreasing by two orders of magnitude. (ii) SL is 4 times lower in the presence of chloramphenicol (-AA+CLP) and 6.5 times lower under +AA+CLP conditions. Changes in the sedimentation rate of DNA determined in alkaline sucrose gradient correlate with the differences in SL: under -AA conditions the sedimentation rate of DNA decreases whereas in the presence of CLP no decrease occurs. The results suggest that certain enzymatic processes take place under -AA conditions which lead to irreparable changes in DNA.

  5. Helium 3 precipitation in AISI 316L stainless steel induced by radioactive decay of tritium: Microstructural study of helium bubble precipitation

    International Nuclear Information System (INIS)

    Brass, A.M.; Chene, J.

    1994-01-01

    The development of the thermonuclear technology has given rise to a renewed interest in the study of the behavior of helium in metals. A great amount of work is still required for the understanding of the role of helium on the mechanical properties of structural materials for fusion technology, especially austenitic stainless steels. This article deals with the study of the influence of thermomechanical heat treatments, aging conditions (temperature and time), and helium concentration of helium bubble precipitation in a 316L austenitic steel. Helium was generated by the radioactive decay of tritium (tritium trick). Helium bubbles impede the grain growth in 316L steel aged at 1,373 K and also the recrystallization reaction at this temperature if cold working is performed prior to aging. Transmission electron microscopy (TEM) observations indicated a weak helium precipitation at 1,073 and 1,223 K, presumably due to the presence of trapping sites for tritium, and no bubble growth after aging up to 100 hours. Precipitation sites are mainly dislocations in the matrix at 1,073 K and grain boundaries and individual dislocations in the matrix at 1,223 K. The large bubble size (50 nm) observed at 1,373 K, even for short aging times (0.083), can partly be attributed to bubble dragging by dislocations toward the grain boundaries. Cold deformation prior to aging leads to a larger bubble size due to growth enhancement during recrystallization. Decreasing the helium content leads to a smaller helium bubble size and density. Tritium trapping at helium bubbles may favor helium 3 accumulation on defects such as grain boundaries, as observed by tritium autoradiography

  6. Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation and in-situ tritium extraction

    International Nuclear Information System (INIS)

    Guo Chunqiu; Xie Jiachun; Liu Xingmin

    2013-01-01

    In-pile irradiation and in-situ tritium extraction experiment is one of associated domestic research projects in ITER special program. According to the technical requirements of in-pile irradiation experiment of helium cooled ceramic breeder (ceramic) pebble bed assembly in a research reactor, the feasibility of the design for the in-pile irradiation and in-situ tritium extraction experiment of ceramic pebble bed assembly was evaluated. By conducting thermal-hydraulic design calculation with different in-pile irradiation channels, locations and structure parameters for ceramic pebble bed assembly, a reasonable design scheme of ceramic pebble bed assembly satisfying the design requirements for in-pile irradiation was obtained. (authors)

  7. Simulation of alpha decay of actinides in iron phosphate glasses by ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dube, Charu L., E-mail: dubecharu@gmail.com; Stennett, Martin C.; Gandy, Amy S.; Hyatt, Neil C.

    2016-03-15

    Highlights: • Alpha decay of actinides in iron phosphate glasses is simulated by employing ion irradiation technique. • FTIR and Raman spectroscopic measurements confirm modification of glass network. • The depolymerisation of glass network after irradiation is attributed to synergetic effect of nuclear and electronic losses. - Abstract: A surrogate approach of ion beam irradiation is employed to simulate alpha decay of actinides in iron phosphate nuclear waste glasses. Bismuth and helium ions of different energies have been selected for simulating glass matrix modification owing to radiolysis and ballistic damage due to recoil atoms. Structural modification and change in coordination number of network former were probed by employing Reflectance Fourier-Transform Infrared (FT-IR), and Raman spectroscopies as a consequence of ion irradiation. Depolymerisation is observed in glass sample irradiated at intermediate energy of 2 MeV. Helium blisters of micron size are seen in glass sample irradiated at low helium ion energy of 30 keV.

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

    International Nuclear Information System (INIS)

    Dolle, L.; Bazin, J.

    1977-01-01

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

  9. Control of postharvest decay in fruits and vegetables by irradiation

    International Nuclear Information System (INIS)

    Ismail, F.A.; Afifi, S.A.

    1976-01-01

    The effect of postharvest gamma-irradiation of strawberries and carrots on their organoleptic quality, storage capacity and chemical composition was investigated. Radiation doses of 25 krad for strawberries and 75 krad for carrots were chosen as optimum doses not causing significant changes in the organoleptic quality. It was found that strawberries and carrots irradiated with a dose of 25 and 75 krad, respectively, retained its microbiological stability even during storage at 12 0 C (strawberries) and at 25 - 30 0 C (carrots), respectively. It was also found that changes in the chemical composition depend on storage temperature and time and are not affected substantially by irradiation. (author)

  10. Photoproduction of tritium

    International Nuclear Information System (INIS)

    Becker, J.A.; Anderson, J.D.; Weiss, M.S.

    1995-01-01

    3 H (Tritium) is required for maintenance of nuclear weapons in the stockpile. The National Defense need for 3 H was historically met by the Savannah River Facility. This facility is no longer safe for operation. 3 H decays with a mean lifetime τ = 17.8 y, and therefore new methods of 3 H production are required to meet US military requirements. Irradiation of 7 Li by low-energy photons produces tritium ( 3 H) via the photodisintegration process. Waste heat from the 7 Li target can be extracted and used for the direct generation of electricity. Other advantages include: negligible residual radioactivity, simple target technology, small low-energy electron accelerators for bremsstrahlung production (the photon source), developed liquid metal technology, modularity, simple extraction of 3 H from a recirculating 7 Li target, abundant supply of 7 Li, and straightforward target-accelerator-bremsstrahlung converter interface. A schematic plant characterized by very low risk is described, and a figure-of-merit is obtained

  11. Kinetics of free radical decay reactions in lactic acid homo and copolymers irradiated to sterilization dose

    International Nuclear Information System (INIS)

    Kantoglu, O.; Ozbey, T.; Gueven, O.

    1995-01-01

    The kinetics of free radical decay reactions of poly(L-Lactic acid), poly(DL-Lactic acid) and random copolymer of lactic and glycolic acid have been investigated for decays taking place in air and in vacuum. The change in ESR lines of γ-irradiated polymers have been followed over a long time period. The decay has been found to follow neither simple first-order nor second-order kinetics. Various kinetic approaches including composite first or second-order mechanisms and diffusion-controlled first or second-order equations were determined to be also unsatisfactory. The decay of radicals in bulk irradiated lactic acid homo and copolymers was found to be best described when the second-order non-classical equation with time dependent rate constant approach was used. (Author)

  12. Systematic preservation of teeth and prophylaxis of dental decay on irradiated patients

    International Nuclear Information System (INIS)

    Horiot, J.C.; Schraub, S.

    1975-01-01

    200 patients were irradiated on the oral cavity and on salivary glands with doses of 3,000Rads and up. Teeth were preserved. A prophylactic treatment of dental decay was made using 5 minutes daily a fluoride gel into a carrier. The results showed an incidence of 5% for dental decay and 1% for osteo-radionecroses. Details are given about the selection of patients, the dental work-up and dental care before, during and after radiation therapy [fr

  13. Systematic preservation of teeth and prophylaxis of dental decay on irradiated patients. Preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Horiot, J C [Centre de Lutte contre le Cancer Georges-Francois-Leclerc, 21 - Dijon (France); Schraub, S [Centre Hospitalier Universitaire, 25 - Besancon (France)

    1975-11-01

    200 patients were irradiated on the oral cavity and on salivary glands with doses of 3,000 Rads and up. Teeth were preserved. A prophylactic treatment of dental decay was made using 5 minutes daily a fluoride gel into a carrier. The results showed an incidence of 5% for dental decay and 1% for osteo-radionecroses. Details are given about the selection of patients, the dental work-up and dental care before, during and after radiation therapy.

  14. Analysis of tritium production in concentric spheres of oralloy and 6LiD irradiated by 14-MeV neutrons

    International Nuclear Information System (INIS)

    Fawcett, L.R. Jr.; Roberts, R.R. II; Hunter, R.E.

    1988-03-01

    Tritium production and activation of radiochemical detector foils in a sphere of 6 LiD with an oralloy core irradiated by a central source of 14-MeV neutrons have been calculated and compared with experimental measurements. The experimental assembly consisted of an oralloy sphere surrounded by three solid 6 LiD concentric shells with ampules of 6 LiH and 7 LiH and activation foils located in several positions throughout the assembly. The Los Alamos Monte Carlo Neutron Photon Transport Code (MCNP) was used to calculate neutron transport throughout the system, tritium production in the ampules, and foil activation. The overall experimentally observed-to-calculated ratios of tritium production were 0.996 +- 2.5% in 6 Li ampules and 0.903 +- 5.2% in 7 Li ampules. Observed-to-calculated ratios for foil activation are also presented. 11 refs., 4 figs., 7 tabs

  15. Improved limit on the mass of ν/sub e/ from the beta decay of molecular tritium

    International Nuclear Information System (INIS)

    Bowles, T.J.; Friar, J.L.; Robertson, R.G.H.; Stephenson, G.J. Jr.; Wark, D.L.; Wilkerson, J.F.; Knapp, D.A.

    1989-01-01

    We report a new upper limit of 13.4 eV (95% confidence level) on the mass of the electron antineutrino from a study of the shape of the beta spectrum of free molecular tritium. This result appears to be inconsistent with a reported value for the mass of 26(5) eV. The electron neutrino is evidently not massive enough to close the universe by itself. 23 refs., 1 fig., 2 tabs

  16. Improved limit on the mass of bar νe from the beta decay of molecular tritium

    International Nuclear Information System (INIS)

    Bowles, T.J.; Robertson, R.G.H.; Wark, D.L.; Wilkerson, J.F.; Stephenson, G.J.; Friar, J.L.; Knapp, D.A.

    1990-01-01

    We report a new upper limit of 13.4 eV (95% confidence level) on the mass of the electron antineutrino from a study of the shape of the beta spectrum of free molecular tritium. This result appears to be inconsistent with a reported value for the mass of 26(5) eV. The electron neutrino is evidently not massive enough to close the universe by itself. 21 refs., 1 fig., 2 tabs

  17. Calculation of radioactive decay chains produced by neutron irradiation

    International Nuclear Information System (INIS)

    Ponti, C.

    1984-01-01

    The calculation of the decay chains of radioactive isotopes may be solved by different computer codes that apply similar numerical methods. Analytical forms of solution of this problem are known since a long time, but their implementation has always been hindered by the possible occurrence of large numerical errors or by the need, to avoid them, of very high numerical precision. In this paper the reason of these difficulties has been pointed out and solved. In particular a new family of mathematical identities has been found

  18. Influence of hot water dip and gamma irradiation on postharvest fungal decay of Galia melons

    International Nuclear Information System (INIS)

    Barkai-Golan, R.; Padova, R.; Ross, I.; Lapidot, M.; Copel, A.; Davidson, H.

    1993-01-01

    Dipping Galia melons in hot water at 52 deg C for 5 min or at 55 deg for 2 min resulted in 12-15% decay (caused by Alternaria alternata, Fusarium spp. and Trichothecium roseum) during prolonged storage (12 d at 6 deg plus 3 d at 18 deg ) compared with 75% decay in untreated fruit or 60% decay in cold-water-dipped fruit. Irradiation at 0.5 or 1 kGy had no significant effect on decay development. However, combination of heat treatment with a 0.5 kGy dose prevented fungal growth, resulting in 5% decay during storage. Combinations of heating with 1 kGy irradiation gave no improvement in anti-fungal effect over treatment with 0.5 kGy and sometimes resulted in a decreased suppressive effect. Reducing the duration of dipping at 55 deg from 2 to 0.5 min, applied alone or in combination with irradiation, considerably reduced the anti-fungal effect of the treatment. The effective combined treatment resulted in 12-15% of slight peel damage, but all the fruits were regarded as marketable. No differences in fruit firmness were recorded among the treatments

  19. Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

    International Nuclear Information System (INIS)

    Le-Guillou, Mael

    2014-01-01

    In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

  20. Evolution of nuclear weapon-produced tritium and its decay product He-3 in the Mediterranean Sea, 1952-2011

    Energy Technology Data Exchange (ETDEWEB)

    Roether, Wolfgang [IUP Univ. Bremen (Germany)

    2015-07-01

    After 1952, tritium in the Mediterranean Sea (Med) rose 100-fold up to a 30 TU peak in 1965 (1 TU = [H-3/H] . 10{sup 18}) and thereafter declined to about 1 TU in 2011. This resulted in a strong supply of its stable daughter He-3, concentrated in the 1960s. Terrigenic He being low in He-3 due to a predominantly crustal origin allowed precise determination of the tritiugenic He-3 (± 0.7% in delta He-3). The highest He-3 was found off the Tyrrhenian Sea in 1983 (delta He-3 ∝ 16%). The principle input into the subsurface waters occurs in the Eastern Med by way of the Levantine Intermediate Water (LIW), which moves as a high-salinity core westward toward the Strait of Gibraltar. The observations demonstrate strong longitudinal flow dispersion. As tritium-He-3 ages are biased toward times of strong input, one finds an age increase with calendar year of observation, attributed to fast flow contributions showing up first. A realistic estimate for transfer from the LIW source up to the Western Med is 22 years. Especially in the East, He-3 lately decreased distinctly, in response to the reduced He-3 generation.

  1. Analysis of decay of radicals induced in irradiated foods during long storage

    International Nuclear Information System (INIS)

    Kishita, Keigo; Kawamura, Shoei; Nakamura, Hideo; Ukai, Mitsuko; Kikuchi, Masahiro; Kobayashi, Yasuhiko

    2013-01-01

    By electron spin resonance (ESR) spectroscopy, we revealed free radicals in γ-ray irradiated foods ; black pepper, green coffee bean, cereal flour and ginseng. We also analyzed the decay behavior of radiation induced free radicals during long storage. The ESR spectrum of experimental irradiated foods consists of a sextet signal centered at g=2.0 and a singlet signal at the same g-value position and a singlet signal at g=4.0. The ESR spectrum of the cereal flour sample showed only singlet signal at g=2.0. The singlet signal at g=2.0 is originated from organic free radicals and its peak intensity showed the dependence of γ-ray radiation dose levels. But the signal intensity was decreased during storage. Only after 3 hours of radiation treatment the peak intensity was decreased fast and after that the intensity was decreased slowly. The analysis of radical decay process using the simulation methods based on the theory of reaction speed, the three decay behavior was showed. It is considered that at least three or more kinds of radicals were induced in irradiated foods and in decay during long time storage. (author)

  2. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1990-01-01

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

  3. Tritium accounting for PHWR plants

    International Nuclear Information System (INIS)

    Nair, P.S.; Duraisamy, S.

    2012-01-01

    Tritium, the radioactive isotope of hydrogen, is produced as a byproduct of the nuclear reactions in the nuclear power plants. In a Pressurized Heavy Water Reactor (PHWR) tritium activity is produced in the Heat Transport and Moderator systems due to neutron activation of deuterium in heavy water used in these systems. Tritium activity build up occurs in some of the water systems in the PHWR plants through pick up from the plant atmosphere, inadvertent D 2 O ingress from other systems or transfer during processes. The tritium, produced by the neutron induced reactions in different systems in the reactor undergoes multiple processes such as escape through leaks, storage, transfer to external locations, decay, evaporation and diffusion and discharge though waste streams. Change of location of tritium inventory takes place during intentional transfer of heavy water, both reactor grade and downgraded, from one system to another. Tritium accounting is the application of accounting techniques to maintain knowledge of the tritium inventory present in different systems of a facility and to construct activity balances to detect any discrepancy in the physical inventories. It involves identification of all the tritium hold ups, transfers and storages as well as measurement of tritium inventories in various compartments, decay corrections, environmental release estimations and evaluation of activity generation during the accounting period. This paper describes a methodology for creating tritium inventory balance based on periodic physical inventory taking, tritium build up, decay and release estimations. Tritium accounting in the PHWR plants can prove to be an effective regulatory tool to monitor its loss as well as unaccounted release to the environment. (author)

  4. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2015-06-15

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

  5. Concentration of Tritium and Members of the Uranium and Thorium Decay Chains in Groundwaters in Slovenia and their Implication for Managing Groundwater Resources

    Energy Technology Data Exchange (ETDEWEB)

    Korun, M.; Kovacic, K.; Kozar-Logar, J. [Jozef Stefan Institute, Ljubljana (Slovenia)

    2013-07-15

    Samples of groundwater were measured in terms of their activity concentration of gamma ray emitters, members of the uranium and thorium decay chains and tritium. The distributions of the number of samples over the measured activity concentrations are presented for {sup 238}U, {sup 226}Ra, {sup 210}Pb, {sup 228}Ra, {sup 228}Th, {sup 40}K and {sup 3}H. The distributions have three distinct shapes: log-normal distributions ({sup 238}U, {sup 226}Ra, {sup 228}Ra, {sup 228}Th), bimodal distributions ({sup 210}Pb, {sup 40}K), and a normal distribution ({sup 3}H). It appears that the log-normal distributions reflect the dilution of the radionuclides dissolved in the water. The bimodal distributions, being the sum of a log-normal distribution and a distribution having its maximum at the activity concentration of the higher mode, indicate influences from the soil surface, e.g. washout from the atmosphere and fertilizing. The normal distribution indicates mixing with rainwater under circumstances that are characterized by several independent variable parameters. (author)

  6. An assembly of tritium production experiment

    International Nuclear Information System (INIS)

    Abe, Toshihiko

    1981-01-01

    An assembly for tritium production experiment, i.e. Tritium Extraction System (TREX) constructed as a small scale test facility for tritium production, and Tritium Removal System (TRS) attached to TREX, and the preliminary results of the experiments with them are described. The radiological safety of the process and operation is also an important consideration. Lithium-aluminum alloy was selected as the most promising target material. The following matters are involved in the scope of production technology: the selection of a target material and target preparation, reactor irradiation, the construction of a facility for the extraction of tritium from the irradiated target, the establishment of the optimum conditions of extraction, the purification, collection and storage of tritium, and the inspection of the product. The tritium production experiment at JAERI is yet on the initial stage; the development is to be continued with the stepwise increase of the scale of tritium production. (J.P.N.)

  7. Study on tritium recovery from breeder materials

    International Nuclear Information System (INIS)

    Moriyama, H.; Moritani, K.

    1997-01-01

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

  8. Tritium conference days; Journees tritium

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  9. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  11. Tritium monitor and collection system

    Science.gov (United States)

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

    1992-01-14

    This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber. Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next on-line getter is valved on. Simultaneously, the first getter is heated and a pure helium purge is employed to carry the tritium from the getter. The tritium loaded gas stream is then routed through an ion chamber which measures the tritium activity. The ion chamber effluent passes through a collection getter that readsorbs the tritium and is removable from the system once it is loaded and is then replaced with a clean getter. Prior to removal of the collection getter, the system switches to a parallel collection getter. The effluent from the collection getter passes through a guard getter to remove traces of tritium prior to exiting the system. The tritium loaded collection getter, once removed, is analyzed by liquid scintillation techniques. The entire sequence is under computer control except for the removal and analysis of the collection getter. 7 figs.

  12. Quantitative analysis of tritium distribution in austenitic stainless steels welds

    International Nuclear Information System (INIS)

    Roustila, A.; Kuromoto, N.; Brass, A.M.; Chene, J.

    1994-01-01

    Tritium autoradiography was used to study the tritium distribution in laser and arc (TIG) weldments performed on tritiated AISI 316 samples. Quantitative values of the local tritium concentration were obtained from the microdensitometric analysis of the autoradiographs. This procedure was used to map the tritium concentration in the samples before and after laser and TIG treatments. The effect of the detritiation conditions and of welding on the tritium distribution in the material is extensively characterized. The results illustrate the interest of the technique for predicting a possible embrittlement of the material associated with a local enhancement of the tritium concentration and the presence of helium 3 generated by tritium decay. ((orig.))

  13. Tritium sources

    International Nuclear Information System (INIS)

    Glodic, S.; Boreli, F.

    1993-01-01

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

  14. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1989-01-01

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

  15. Gamma irradiation, hot water and imazalil treatments on decay organisms and physical quality of stored netted muskmelon fruit

    International Nuclear Information System (INIS)

    Lester, G.

    1989-01-01

    Nonchemical treatments of gamma irradiation (2 Kilograys) and hot water (57°C) and the fungicide imazalil (1000 ppm) were compared with and without shrink-film wrap for effects on decay and physical quality of netted muskmelon fruit stored at 4°C for 0 through 60 days. Gamma irradiation was ineffective in controlling decay and surface molds, and injurious to physical quality by decreasing firmness, increasing fresh weight loss, membrane leakage and vein track browning. Hot water treatment coupled with shrink-film wrap was effective in controlling decay activity and maintaining physical quality up to 20 days storage. Imazalil coupled with shrink-film wrap controlled the incidence and severity of decay and maintained fruit firmness, moisture loss, membrane permeability and vein track browning for almost 60 days storage. (author)

  16. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

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

  17. Irradiation behaviour of a tritium breeding material, γ-LiAlO 2- results of two in-pile experiments: ALICE I and ALICE II

    Science.gov (United States)

    Botter, F.; Rasneur, B.; Roth, E.

    1988-11-01

    γ-LiAlO 2 has been studied at CEA as potential breeder material for fusion reactors within the scope of the EEC fusion technology program. Radiation damage was investigated by irradiating unclad aluminate samples in the core of the OSIRIS reactor at Saclay. As part of the international breeder material comparison program named BEATRIX, US samples were irradiated along with those prepared in Saclay; samples of natural 6Li content and 96% enriched ones were irradiated. Shapes were chosen to enable postirradiation examinations (PIE), and microstructures were optimized for tritium release. The ALICE 1 experiment was carried out during 25.7 full power days (FPD), ALICE II lasted 36.3 FPD. Temperatures ranged from 400 to 600°C in the first, from 750 to 850°C in the second ALICE irradiation (sample core temperatures). In both cases the maximum flux on the samples was 2.1 × 10 18n m -2 s -1 fast, and 0.7 × 10 18n m -2 s -2 thermal Power dissipated was up to 100 W/cm 3, higher than the average in most reactor blanket designs by a factor 3 to 10, thus enabling the highest burn-ups to correspond to more than two years of possible operation in a full-scale reactor. In the lower temperature range of irradiation no significant damage was observed. In the higher one shrinkage due to sintering was induced. Whatever the microstructure, the flux and temperature, all samples (but one) not exceeding 5 mm diameter and length were mechanically intact. Above those dimensions cracking, which can be assigned to excessive thermal stress, could be observed. Given anticipated operating conditions of blankets being designed, the behaviour of γ-LiAlO 2 under irradiation is that of a very promising material.

  18. Optimization of irradiation decay and counting times in nuclear activation analysis using short-lived nuclides

    International Nuclear Information System (INIS)

    Bjoernstad, T.

    This work describes a method and outlines a procedure for optim- ization of an activation analysis with respect to the experimental times, irradiation time, t(subi), decay time and counting time. The method is based on the 'minimum relative standard deviation criterion', and specially designed for the use on short-lived nuclides. A computer program, COMB1, is written in the BASIC language in order to make the calculations easier and faster. It is intended to be understandable, and easily applicable on a computer of modest size. Time and cost are important factors, especially for routine analysis on a service basis. In such cases one can often allow a controlled reduction in the analysis quality (through a higher relative standard deviation). The procedure outlined can therefore help find acceptable conditions by calculation of the 'best practical' (or reasonable) experimental time values, and the minimum number of accumulation cycles necessary to fulfil the requirements given. (Auth.)

  19. DP-THOT - a calculational tool for bundle-specific decay power based on actual irradiation history

    International Nuclear Information System (INIS)

    Johnston, S.; Morrison, C.A.; Albasha, H.; Arguner, D.

    2005-01-01

    A tool has been created for calculating the decay power of an individual fuel bundle to take account of its actual irradiation history, as tracked by the fuel management code SORO. The DP-THOT tool was developed in two phases: first as a standalone executable code for decay power calculation, which could accept as input an entirely arbitrary irradiation history; then as a module integrated with SORO auxiliary codes, which directly accesses SORO history files to retrieve the operating power history of the bundle since it first entered the core. The methodology implemented in the standalone code is based on the ANSI/ANS-5.1-1994 formulation, which has been specifically adapted for calculating decay power in irradiated CANDU reactor fuel, by making use of fuel type specific parameters derived from WIMS lattice cell simulations for both 37 element and 28 element CANDU fuel bundle types. The approach also yields estimates of uncertainty in the calculated decay power quantities, based on the evaluated error in the decay heat correlations built-in for each fissile isotope, in combination with the estimated uncertainty in user-supplied inputs. The method was first implemented in the form of a spreadsheet, and following successful testing against decay powers estimated using the code ORIGEN-S, the algorithm was coded in FORTRAN to create an executable program. The resulting standalone code, DP-THOT, accepts an arbitrary irradiation history and provides the calculated decay power and estimated uncertainty over any user-specified range of cooling times, for either 37 element or 28 element fuel bundles. The overall objective was to produce an integrated tool which could be used to find the decay power associated with any identified fuel bundle or channel in the core, taking into account the actual operating history of the bundles involved. The benefit is that the tool would allow a more realistic calculation of bundle and channel decay powers for outage heat sink planning

  20. Radiolytic unsaturation decay in polyethylene. Part II--the effect of irradiation temperature, thermal history and orientation

    International Nuclear Information System (INIS)

    Lyons, Bernard J.

    2004-01-01

    The decay rate of vinyl unsaturation in high-density polyethylenes irradiated at temperatures from about 310 to 450 K, changes significantly in the melting range up to the crystalline melting point as does free radical mobility and the polymer crystallinity. However, orienting the polymer, or slow cooling or quenching from the melt, prior to irradiation, do not alter the decay process or its rate, although they do alter the rate of increase of insoluble gel and of elastic modulus in the molten state. It is suggested that, below ∼340 K, the marked deviations from a first-order decay result from the limited mobility of polymeric free radicals in the crystalline phase and from scavenging, by vinyl groups, segregated into the amorphous phase, of radiolytic hydrogen atoms (H). In the melting range, the mobility of polymeric free radicals increases as the crystallinity decreases, reducing the importance of scavenging, so vinyl decay approximates more closely to a first-order relation. In the melt, the vinyl decay relation is not changed qualitatively by H atom scavenging, but the effective vinyl concentration is lower, so the decay rate drops sharply

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

    International Nuclear Information System (INIS)

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

    1990-02-01

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

  2. Preliminary decay heat calculations for the fuel loaded irradiation loop device of the RMB multipurpose Brazilian reactor

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel; Costa, Antonio Carlos L. da; Andrade, Edison P., E-mail: campolina@cdtn.br, E-mail: aclp@cdtn.br, E-mail: epa@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2017-07-01

    The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work preliminary neutronic calculations have been carried out in order to determine parameters to the cooling system of the Loop basic design. The heat released as a result of radioactive decay of fuel samples was calculated using ORIGEN-ARP and it resulted less than 200 W after 1 hour of irradiation interruption. (author)

  3. Experiments on tritium behavior in beryllium, (1)

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi; Ishizuka, Etsuo; Matsumoto, Mikio; Inada, Seiji; Sezaki, Katsuji; Saito, Minoru; Kato, Mineo.

    1989-06-01

    In JMTR, it was observed that the tritium concentration of the primary coolant increases with the reactor operation at 50 MW. As one of the tritium generation sources, we paid attention to a neutron reflector made of beryllium because the tritium generation rate in the beryllium is bigger than other components in the reactor core. On the other hand, the irradiation test of blanket materials (i.e. tritium breeding materials and neutron multipling materials) are planned for development of the fusion reactor in JMTR and the beryllium will be also irradiated as a neutron multiplier with tritium breeding materials. Therefore, as the irradiated specimens, we used a hot-pressed beryllium disk fabricated by the same method as the neutron reflector or the neutron multiplier and conducted the irradiation tests in JMTR. The purpose of these tests are to clarify the tritium behavior in the hot-pressed beryllium. In this paper, from a viewpoint of the fabrication of capsules for neutron irradiation, the specifications of the irradiated specimens and capsules are summarized. Additionally, the results on the puncture test of the container of the irradiation specimens are described. (author)

  4. Tritium immobilisation

    International Nuclear Information System (INIS)

    Bridger, N.J.

    1982-01-01

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

  5. 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)

  6. Neutronics experiments for uncertainty assessment of tritium breeding in HCPB and HCLL blanket mock-ups irradiated with 14 MeV neutrons

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Pillon, M.; Villari, R.; Fischer, U.; Klix, A.; Leichtle, D.; Kodeli, I.; Pohorecki, W.

    2012-01-01

    Two neutronics experiments have been carried out at 14 MeV neutron sources on mock-ups of the helium cooled pebble bed (HCBP) and the helium cooled lithium lead (HCLL) variants of ITER test blanket modules (TBMs). These experiments have provided an experimental validation of the calculations of the tritium production rate (TPR) in the two blanket concepts and an assessment of the uncertainties due to the uncertainties on nuclear data. This paper provides a brief summary of the HCPB experiment and then focuses in particular on the final results of the HCLL experiment. The TPR has been measured in the HCLL mock-up irradiated for long times at the Frascati 14 MeV Neutron Generator (FNG). Redundant and well-assessed experimental techniques have been used to measure the TPR by different teams for inter-comparison. Measurements of the neutron and gamma-ray spectra have also been performed. The analysis of the experiment, carried out by the MCNP code with FENDL-2.1 and JEFF-3.1.1 nuclear data libraries, and also including sensitivity/uncertainty analysis, shows good agreement between measurements and calculations, within the total uncertainty of 5.9% at 1σ level. (paper)

  7. Tritium Storage Material

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  8. Direct measurement of tritium production rate in LiPb with removed parasitic activities: Preliminary experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kuc, Tadeusz, E-mail: kuc@agh.edu.pl; Pohorecki, Władysław; Ostachowicz, Beata

    2014-10-15

    Liquid scintillation (LS) technique applied to direct measurement of tritium activity produced in LiPb eutectic in Frascati HCLL TBM mock-up neutronic experiment has been tested so far in the case of LS measurement after long period since irradiation. LiPb samples irradiated in neutron filed show, except of tritium, meaningful activity of other radioisotopes (parasitic). Parasitic activity, mainly from isotopes of lead ({sup 209}Pb, {sup 204m}Pb, {sup 203}Pb) calculated with the use of FISPACT, exceeds ca 5 times tritium activity 1.4 h after irradiation. We propose to remove disturbing radioisotopes in a chemical way to avoid long “cooling” of the irradiated samples before tritium measurement. Samples (1 g of LiPb) irradiated in reactor fast neutron flux were diluted and metallic cations removed by chemical precipitation. For this purpose we used: potassium iodide (KJ), strontium chloride (SrCl{sub 2}), APDC (C{sub 5}H{sub 8}NS{sub 2}·NH{sub 4}), NaDDTC (C{sub 5}H{sub 10}NNaS{sub 2}·3H{sub 2}O), and PAN (C{sub 15}H{sub 11}N{sub 3}O). Precipitation procedure in each case lasted ca 5–25 min, and the following filtration next 10–20 min. In each filtrate (ca 120 ml) we measured Pb concentration in total reflection X-ray fluorescence (TXRF) analyzer and parasitic activity (left after 21-day “cooling”) applying HPGe gamma spectrometer. Pb cations precipitated by SrCl{sub 2} and than by PAN lowered activity of Pb isotopes to less than 1% of the initial tritium activity. Another combination of reagents: NaDDTC followed by SrCl{sub 2} in a single and double step filtration reduced Pb concentration 10{sup 2} and 10{sup 4} times, respectively. Reduction of this order allows tritium radiometric measurement ca 3 h after irradiation with acceptable accuracy. This time can be shortened by applying correction for decay of known parasitic activity. Input of {sup 76}As and other less abundant radioisotopes can be eliminated using high purity LiPb. Tritium activity of

  9. Management of Tritium in ITER Waste

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  10. Code ACTIVE for calculation of the transmutation, induced activity and decay heat in neutron irradiation

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Harada, Yuhei; Asami, Naoto.

    1976-03-01

    The computer code ACTIVE has been prepared for calculation of the transmutation rate, induced activity and decay heat. Calculations are carried out with activation chain and spatial distribution of neutron energy spectrum. The spatial distribution of secondary gamma-ray source due to the unstable nuclides is also obtainable. Special attension is paid to the short life decays. (auth.)

  11. Toxicity of tritium

    International Nuclear Information System (INIS)

    Dobson, R.L.

    1979-01-01

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

  12. A computer program to calculate nuclide yields in complex decay chain for selection of optimum irradiation and cooling condition

    International Nuclear Information System (INIS)

    Takeda, Tsuneo

    1977-11-01

    This report is prepared as a user's input manual for a computer code CODAC-No.5 and provides a general description of the code and instructions for its use. The code represents a modified version of the CODAC-No.4 code. The code developed is capable of calculating radioactive nuclide yields in an any given complex decay and activation chain independent of irradiation history. In this code, eighteen kinds of valuable tables and graphs can be prepared for output. They are available for selection of optimum irradiation and cooling conditions and for other intentions in accordance with irradiation and cooling. For a example, the ratio of a nuclide yield to total nuclide yield depending on irradiation and cooling times is obtained. In these outputs, several kinds of complex and intricate equations and others are included. This code has almost the same input forms as that of CODAC-No.4 code excepting input of irradiation history data. Input method and formats used for this code are very simple for any kinds of nuclear data. List of FORTRAN statements, examples of input data and output results and list of input parameters and its definitions are given in this report. (auth.)

  13. Tritium transfer studies in cellulose-HTO system

    International Nuclear Information System (INIS)

    Jayaraman, A.P.; Misra, B.M.

    1986-01-01

    This paper describes some aspects of studies on transfer of tritium to cellulose from tritiated water at six different specific activities and discusses the generalized tritiation pattern. Cellulose was irradiated in steps to 10 M Rads and the tritium transfer was determined at each stage. Experimental results signify substantial increase of tritiation in cellulose at higher dose of irradiation. (author). 8 refs

  14. Tritium target performance during an LBLOCA in a PWR

    International Nuclear Information System (INIS)

    Reid, B.D.

    1996-01-01

    In December 1995, the U.S. Department of Energy (DOE) announced a preferred strategy for acquiring a new supply of tritium. That strategy is based on pursuing the two most promising production alternatives. These alternatives include either constructing an accelerator-produced tritium system for tritium production or procuring an existing commercial light water reactor or irradiation services from such a reactor to irradiate tritium targets. This paper discusses the safety performance of a tritium target in a commercial pressurized water reactor (PWR). The current conceptual design for the light water tritium targets is quite similar, in terms of external dimensions and materials, to early designs for stainless steel clad discrete burnable absorbers used in PWRs. The tritium targets nominally consist of an annular lithium aluminate pellet wrapped in a Zircaloy-4 getter and clad with Type 316 stainless steel

  15. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  16. Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

    Energy Technology Data Exchange (ETDEWEB)

    Bourganel, St

    2002-11-01

    After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

  17. Removal of contaminating tritium and tritium pressure measurement by a secondary electron multiplier

    International Nuclear Information System (INIS)

    Ichimura, K.; Watanabe, K.; Nishizawa, K.; Fujita, J.

    1984-01-01

    A ceramic secondary electron multiplier (SEM), Ceratron, was used to study impairment of the SEM performance due to adsorbed tritium, its decontamination, and the applicability of the SEM to measure tritium pressure. The background level of the SEM increased significantly, up to its counting limit, due to tritium adsorption. Heating it to 300 0 C in vacuo and/or in the presence of reactive gases such as D 2 and CO at 1 x 10 -4 Pa was not effective to decontaminate the SEM, whereas photon irradiation was extremely powerful for the decontamination. The tritium (HT) pressure in a range of 1 x 10 -6 - 1 x 10 -3 Pa could be measured with no significant impairment of the SEM performance with the aid of photon irradiation. It is revealed that a particle flux as low as 1 particle/s will be able to measure in the presence of tritium if suitable photon sources are installed in the systems. (orig.)

  18. Magmatic tritium

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  19. Tritium, carbon-14, and iodine-129 as indicators for localized vertical recharge along an anticline in the Columbia River Basalts using a decay-corrected mixing model

    Energy Technology Data Exchange (ETDEWEB)

    Hall, S.H.; Johnson, V.G.; Early, T.O.

    1987-11-01

    Tritium, /sup 14/C, and /sup 129/I in groundwater samples are used to demonstrate vertical recharge and measure flow velocity in the fractured and faulted Umtanum Ridge-Gable Mountain acticline, within the Columbia River Basalts, at a sampling site about 6 mi northeast of the proposed high-level nuclear waste repository at the Hanford Site, Washington State. Mixing model calculations yield an apparent downward migration rate of 15 to 19 ft/yr through a sequence of aquifers in the Wanapum Basalt that range in depth from 698 to 1373 ft. Estimates of the vertical flow rate in the overlying Saddle Mountains Basalt are somewhat higher. Hydrographs from neighboring wells, hydrostatic heads, pump test data, and the chemical composition of groundwater samples from the sampling well are consistent with interaquifer communication. Some hydrologic evidence from aquifers in this region suggests that, in the past, flow may have been upward. This possible reversal of flow may be associated with water table mounding in the unconfined aquifer, caused by waste disposal activities at the Hanford Site since World War II. 17 refs., 12 figs., 3 tabs.

  20. Tritium, carbon-14, and iodine-129 as indicators for localized vertical recharge along an anticline in the Columbia River Basalts using a decay-corrected mixing model

    International Nuclear Information System (INIS)

    Hall, S.H.; Johnson, V.G.; Early, T.O.

    1987-11-01

    Tritium, 14 C, and 129 I in groundwater samples are used to demonstrate vertical recharge and measure flow velocity in the fractured and faulted Umtanum Ridge-Gable Mountain acticline, within the Columbia River Basalts, at a sampling site about 6 mi northeast of the proposed high-level nuclear waste repository at the Hanford Site, Washington State. Mixing model calculations yield an apparent downward migration rate of 15 to 19 ft/yr through a sequence of aquifers in the Wanapum Basalt that range in depth from 698 to 1373 ft. Estimates of the vertical flow rate in the overlying Saddle Mountains Basalt are somewhat higher. Hydrographs from neighboring wells, hydrostatic heads, pump test data, and the chemical composition of groundwater samples from the sampling well are consistent with interaquifer communication. Some hydrologic evidence from aquifers in this region suggests that, in the past, flow may have been upward. This possible reversal of flow may be associated with water table mounding in the unconfined aquifer, caused by waste disposal activities at the Hanford Site since World War II. 17 refs., 12 figs., 3 tabs

  1. Comparison of the effect of gamma irradiation, heat-radiation combination, and sulphur dioxide generating pads on decay and quality of grapes

    International Nuclear Information System (INIS)

    Thomas, Paul; Brij Bhushan; Joshi, M.R.

    1995-01-01

    Effect of gamma irradiation, heat-radiation combination and in-package sulphur dioxide fumigation on fungal decay, and quality of seedless grape cultivars, Thompson, Sonaka and Tas-A-Ganesh was evaluated under different storage regimes. Irradiation at 2 kGy or a combination of hot water dip (50 degC, 5 min), plus irradiation (1 kGy) showed less spoilage due to Rhizopus spp. and Botrytis spp. in grapes packaged in tissue paper lined boxes and stored at 4 deg, 15 deg and 25 degC. Storage in polyethylene lined boxes increased the fungal rot. In-package sulphur dioxide generating pad was most effective for control of decay in polyethylene lined boxes stored at 10 deg and 20 degC, but caused berry bleaching. Irradiation at 2.5 or 3.5 kGy controlled decay at 10 degC, but not so effectively at 20 degC. Organoleptic quality, berry firmness, and soluble solids were not affected by irradiation, but decreases in titratable acids and ascorbic acid were recorded. Packaging in polyethylene lined boxes retained berry turgidity, while slight shrivelling occurred in tissue paper lined boxes. The results indicate that gamma irradiation has potential as an alternative to sulphur dioxide fumigation for decay control during shipping and storage. (author). 20 refs., 4 tabs

  2. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

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

  3. Decay of organic free radicals in γ-ray irradiated pepper during thermal treatment as detected by electron spin resonance spectroscopy

    International Nuclear Information System (INIS)

    Ukai, Mitsuko; Shimoyama, Yuhei

    2004-01-01

    Using electron spin resonance (ESR) spectroscopy we analysed the thermal decay process of radicals in γ-Irradiated pepper Upon irradiation, the satellite signals were newly induced and appeared at the symmetric positions of the organic free radical, i.e., the g=2.0 signal. Heat treatment decreased the satellite signals exponentially. The ESR signal of the pepper heated for more than 10 min was essentially the same as that before irradiation. To evaluate the radical decay by heat-treatment, we formulated a time-dependent master equation. We could evaluate the time constant of the radical decay based upon the general solution of the equation together with the nonlinear least-squares method

  4. Development of a compact tritium activity monitor and first tritium measurements

    Energy Technology Data Exchange (ETDEWEB)

    Röllig, M., E-mail: marco.roellig@kit.edu; Ebenhöch, S.; Niemes, S.; Priester, F.; Sturm, M.

    2015-11-15

    Highlights: • We report about experimental results of a new tritium activity monitoring system using the BIXS method. • The system is compact and easy to implement. It has a small dead volume of about 28 cm{sup 3} and can be used in a flow-through mode. • Gold coated surfaces are used to improve significantly count rate stability of the system and to reduce stored inventory. - Abstract: To develop a convenient tool for in-line tritium gas monitoring, the TRitium Activity Chamber Experiment (TRACE) was built and commissioned at the Tritium Laboratory Karlsruhe (TLK). The detection system is based on beta-induced X-ray spectrometry (BIXS), which observes the bremsstrahlung X-rays generated by tritium decay electrons in a gold layer. The setup features a measuring chamber with a gold-coated beryllium window and a silicon drift detector. Such a detection system can be used for accountancy and process control in tritium processing facilities like the Karlsruhe Tritium Neutrino Experiment (KATRIN). First characterization measurements with tritium were performed. The system demonstrates a linear response between tritium partial pressure and the integral count rate in a pressure range of 1 Pa up to 60 Pa. Within 100 s measurement time the lower detection limit for tritium is (143.63 ± 5.06) · 10{sup 4} Bq. The system stability of TRACE is limited by a linear decrease of integral count rate of 0.041 %/h. This decrease is most probably due to exchange interactions between tritium and the stainless steel walls. By reducing the interaction surface with stainless steel, the decrease of the integral count rate was reduced to 0.008 %/h. Based on the first results shown in this paper it can be concluded that TRACE is a promising complement to existing tritium monitoring tools.

  5. Tritium accountancy

    International Nuclear Information System (INIS)

    Avenhaus, R.; Spannagel, G.

    1995-01-01

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

  6. The paper irradiation for the control of decay produced by biological agents

    International Nuclear Information System (INIS)

    Calvo, Ana M.; Gonzalez, Maria E.

    2001-01-01

    Contamination of books and documents by insects and mould is a source of deterioration for the materials and a health risk for the library and archive staff and for the readers. The use of gamma radiation for decontamination is not widely used due to the damage it might cause to the paper and to some misconceptions concerning radiation technology that prevail among conservators. This paper reports the effects a dose of 15 KGy had on some properties of two kinds of paper in use in the library, one common paper for photocopies, printing and writing (Executive Ledesma) and the other a free of acid paper (Permalife) used in conservation and document restoration. The tenacity (MPa), tensile energy absorption (N/mm) and percent deformation were measured. No significant differences were found between irradiated and control samples of both kinds of paper. Light microscopy after the treatment with a reactive for the detection of lignin showed no differences after irradiation of the papers. No changes in the structure of the cellulosic fibres were detected by scanning electron microscopy after the irradiation treatment. Further studies have been undertaken regarding the ageing of the irradiated paper in relation to the non-irradiated controls. (author)

  7. Genetic effects of decay of tritium incorporated into cells of yeast Saccharomyces cerevisiae. 5. Lethal and mutagenic effects and the nature of mutations induced by /sup 3/H decay in the 6-th position of thymine

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, E.L.; Korolev, V.G. (AN SSSR, Leningrad. Inst. Yadernoj Fiziki)

    1982-03-01

    Lethal and mutagenous effects as well as nature of mutations induced with /sup 3/H decay in the sixth position of thymine (6-/sup 3/H-T) have been studied. Inactivation probability of haploid yeasts constituted ..cap alpha..=(6.1+-1.0)x10/sup -3/ decay/sup -1/ or ..cap alpha..=(7.6+-1.3)x10/sup -5/ rad/sup -1/, and probability of mutation appearance in genes ade 1, ade -K is (2.8+-1.7)x10/sup -8/ decay/sup -1/ or K=(3.5+-2.1)x10/sup -10/ rad/sup -1/. Lethal and mutageneous effects of 6-/sup 3/H-T don't differ considerably from those for /sup 3/H decay in the fifth position of thymine (5-/sup 3/H-T). From the point of view of frequency of transversions and mutations of read-out frame shift type induced in ade 2 gene, 6-/sup 3/H-T doesn't differ from 5-/sup 3/H-T. However, in comparison with the latter 6-/sup 3/H-T causes appearance of a larger amount of AT ..-->.. GTs transitions. A scheme, according to which 5 methyl barbituric acid (5MBK) is a finite product of /sup 3/H decay in the sixth position of thymine, is suggested. The results obtained point to that fact that 5MBK represents weak mutageneous damage of thymine causing the exchange of AT pair.

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

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.

    1991-01-01

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

  9. Introduction to Wolsong Tritium Removal Facility (WTRF)

    International Nuclear Information System (INIS)

    Song, K. M.; Sohn, S. H.; Kang, D. W.; Chung, H. S.

    2005-01-01

    Four CANDU 6 reactors have been operated at Wolsong site. Tritium is primarily produced in heavywater-moderated-power reactors by neutron capture of deuterium nuclei in the heavy water moderator and coolant. The concentration of tritium in the reactor moderator and coolant systems increases with time of reactor operation. For CANDU 6 reactors, the estimated equilibrium values are ∼3 TBq/kg-D 2 O in the moderator and ∼74 GBq/kg-D 2 O in the coolant, where the production rate is balanced by tritium decay and water makeup and loss process. The tritium level in the moderator heavy water of Wolsong Unit-1 is getting higher for about 20-year operation and is over 2.22x10 12 Bq/kg at the end of 2003. It was known that the tritium levels in the moderators of the other units would be also steadily increased. In order to reduce the tritium activity, KHNP has committed to construct a Tritium Removal Facility (TRF) at the Wolsong site

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

    International Nuclear Information System (INIS)

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

    1983-08-01

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

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

    International Nuclear Information System (INIS)

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

    1983-10-01

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

  12. Efigie: a computer program for calculating end-isotope accumulation by neutron irradiation and radioactive decay

    International Nuclear Information System (INIS)

    Ropero, M.

    1978-01-01

    Efigie is a program written in Fortran V which can calculate the concentration of radionuclides produced by neutron irradiation of a target made of either a single isotope or several isotopes. The program includes optimization criteria that can be applied when the goal is the production of a single nuclide. The effect of a cooling time before chemical processing of the target is also accounted for.(author) [es

  13. STAR facility tritium accountancy

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  14. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

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

  15. Calibrations of a tritium extraction facility

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Oliver, B.M.; Farrar, H. IV.

    1983-01-01

    A tritium extraction facility has been built for the purpose of measuring the absolute tritium concentration in neutron-irradiated lithium metal samples. Two independent calibration procedures have been used to determine what fraction, if any, of tritium is lost during the extraction process. The first procedure compares independently measured 4 He and 3 H concentrations from the 6 Li(n,α)T reaction. The second procedure compared measured 6 Li(n,α)T/ 197 Au (n,γ) 198 Au thermal neutron reaction rate ratios with those obtained from Monte Carlo calculations using well-known cross sections. Both calibration methods show that within experimental errors (approx. 1.5%) no tritium is lost during the extraction process

  16. Tritium target manufacturing for use in accelerators

    Science.gov (United States)

    Bach, P.; Monnin, C.; Van Rompay, M.; Ballanger, A.

    2001-07-01

    As a neutron tube manufacturer, SODERN is now in charge of manufacturing tritium targets for accelerators, in cooperation with CEA/DAM/DTMN in Valduc. Specific deuterium and tritium targets are manufactured on request, according to the requirements of the users, starting from titanium target on copper substrate, and going to more sophisticated devices. A wide range of possible uses is covered, including thin targets for neutron calibration, thick targets with controlled loading of deuterium and tritium, rotating targets for higher lifetimes, or large size rotating targets for accelerators used in boron neutron therapy. Activity of targets lies in the 1 to 1000 Curie, diameter of targets being up to 30 cm. Special targets are also considered, including surface layer targets for lowering tritium desorption under irradiation, or those made from different kinds of occluders such as titanium, zirconium, erbium, scandium, with different substrates. It is then possible to optimize either neutron output, or lifetime and stability, or thermal behavior.

  17. Oxidative Tritium Decontamination System

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Tritium activities in Canada

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1995-01-01

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

  19. Long-lived radicals produced by γ-irradiation or vital activity in plants, animals, cells, and protein solution: their observation and inhomogeneous decay dynamics

    International Nuclear Information System (INIS)

    Miyazaki, Tetsuo; Morikawa, Akiyuki; Kumagai, Jun; Ikehata, Masateru; Koana, Takao; Kikuchi, Shoshi

    2002-01-01

    Long-lived radicals produced by γ-irradiation or vital activity in plants, animals, cells, and protein (albumin) solution were studied by electron spin resonance spectroscopy. Long-lived radicals produced by vital activity exist in biological systems, such as plants, animals, and cells, in the range of 0.1-20 nmol g -1 . Since vital organs keep the radicals at a constant concentration, the radicals are probably related to life conservation. Long-lived radicals are also produced by γ-irradiation of cells or protein solution. The radicals decay after death of living things or after γ-irradiation. We found that the decay dynamics in all biological systems can be expressed by the same kinetic equation of an inhomogeneous reaction

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Science.gov (United States)

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

    2010-01-01

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

  2. The Tritium White Paper

    International Nuclear Information System (INIS)

    2009-01-01

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

  3. Tritium-caused background currents in electron multipliers

    International Nuclear Information System (INIS)

    Malinowski, M.E.

    1979-05-01

    One channel electron multiplier (Galileo No. 4501) and one 14 stage Be/Cu multiplier (Dumont No. SPM3) were exposed to tritium pressures between approx. 10 -7 Torr to 10 -3 Torr in amounts from approx. 10 -5 Torr-s to 60 Torr-s and the β-decay caused currents in the multipliers measured. The background currents in both multipliers consisted of two components: (1) a high, reversible current which was proportional to the tritium exposure pressure; and (2) a lower, irreversible background current which increased with increasing cumulative tritium exposure. The β-decay caused currents in each multiplier increased the same way with exposure, suggesting the detected electrons arose from decaying tritium adsorbed on surfaced external to the multipliers

  4. ZEPHYR tritium system

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  5. TFTR tritium handling concepts

    International Nuclear Information System (INIS)

    Garber, H.J.

    1976-01-01

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

  6. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

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

  7. Helium effects on tritium storage materials

    International Nuclear Information System (INIS)

    Moysan, I.; Contreras, S.; Demoment, J.

    2008-01-01

    For ten years French Tritium laboratories have been using metal hydride storage beds with LaNi 4 Mn for process gas (HDT mixture) absorption, desorption and for both short and long term storage. This material has been chosen because of its low equilibrium pressure and of its ability to retain decay helium 3 in its lattice. Aging effects on the thermodynamic behavior of LaNi 4 Mn have been investigated. Aging, due to formation of helium 3 in the lattice, decreases the desorption isotherm plateau pressure and shifts the α phase to the higher stoichiometries. Life time of the two kinds of tritium (and isotopes) storage vessels managed in the laboratory depends on these aging changes. The Tritium Long Term Storage (namely STLT) and the hydride storage vessel (namely FSH 400) are based on LaNi 4 Mn even though they are not used for the same applications. STLT contains LaNi 4 Mn in an aluminum vessel and is designed for long term pure tritium storage. The FSH 400 is composed of LaNi 4 Mn included within a stainless steel container. This design is aimed at storing low tritium content mixtures (less than 3% of tritium) and for supplying processes with HDT gas. Life time of the STLT can reach 12 years. Life time of the FSH 400 varies from 1.2 years to more than 25 years depending on the application. (authors)

  8. Helium effects on tritium storage materials

    Energy Technology Data Exchange (ETDEWEB)

    Moysan, I.; Contreras, S.; Demoment, J. [CEA Valduc, Service HDT, 21 - Is-sur-Tille (France)

    2008-07-15

    For ten years French Tritium laboratories have been using metal hydride storage beds with LaNi{sub 4}Mn for process gas (HDT mixture) absorption, desorption and for both short and long term storage. This material has been chosen because of its low equilibrium pressure and of its ability to retain decay helium 3 in its lattice. Aging effects on the thermodynamic behavior of LaNi{sub 4}Mn have been investigated. Aging, due to formation of helium 3 in the lattice, decreases the desorption isotherm plateau pressure and shifts the {alpha} phase to the higher stoichiometries. Life time of the two kinds of tritium (and isotopes) storage vessels managed in the laboratory depends on these aging changes. The Tritium Long Term Storage (namely STLT) and the hydride storage vessel (namely FSH 400) are based on LaNi{sub 4}Mn even though they are not used for the same applications. STLT contains LaNi{sub 4}Mn in an aluminum vessel and is designed for long term pure tritium storage. The FSH 400 is composed of LaNi{sub 4}Mn included within a stainless steel container. This design is aimed at storing low tritium content mixtures (less than 3% of tritium) and for supplying processes with HDT gas. Life time of the STLT can reach 12 years. Life time of the FSH 400 varies from 1.2 years to more than 25 years depending on the application. (authors)

  9. Tritium conference days

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. Sources of tritium

    International Nuclear Information System (INIS)

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

    1980-12-01

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

  11. 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)

  12. Tritium pellet injector results

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Environmental aspects of tritium

    International Nuclear Information System (INIS)

    Quisenberry, D.R.

    1979-01-01

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

  14. Development of ITER Tritium Storage Material

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. C.; Kim, K. R.; Paek, S. W.; Shim, M.; Noh, B

    2007-01-15

    The ZrCo getter beds are built of a primary vessel which contains the ZrCo powder and of a secondary outer vessel. The purpose of the secondary outer vessel is to capture permeated or leaked tritium and to present a good thermal insulation when properly evacuated. A third volume, a helium filled loop, is installed in the primary volume to remove the decay heat and is used to perform tritium accountancy measurements. In this report the authors verified that ZrCo can be used safely under a low pressure and temperature.

  15. Organically bound tritium

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.

    1993-01-01

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

  16. Tritium sampling and measurement

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Oestlund, H.G.; Mason, A.S.

    1985-04-01

    Tritium in the form of water, HTO, from the atmospheric testing of nuclear devices in the 60s has now mainly disappeared from the atmosphere and entered the ocean. The additions of such tritium from Chinese and French tests in the 70s were observed but did not make a big impression on the diminishing inventory of atmospheric HTO. Tritium in elemental form, HT, went through a maximum in the mid 70s, apparently primarily as a results of some underground testing of large nuclear devices and releases from civilian and military nuclear industry. The mid 70s maximum was 1.3 kg of tritium in this form, and in 1984 0.5 kg remain. The disappearance is slower than the decay rate of tritium, so sources must still have been present during this time. The global distribution shows, not unexpectedly, smaller inventory in the Southern Hemisphere across the equator and thus southward transport of HT. The chemical lifetime of hydrogen gas in the atmosphere, assuming the elemental tritium being in the form of HT, not T 2 , has been estimated between 6 and 10 years. It is to be expected that increasing activity of nuclear fuel reprocessing would in the near future again increase the global tritium gas inventory. Tritium in the form of light hydrocarbons, primarily methane, has also been measured, and in this form a quantity of 200 g of tritium resided in the global atmosphere 1956 to 1976. By 1982 it had decreased to 50 g. 25 refs., 5 figs., 11 tabs

  18. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    International Nuclear Information System (INIS)

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor

  19. Experimental method for investigating helium effects in irradiated vanadium

    International Nuclear Information System (INIS)

    Smith, D.L.; Matsui, H.; Greenwood, L.; Loomis, B.

    1987-10-01

    Analyses have been performed which indicate that an effective method for experimentally investigating helium effects in neutron irradiated vanadium base alloys can be developed. The experimental procedure involves only modest modifications to existing procedures currently used for irradiation testing of vanadium-base alloys in the FFTF reactor. Helium is generated in the vanadium alloy by decay of tritium which is either preinjected or generated within the test capsule. Calculations indicate that nearly constant He/dpa ratios of desired magnitude can be attained by proper selection of experimental parameters. The proposed method could have a major impact on the development of vanadium base alloys for fusion reactor applications. 8 refs., 4 figs

  20. Confinement and Tritium Stripping Systems for APT Tritium Processing

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Heung, L.K.

    1997-10-20

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

  1. Confinement and Tritium Stripping Systems for APT Tritium Processing

    International Nuclear Information System (INIS)

    Hsu, R.H.; Heung, L.K.

    1997-01-01

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

  2. Environmental tritium in trees

    International Nuclear Information System (INIS)

    Brown, R.M.

    1979-01-01

    The distribution of environmental tritium in the free water and organically bound hydrogen of trees growing in the vicinity of the Chalk River Nuclear Laboratories (CRNL) has been studied. The regional dispersal of HTO in the atmosphere has been observed by surveying the tritium content of leaf moisture. Measurement of the distribution of organically bound tritium in the wood of tree ring sequences has given information on past concentrations of HTO taken up by trees growing in the CRNL Liquid Waste Disposal Area. For samples at background environmental levels, cellulose separation and analysis was done. The pattern of bomb tritium in precipitation of 1955-68 was observed to be preserved in the organically bound tritium of a tree ring sequence. Reactor tritium was discernible in a tree growing at a distance of 10 km from CRNL. These techniques provide convenient means of monitoring dispersal of HTO from nuclear facilities. (author)

  3. Tritium monitoring techniques

    International Nuclear Information System (INIS)

    DeVore, J.R.; Buckner, M.A.

    1996-05-01

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

  4. Tritium in metals

    International Nuclear Information System (INIS)

    Schober, T.

    1990-01-01

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

  5. Tritium sources; Izvori tricijuma

    Energy Technology Data Exchange (ETDEWEB)

    Glodic, S [Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Boreli, F [Elektrotehnicki fakultet, Belgrade (Yugoslavia)

    1993-07-01

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

  6. High-pressure tritium

    International Nuclear Information System (INIS)

    Coffin, D.O.

    1976-01-01

    Some solutions to problems of compressing and containing tritium gas to 200 MPa at 700 0 K are discussed. The principal emphasis is on commercial compressors and high-pressure equipment that can be easily modified by the researcher for safe use with tritium. Experience with metal bellows and diaphragm compressors has been favorable. Selection of materials, fittings, and gauges for high-pressure tritium work is also reviewed briefly

  7. Study of tritium decontamination of stainless steel, copper, aluminum metals by tritium dry desorption

    International Nuclear Information System (INIS)

    Xie Yun; Shi Zhengkun; Wu Tao

    2014-01-01

    In order to study the decontamination efficiency of stainless steel, copper, aluminum metals contaminated by tritium, the metals were decontaminated by exposing to UV, ozone, heating, and the combination of heating, UV and ozone. The result indicates that the elevation of temperature can obviously improve decontamination. While irradiated by 172 nm UV, the decontamination efficiency is low, but it is better while heated and irradiated by 172 nm UV. If the stainless steel is irradiated by 172 nm UV and heated at 500℃ for 4 h, the decontamination efficiency is 99.2%. There is better decontamination efficiency of copper while exposed to ozone. While exposed to ozone and heated at 500℃, the decontamination efficiencies of stainless steel, copper and aluminum are higher than 99.2%. The decontamination efficiency can more obviously improve when metal is heated at high temperature (500℃) than low temperature (300℃). The surface tritium of metal placed at 30 d after decontamination increases because of diffusion and penetration of the tritium. Resolution spectra of tritium show that there are four kinds of contamination adsorbed tritium of stainless steel. (authors)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  9. EXOTIC: Development of ceramic tritium breeding materials

    International Nuclear Information System (INIS)

    Kwast, H.; Conrad, R.

    1989-09-01

    In this fifth EXOTIC annual progress report the work carried out in 1988 is reported. For EXOTIC-1, -2 and -3 the post-irradiation examinations have been continued with tritium retention measurements, annealng experiments, determination of physical and mechanical properties and X-ray diffraction analysis. Irradiation of EXOTIC-4 has been performde and the post-irradiation examination has started. Transient tritium release curves are given and analysed. The resulting tritium residence times show that for the Li-zirconates a residence time of less than one day can be achieved in the temperature region of 350-600 C. The loading scheme, the objectives and some fabrication data of EXOTIC-5 are give. Moreover, the fabrication of laboratory scale batches has started to investigate the effect of microstructural parameters on tritium release. Finally, an investigation was started on the system Li 2 O-ZrO 2 , with emphasis on the lithia-rich compositions. 40 figs., 9 refs., 10 tabs

  10. Tritium and helium-3 in metals

    International Nuclear Information System (INIS)

    Lasser, R.

    1989-01-01

    The book surveys recent results on the behaviour of tritium and its decay product helium-3 metals. In contrast to many earlier books which discuss the properties of the stable hydrogen isotopes without mentioning tritium, this book reviews mainly the results on tritium in metals. Due to the difficulties in preparing metal tritide samples, very important quantities such as diffusivity, superconductivity, solubility, etc. have only been determined very recently. The book not only presents the measured tritium data, but also the isotopic dependency of the different physical properties by comparing H, D and T results. A chapter is devoted to helium-3 in metals. Aspects such as helium release, generation of helium bubbles, swelling, and change of the lattice parameter upon aging are discussed. The book provides the reader with up-to-date information and deep insight into the behaviour of H, D, T and He-3 in metals. Further important topics such a tritium production, its risks, handling and discharge to the environment are also addressed

  11. 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.

  12. Tritium in the Savannah River Site environment

    International Nuclear Information System (INIS)

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

    1991-05-01

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

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

    International Nuclear Information System (INIS)

    Klein, J.E.; Mallory, M.K.; Nobile, A. Jr.

    1991-01-01

    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 3 He. 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

  14. The ITER tritium systems

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Radionuclide Basics: Tritium

    Science.gov (United States)

    Tritium is a hydrogen atom that has two neutrons in the nucleus and one proton. It is radioactive and behaves like other forms of hydrogen in the environment. Tritium is produced naturally in the upper atmosphere and as a byproduct of nuclear fission.

  16. In-vessel tritium

    International Nuclear Information System (INIS)

    Ueda, Yoshio; Ohya, Kaoru; Ashikawa, Naoko; Ito, Atsushi M.; Kato, Daiji; Kawamura, Gakushi; Takayama, Arimichi; Tomita, Yukihiro; Nakamura, Hiroaki; Ono, Tadayoshi; Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Nakano, Tomohide; Nakamura, Makoto; Hoshino, Kazuo; Kenmotsu, Takahiro; Wada, Motoi; Saito, Seiki; Takagi, Ikuji; Tanaka, Yasunori; Tanabe, Tetsuo; Yoshida, Masafumi; Toma, Mitsunori; Hatayama, Akiyoshi; Homma, Yuki; Tolstikhina, Inga Yu.

    2012-01-01

    The in-vessel tritium research is closely related to the plasma-materials interaction. It deals with the edge-plasma-wall interaction, the wall erosion, transport and re-deposition of neutral particles and the effect of neutral particles on the fuel recycling. Since the in-vessel tritium shows a complex nonlinear behavior, there remain many unsolved problems. So far, behaviors of in-vessel tritium have been investigated by two groups A01 and A02. The A01 group performed experiments on accumulation and recovery of tritium in thermonuclear fusion reactors and the A02 group studied theory and simulation on the in-vessel tritium behavior. In the present article, outcomes of the research are reviewed. (author)

  17. Protection against tritium radiations

    International Nuclear Information System (INIS)

    Bal, Georges

    1964-05-01

    This report presents the main characteristics of tritium, describes how it is produced as a natural or as an artificial radio-element. It outlines the hazards related to this material, presents how materials and tools are contaminated and decontaminated. It addresses the issue of permissible maximum limits: factors of assessment of the risk induced by tritium, maximum permissible activity in body water, maximum permissible concentrations in the atmosphere. It describes the measurement of tritium activity: generalities, measurement of gas activity and of tritiated water steam, tritium-induced ionisation in an ionisation chamber, measurement systems using ionisation chambers, discontinuous detection of tritium-containing water in the air, detection of surface contamination [fr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  19. EXOTIC: Development of ceramic tritium breeding materials

    International Nuclear Information System (INIS)

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

    1989-03-01

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

  20. Fluorine 18 in tritium generator ceramic materials

    International Nuclear Information System (INIS)

    Jimenez-Becerril, J.; Bosch, P.; Bulbulian, S.

    1992-01-01

    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 LiAlO 2 and the Li 2 O. 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)

  1. Tritium breeding in fusion reactors

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1982-10-01

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

  2. Process for recovering tritium from molten lithium metal

    Science.gov (United States)

    Maroni, Victor A.

    1976-01-01

    Lithium tritide (LiT) is extracted from molten lithium metal that has been exposed to neutron irradiation for breeding tritium within a thermonuclear or fission reactor. The extraction is performed by intimately contacting the molten lithium metal with a molten lithium salt, for instance, lithium chloride - potassium chloride eutectic to distribute LiT between the salt and metal phases. The extracted tritium is recovered in gaseous form from the molten salt phase by a subsequent electrolytic or oxidation step.

  3. Tritium release from lithium ceramics at constant temperature

    International Nuclear Information System (INIS)

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

    1992-02-01

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

  4. Fusion reactor start-up without an external tritium source

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, S., E-mail: Shanliang.Zheng@ccfe.ac.uk; King, D.B.; Garzotti, L.; Surrey, E.; Todd, T.N.

    2016-02-15

    Highlights: • Investigated the feasibility (including plasma physics, neutronics and economics) of starting a fusion reactor from running pure D–D fusion reactor to gradually move towards the D–T operation. • Proposed building up tritium from making use of neutrons generated by D–D fusion reactions. • Studied plasma physics feasibility for pure D–D operation and provided consistent fusion power and neutron yield in the plasma with different mixture of deuterium and tritium. • Discussed the economics aspect for operating a pure D–D fusion reactor towards a full-power D–T fusion reactor. - Abstract: It has long been recognised that the shortage of external tritium sources for fusion reactors using D–T, the most promising fusion fuel, requires all such fusion power plants (FPP) to breed their own tritium. It is also recognised that the initial start-up of a fusion reactor will require several kilograms of tritium within a scenario in which radioactive decay, ITER and subsequent demonstrator reactors are expected to have consumed most of the known tritium stockpile. To circumvent this tritium fuel shortage and ultimately achieve steady-state operation for a FPP, it is essential to first accumulate sufficient tritium to compensate for loss due to decay and significant retention in the materials in order to start a new FPP. In this work, we propose to accumulate tritium starting from D–D fusion reactions, since D exists naturally in water, and to gradually build up the D–T plasma targeted in fusion reactor designs. There are two likely D–D fusion reaction channels, (1) D + D → T + p, and (2) D + D → He3 + n. The tritium can be generated via the reaction channel ‘(1)’ and the 2.45 MeV neutrons from ‘(2)’ react with lithium-6 in the breeding blanket to produce more tritium to be fed back into plasma fuel. Quantitative evaluations are conducted for two blanket concepts to assess the feasibility and suitability of this approach to FPP

  5. Modeling of tritium behavior in Li2O

    International Nuclear Information System (INIS)

    Billone, M.C.; Attaya, H.; Kopasz, J.P.

    1992-08-01

    The TIARA and DISPL2 codes are being developed at Argonne National Laboratory to predict tritium retention and release from lithium ceramics under steady-state and transient conditions, respectively. Tritium retention and release are important design and safety issues for tritium-breeding blankets of fusion reactors. Emphasis has been placed on tritium behavior in Li 2 O because of the selection of this ceramic as a first option for the ITER driver blanket and because of the relatively good material properties data base for Li 2 O. Models and correlations for diffusion, surface desorption/adsorption, and solubility/precipitation of tritium in Li 2 0 have been developed based on well-controlled laboratory data from as-fabricated and irradiated samples. With the models and correlations, the codes are validated to the results of in-reactor purge flow tests. The results of validation of TIARA to tritium retention data from VOM-15H, EXOTIC-2, and CRITIC-1 are presented, along with predictions of tritium retention in BEATRIX-II. For DISPL2, results are presented for tritium release predictions vs. data for MOZART, CRITIC-1, and BEATRIX-II. Recommendations are made for improving both the data base and the modeling to allow extrapolation with reasonable uncertainty levels to fusion reactor design conditions

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

    International Nuclear Information System (INIS)

    1995-10-01

    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

  7. Cellular- and micro-dosimetry of heterogeneously distributed tritium.

    Science.gov (United States)

    Chao, Tsi-Chian; Wang, Chun-Ching; Li, Junli; Li, Chunyan; Tung, Chuan-Jong

    2012-01-01

    The assessment of radiotoxicity for heterogeneously distributed tritium should be based on the subcellular dose and relative biological effectiveness (RBE) for cell nucleus. In the present work, geometry-dependent absorbed dose and RBE were calculated using Monte Carlo codes for tritium in the cell, cell surface, cytoplasm, or cell nucleus. Penelope (PENetration and Energy LOss of Positrins and Electrons) code was used to calculate the geometry-dependent absorbed dose, lineal energy, and electron fluence spectrum. RBE for the intestinal crypt regeneration was calculated using a lineal energy-dependent biological weighting function. RBE for the induction of DNA double strand breaks was estimated using a nucleotide-level map for clustered DNA lesions of the Monte Carlo damage simulation (MCDS) code. For a typical cell of 10 μm radius and 5 μm nuclear radius, tritium in the cell nucleus resulted in much higher RBE-weighted absorbed dose than tritium distributed uniformly. Conversely, tritium distributed on the cell surface led to trivial RBE-weighted absorbed dose due to irradiation geometry and great attenuation of beta particles in the cytoplasm. For tritium uniformly distributed in the cell, the RBE-weighted absorbed dose was larger compared to tritium uniformly distributed in the tissue. Cellular- and micro-dosimetry models were developed for the assessment of heterogeneously distributed tritium.

  8. BEATRIX-II: In-situ tritium recovery data correction

    International Nuclear Information System (INIS)

    Slagle, O.D.; Hollenberg, G.W.; Kurasawa, T.; Verrall, R.A.

    1993-09-01

    BEATRIX-II was an in-situ tritium recovery experiment in a fast reactor to characterize the irradiation behavior of fusion ceramic breeder materials. Correcting and compiling the in-situ tritium recovery data involved correcting the ion chamber response for the effect of sweep gas composition or amount of hydrogen in the helium sweep gas and for the buildup of background. The effect of sweep gas composition was addressed in the previous workshop. During the operation of Phase I of the experiment the backgrounds of the ion chambers were found to reach significant levels relative to the tritium recovery concentrations in the sweep gas from the specimen canisters. The measured tritium concentrations were corrected for background by comparing the tritium recovery rate during reference conditions with the predicted tritium generation rate. Background increases were found to be associated with tritium recovery peaks and elevated levels of moisture in the sweep gas. These conditions typically occurred when the hydrogen concentration in the sweep gas was increased to 0.1% after extended operation in He or He-0.01% H 2 . Three examples of this increase in ionization chamber background are described. The final corrected BEATRIX-II, Phase I tritium recovery data provide a valuable resource to be used for predicting the performance of Li 2 O in a fusion blanket application

  9. Tritium permeation through iron

    International Nuclear Information System (INIS)

    Hagi, Hideki; Hayashi, Yasunori

    1989-01-01

    An experimental method for measuring diffusion coefficients and permeation rates of tritium in metals around room temperature has been established, and their values in iron have been obtained by using the method. Permeation rates of tritium and hydrogen through iron were measured by the electrochemical method in which a tritiated aqueous solution was used as a cathodic electrolyte. Tritium and hydrogen were introduced from one side of a membrane specimen by cathodic polarization, while at the other side of the specimen the permeating tritium and hydrogen were extracted by potentiostatical ionization. The amount of permeated hydrogen was obtained by integrating the anodic current, and that of tritium was determined by measuring the radioactivity of the electrolyte sampled from the extraction side. Diffusion coefficients of tritium (D T ) and hydrogen (D H ) were determined from the time lag of tritium and hydrogen permeation. D T =9x10 -10 m 2 /s and D H =4x10 -9 m 2 /s at 286 K for annealed iron specimens. These values of D T and D H were compared with the previous data of the diffusion coefficients of hydrogen and deuterium, and the isotope effect in diffusion was discussed. (orig.)

  10. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    Directory of Open Access Journals (Sweden)

    Stankunas Gediminas

    2017-01-01

    Full Text Available This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-cooled lithium-lead has the highest total decay heat at longer decay times in comparison to the helium-cooled design which has the lowest total decay heat. In addition, major nuclides were identified for water-cooled lithium-lead in W armour, Eurofer, and LiPb. In addition, great attention has been dedicated to the analysis of the decay heat and activity both from the different water-cooled lithium-lead blanket modules for the entire reactor and from each water-cooled lithium-lead blanket module separately. The neutron induced activation and decay heat at shutdown were calculated by the FISPACT code, using the neutron flux densities and spectra that were provided by the preceding MCNP neutron transport calculations.

  11. 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)

  12. Tritium technology. A Canadian overview

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    2002-01-01

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

  13. Metabolism and dosimetry of tritium

    International Nuclear Information System (INIS)

    Hill, R.L.; Johnson, J.R.

    1993-01-01

    This document was prepared as a review of the current knowledge of tritium metabolism and dosimetry. The physical, chemical, and metabolic characteristics of various forms of tritium are presented as they pertain to performing dose assessments for occupational workers and for the general public. For occupational workers, the forms of tritium discussed include tritiated water, elemental tritium gas, skin absorption from elemental tritium gas-contaminated surfaces, organically bound tritium in pump oils, solvents and other organic compounds, metal tritides, and radioluminous paints. For the general public, age-dependent tritium metabolism is reviewed, as well as tritiated water, elemental tritium gas, organically bound tritium, organically bound tritium in food-stuffs, and tritiated methane. 106 refs

  14. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Carson, S.D. [Sandia National Laboratories, New Mexico, NM (United States); Peterson, P.K. [Sandia National Laboratories, New Mexico, NM (United States)

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  15. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    International Nuclear Information System (INIS)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R.; Carson, S.D.; Peterson, P.K.

    1997-01-01

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  17. Tritium uptake in cultivated plants after short-term exposure to atmospheric tritium

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.; Paunescu, N.

    1998-01-01

    The tritium behavior in crop plants is of particular interest for the prediction of doses to humans due to ingestion. Tritium is present in plants in two forms: tritium free water tissue (TWT) and organically bound tritium (OBT). The both forms are to be considered in models calculating the ingestion dose. Potato plants belong to the major food crops in many countries and were chosen as representatives of crops whose edible parts grow under ground. Green bean were chosen as representatives of vegetables relevant in human diet. This vegetable may be consumed as green pod and it may be conserved over a long period of time. Green bean and potato plants were exposed to tritiated water vapor in the atmosphere during their generative phase of development. The uptake of tritium and the conversion into organic matter was studied under laboratory conditions at two different light intensities. The tritium concentrations in plants were followed until harvest. In leaves, the tritium uptake into tissue water under night conditions was 5-6 times lower than under day-time conditions. The initial incorporation into organic matter under night conditions was 0.7% of the tissue water concentration in leaves of both plant species. However, under light irradiation, this value increased to only 1.8% in bean leaves and 0.9% in potato leaves, which indicates a participation of processes other than photosynthesis in tritium incorporation into organic material. Organically bound tritium (OBT) was translocated into pods and tubers which represented a high percentage of the total organically bound tritium at harvest. The behavior of total OBT in all plants under study showed that OBT, once generated, is lost very slowly until harvest, in particular when storage organs of plants were in their phase of development at the time of exposure. OBT is translocated into the storage organs which may be used in the human diet and thus may contribute to the ingestion dose for a long time after the

  18. Experience in handling concentrated tritium

    International Nuclear Information System (INIS)

    Holtslander, W.J.

    1985-12-01

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

  19. Problems of anthropogenic tritium limitation

    Directory of Open Access Journals (Sweden)

    Kochetkov О.A.

    2013-12-01

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

  20. Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

    OpenAIRE

    Stankunas Gediminas; Tidikas Andrius

    2017-01-01

    This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-c...

  1. Tritium waste management on the La Hague AREVA NC site: associated impact and monitoring

    International Nuclear Information System (INIS)

    Devin, P.; Deguette, H.

    2009-01-01

    The authors propose an analysis of tritium behaviour in the nuclear fuel processed in the AREVA NC plant in La Hague, of its presence in the plant and in its wastes, and of the impact of these wastes and the tritium monitoring in the environment. First, they present the AREVA NC plant and evoke the legal context concerning the waste management. They report and discuss the analysis of the presence and behaviour of tritium in irradiated fuel, of its behaviour during spent fuel processing, the evolution of tritium releases (legal limitations, evolutions since 1992), of measurement of activity in effluents, and discuss a study of possible reductions of tritium releases by La Hague plants (mainly in sea waters). They also report the computational assessment of the dosimetric impact of tritium on neighbouring population. They describe how the presence of tritium in the environment is monitored within the annual radioactivity monitoring programme

  2. Tritium labeling by thermally generated tritons

    International Nuclear Information System (INIS)

    Morimoto, H.; Williams, P.G.; Saljoughian, M.

    1988-06-01

    The predominant effect of thermal atom irradiation on solid molecules is saturation of their aromatic functions. Only low level of tritium exchange is observed for aliphatic solids. In contrast, liquids whose frozen surface can be rendered somewhat mobile at appropriate temperatures exhibit more exchange than addition. The rank order of effectiveness of several metals in promoting exchange/addition appears similar to the rank order for heterogeneous catalytic hydrogenation. 3 refs., 8 figs

  3. Tritium permeation and recovery

    International Nuclear Information System (INIS)

    Bond, R.A.; Hamilton, A.M.

    1987-01-01

    The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. In this appendix, tritium transport in the DEMO breeding blanket is considered with emphasis on the permeation rate from the lithium-lead breeder into the coolant. A computational model used to calculate the tritium transport in the breeder blanket is described. Results are reported for the tritium transport in the NET/INTOR type blanket as well as the DEMO blanket in order to provide a comparison. In addition, results are presented for the helium coolant tritium extraction analysis. (U.K.)

  4. Tritium protective clothing

    International Nuclear Information System (INIS)

    Fuller, T.P.; Easterly, C.E.

    1979-06-01

    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

  5. Tritium breeding blanket

    International Nuclear Information System (INIS)

    Smith, D.; Billone, M.; Gohar, Y.; Baker, C.; Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H.; Raffray, A.; Sviatoslavsky, I.; Simbolotti, G.; Shatalov, G.

    1991-01-01

    The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R ampersand D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs

  6. Tritium in plants

    International Nuclear Information System (INIS)

    Vichot, L.; Losset, Y.

    2009-01-01

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

  7. Tritium-v. 2

    International Nuclear Information System (INIS)

    1987-01-01

    Several bibliographical references about tritium are shown. The following aspects are presented: properties, analysis, monitoring, dosimetry reactions, labelling, industrial production, radiological protection, applications to science, technology and industry and some processes to obtain the element. (E.G.) [pt

  8. Tritium waste package

    Science.gov (United States)

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

    1995-01-01

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

  9. Tritium protective clothing

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, T. P.; Easterly, C. E.

    1979-06-01

    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.

  10. Future use of tritium in mapping pre-bomb groundwater volumes.

    Science.gov (United States)

    Eastoe, C J; Watts, C J; Ploughe, M; Wright, W E

    2012-01-01

    The tritium input to groundwater, represented as volume-weighted mean tritium concentrations in precipitation, has been close to constant in Tucson and Albuquerque since 1992, and the decrease in tritium concentrations at the tail end of the bomb tritium pulse has ceased. To determine the future usefulness of tritium measurements in southwestern North America, volume-weighted mean tritium levels in seasonal aggregate precipitation samples have been gathered from 26 sites. The averages range from 2 to 9 tritium units (TU). Tritium concentrations increase with site latitude, and possibly with distance from the coast and with site altitude, reflecting local ratios of combination of low-tritium moisture advected from the oceans with high-tritium moisture originating near the tropopause. Tritium used alone as a tool for mapping aquifer volumes containing only pre-bomb recharge to groundwater will become ambiguous when the tritium in precipitation at the end of the bomb tritium pulse decays to levels close to the analytical detection limit. At such a time, tritium in precipitation from the last one to two decades of the bomb pulse will become indistinguishable from pre-bomb recharge. The threshold of ambiguity has already arrived in coastal areas with a mean of 2 TU in precipitation and will follow in the next three decades throughout the study region. Where the mean tritium level is near 5 TU, the threshold will occur between 2025 and 2030, given a detection limit of 0.6 TU. Similar thresholds of ambiguity, with different local timing possible, apply globally. © 2011, The Author(s). Ground Water © 2011, National Ground Water Association.

  11. 2017 Accomplishments – Tritium Aging Studies on Stainless Steel Weldments and Heat-Affected Zones

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Michael J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hitchcock, Dale [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Krentz, Tim [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McNamara, Joy [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Duncan, Andrew [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2018-01-31

    In this study, the combined effects tritium and decay helium in forged and welded Types 304L and 21-6-9 stainless steels were studied. To measure these effects, fracture mechanic specimens were thermally precharged with tritium and aged for approximately 17 years to build in decay helium from tritium decay prior to testing. The results are compared to earlier measurements on the same alloys and weldments (4-5, 8-9). In support of Enhanced Surveillance, “Tritium Effects on Materials”, the fracture toughness properties of long-aged tritium-charged stainless-steel base metals and weldments were measured and compared to earlier measurements. The fracture-toughness data were measured by thermally precharging as-forged and as-welded specimens with tritium gas at 34.5 MPa and 350°C and aging for approximately 17 years to build-in decay helium prior to testing. These data result from the longest aged specimens ever tested in the history of the tritium effects programs at Savannah River and the fracture toughness values measured were the lowest ever recorded for tritium-exposed stainless steel. For Type 21-6-9 stainless steel, fracture toughness values were reduced to less than 2-4% of the as-forged values to 41 lbs / in specimens that contained more than 1300 appm helium from tritium decay. The fracture toughness properties of long-aged weldments were also measured. The fracture toughness reductions were not as severe because the specimens did not retain as much tritium from the charging and aging as did the base metals. For Type 304L weldments, the specimens in this study contained approximately 600 appm helium and their fracture toughness values averaged 750 lbs / in. The results for other steels and weldments are reported and additional tests will be conducted during FY18.

  12. Thin film tritium dosimetry

    Science.gov (United States)

    Moran, Paul R.

    1976-01-01

    The present invention provides a method for tritium dosimetry. A dosimeter comprising a thin film of a material having relatively sensitive RITAC-RITAP dosimetry properties is exposed to radiation from tritium, and after the dosimeter has been removed from the source of the radiation, the low energy electron dose deposited in the thin film is determined by radiation-induced, thermally-activated polarization dosimetry techniques.

  13. Management of tritium wastes

    International Nuclear Information System (INIS)

    Kisalu, J.; Mellow, D.G.; Pennington, J.D.; Thompson, H.M.; Wood, E.

    1991-07-01

    This work provides a review of the management of tritium wastes with particular reference to current practice, possible alternatives and to the implications of any alternatives considered. It concludes that reduction in UK emissions from nuclear industry is feasible but at a cost out of all proportion to the reduction in dose commitment achievable. Commercial usage of tritium involves importation at several times the UK nuclear production level although documentation is sparse. (author)

  14. PRODUCTION OF TRITIUM

    Science.gov (United States)

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

    1963-02-26

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

  15. Technology benefits resulting from accelerator production of tritium

    International Nuclear Information System (INIS)

    1998-01-01

    One of the early and most dramatic uses of nuclear transformations was in development of the nuclear weapons that brought World War II to an end. Despite that difficult introduction, nuclear weapons technology has been used largely as a deterrent to war throughout the latter half of the twentieth century. The Accelerator Production of Tritium (APT) offers a clean, safe, and reliable means of producing the tritium (a heavy form of hydrogen) needed to maintain the nuclear deterrent. Tritium decays away naturally at a rate of about 5.5% per year; therefore, the tritium reservoirs in nuclear weapons must be periodically replenished. In recent years this has been accomplished by recycling tritium from weapons being retired from the stockpile. Although this strategy has served well since the last US tritium production reactor was shut down in 1988, a new tritium production capability will be required within ten years. Some benefits will result from direct utilization of some of the APT proton beam; others could result from advances in the technologies of particle accelerators and high power spallation targets. The APT may save thousands of lives through the production of medical isotopes, and it may contribute to solving the nation's problem in disposing of long-lived nuclear wastes. But the most significant benefit may come from advancing the technology, so that the great potential of accelerator applications can be realized during our lifetimes

  16. Tritium in nuclear power plants

    International Nuclear Information System (INIS)

    Badyaev, V.V.; Egorov, Yu.A.; Sklyarov, V.P.; Stegachev, G.V.

    1981-01-01

    The problem of tritium formation during NPP operation is considered on the basis of available published data. Tritium characteristics are given, sources of the origin of natural and artificial tritium are described. NPP contribution to the total tritium amount in the environment is determined, as well as contribution of each process in the reactor to the quantity of tritium, produced at the NPP. Thermal- and fast-neutron reactions with tritium production are shown, their contribution to the total amount of tritium in a coolant is estimated, taking into account the type of reactor. Data on tritium content in NPP wastes and in the air of working premises are presented. Methods for sampling and sample preparation to measurements as well as the appropriate equipment are considered. Design of the gas-discharge counter of internal filling, used for measuring tritium activity in samples is described [ru

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

    Nobuta, Y.; Hatano, Y.; Matsuyama, M.; Abe, S.; Akamaru, S.; Yamauchi, Y.; Hino, T.; Suzuki, S.; Akiba, M.

    2014-01-01

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

  19. Japanese university program on tritium radiobiology and environmental tritium

    International Nuclear Information System (INIS)

    Okada, Shigefumi

    1989-01-01

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

  20. Tritium Assay and Dispensing of KEPRI Tritium Lab

    International Nuclear Information System (INIS)

    Sohn, S. H.; Song, K. M.; Lee, S. K.; Lee, K.W.; Ko, B. W.

    2009-01-01

    The Wolsong Tritium Removal Facility(WTRF) has been constructed to reduce tritium levels in the heavy water systems and environmental emissions at the site. The WTRF was designed to process 100 kg/h of heavy water with the overall tritium extraction efficiency of 97% per single pass and to produce ∼700 g of tritium as T2 per year at the feed concentration of 0.37 TBq/kg. The high purity tritium greater than 99% is immobilized as a metal hydride to secure its long term storage. The recovered tritium will be made available for industrial uses and some research applications in the future. Then KEPRI is constructing the tritium lab. to build-up infrastructure to support tritium research activities and to support tritium control and accountability systems for tritium export. This paper describes the initial phases of the tritium application program including the laboratory infrastructure to support the tritium related R and D activities and the tritium controls in Korea

  1. Comparison of tritium production facilities

    International Nuclear Information System (INIS)

    He Kaihui; Huang Jinhua

    2002-01-01

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

  2. Overview of R and D activities on tritium processing and handling technology in JAEA

    Energy Technology Data Exchange (ETDEWEB)

    Yamanishi, Toshihiko, E-mail: yamanishi.toshihiko@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Oyaidsu, Makoto; Yamada, Masayuki; Suzuki, Takumi; Hayashi, Takumi [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The tritium technologies have been studied at Tritium Process Laboratory of JAEA. Black-Right-Pointing-Pointer A monitoring method for the blanket system of a fusion reactor have been studied. Black-Right-Pointing-Pointer Basic studies on the tritium behavior in confinement system have been carried out. Black-Right-Pointing-Pointer Studies on the detritiation have been carried out as another significant activity. - Abstract: In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main R and D activities are: the tritium processing technology for the blanket recovery systems; the basic tritium behavior in confinement materials; and detritiation and decontamination. The R and D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.

  3. Tritium diffusion in polycrystalline lithium tungstate

    International Nuclear Information System (INIS)

    Krutyakov, A.N.; Shadrin, A.A.; Saunin, E.I.; Gromov, V.V.; Shafiev, A.I.

    1984-01-01

    Using radiometric method the investigation of tritium separation from neutron irradiated (neutron flux density 1.2x10 13 n/cm 2 xs) polycrystalline Li 2 WO 4 in the temperature range 200-680 deg C has been carried out. It is established that the use of helium as gas-carrier of flow-type gas-discharge counter permits to conduct continuous stable measurements of concentrations of tritium extracted depending on its chemical state. It is shown that volume diffusion is the process, limiting tritiated particle separation rate from Li 2 WO 4 . It is found that the process of tritium volume diffusion in Li 2 WO 4 corresponds to two different mechanisms respectively in low- (200-300 deg C) and high-temperature (350-680 deg C) ranges. A supposition is made that in the low-temperature range the process of diffusion is conditioned by the dissociation of the radiation defect-tritiated particle complex, which is confirmed by the data on radiation defect annealing in Li 2 WO 4 . The value of activation energy of tritium separation process in the range 350-680 deg C, proved to be equal to 13.3 kJ/mol. Possible role of crystal structure peculiarities of Li 2 WO 4 for diffusion process is pointed out

  4. 2009 EVALUATION OF TRITIUM REMOVAL AND MITIGATION TECHNOLOGIES FOR WASTEWATER TREATMENT

    Energy Technology Data Exchange (ETDEWEB)

    LUECK KJ; GENESSE DJ; STEGEN GE

    2009-02-26

    Since 1995, a state-approved land disposal site (SALDS) has received tritium contaminated effluents from the Hanford Site Effluent Treatment Facility (ETF). Tritium in this effluent is mitigated by storage in slow moving groundwater to allow extended time for decay before the water reaches the site boundary. By this method, tritium in the SALDS is isolated from the general environment and human contact until it has decayed to acceptable levels. This report contains the 2009 update evaluation of alternative tritium mitigation techniques to control tritium in liquid effluents and groundwater at the Hanford site. A thorough literature review was completed and updated information is provided on state-of-the-art technologies for control of tritium in wastewaters. This report was prepared to satisfy the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-026-07B (Ecology, EPA, and DOE 2007). Tritium separation and isolation technologies are evaluated periodically to determine their feasibility for implementation to control Hanford site liquid effluents and groundwaters to meet the Us. Code of Federal Regulations (CFR), Title 40 CFR 141.16, drinking water maximum contaminant level (MCL) for tritium of 20,000 pOll and/or DOE Order 5400.5 as low as reasonably achievable (ALARA) policy. Since the 2004 evaluation, there have been a number of developments related to tritium separation and control with potential application in mitigating tritium contaminated wastewater. These are primarily focused in the areas of: (1) tritium recycling at a commercial facility in Cardiff, UK using integrated tritium separation technologies (water distillation, palladium membrane reactor, liquid phase catalytic exchange, thermal diffusion), (2) development and demonstration of Combined Electrolysis Catalytic Exchange (CECE) using hydrogen/water exchange to separate tritium from water, (3) evaporation of tritium contaminated water for dispersion in the

  5. Tritium emissions reduction facility (TERF)

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. 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.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. Unclassified information on tritium extraction and purification technology: attachment 1

    International Nuclear Information System (INIS)

    McNorrill, P.L.

    1976-01-01

    Several tritium recovery and purification techniques developed at non-production sites are described in the unclassified and declassified literature. Heating of irradiated Li-Al alloy under vacuum to release tritium is described in declassified reports of Argonne National Laboratory. Use of palladium membranes to separate hydrogen isotopes from other gases is described by Argonne, KAPL, and others. Declassified KAPL reports describe tritium sorption on palladium beds and suggest fractional absorption as a means of isotope separation. A thermal diffusion column for tritium enrichment is described in a Canadian report. Mound Laboratory reports describe theoretical and experimental studies of thermal diffusion columns. Oak Ridge reports tabulate ''shape factors'' for thermal diffusion columns. Unclassified journals contain many articles on thermal diffusion theory, experiments, and separation of gas mixtures by thermal diffusion columns; much of these data can be readily extended to the separation of hydrogen-tritium mixtures. Cryogenic distillation for tritium recovery is described in the Mound Laboratory reports. Process equipment such as pumps, valves, Hopcalite beds, and uranium beds are described in reports by ANL, KAPL, and MLM, and in WASH-1269, Tritium Control Technology

  9. Tritium biological effects and perspective of the biological study

    International Nuclear Information System (INIS)

    Komatsu, Kenshi

    1998-01-01

    Since tritium is an emitter of weak β-rays (5.7keV) and is able to bind to DNA, i.e., the most important genome component, the biological effects should be expected to be more profound than that of X-rays and γ-rays. When carcinogenesis, genetical effects and the detriments for fetus and embryo were used as a biological endpoint, most of tritium RBE (relative biological effectiveness) ranged from 1 to 2. The tritium risk in man could be calculated from these RBEs and γ-ray risk for human exposure, which are obtained mainly from the data on Atomic Bomb survivors. However, the exposure modality from environmental tritium should be a chronic irradiation with ultra low dose rate or a fractionated irradiation. We must estimate the tritium effect in man based on biological experiments alone, due to lack of such epidemiological data. Low dose rate experiment should be always accompanied by the statistical problem of data, since their biological effects are fairy low, and they should involve a possible repair system, such as adaptive response (or hormesis effect) and 'Kada effect' observed in bacteria. Here we discuss future works for the tritium assessment in man, such as (1) developing a high radiation sensitive assay system with rodent hybrid cells containing a single human chromosome and also (2) study on mammal DNA repair at molecular levels using a radiosensitive hereditary disease, Nijmegen Breakage Syndrome. (author)

  10. Tritium production distribution in the accelerator production of tritium device

    International Nuclear Information System (INIS)

    Kidman, R.B.

    1997-11-01

    Helium-3 ( 3 He) gas is circulated throughout the accelerator production of tritium target/blanket (T/B) assembly to capture neutrons and convert 3 He to tritium. Because 3 He is very expensive, it is important to know the tritium producing effectiveness of 3 He at all points throughout the T/B. The purpose of this paper is to present estimates of the spatial distributions of tritium production, 3 He inventory, and the 3 He FOM

  11. Tritium Systems Test Facility

    International Nuclear Information System (INIS)

    Cafasso, F.A.; Maroni, V.A.; Smith, W.H.; Wilkes, W.R.; Wittenberg, L.J.

    1978-01-01

    This TSTF proposal has two principal objectives. The first objective is to provide by mid-FY 1981 a demonstration of the fuel cycle and tritium containment systems which could be used in a Tokamak Experimental Power Reactor for operation in the mid-1980's. The second objective is to provide a capability for further optimization of tritium fuel cycle and environmental control systems beyond that which is required for the EPR. The scale and flow rates in TSTF are close to those which have been projected for a prototype experimental power reactor (PEPR/ITR) and will permit reliable extrapolation to the conditions found in an EPR. The fuel concentrations will be the same as in an EPR. Demonstrations of individual components of the deuterium-tritium fuel cycle and of monitoring, accountability and containment systems and of a maintenance methodology will be achieved at various times in the FY 1979-80 time span. Subsequent to the individual component demonstrations--which will proceed from tests with hydrogen (and/or deuterium) through tracer levels of tritium to full operational concentrations--a complete test and demonstration of the integrated fuel processing and tritium containment facility will be performed. This will occur near the middle of FY 1981. Two options were considered for the TSTF: (1) The modification of an existing building and (2) the construction of a new facility

  12. Tritium analysis at TFTR

    International Nuclear Information System (INIS)

    Voorhees, D.R.; Rossmassler, R.L.; Zimmer, G.

    1995-01-01

    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

  13. Recommendations for Tritium Science and Technology Research and Development in Support of the Tritium Readiness Campaign, TTP-7-084

    Energy Technology Data Exchange (ETDEWEB)

    Senor, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-10-30

    Between 2006 and 2012 the Tritium Readiness Campaign Development and Testing Program produced significant advances in the understanding of in-reactor TPBAR performance. Incorporating these data into existing TPBAR performance models has improved permeation predictions, and the discrepancy between predicted and observed tritium permeation in the WBN1 coolant has been decreased by about 30%. However, important differences between predicted and observed permeation still remain, and there are significant knowledge gaps that hinder the ability to reliably predict other aspects of TPBAR performance such as tritium distribution, component integrity, and performance margins. Based on recommendations from recent Tritium Readiness Campaign workshops and reviews coupled with technical and programmatic priorities, high-priority activities were identified to address knowledge gaps in the near- (3-5 year), middle- (5-10 year), and long-term (10+ year) time horizons. It is important to note that there are many aspects to a well-integrated research and development program. The intent is not to focus exclusively on one aspect or another, but to approach the program in a holistic fashion. Thus, in addition to small-scale tritium science studies, ex-reactor tritium technology experiments such as TMED, and large-scale in-reactor tritium technology experiments such as TMIST, a well-rounded research and development program must also include continued analysis of WBN1 performance data and post-irradiation examination of TPBARs and lead use assemblies to evaluate model improvements and compare separate-effects and integral component behavior.

  14. Tritium - is it underestimated

    International Nuclear Information System (INIS)

    Whitlock, G.D.

    1980-01-01

    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 3 H 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.)

  15. Tritium in HTR systems

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1987-07-01

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

  16. Atmospheric tritium. Measurement and application

    International Nuclear Information System (INIS)

    Frejaville, Gerard

    1967-02-01

    The possible origins of atmospheric tritium are reviewed and discussed. A description is given of enrichment (electrolysis and thermal diffusion) and counting (gas counters and liquid scintillation counters) processes which can be used for determining atmospheric tritium concentrations. A series of examples illustrates the use of atmospheric tritium for resolving a certain number of hydrological and glaciological problems. (author) [fr

  17. Handling of tritium at TFTR

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. Tritium effluent removal system

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  19. Development of tritium technology at the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    Anderson, J.L.; Bartlit, J.R.

    1982-01-01

    The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for large scale fusion reactor systems starting with the Fusion Engineering Device (FED) or the International Tokamak Reactor (INTOR). This paper briefly describes the fuel cycle and safety systems at TSTA including the Vacuum Facility, Fuel Cleanup, Isotope Separation, Transfer Pumping, Emergency Tritium Cleanup, Tritium Waste Treatment, Tritium Monitoring, Data Acquisition and Control, Emergency Power and Gas Analysis systems. Discussed in further detail is the experimental program proposed for the startup and testing of these systems

  20. In situ measurement of tritium permeation through stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Luscher, Walter G., E-mail: walter.luscher@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); Senor, David J., E-mail: david.senor@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); Clayton, Kevin K., E-mail: kevin.clayton@inl.gov [Idaho National Laboratory, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States); Longhurst, Glen R., E-mail: glenlonghurst@suu.edu [Idaho National Laboratory, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States)

    2013-06-15

    Highlights: ► In situ tritium permeation measurements collected over broad pressure range. ► Test conditions relevant to 316 SS in commercial light water reactors. ► Comparisons between in- and ex-reactor measurements provided. ► Correlation between tritium permeation, temperature, and pressure developed. -- Abstract: 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 {sup 4}He carrier gas mixed with a specified quantity of tritium (T{sub 2}) 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 {sup 3}He to T, was also evaluated by introducing a {sup 4}He carrier gas mixed with {sup 3}He at a partial pressure of 1013 Pa at 330 °C. Less than 3% of the tritium resulting from {sup 3}He transmutation contributed to tritium permeation.

  1. Monitoring of tritium

    Science.gov (United States)

    Corbett, James A.; Meacham, Sterling A.

    1981-01-01

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

  2. Results of tritium experiments on ceramic electrolysis cells and palladium diffusers for application to fusion reactor fuel cleanup systems

    International Nuclear Information System (INIS)

    Carlson, R.V.; Binning, K.E.; Konishi, S.; Yoshida, H.; Naruse, Y.

    1987-01-01

    Tritium tests at the Tritium Systems Test Assembly have demonstrated that ceramic electrolysis cells and palladium alloy diffuser developed in Japan are possible components for a fusion reactor fuel cleanup system. Both components have been successfully operated with tritium for over a year. A failure of the first electrolysis cell was most likely the result of an over voltage on the ceramic. A simple circuit was developed to eliminate this mode of failure. The palladium diffusers tubes exhibited some degradation of mechanical properties as a result of the build up of helium from the tritium decay, after 450 days of operation with tritium, however the effects were not significant enough to affect the performance. New models of the diffuser and electrolysis cell, providing higher flow rates and more tritium compatible designs are currently being tested with tritium. 8 refs., 5 figs

  3. EPR studies of free radicals decay and survival in gamma irradiated aminoglycoside antibiotics: sisomicin, tobramycin and paromomycin.

    Science.gov (United States)

    Wilczyński, Sławomir; Pilawa, Barbara; Koprowski, Robert; Wróbel, Zygmunt; Ptaszkiewicz, Marta; Swakoń, Jan; Olko, Paweł

    2012-02-14

    Radiation sterilization technology is more actively used now that any time because of its many advantages. Gamma radiation has high penetrating power, relatively low chemical reactivity and causes small temperature rise. But on the other hand radiosterilization can lead to radiolytic products appearing, in example free radicals. Free radicals in radiative sterilized sisomicin, tobramycin and paromomycin were studied by electron paramagnetic resonance (EPR) spectroscopy. Dose of gamma irradiation of 25kGy was used. Concentrations and properties of free radicals in irradiated antibiotics were studied. EPR spectra were recorded for samples stored in air and argon. For gamma irradiated antibiotics strong EPR lines were recorded. One- and two-exponential functions were fitted to experimental points during testing and researching of time influence of the antibiotics storage to studied parameters of EPR lines. Our study of free radicals in radiosterilized antibiotics indicates the need for characterization of medicinal substances prior to sterilization process using EPR values. We propose the concentration of free radicals and other spectroscopic parameters as useful factors to select the optimal type of sterilization for the individual drug. The important parameters are i.a. the τ time constants and K constants of exponential functions. Time constants τ give us information about the speed of free radicals concentration decrease in radiated medicinal substances. The constant K(0) shows the free radicals concentration in irradiated medicament after long time of storage. Copyright © 2011 Elsevier B.V. All rights reserved.

  4. Tritium Sequestration in Gen IV NGNP Gas Stream via Proton Conducting Ceramic Pumps

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Fanglin Frank [Univ. of South Carolina, Columbia, SC (United States); Adams, Thad M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brinkman, Kyle [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reifsnider, Kenneth [Univ. of South Carolina, Columbia, SC (United States)

    2011-09-30

    Several types of high-temperature proton conductors based on SrCeO3 and BaCeO3 have been systematically investigated in this project for tritium separation in NGNP applications. One obstacle for the field application is the chemical stability issues in the presence of steam and CO2 for these proton conductors. Several strategies to overcome such issues have been evaluated, including A site doping and B site co-doping method for perovskite-structured proton conductors. Novel zirconium-free proton conductors have also been developed with improved electrical conductivity and enhanced chemical stability. Novel catalytic materials for the proton-conducting separation membranes have been investigated. A tubular geometry proton-conducting membrane has been developed for the proton separation membranes. Total dose rate estimated from tritium decay (beta emission) under realistic membrane operating conditions, combined with electron irradiation experiments, indicates that proton ceramic materials possess the appropriate radiation stability for this application.

  5. Proton-induced accelerated decay of the fungicide, vinclozolin, on TiO2 surface under solar irradiation: Experimental and DFT study.

    Science.gov (United States)

    Osipov, Ivan; Gorbachev, Mikhail Y; Gorinchoy, Natalia N

    2018-04-03

    The photochemical degradation of vinclozolin by addition of titanium dioxide on silica support has been examined both experimental and quantum-chemically. Solar irradiation of vinclozolin on silica with and without addition of titanium dioxide for 6 h resulted in 21% and 97.8% vinclozolin residues, respectively. In both these cases, phototransformation leads to the formation of (3,5-dichlorophenyl isocyanate) and (3,5-dichloroaniline). The presence of the intermediary product resulted from opening of the 2,4-oxazolidine-dione ring is also confirmed by GS-MS and LC-MS chromatography. The proton-induced mechanism of vinclozolin decay at the above experimental conditions is clarified on the base of DFT calculations.

  6. Tritium in rad waste management

    International Nuclear Information System (INIS)

    Gandhi, P.M.; Ali, S.S.; Mathur, R.K.; Rastogi, R.C.

    1990-01-01

    Radioactive waste arising from PHWR's are invariably contaminated with tritium activity. Their disposal is crucial as it governs the manner and extent of radioactive contamination of human environment. The technique of tritium measurement and its application plays an important role in assessing the safety of the disposal system. Thus, typical applications involving tritium measurements include the evaluation of a site for solid waste burial facility and evaluation of a water body for liquid waste dispersal. Tritium measurement is also required in assessing safe air route dispersal of tritium. (author)

  7. ARIES-I tritium system

    International Nuclear Information System (INIS)

    Sze, D.K.; Tam, S.W.; Billone, M.C.; Hassanein, A.M.; Martin, R.

    1990-09-01

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

  8. Tritium recycling and inventory in eroded debris of plasma-facing materials

    International Nuclear Information System (INIS)

    Hassanein, A.

    1999-01-01

    Damage to plasma-facing components (PFCs) and structural materials due to loss of plasma confinement in magnetic fusion reactors remains one of the most serious concerns for safe, successful, and reliable tokamak operation. High erosion losses due to surface vaporization, spallation, and melt-layer splashing are expected during such an event. The eroded debris and dust of the PFCs, including trapped tritium, will be contained on the walls or within the reactor chamber therefore, they can significantly influence plasma behavior and tritium inventory during subsequent operations. Tritium containment and behavior in PFCS and in the dust and debris is an important factor in evaluating and choosing the ideal plasma-facing materials (PFMs). Tritium buildup and release in the debris of candidate materials is influenced by the effect of material porosity on diffusion and retention processes. These processes have strong nonlinear behavior due to temperature, volubility, and existing trap sites. A realistic model must therefore account for the nonlinear and multidimensional effects of tritium diffusion in the porous-redeposited and neutron-irradiated materials. A tritium-transport computer model, TRAPS (Tritium Accumulation in Porous Structure), was developed and used to evaluate and predict the kinetics of tritium transport in porous media. This model is coupled with the TRICS (Tritium In Compound Systems) code that was developed to study the effect of surface erosion during normal and abnormal operations on tritium behavior in PFCS

  9. Tritium breeding materials

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Johnson, C.E.; Abdou, M.

    1984-03-01

    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

  10. Tritium breeding materials

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Johnson, C.E.; Abdou, M.A.

    1984-01-01

    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

  11. Properties of tritium and its compounds

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  12. Use of tritium and sources

    International Nuclear Information System (INIS)

    Noguchi, Hiroshi

    1997-01-01

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

  13. Tritium calibration of the LUX dark matter experiment

    Science.gov (United States)

    Akerib, D. S.; Araújo, H. M.; Bai, X.; Bailey, A. J.; Balajthy, J.; Beltrame, P.; Bernard, E. P.; Bernstein, A.; Biesiadzinski, T. P.; Boulton, E. M.; Bradley, A.; Bramante, R.; Cahn, S. B.; Carmona-Benitez, M. C.; Chan, C.; Chapman, J. J.; Chiller, A. A.; Chiller, C.; Currie, A.; Cutter, J. E.; Davison, T. J. R.; de Viveiros, L.; Dobi, A.; Dobson, J. E. Y.; Druszkiewicz, E.; Edwards, B. N.; Faham, C. H.; Fiorucci, S.; Gaitskell, R. J.; Gehman, V. M.; Ghag, C.; Gibson, K. R.; Gilchriese, M. G. D.; Hall, C. R.; Hanhardt, M.; Haselschwardt, S. J.; Hertel, S. A.; Hogan, D. P.; Horn, M.; Huang, D. Q.; Ignarra, C. M.; Ihm, M.; Jacobsen, R. G.; Ji, W.; Kazkaz, K.; Khaitan, D.; Knoche, R.; Larsen, N. A.; Lee, C.; Lenardo, B. G.; Lesko, K. T.; Lindote, A.; Lopes, M. I.; Malling, D. C.; Manalaysay, A. G.; Mannino, R. L.; Marzioni, M. F.; McKinsey, D. N.; Mei, D.-M.; Mock, J.; Moongweluwan, M.; Morad, J. A.; Murphy, A. St. J.; Nehrkorn, C.; Nelson, H. N.; Neves, F.; O'Sullivan, K.; Oliver-Mallory, K. C.; Ott, R. A.; Palladino, K. J.; Pangilinan, M.; Pease, E. K.; Phelps, P.; Reichhart, L.; Rhyne, C.; Shaw, S.; Shutt, T. A.; Silva, C.; Solovov, V. N.; Sorensen, P.; Stephenson, S.; Sumner, T. J.; Szydagis, M.; Taylor, D. J.; Taylor, W.; Tennyson, B. P.; Terman, P. A.; Tiedt, D. R.; To, W. H.; Tripathi, M.; Tvrznikova, L.; Uvarov, S.; Verbus, J. R.; Webb, R. C.; White, J. T.; Whitis, T. J.; Witherell, M. S.; Wolfs, F. L. H.; Young, S. K.; Zhang, C.; LUX Collaboration

    2016-04-01

    We present measurements of the electron-recoil (ER) response of the LUX dark matter detector based upon 170 000 highly pure and spatially uniform tritium decays. We reconstruct the tritium energy spectrum using the combined energy model and find good agreement with expectations. We report the average charge and light yields of ER events in liquid xenon at 180 and 105 V /cm and compare the results to the NEST model. We also measure the mean charge recombination fraction and its fluctuations, and we investigate the location and width of the LUX ER band. These results provide input to a reanalysis of the LUX run 3 weakly interacting massive particle search.

  14. Tritium determination in water

    International Nuclear Information System (INIS)

    Gavini, Ricardo M.

    2008-01-01

    An analytical procedure for the determination of tritium in water is described in this paper. The determination is carried out in presence of other radionuclides, such as Fe-55, Ni-63, Mn-54, Zn-65, Co-60, Cd-109, Sr-90, Cs-134 and Cs-137. The method consists in a simple distillation stage prior to measurement by liquid scintillation counting. The samples containing beta and gamma emitters are conditioned with a (NO 3 ) 2 Pb solution and Na(OH) up to pH = 7 - 8. This produces lead hydroxide precipitation that allows fixing volatile elements, which could be transported together with tritium, and may increase the extinction degree of the sample or interfere with the counting process. Special attention must be paid if presence of Fe-55 (E max ∼ 5.95 keV) is suspected as it might not be distinguished from tritium (E max ∼ 18 keV), leading to an overestimation of tritium activity. Different tests were carried to obtain the optimum method conditions, to achieve the purification of the tritium and a pH near to 7 in the distilled. The detection limit (2σ) was 8.0 Bq/l and the distillation performance was 98.3 %. This technique was applied to water samples containing Fe-55 and other gamma radionuclides in 1M hydrochloric acid media in successive Environmental Measurements Laboratory (EML), U.S. Department of Energy (DOE) intercomparison programs. The results obtained were very satisfactory and are presented in this paper. (author)

  15. Shipment and Storage Containers for Tritium Production Transportation Casks

    International Nuclear Information System (INIS)

    Massey, W.M.

    1998-01-01

    A shipping and storage container for the Tritium production transportation casks may be required but requirements for protection of the irradiated rods and radioactive contamination have not been finalized. This report documents the various possibilities for the container depending on the final requirements

  16. Tritium isolation from lithium inorganic compounds applicable to thermonuclear reactor breeding blanket

    International Nuclear Information System (INIS)

    Vasil'ev, V.G.; Ershova, Z.V.; Nikiforov, A.S.

    1982-01-01

    Tritium separation from inorganic lithium compounds: Li 2 O, LiAlO 2 , Li 2 SiO 3 , Li 4 SiO 4 , LiF, LiBeF 3 , Li 2 BeF 4 irradiated with a beam of a gamma facility and a nuclear reactor, has been studied. In the first case the gas phase is absent. In the latter one- the tritium amount in the gas does not exceed 1-2% of its total amount in the salt. Based on the EPR spectra of irradiated salts the concentrations of paramagnetic centres are calculated. It is shown that during thermal annealing the main portion of tritium in the gas phase is in the form of oxide (HTO, T 2 O). Tritium is separated from lithium fluoroberyllates in the form of hydrogen (HT, T 2 ). The kinetics of tritium oxide isolation from irradiated lithium oxide aluminate, metha- and orthosilicates, lithium sulphate has been studied. The activation energies of tritium oxide separation process are presented. A supposition is made that chemical reaction of the HTO (T 2 O) or HT(T 2 ) or HF(TF) formation is a limiting stage. Clarification of the process stage limiting the rate of tritium recovery will permit to evaluate conditions for the optimum work of lithium material in the blanket, lithium zone to select the lithium element structure and temperature regime of irradiation

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  18. Tritium in the aquatic environment

    International Nuclear Information System (INIS)

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

    1986-02-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  20. Tritium breeders and tritium permeation barrier coatings for fusion reactor

    International Nuclear Information System (INIS)

    Yamawaki, Michio; Kawamura, Hiroshi; Tsuchiya, Kunihiko

    2004-01-01

    A state of R and D of tritium breeders and tritium permeation barrier coatings for fusion reactor is explained. A list of candidate for tritium breeders consists of ceramics containing lithium, for examples, Li 2 O, Li 2 TiO 3 , Li 2 ZrO 3 , Li 4 SiO 4 and LiAlO 2 . The characteristics and form are described. The optimum particle size is from 1 to 10 μm. The production technologies of tritium breeders in the world are stated. Characteristics of ceramics with lithium as tritium breeders are compared. TiC, TiN/TiC, Al 2 O 3 and Cr 2 O 3 -SiO 2 -P 2 O 5 are tritium permeation barrier coating materials. These production methods and evaluation of characteristics are explained. (S.Y.)

  1. Universal tritium transmitter

    International Nuclear Information System (INIS)

    Cordaro, J. V.; Wood, M.

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-05-01

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

  3. Tritium trapping states induced by lithium-depletion in Li{sub 2}TiO{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Makoto, E-mail: kobayashi.makoto@LHD.nifs.ac.jp [National Institute for Fusion Science (Japan); Oya, Yasuhisa; Okuno, Kenji [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan)

    2017-04-15

    Identifications of tritium trapping states in neutron-irradiated Li{sub 1.8}TiO{sub 2.9} (lithium-depleted Li{sub 2}TiO{sub 3}) were carried out by the out-of-pile tritium release behavior. Tritium release behaviors for neutron-irradiated Li{sub 2}TiO{sub 3} and tritium gas-exposed TiO{sub 2} were also measured for comparison. Among the tritium release spectra for these samples, three tritium release peaks were appeared. By the kinetic analyses of tritium release behaviors, the Arrhenius parameters for three peaks were evaluated. Especially for Li{sub 1.8}TiO{sub 2.9}, there were two tritium release peaks, and the peak in lower temperature region was assigned to the tritium release controlled by the diffusion process in Li{sub 2}TiO{sub 3} structure. The other tritium release peak, which was hardly appeared for Li{sub 2}TiO{sub 3}, was assigned to the release of tritium trapped as hydroxyl groups in Li{sub 1.8}TiO{sub 2.9}, indicating that lithium-depletion would result in the formation of hydroxyl groups in Li{sub 2}TiO{sub 3}. Lithium vacancies existed in Li{sub 2}TiO{sub 3} crystal structure would promote the tritium trapping as hydroxyl groups due to the decreased charge repulsion between lithium ions and tritium ion, resulting in the difficulty of recovering tritium from Li{sub 2}TiO{sub 3} effectively.

  4. Environmental monitoring for tritium in tritium separation facility

    International Nuclear Information System (INIS)

    Varlam, Carmen; Stefanescu, Ioan; Steflea, Dumitru; Lazar, Roxana Elena

    2001-01-01

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

  5. Environmental monitoring for tritium at tritium separation facility

    International Nuclear Information System (INIS)

    Varlam, C.; Stefanescu, I.; Steflea, D.; Lazar, R.E.

    2001-01-01

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

  6. Tritium concentration monitor

    International Nuclear Information System (INIS)

    Shono, Kosuke.

    1991-01-01

    A device for measuring the concentration of tritium in gaseous wastes in a power plant and a nuclear fuel reprocessing plant is reduced in the size and improved in performance. The device of the present invention pressurizes a sampling gas and cools it to a dew point. Water content in the sampling gas cooled to the dew point is condensated and recovered to a fine tube-like water content recovering container. The concentration of the recovered condensates is measured by a tritium density analyzer. With such procedures, since the specimen is pressurized, the dew point can be elevated. Accordingly, the size of the cooling device can be decreased, enabling to contribute to the reduction of the size of the entire device. Further, since the water content recovering device is formed as a fine tube, the area of contact between the specimen gas and the liquid condensated water can be reduced. Accordingly, evaporation of the liquid condensates can be prevented. (I.S.)

  7. Metabolism of organically bound tritium

    International Nuclear Information System (INIS)

    Travis, C.C.

    1984-01-01

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

  8. A prototype wearable tritium monitor

    International Nuclear Information System (INIS)

    Surette, R. A.; Dubeau, J.

    2008-01-01

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

  9. Effects of tritium in elastomers

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1982-01-01

    Elastomers are used as flange gaskets in the piping system of the Savannah River Plant tritium facilities. A number of elastomers is being examined to identify those compounds more radiation-resistant than the currently specified Buna-N rubber and to study the mechanism of tritium radiation damage. Radiation resistance is evaluated by compression set tests on specimens exposed to about 1 atm tritium for several months. Initial results show that ethylene-propylene rubber and three fluoroelastomers are superior to Buna-N. Off-gassing measurements and autoradiography show that retained surface absorption of tritium varies by more than an order of magnitude among the different elastomer compounds. Therefore, tritium solubility and/or exchange may have a role in addition to that of chemical structure in the damage process. Ongoing studies of the mechanism of radiation damage include: (1) tritium absorption kinetics, (2) mass spectroscopy of radiolytic products, and (3) infrared spectroscopy

  10. Biological effects of tritium

    International Nuclear Information System (INIS)

    Nieto, M.

    1985-01-01

    The aim of this project is to study the thermal effects on proliferation activity in the intestinal epithelium of the goldfish acclimated at different temperatures (stationary state). The cell division occurs only at certain phases of the circadian cycle when the proliferative activity is synchronized or trained by an environmental factor such as light-dark cycle. Another aspect of the project is the study of the biological effects, non-stochastic, on cell kinetics in animals chronically exposed to low dose rates or tritium and gamma rays from 60 CO, used as a standard radiation. The influence on the accumulated dose per cell and cycle cell in function of the duration of the cell cycle at different acclimation temperatures should be considered. To calculate the risk of tritium contamination from nuclear power plants (radiation exposure), the organic tissue-bond is of decisive importance due to the long turnover of the organic tissue-bond in organisms favouring transport of tritium to other organisms of the ecosystem and to man. (author)

  11. Tritium removal and retention device

    International Nuclear Information System (INIS)

    Boyle, R.F.; Durigon, D.D.

    1980-01-01

    A device is provided for removing and retaining tritium from a gaseous medium, and also a method of manufacturing the device. The device, consists of an inner core of zirconium alloy, preferably Zircaloy-4, and an outer adherent layer of nickel which acts as a selective and protective window for passage of tritium. The tritium then reacts with or is absorbed by the zirconium alloy, and is retained until such time as it is desirable to remove it during reprocessing. (auth)

  12. An overview of tritium production

    International Nuclear Information System (INIS)

    He Kaihui; Huang Jinghua; Feng Kaiming

    2002-01-01

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

  13. Tritium safety issues for TFCX

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  14. Tritium inventory tracking and management

    International Nuclear Information System (INIS)

    Eichenberg, T.W.; Klein, A.C.

    1990-01-01

    This investigation has identified a number of useful applications of the analysis of the tracking and management of the tritium inventory in the various subsystems and components in a DT fusion reactor system. Due to the large amounts of tritium that will need to be circulated within such a plant, and the hazards of dealing with the tritium an electricity generating utility may not wish to also be in the tritium production and supply business on a full time basis. Possible scenarios for system operation have been presented, including options with zero net increase in tritium inventory, annual maintenance and blanket replacement, rapid increases in tritium creation for the production of additional tritium supplies for new plant startup, and failures in certain system components. It has been found that the value of the tritium breeding ratio required to stabilize the storage inventory depends strongly on the value and nature of other system characteristics. The real operation of a DT fusion reactor power plant will include maintenance and blanket replacement shutdowns which will affect the operation of the tritium handling system. It was also found that only modest increases in the tritium breeding ratio are needed in order to produce sufficient extra tritium for the startup of new reactors in less than two years. Thus, the continuous operation of a reactor system with a high tritium breeding ratio in order to have sufficient supplies for other plants is not necessary. Lastly, the overall operation and reliability of the power plant is greatly affected by failures in the fuel cleanup and plasma exhaust systems

  15. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1983-08-01

    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 233 U 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 3 He 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 3 He 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

  16. TFTR tritium operations lessons learned

    International Nuclear Information System (INIS)

    Gentile, C.A.; Raftopoulos, S.; LaMarche, P.

    1996-01-01

    The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980's The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America's first operational D-T fusion reactor. This paper will discuss negative pressure employing 'elephant trunks' in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and Δ pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs

  17. 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.

  18. Overview of tritium fast-fission yields

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1981-03-01

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

  19. Technology developments for improved tritium management

    International Nuclear Information System (INIS)

    Miller, J.M.; Spagnolo, D.A.

    1994-06-01

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

  20. 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.

  1. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. Calibration of an experimental model of tritium storage bed designed for 'in situ' accountability

    International Nuclear Information System (INIS)

    Bidica, Nicolae; Stefanescu, Ioan; Bucur, Ciprian; Bulubasa, Gheorghe; Deaconu, Mariea

    2009-01-01

    Full text: Objectives: Tritium accountancy of the storage beds in tritium facilities is an important issue for tritium inventory control. The purpose of our work was to perform calibration of an experimental model of tritium storage bed with a special design, using electric heaters to simulate tritium decay, and to evaluate the detection limit of the accountancy method. The objective of this paper is to present an experimental method used for calibration of the storage bed and the experimental results consisting of calibration curves and detection limit. Our method is based on a 'self-assaying' tritium storage bed. The basic characteristics of the design of our storage bed consists, in principle, of a uniform distribution of the storage material on several copper thin fins (in order to obtain a uniform temperature field inside the bed), an electrical heat source to simulate the tritium decay heat, a system of thermocouples for measuring the temperature field inside the bed, and good thermal isolation of the bed from the external environment. Within this design of the tritium storage bed, the tritium accounting method is based on determining the decay heat of tritium by measuring the temperature increase of the isolated storage bed. Experimental procedure consisted in measuring of temperature field inside the bed for few values of the power injected with the aid of electrical heat source. Data have been collected for few hours and the temperature increase rate was determined for each value of the power injected. Graphical representation of temperature rise versus injected powers was obtained. This accounting method of tritium inventory stored as metal tritide is a reliable solution for in-situ tritium accountability in a tritium handling facility. Several improvements can be done regarding the design of the storage bed in order to improve the measurement accuracy and to obtain a lower detection limit as for instance use of more accurate thermocouples or special

  3. IN-SITU TRITIUM BETA DETECTOR

    International Nuclear Information System (INIS)

    Berthold, J.W.; Jeffers, L.A.

    1998-01-01

    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

  4. 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

  5. JET experiments with tritium and deuterium–tritium mixtures

    NARCIS (Netherlands)

    Horton, L.; Batistoni, P.; Boyer, H.; Challis, C.; Ciric, D.; Donne, A. J. H.; Eriksson, L. G.; Garcia, J.; Garzotti, L.; Gee, S.; Hobirk, J.; Joffrin, E.; Jones, T.; King, D. B.; Knipe, S.; Litaudon, X.; Matthews, G. F.; Monakhov, I.; Murari, A.; Nunes, I.; Riccardo, V.; Sips, A. C. C.; Warren, R.; Weisen, H.; Zastrow, K. D.

    2016-01-01

    Extensive preparations are now underway for an experiment in the Joint European Torus (JET) using tritium and deuterium–tritium mixtures. The goals of this experiment are described as well as the progress that has been made in developing plasma operational scenarios and physics reference pulses for

  6. Tritium ions in the Source and Transport Section (STS) of KATRIN

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Manuel [Karlsruhe Institute of Technology (Germany); Collaboration: KATRIN-Collaboration

    2016-07-01

    The KArlsruhe TRItium Neutrino (KATRIN) experiment aims at the model independent measurement of the electron neutrino mass. It is designed for a neutrino mass sensitivity of 0.2 eV (90% CL) after three years of measurement time. KATRIN measures the end point of the tritium beta decay spectrum using a MAC-E filter and a Windowless Gaseous Tritium Source (WGTS). While neutral tritium gas molecules are pumped through the WGTS, the decay electrons are guided to the detector with a magnetic field. Tritium ions, however, also leave the WGTS following the magnetic field lines. For KATRIN measurements it is imperative to prevent tritium ions from reaching the detector or the spectrometers, where they could decay and cause an indistinguishable background. Ion blocking measures are implemented by electric blocking potentials and electric dipoles to drift out trapped ions. Their effective operation will be tested during KATRIN commissioning measurements: The ion flux between STS and spectrometers can be measured with the Forward Beam Monitor (FBM). It offers a manipulator arm to introduce a detector into the flux tube. For ion detection, a Faraday Cup for the FBM is being designed and constructed.

  7. Tritium contaminated waste management at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Jalbert, R.A.; Carlson, R.V.

    1987-01-01

    The Tritium Systems Test Assembly (TSTA) at Los Alamos continues to move toward full operation of an integrated, full-sized, computer-controlled fusion fuel processing loop. Concurrent nonloop experiments further the development of advanced tritium technologies and handling methods. Since tritium operations began in June 1984, tritium contaminated wastes have been produced at TSTA that are roughly typical in kind and amount of those to be produced by tritium fueling operations at fusion reactors. Methods of managing these wastes are described, including information on some methods of decontamination so that equipment can be reused. Data are given on the kinds and amounts of wastes and the general level of contamination. Also included are data on environmental emissions and doses to personnel that have resulted from TSTA operations. Particular problems in waste managements are discussed

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

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

    2005-01-01

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

  9. Recovery of tritium from CANDU reactors, its storage and monitoring of its migration in the environment

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Osborne, R.V.

    1979-07-01

    Tritium is produced in CANDU heavy water reactors mainly by neutron activation of deuterium. The typical production rate is 2.4 kCi per megawatt-year (89 TBq. per megawatt-year. In Pickering Generating Station the average concentration of tritium in the moderators has reached 16 Ci.kg -1 (0.6 TBq.kg -1 ) and in coolants, 0.5 Ci.kg -1 (0.02 TBq.kg -1 ). Concentrations will continue to increase towards an equilibrium determined by the production rate, the tritium decay rate and heavy water replacement. Tritium removal methods that are being considered for a pilot plant design are catalytic exchange of DTO with D 2 and electrolysis of D 2 O/DTO to provide feed for cryogenic distillation of D 2 /DT/T 2 . Storage methods for the removed tritium - as elemental gas, as metal hydrides and in cements - are also being investigated. Transport of tritiated wastes should not be a particularly difficult problem in light of extensive experience in transporting tritiated heavy water. Methods for determining the presence of tritium in the environment of any tritium handling facility are well established and have the capability of measuring concentrations of tritium down to current ambient values. (author)

  10. Tritium transport and control in the FED

    International Nuclear Information System (INIS)

    Rogers, M.L.

    1981-01-01

    The tritium systems for the FED have three primary purposes. The first is to provide tritium and deuterium fuel for the reactor. This fuel can be new tritium or deuterium delivered to the plant site, or recycled DT from the reactor that must be processed before it can be recycled. The second purpose of the FED tritium systems is to provide state-of-the-art tritium handling to limit worker radiation exposure and to minimize tritium losses to the environment. The final major objective of the FED tritium systems is to provide an integrated system test of the tritium handling technology necessary to support the fusion reactor program. Every effort is being made to incorporate available information from the Tritium System Test Assembly (TSTA) at Los Alamos National Laboratory, the Tokamak Fusion Test Reactor (TFTR) tritium systems, and the tritium handling information generated within DOE for the past 20 years

  11. Ontario Hydro diversifies into tritium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    A report is given on a plant which is to be built at the Darlington Candu reactor site in Canada for the extraction of tritium from heavy water. As tritium is used as a fuel in fusion research the market for it is expected to grow. The design of the system is outlined with the help of a flow diagram. (U.K.)

  12. Tritium containment in fusion facilities

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1978-01-01

    The key environmental control systems that have been identified and are being developed are listed. A brief description of each of the following systems is given: primary process materials, permeation barriers, secondary containment, tritium waste treatment, emergency tritium cleanup, maintenance procedures, and tertiary containment

  13. 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)

  14. substantiation of the standards for tritium amino acids intake by human organism

    International Nuclear Information System (INIS)

    Zhesko, T.V.; Balonov, M.I.

    1984-01-01

    Calculated values of β-irradiation tissue doses of bound tritium and tritium oxide absorbed by animals treated with different amino acids are given. The obtained dose values are compared with tritium water doses. The data obtained in animal studies are extrapolated to man in order to determine the dose equivalent to the incorporated 3 H-amino acids dose. It is assumed that the dose equivalent to 3 H-amino acids radiation is three times as high as the equivalent amount of tritium oxide. 9 mCi/yr is established as the maximum allowable blood intake of 3 H-amino acids. Due to their metabolic characteristics, air permissible concentrations of 3 H-amino acids and tritium oxide have approximately the same values. The value of 30 nCi/cm 2 is recommended as a standard for work clothing contamination with 3 H-amino acids

  15. Tritium breeding experiments with lithium titanate in thermal-type mockups

    International Nuclear Information System (INIS)

    Klix, Axel; Takahashi, Akito; Ochiai, Kentaro; Nishitani, Takeo

    2004-01-01

    Lithium titanate, an advanced tritium breeding material, is currently investigated in integral mock-up experiments at FNS. A method was developed which allows to measure low tritium concentrations directly in this material. The local tritium production rate was obtained by small lithium titanate pellet detectors inserted into the breeding layers which are dissolved after irradiation of the assemblies, and the accumulated tritium was counted by liquid scintillation techniques. The measurement method was applied in mock0-up experiments with candidate materials for the future DEMO reactor breeding blanket. Experimental assemblies consisted of sheets of low activation ferritic steel F82H, lithium titanate, and berylium. Tritium production rate profiles were obtained and compared with results from calculations with the Monte Carlo neutron transport code MCNP-4C. In case of the mock-ups with 95% enriched lithium titanate, the C/E ratios were within the error estimate while larger discrepancies were observed in case of 40% enriched lithium titanate. (author)

  16. Recovery of tritium from lithium-sintered aluminium product (SAP) and lithium-aluminium alloys

    International Nuclear Information System (INIS)

    Talbot, J.B.; Wiffen, F.W.

    1979-01-01

    The tritium release rates of irradiated samples of lithium-containing aluminium (Li-Al) and sintered aluminium product (Li-SAP) were investigated to evaluate the potential application of both materials in fusion reactors. The observed release rates followed the pattern expected for bulk diffusion of tritium in a solid. Therefore, diffusion coefficients for tritium in Li-SAP were determined over a temperature range of 383 and 500 0 C and tritium in Li-Al at 450 0 C. At 450 0 C, the diffusion coefficients of tritium in Li-SAP and Li-Al are 2.988 x 10 -10 cm 2 sec -1 and 1.462 x 10 -6 cm 2 sec -1 , respectively. (author)

  17. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

    Halle, A. von; Anderson, J.L.; Gentile, C.; Grisham, L.; Hosea, J.; Kamperschroer, J.; LaMarche, P.; Oldaker, M.; Nagy, A.; Raftopoulos, S.; Stevenson, T.

    1995-01-01

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

  18. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

    von Halle, A.; Gentile, C.

    1994-01-01

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

  19. Tritium in the food chain

    International Nuclear Information System (INIS)

    Koenig, L.A.

    1988-01-01

    Tritium is a hydrogen isotope taking part in the global hydrogen cycle as well as in all metabolic processes. The resultant problems with respect to the food chain are summarized briefly with emphasis on 'organically bound tritium'. However, only a small number of the numerous publications on this topic can be taken into consideration. Publications describing experiments under defined conditions are reported, thus allowing a semiempirical interpretation to be made. Tritium activity measurements of food grown in the vicinity of the Karlsruhe Nuclear Research Center have been carried out. A list of the results is given. A dose assessment is performed under simplifying assumptions. Even when the organically bound tritium is taken into account, a radiation exposure of less than 1% of that of K-40 is obtained under these conditions. To avoid misinterpretation, the specific activity (Bq H-3/g H) of water-bound and organically bound tritium has to be considered separately. (orig.) [de

  20. Tritium metabolism in rat tissues

    International Nuclear Information System (INIS)

    Takeda, H.

    1982-01-01

    As part of a series of studies designed to evaluate the relative radiotoxicity of various tritiated compounds, metabolism of tritium in rat tissues was studied after administration of tritiated water, leucine, thymidine, and glucose. The distribution and retention of tritium varied widely, depending on the chemical compound administered. Tritium introduced as tritiated water behaved essentially as body water and became uniformly distributed among the tissues. However, tritium administered as organic compounds resulted in relatively high incorporation into tissue constituents other than water, and its distribution differed among the various tissues. Moreover, the excretion rate of tritium from tissues was slower for tritiated organic compounds than for tritiated water. Administrationof tritiated organic compounds results in higher radiation doses to the tissues than does administration of tritiated water. Among the tritiated compounds examined, for equal radioactivity administered, leucine gave the highest radiation dose, followed in turn by thymidine, glucose, and water. (author)

  1. Tritium behaviour in higher plants

    International Nuclear Information System (INIS)

    Guenot, J.

    1984-05-01

    Vine grapes and potato seedlings have been exposed in situ to tritiated water vapor and 14 C labeled carbon dioxide. Leaves sampling was done during and after the exposition. Measurements allowed to distinguish the three forms of tritium in leaves, i.e. tissue free water tritium (TFWT) and organically bound tritium (OBT), in exchangeable position or not. The results lead to a description of the dynamical behaviour of tritium between these three compartments. It has been shown that 20% of organically bound hydrogen is readily exchangeable thus being in permanent isotopic equilibium with tissue free water. Moreover, the activity of nonexchangeable OBT appears to be strongly related to the organic 14 C, which shows that photosynthesis is responsible of tritium incorporation in organic nonexchangeable position, and occurs with a 20% discrimination in favor of protium. In contrast with the other two compartments, this fixation is almost irreversible, which is a fact of importance from a radiological point of view [fr

  2. Tritium practices past and present

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  3. Tritium-surface interactions

    International Nuclear Information System (INIS)

    Kirkaldy, J.S.

    1983-06-01

    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

  4. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    Energy Technology Data Exchange (ETDEWEB)

    Dautel, W.A.

    1996-10-01

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  5. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    International Nuclear Information System (INIS)

    Dautel, W.A.

    1996-01-01

    The Department of Energy is currently engaged in a dual-track strategy to develop an accelerator and a commercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle'costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Department's purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work together 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay after 2005

  6. Torus evacuation and tritium handling on NET

    International Nuclear Information System (INIS)

    Dinner, P.; Chazalon, M.; Iseli, M.

    1986-08-01

    The use of tritium as a fuel affects the design of many systems, as well as requiring several new systems not needed on non DT-burning Tokamaks. This paper summarizes: major tritium process interconnections, tritium flows and inventories; primary requirements, preferred design alternatives, and related development issues; design philosophy for tritium and primary vacuum systems. 14 refs

  7. Contribution to the tritium continental effect

    International Nuclear Information System (INIS)

    Lewis, R.R.; Froehlich, K.; Hebert, D.

    1987-01-01

    The results of tritium measurements of atmospheric water vapour and precipitation samples for 1982 and 1983 are presented. The data were used to establish a simple model describing the tritium continental effect taking into account re-evaporation of tritium from the continental land surfaces and man-made tritium. (author)

  8. Contribution to the tritium continental effect

    International Nuclear Information System (INIS)

    Lewis, R.R.; Froehlich, K.; Hebert, D.

    1987-01-01

    The results of tritium measurements of atmospheric water vapour and precipitation samples for 1982 and 1983 are presented. The data were used to establish a simple model describing the tritium continental effect taking into account re-evaporation of tritium from the continental land surfaces. Some comments on man made tritium are given. (author)

  9. Survey of tritium wastes and effluents in near-term fusion-research facilities

    International Nuclear Information System (INIS)

    Bickford, W.E.; Dingee, D.A.; Willingham, C.E.

    1981-08-01

    The use of tritium control technology in near-term research facilities has been studied for both the magnetic and inertial confinement fusion programs. This study focused on routine generation of tritium wastes and effluents, with little referene to accidents or facility decommissioning. This report serves as an independent review of the effectiveness of planned control technology and radiological hazards associated with operation. The facilities examined for the magnetic fusion program included Fusion Materials Irradiation Testing Facility (FMIT), Tritium Systems Test Assembly (TSTA), and Tokamak Fusion Test Reactor (TFTR) in the magnetic fusion program, while NOVA and Antares facilities were examined for the inertial confinement program

  10. Autoradiographical Detection of Tritium in Cu-Ni Alloy by Scanning Electron Microscopy

    OpenAIRE

    高安, 紀; 中野, 美樹; 竹内, 豊三郎

    1981-01-01

    The autoradiograph of tritium dispersed in Cu-Ni alloy sheet by 6Li(n,α)3H reaction was obtained by a scanning electron microscope. Prior to the irradiation of neutrons 6Li was deposited on the sheet by evaporation. The liquid emulsion, Fuji-ER, was used in this study. The distribution of tritium was detected by the dispersion of silver grains remaining in the emulsion after the development was carried out.

  11. Consideration of disposal alternatives for tritium-contaminated wastewater streams at Hanford

    International Nuclear Information System (INIS)

    Waters, E.D.

    1988-03-01

    Small quantities of tritium are produced as an undesirable by-product of the operation of light-water reactors. At the US Department of Energy Hanford Site in Washington State, some tritium has been discharged to the environment in low-level liquid and gaseous wastes from the N Reactor plant, but more than 97% of the tritium stays typically within the irradiated fuel as it is delivered for reprocessing. During fuel reprocessing, the tritium is distributed in the process streams, and most of the tritium is presently released to the soil column with excess process condensates from the Plutonium-Uranium Extraction (PUREX) Plant. On an annual basis, approximately 1 g of tritium is discharged in more than 1 x 10 6 L of process condensate water. Principal tritium release points and quantities are presented in section 4.0. The present study is intended to identify and evaluate alternate methods of tritium control and disposal that might merit additional study or development for potential application to Hanford Site effluents. 30 refs., 15 figs., 5 tabs

  12. Final programmatic environmental impact statement for tritium supply and recycling. Volume III

    International Nuclear Information System (INIS)

    1995-10-01

    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

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

    International Nuclear Information System (INIS)

    1995-10-01

    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. 550 refs

  14. 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)

  15. Current status of tritium calorimetry at TLK

    International Nuclear Information System (INIS)

    Buekki-Deme, A.; Alecu, C.G.; Kloppe, B.; Bornschein, B.

    2015-01-01

    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)

  16. Overview of the tritium system of Ignitor

    International Nuclear Information System (INIS)

    Rizzello, C.; Tosti, S.

    2008-01-01

    Among the recent design activities of the Ignitor program, the analysis of the tritium system has been carried out with the aim to describe the main equipments and the operations needed for supplying the deuterium-tritium mixtures and recovering the plasma exhaust. In fact, the tritium system of Ignitor provides for injecting deuterium-tritium mixtures into the vacuum chamber in order to sustain the fusion reaction: furthermore, it generally manages and controls the tritium and the tritiated materials of the machine fuel cycle. Main functions consist of tritium storage and delivery, tritium injection, tritium recovery from plasma exhaust, treatment of the tritiated wastes, detritiation of the contaminated atmospheres, tritium analysis and accountability. In this work an analysis of the designed tritium system of Ignitor is summarized

  17. Tritium, biography of an element

    International Nuclear Information System (INIS)

    Keller, C.

    1980-01-01

    Tritium is the lightest radioactive atom, an isotope of hydrogen. In science it has many uses, particularly for marking organic molecules in order to find out about biochemical and medical processes. But also the traces of tritium contained in rain or sea water are used for investigations; they range from establishing the vintage of old wines to ascertaining sea water mixtures. Tritium will become important in large-scale technology if it should become possible to construct fusion reactors, since it is one of the fuels. (orig.) [de

  18. Tritium monitor for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jalbert, R.A.

    1982-08-01

    This report describes the design, operation, and performance of a flow-through ion-chamber instrument designed to measure tritium concentrations in air containing /sup 13/N, /sup 16/N, and /sup 41/Ar produced by neutrons generated by D-T fusion devices. The instrument employs a chamber assembly consisting of two coaxial ionization chambers. The inner chamber is the flow-through measuring chamber and the outer chamber is used for current subtraction. A thin wall common to both chambers is opaque to the tritium betas. Currents produced in the two chambers by higher energy radiation are automatically subtracted, leaving only the current due to tritium.

  19. The organically bound tritium: an analyst vision

    International Nuclear Information System (INIS)

    Ansoborlo, E.; Baglan, N.

    2009-01-01

    The authors report the work of a work group on tritium analysis. They recall the different physical forms of tritium: gas (HT, hydrogen-tritium), water vapour (HTO or tritiated water) or methane (CH3T), but also in organic compounds (OBT, organically bound tritium) which are either exchangeable or non-exchangeable. They evoke measurement techniques and methods, notably to determine the tritium volume activity. They discuss the possibilities to analyse and distinguish exchangeable and non-exchangeable OBTs

  20. Tritium monitoring at the Sandia Tritium Research Laboratory

    International Nuclear Information System (INIS)

    Devlin, T.K.

    1978-10-01

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

  1. 2016 Accomplishments. Tritium aging studies on stainless steel. Forging process effects on the fracture toughness properties of tritium-precharged stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Michael J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-01

    Forged austenitic stainless steels are used as the materials of construction for pressure vessels designed to contain tritium at high pressure. These steels are highly resistant to tritium-assisted fracture but their resistance can depend on the details of the forging microstructure. During FY16, the effects of forging strain rate and deformation temperature on the fracture toughness properties of tritium-exposed-and-aged Type 304L stainless steel were studied. Forgings were produced from a single heat of steel using four types of production forging equipment – hydraulic press, mechanical press, screw press, and high-energy-rate forging (HERF). Each machine imparted a different nominal strain rate during the deformation. The objective of the study was to characterize the J-Integral fracture toughness properties as a function of the industrial strain rate and temperature. The second objective was to measure the effects of tritium and decay helium on toughness. Tritium and decay helium effects were measured by thermally precharging the as-forged specimens with tritium gas at 34.5 MPa and 350°C and aging for up to five years at -80°C to build-in decay helium prior to testing. The results of this study show that the fracture toughness properties of the as-forged steels vary with forging strain rate and forging temperature. The effect is largely due to yield strength as the higher-strength forgings had the lower toughness values. For non-charged specimens, fracture toughness properties were improved by forging at 871°C versus 816°C and Screw-Press forgings tended to have lower fracture toughness values than the other forgings. Tritium exposures reduced the fracture toughness values remarkably to fracture toughness values averaging 10-20% of as-forged values. However, forging strain rate and temperature had little or no effect on the fracture toughness after tritium precharging and aging. The result was confirmed by fractography which indicated that fracture modes

  2. Booster Long 13 irradiation studies

    Energy Technology Data Exchange (ETDEWEB)

    Leveling, A.; Mokhov, N.; Moore, C.D.; /Fermilab

    1998-06-01

    Extraction from the Booster to the Main Ring occurred at Long Straight 13. The nature of the extraction process was such that 1% to 2% of the beam was lost in this region. There was an appreciable amount of beam extracted as shown in Table 1, which gives the yearly integrated intensities from 1973 to 1997. A simple model of the extraction losses was set up by Chandra Bhat utilizing the program CASIM. A sample output I shown in figure 1 which gives contours of stars/cm3 in the dirt, also schematically depicted are the three six feet deep sampling holes which were drilled to map out this cascade. One aspect of this study has been the study of the production of non-migrating nuclides and further study may in fact yield better values for the K parameter, the probability per star that an atom of the particular nuclide will be produced. Also the results of this study can give experimental numbers for the production of other nuclides when the amount of Na22 has been calculated. However, the most important part of this study has been the determination of the amount of tritium produced by extraction from the Booster and the experimentally determined migration rate. If we look at the top sample result in hole S2 of 777 pCi/ml of tritium and use the experimentally determined rate of migration and the depth to the aquifer of 13.1m, they calculate that the concentration will have decayed away to 1.1E-8 pCi/ml. If we look at the bottom sample, which is 11.3 m away from the aquifer, they calculate that the 116 pCi/ml will have decayed to 5.2E-8 pCi/ml. The conclusions is that the rate of migration determined over the 24 year irradiation history of the Booster extraction point is small enough that there is no problem with migration of tritium to the aquifer.

  3. JET experiments with tritium and deuterium–tritium mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Lorne, E-mail: Lorne.Horton@jet.uk [JET Exploitation Unit, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Batistoni, P. [Unità Tecnica Fusione - ENEA C. R. Frascati - via E. Fermi 45, Frascati (Roma), 00044, Frascati (Italy); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Boyer, H.; Challis, C.; Ćirić, D. [CCFE, Culham Science Centre, Abingdon OX14 3DB, Oxon (United Kingdom); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Donné, A.J.H. [EUROfusion Programme Management Unit, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); FOM Institute DIFFER, PO Box 1207, NL-3430 BE Nieuwegein (Netherlands); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Eriksson, L.-G. [European Commission, B-1049 Brussels (Belgium); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Garcia, J. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Garzotti, L.; Gee, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB, Oxon (United Kingdom); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Hobirk, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Joffrin, E. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); and others

    2016-11-01

    Highlights: • JET is preparing for a series of experiments with tritium and deuterium–tritium mixtures. • Physics objectives include integrated demonstration of ITER operating scenarios, isotope and alpha physics. • Technology objectives include neutronics code validation, material studies and safety investigations. • Strong emphasis on gaining experience in operation of a nuclear tokamak and training scientists and engineers for ITER. - Abstract: Extensive preparations are now underway for an experiment in the Joint European Torus (JET) using tritium and deuterium–tritium mixtures. The goals of this experiment are described as well as the progress that has been made in developing plasma operational scenarios and physics reference pulses for use in deuterium–tritium and full tritium plasmas. At present, the high performance plasmas to be tested with tritium are based on either a conventional ELMy H-mode at high plasma current and magnetic field (operation at up to 4 MA and 4 T is being prepared) or the so-called improved H-mode or hybrid regime of operation in which high normalised plasma pressure at somewhat reduced plasma current results in enhanced energy confinement. Both of these regimes are being re-developed in conjunction with JET's ITER-like Wall (ILW) of beryllium and tungsten. The influence of the ILW on plasma operation and performance has been substantial. Considerable progress has been made on optimising performance with the all-metal wall. Indeed, operation at the (normalised) ITER reference confinement and pressure has been re-established in JET albeit not yet at high current. In parallel with the physics development, extensive technical preparations are being made to operate JET with tritium. The state and scope of these preparations is reviewed, including the work being done on the safety case for DT operation and on upgrading machine infrastructure and diagnostics. A specific example of the latter is the planned calibration at

  4. Thought experiment with tritium

    International Nuclear Information System (INIS)

    Anderson, H.F.; Everhart, J.L.; Hobrock, D.L.; Seabaugh, P.W.

    1995-01-01

    An experiment is proposed in which a minimum of thirty (30) grams of tritium is packaged as lithium tritide in a steel container weighing several kilograms. After decontamination of the outside surface, calorimetry measurements would be made, and the unit would be weighed very accurately. After several decades, the calorimeter and weight measurements would be repeated. If the weight measurements could be made with the required accuracy, it would be possible to correlate the observed change in mass with the total energy emitted (calculated from the mean energy measured by calorimetry) over the time interval. If successful, this experiment would, in the opinion of the authors, be the first laboratory experiment to directly verify the equivalency of mass and energy. 4 refs., 2 figs., 3 tabs

  5. New facility for post irradiation examination of neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi

    1995-01-01

    Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

  6. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

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

  7. Lifetime and shelf life of sealed tritium-filled plasma focus chambers with gas generator

    Directory of Open Access Journals (Sweden)

    B.D. Lemeshko

    2017-11-01

    Full Text Available The paper describes the operation features of plasma focus chambers using deuterium–tritium mixture. Handling tritium requires the use of sealed, vacuum-tight plasma focus chambers. In these chambers, there is an accumulation of the impurity gases released from the inside surfaces of the electrodes and the insulator while moving plasma current sheath inside chambers interacting with β-electrons generated due to the decay of tritium. Decay of tritium is also accompanied by the accumulation of helium. Impurities lead to a decreased yield of neutron emission from plasma focus chambers, especially for long term operation. The paper presents an option of absorption type gas generator in the chamber based on porous titanium, which allows to significantly increase the lifetime and shelf life of tritium chambers. It also shows the results of experiments on the comparison of the operation of sealed plasma focus chambers with and without the gas generator. Keywords: Plasma focus, Neutron yield, Tritium-filled plasma focus chambers, PACS Codes: 29.25.-v, 52.58.Lq

  8. Tritium transport around nuclear facilities

    International Nuclear Information System (INIS)

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

    1981-01-01

    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

  9. TFTR tritium inventory accountability system

    International Nuclear Information System (INIS)

    Saville, C.; Ascione, G.; Elwood, S.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Stencel, J.; Voorhees, D.; Tilson, C.

    1995-01-01

    This paper discusses the program, PPPL (Princeton Plasma Physics Laboratory) Material Control and Accountability Plan, that has been implemented to track US Department of Energy's tritium and all other accountable source material. Specifically, this paper details the methods used to measure tritium in various systems at the Tokamak Fusion Test Reactor; resolve inventory differences; perform inventory by difference inside the Tokamak; process and measure plasma exhaust and other effluent gas streams; process, measure and ship scrap or waste tritium on molecular sieve beds; and detail organizational structure of the Material Control and Accountability group. In addition, this paper describes a Unix-based computerized software system developed at PPPL to account for all tritium movements throughout the facility. 5 refs., 2 figs

  10. Tritium management for fusion reactors

    International Nuclear Information System (INIS)

    Rouyer, J.L.; Djerassi, H.

    1985-01-01

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

  11. Radiotoxicity of tritium in mammals

    International Nuclear Information System (INIS)

    Silini, G.; Metalli, P.; Vulpis, G.

    1972-12-01

    Basic data relative to tritium, its physicochemical behaviour in environment, its major sources of contamination and its metabolism through the mammalian organisms are reviewed. After considering the radiotoxicity of tritium particularly at the cellular and whole-body level the conclusion is drawn that the major uncertainties regard the fraction of tritium incorporated into the nuclei of some tissues. This fraction is eliminated very slowly and is capable of modifying the genetic structures of the nucleus. A more refined analysis of radiobiological phenomena and a better knowledge of the dose effect relationship should permit the extrapolation of the data to the low doses of tritium contamination. This extrapolation is of great interest in the field of public health for the elaboration of the relevant radioprotection standards

  12. Tritium transport around nuclear faciliteis

    International Nuclear Information System (INIS)

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

    1982-01-01

    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.)

  13. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.; Talbot, J.B.

    1980-01-01

    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

  14. Remediation of ground water containing volatile organic compounds and tritium

    International Nuclear Information System (INIS)

    Shukla, S.N.; Folsom, E.N.

    1994-03-01

    The Trailer 5475 (T-5475) East Taxi Strip Area at Lawrence Livermore National Laboratory (LLNL), Livermore, California was used as a taxi strip by the US Navy to taxi airplanes to the runway from 1942 to 1947. Solvents were used in some unpaved areas adjacent to the East Taxi Strip for cleaning airplanes. From 1953 through 1976, the area was used to store and treat liquid waste. From 1962 to 1976 ponds were constructed and used for evaporation of liquid waste. As a result, the ground water in this area contains volatile organic compounds (VOCs) and tritium. The ground water in this area is also known to contain hexavalent chromium that is probably naturally occurring. Therefore, LLNL has proposed ''pump-and-treat'' technology above grade in a completely closed loop system. The facility will be designed to remove the VOCs and hexavalent chromium, if any, from the ground water, and the treated ground water containing tritium will be reinjected where it will decay naturally in the subsurface. Ground water containing tritium will be reinjected into areas with equal or higher tritium concentrations to comply with California regulations

  15. Cosmogenic production of tritium in dark matter detectors

    Science.gov (United States)

    Amaré, J.; Castel, J.; Cebrián, S.; Coarasa, I.; Cuesta, C.; Dafni, T.; Galán, J.; García, E.; Garza, J. G.; Iguaz, F. J.; Irastorza, I. G.; Luzón, G.; Martínez, M.; Mirallas, H.; Oliván, M. A.; Ortigoza, Y.; Ortiz de Solórzano, A.; Puimedón, J.; Ruiz-Chóliz, E.; Sarsa, M. L.; Villar, J. A.; Villar, P.

    2018-01-01

    The direct detection of dark matter particles requires ultra-low background conditions at energies below a few tens of keV. Radioactive isotopes are produced via cosmogenic activation in detectors and other materials and those isotopes constitute a background source which has to be under control. In particular, tritium is specially relevant due to its decay properties (very low endpoint energy and long half-life) when induced in the detector medium, and because it can be generated in any material as a spallation product. Quantification of cosmogenic production of tritium is not straightforward, neither experimentally nor by calculations. In this work, a method for the calculation of production rates at sea level has been developed and applied to some of the materials typically used as targets in dark matter detectors (germanium, sodium iodide, argon and neon); it is based on a selected description of tritium production cross sections over the entire energy range of cosmic nucleons. Results have been compared to available data in the literature, either based on other calculations or from measurements. The obtained tritium production rates, ranging from a few tens to a few hundreds of nuclei per kg and per day at sea level, point to a significant contribution to the background in dark matter experiments, requiring the application of specific protocols for target material purification, material storing underground and limiting the time the detector is on surface during the building process in order to minimize the exposure to the most dangerous cosmic ray components.

  16. Deuterium-tritium fuel self-sufficiency in fusion reactors

    International Nuclear Information System (INIS)

    Abdou, M.A.; Vold, E.L.; Gung, C.Y.; Youssef, M.Z.; Shin, K.

    1986-01-01

    Conditions necessary to achieve deuterium-tritium fuel self-sufficiency in fusion reactors are derived through extensive modeling and calculations of the required and achievable tritium breeding ratios as functions of the many reactor parameters and candidate design concepts. It is found that the excess margin in the breeding potential is not sufficient to cover all present uncertainties. Thus, the goal of attaining fuel self-sufficiency significantly restricts the allowable parameter space and design concepts. For example, the required breeding ratio can be reduced by (A) attaining high tritium fractional burnup, >5%, in the plasma, (B) achieving very high reliability, >99%, and very short times, <1 day, to fix failures in the tritium processing system, and (C) ensuring that nonradioactive decay losses from all subsystems are extremely low, e.g., <0.1% for the plasma exhaust processing system. The uncertainties due to nuclear data and calculational methods are found to be significant, but they are substantially smaller than those due to uncertainties in system definition

  17. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

    Wall, W.R.; Cruz, S.L.

    1985-08-01

    This instruction describes the tritium accountability procedures practiced by the Tritium Research Laboratory, at Sandia National Laboratories, Livermore. The accountability procedures are based upon the Sandia National Laboratories, Livermore, Nuclear Materials Operations Manual, SAND83-8036. The Nuclear Materials Operations Manual describes accountability techniques which are in compliance with the Department of Energy 5630 series Orders, Code of Federal Regulations, and Sandia National Laboratories Instructions

  18. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

    Gouge, M.J.; Baylor, L.R.; Cole, M.J.; Combs, S.K.; Dyer, G.R.; Fehling, D.T.; Fisher, P.W.; Foust, C.R.; Langley, R.A.; Milora, S.L.; Qualls, A.L.; Wilgen, J.B.; Schmidt, G.L.; Barnes, G.W.; Persing, R.G.

    1992-01-01

    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

  19. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

    Wall, W.R.

    1981-03-01

    This instruction describes the tritium accountability procedures practiced by the Tritium Research Laboratory, Building 968 at Sandia National Laboratories, Livermore. The accountability procedures are based upon the Sandia National Laboratories, Livermore, Nuclear Materials Operations Manual, SAND78-8018. The Nuclear Materials Operations Manual describes accountability techniques which are in compliance with the Department of Energy Manual, Code of Federal Regulations, and Sandia National Laboratories Instructions

  20. Design and test about de tritium system to filling tritium glove box

    International Nuclear Information System (INIS)

    Lei, Jiarong; Du, Yang; Yang, Yong

    2008-01-01

    In order to deal tritium permeated from inflating tritium system at the scene of inflating tritium, dealing waste tritium gas system was designed according to demand and action of dealing waste tritium gas from inflating tritium, and the data of character and volume about appliance of catalyst reaction and drying agent was calculated. Through the test at the scene of inflating tritium, it is result that dealing waste tritium gas system's efficiency reaches above 85% average in circulatory system, so that it can be used in practice extensively. (author)

  1. A laboratory information management system for the analysis of tritium (3H) in environmental waters.

    Science.gov (United States)

    Belachew, Dagnachew Legesse; Terzer-Wassmuth, Stefan; Wassenaar, Leonard I; Klaus, Philipp M; Copia, Lorenzo; Araguás, Luis J Araguás; Aggarwal, Pradeep

    2018-07-01

    Accurate and precise measurements of low levels of tritium ( 3 H) in environmental waters are difficult to attain due to complex steps of sample preparation, electrolytic enrichment, liquid scintillation decay counting, and extensive data processing. We present a Microsoft Access™ relational database application, TRIMS (Tritium Information Management System) to assist with sample and data processing of tritium analysis by managing the processes from sample registration and analysis to reporting and archiving. A complete uncertainty propagation algorithm ensures tritium results are reported with robust uncertainty metrics. TRIMS will help to increase laboratory productivity and improve the accuracy and precision of 3 H assays. The software supports several enrichment protocols and LSC counter types. TRIMS is available for download at no cost from the IAEA at www.iaea.org/water. Copyright © 2018 Elsevier Ltd. All rights reserved.

  2. Separation of Tritium from Wastewater

    International Nuclear Information System (INIS)

    JEPPSON, D.W.

    2000-01-01

    A proprietary tritium loading bed developed by Molecular Separations, Inc (MSI) has been shown to selectively load tritiated water as waters of hydration at near ambient temperatures. Tests conducted with a 126 (micro)C 1 tritium/liter water standard mixture showed reductions to 25 (micro)C 1 /L utilizing two, 2-meter long columns in series. Demonstration tests with Hanford Site wastewater samples indicate an approximate tritium concentration reduction from 0.3 (micro)C 1 /L to 0.07 (micro)C 1 /L for a series of two, 2-meter long stationary column beds Further reduction to less than 0.02 (micro)C 1 /L, the current drinking water maximum contaminant level (MCL), is projected with additional bed media in series. Tritium can be removed from the loaded beds with a modest temperature increase and the beds can be reused Results of initial tests are presented and a moving bed process for treating large quantities of wastewaters is proposed. The moving bed separation process appears promising to treat existing large quantities of wastewater at various US Department of Energy (DOE) sites. The enriched tritium stream can be grouted for waste disposition. The separations system has also been shown to reduce tritium concentrations in nuclear reactor cooling water to levels that allow reuse. Energy requirements to reconstitute the loading beds and waste disposal costs for this process appear modest

  3. Tritium research activities in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Jung, E-mail: kjjung@nfri.re.kr [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Yun, Sei-Hun, E-mail: shyun@nfri.re.kr [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Chang, Min Ho; Kang, Hyun-Goo; Chung, Dongyou; Cho, Seungyon; Lee, Hyeon Gon [National Fusion Research Institute, Yusung-gu, Daejeon 305-333 (Korea, Republic of); Chung, Hongsuk; Choi, Woo-Seok [Korea Atomic Energy Research Institute, Yusung-gu, Daejeon 305-353 (Korea, Republic of); Song, Kyu-Min; Moon, Chang-Bae [Korea Hydro & Nuclear Power Central Research Institute, Yusung-gu, Daejeon 305-343 (Korea, Republic of); Lee, Euy Soo [Dongguk University, Jung-gu, Seoul, 100-715 (Korea, Republic of); Cho, Jungho; Kim, Dong-Sun [Kongju National University, Cheonan, Chungnam, 330-717 (Korea, Republic of); Moon, Hung-Man [Daesung Industrial Gases Co., Ltd., Danwon-gu, Ansan-si, Gyeonggi-do, 425-090 (Korea, Republic of); Noh, Seung Jeong [Dankook University, Suji-gu, Yongin-si, Gyeonggi-do, 448-701 (Korea, Republic of); Ju, Hyunchul [Inha University, Nam-gu, Incheon, 402-751 (Korea, Republic of); Hong, Tae-Whan [Korea National University of Transportation, Chungju, Chungbuk, 380-702 (Korea, Republic of)

    2016-12-15

    Highlights: • NFRI, KAERI and KHNP CRI are major leading group for the ITER tritium SDS design; studying engineering, simulation of hydride bed, risk analysis (on safety, HAZOP), basic study, control logic & sequential operation, and others. KHNP has WTRF which gives favorable experiences for collaboration researchers. • Supplementary research partners: Five Universities (Dongguk University and POSTECH, Inha University, Dankook University, Korea National Transport University, and Kongju National University) and one industrial company (Daesung Industrial Gases Co., Ltd.); studying on basic and engineering, programming & simulation on the various topics for ITER tritium SDS, TEP, ISS, ADS, and etc. - Abstract: Major progress in tritium research in the Republic of Korea began when Korea became responsible for ITER tritium Storage and Delivery System (SDS) procurement package which is part of the ITER Fuel Cycle. To deliver the tritium SDS package, a variety of research institutes, universities and industry have respectively taken roles and responsibilities in developing technologies that have led to significant progress. This paper presents the current work and status of tritium related technological research and development (R&D) in Korea and introduces future R&D plans in the area of fuel cycle systems for fusion power generation.

  4. Mean residence times for tritium in some terrestrial plants

    International Nuclear Information System (INIS)

    Gogate, S.S.; Krishnamoorthy, T.M.; Soman, S.D.

    1975-01-01

    Uptake of tritiated water from a single or multiple exposure, its fixation and elimination from 3 terrestrial plants, Raphanus sativus L., Amaranthus viridis L. and Phyllanthus fraternus Webster under experimental field conditions are described. Tissue free water tritium (TFWT) attains peak concentration within 4 hr after a single exposure in R. sativus and then decreases with a single component from both leaves and root, with a biological half-time of 36.4 +- 4.2 and 48.5 +- 7.2 hr respectively. TFWT in leaves of A. viridis reaches its peak value within 0.5 hr of exposure and its decay exhibits 2 component fall, one having a very short half-life of 2.4 +- 0.7 hr and the other a long half-time of 86.1 +- 2.0 hr. The two component decay of tritium in TFWT is well represented in P. fraternus. The long-lived components are nearly 8 times of the short-lived ones, both in leaves and stems respectively. Generally, the long -lived component accounts for 15% of the peak TFWT. Tissue bound tritium (TBT) reaches to 4% of TFWT in Phyllanthus sp. TBT elimination time is many times longer than the experimental periods employed in the present study. (author)

  5. Tritium concentrations in tree ring cellulose

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  6. 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)

  7. Sublethal effects of tritium on aquatic systems, effects of low-level chronic irradiation on embryonic development, effects of beryllium and lithium on aquatic systems, and teratogenic effects of low-level magnetic fields

    International Nuclear Information System (INIS)

    Strand, J.A.; Roesijadi, G.; Emery, R.M.

    1978-01-01

    It is conceivable, and possible, that with increasing applications of nuclear energy, greater quantities of potentially harmful radionuclides will be released to the environment. This may be particularly true for advanced reactor designs such as the fusion reactor where radionuclides, principally tritium, may be expected to escape from the plant both in gaseous and liquid effluents in quantities significantly greater than for present PWR or BWR designs. Foreseeing such contingencies, the research programs described herein are in response to the need to measure the potential radiation effects of tritium releases on individuals, and ultimately populations and biotic communities. As a first approach, our efforts are directed to determine effects of low-level chronic exposures on developing embryo and larval stages, clearly the most radiosensitive. The anticipated increase in the release of beryllium and lithium from mining, refining, and fabrication of materials used during construction of fusion reactors has also caused concern as to potential adverse effects on the environment. Accordingly, FY-78 fusion related research will include efforts to study the metabolism of each metal in the living organism, and to determine at what levels toxicity may be expected. Fusion related research will also include preliminary experiments on the effects of low-level magnetic fields. It is conceivable that magnetic fields of 70 to 450 gauss will be encountered by attendant personnel working in the transport and hot cell areas of fusion reactors. Also, those personnel assigned to the areas immediately surrounding the reactor may be exposed

  8. Exploration for tritium-free water

    International Nuclear Information System (INIS)

    Hussain, S.D.

    1982-10-01

    Tritium-free water is generally required in large quantities for the preparation of laboratory tritium standards as well as blanks which are used to determine background count rate in the measurement of low level tritium concentrations in water samples by liquid scintillation counting method. In order to meet the requirements of tritium-free water and save the recurring expenditure on its import from abroad, exploration for locating its source in the country was undertaken. Water samples collected from a few possible sources were analysed precisely for their tritium content at the International Atomic Energy Agency, Vienna, Austria and a source of tritium-free water was determined. (authors)

  9. Tritium problems in fusion reactor systems

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1975-01-01

    A brief introduction is given to the role tritium will play in the development of fusion power. The biological and worldwide environmental behavior of tritium is reviewed. The tritium problems expected in fusion power reactors are outlined. A few thoughts on tritium permeation and recent results for tritium cleanup and CT 4 accumulation are presented. Problems involving the recovery of tritium from the breeding blanket in fusion power reactors are also considered, including the possible effect of impurities in lithium blankets and the use of lithium as a regenerable getter pump. (auth)

  10. Studies on chemical phenomena of high concentration tritium water and organic compounds of tritium from viewpoint of the tritium confinement

    International Nuclear Information System (INIS)

    Yamanishi, Toshihiko; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Hara, Masanori; Sugiyama, Takahiko; Okuno, Kenji

    2009-01-01

    As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated two research programs on chemical phenomena of high concentration tritium water and organic compounds of tritium from view point of the tritium confinement have been conducted by the C01 team. The results are summarized as follows: (1) Chemical effects of the high concentration tritium water on stainless steels as structural materials of fusion reactors were investigated. Basic data on tritium behaviors at the metal-water interface and corrosion of metal in tritium water were obtained. (2) Development of the tritium confinement and extraction system for the circulating cooling water in the fusion reactor was studied. Improvement was obtained in the performance of a chemical exchange column and catalysts as major components of the water processing system. (J.P.N.)

  11. Experiment designed to measure the RBE of tritium for the induction of myeloid leukaemia in animals

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J R; Myers, D K; Gragtmans, N J

    1986-01-01

    The range in RBE vales measured for tritium 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 utilise 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 k Vp 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. Acute myeloid leukaemia is one of the earliest recognised examples of radiogenic cancer in humans and this endpoint has also been the subject of animal studies. A brief review is given of these animal studies to see if this endpoint is suitable for an experiment to measure the tritium RBE relative to 200 k Vp X rays. It was concluded that the male CBA/H mouse would be a suitable species and an experiment involving 5000 animals in four to five year study would be required to provide a useful estimate of the RBE for tritium.

  12. Tritium solid targets for intense D-T neutron production and its related problems

    International Nuclear Information System (INIS)

    Sumita, Kenji

    1988-01-01

    This review paper is divided into three parts. Firstly, to attain an intense neutron production rate, the construction of a design with a higher tritium-containing surface and an effective cooling system like a rotating target device are discussed. The maximum attainable intensity based on tritium solid targets shall be estimated regarding planning for future D-T sources. Secondly, on the way to carry out some experiments, an absolute intensity calibration and an angular dependent neutron energy spectrum of the neutron source are essential parameters to analyse the results of the experiments. Sometimes the space dependent neutron spectrum is required as well as the space dependent neutron flux near the targets and irradiation samples. The measurement methods and their examples are reviewed for tritium solid targets. The third part is devoted to discuss the protection to tritium contamination problems due to unavoidable release of tritium gas from targets. Performance and effectiveness of tritium collection systems for intense D-T neutron sources shall be discussed in some examples. Tritium contamination incidents due to the faulted film powder of target surface are also reported in some real incident cases. (author). Abstract only

  13. MISTRAL: A comprehensive model for tritium transport in lithium-base ceramics. Pt. 2

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.; Abdou, M.A.

    1990-01-01

    A new tritium transport model called MISTRAL (Model for Investigative Studies of Tritium Release in Lithium Ceramics) has been developed to describe and predict the kinetics of tritium release in lithium ceramic materials for tritium breeding applications in fusion blankets. The model has transient capabilities and has been developed to analyze the full range of transient conditions produced in in-pile tritium recovery experiments and expected in fusion blankets. Calibration of the model against experiments has been done in parallel with its development in order to assess its predictive capabilities and to identify the ranges of potential applicability. The comparisons of the results available for lithium metasilicate and aluminate samples irradiated respectively in the two in-pile tritium recovery experiments LISA1 and MOZART are presented and discussed in this paper. They have been selected for the calibration of the codes as being good examples of various features relevant for tritium release analysis in ceramic breeders under different transient conditions such as change in temperature, purge gas composition and reactor power. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  15. Simulating tritium retention in tungsten with a multiple trap model in the TMAP code

    International Nuclear Information System (INIS)

    Merrill, Brad J.; Shimada, Masashi; Humrickhouse, Paul W.

    2013-01-01

    Accurately predicting the quantity of tritium retained in plasma facing components is a key safety issue for licensing future fusion power reactors. Retention of tritium in the lattice damage caused when high energy neutrons collide with atoms in the structural material of the reactor's plasma facing components (PFCs) is an area of ongoing experimental research at the Idaho National Laboratory (INL) under the US/Japan TITAN collaboration. Recent experiments with the Tritium Plasma Experiment (TPE), located in the INL's Safety and Tritium Applied Research (STAR) facility, demonstrate that this damage can only be simulated by computer codes like the Tritium Migration Analysis Program (TMAP) if one assumes that the lattice damage produced by these neutrons results in multiple types of hydrogen traps (energy wells) within the material, each possessing a different trap energy and density. Previous attempts to simulate the quantity of deuterium released from neutron irradiated TPE tungsten targets indicated that at least six different traps are required by TMAP to model this release. In this paper we describe a recent extension of the TMAP trap site model to include as many traps as required by the user to simulate retention of tritium in neutron damaged tungsten. This model has been applied to data obtained for tungsten irradiated to a damage level of 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) after exposure to a plasma in TPE. (author)

  16. Radioactive Decay

    Science.gov (United States)

    Radioactive decay is the emission of energy in the form of ionizing radiation. Example decay chains illustrate how radioactive atoms can go through many transformations as they become stable and no longer radioactive.

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

    International Nuclear Information System (INIS)

    Sherwood, A.E.

    1980-01-01

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

  18. Turkey Point tritium. Progress report

    International Nuclear Information System (INIS)

    Ostlund, H.G.; Dorsey, H.G.

    1976-01-01

    In 1972-73 the Florida Power and Light Company (FPL) began operation of two nuclear reactors at Turkey Point on lower Biscayne Bay. One radioactive by-product resulting from the operation of the nuclear reactors, tritium, provides a unique opportunity to study transport and exchange processes on a local scale. Since the isotope in the form of water is not removed from the liquid effluent, it is discharged to the cooling canal system. By studying its residence time in the canal and the pathways by which it leaves the canals, knowledge of evaporative process, groundwater movement, and bay exchange with the ocean can be obtained. Preliminary results obtained from measurement of tritium levels, both in the canal system and in the surrounding environment are discussed. Waters in lower Biscayne Bay and Card and Barnes Sounds receive only a small portion of the total tritium produced by the nuclear plant. The dominating tritium loss most likely is through evaporation from the canals. The capability of measuring extremely low HTO levels allows the determination of the evaporation rate experimentally by measuring the tritium levels of air after having passed over the canals

  19. Tritium Management Loop Design Status

    Energy Technology Data Exchange (ETDEWEB)

    Rader, Jordan D. [ORNL; Felde, David K. [ORNL; McFarlane, Joanna [ORNL; Greenwood, Michael Scott [ORNL; Qualls, A L. [ORNL; Calderoni, Pattrick [Idaho National Laboratory (INL)

    2017-12-01

    This report summarizes physical, chemical, and engineering analyses that have been done to support the development of a test loop to study tritium migration in 2LiF-BeF2 salts. The loop will operate under turbulent flow and a schematic of the apparatus has been used to develop a model in Mathcad to suggest flow parameters that should be targeted in loop operation. The introduction of tritium into the loop has been discussed as well as various means to capture or divert the tritium from egress through a test assembly. Permeation was calculated starting with a Modelica model for a transport through a nickel window into a vacuum, and modifying it for a FLiBe system with an argon sweep gas on the downstream side of the permeation interface. Results suggest that tritium removal with a simple tubular permeation device will occur readily. Although this system is idealized, it suggests that rapid measurement capability in the loop may be necessary to study and understand tritium removal from the system.

  20. Tritium assay of Li2O pellets in the LBM/LOTUS experiments

    International Nuclear Information System (INIS)

    Quanci, J.; Azam, S.; Bertone, P.

    1986-01-01

    One of the objectives of the Lithium Blanket Module (LBM) program is to test the ability of advanced neutronics codes to model the tritium breeding characteristics of a fusion blanket exposed to a toroidal fusion neutron source. The LBM consists of over 20,000 cylindrical lithium oxide pellets and numerous diagnostic pellets and wafers. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a Haefely sealed neutron generator that gives a point deuterium-tritium neutron source up to 5 x 10 12 14-MeV n/s. Both Princeton Plasma Physics Lab. (PPPL) and EPFL assayed the tritium bred at various positions in the LBM. EPFL employed a dissolution technique while PPPL recovered the tritium by a thermal extraction method

  1. Study of measurement method of tritium induced in concrete of high-energy proton accelerator facilities

    International Nuclear Information System (INIS)

    Ohtsuka, N.; Ishihama, S.; Kunifuda, T.; Hayasaka, N.; Miura, T.

    2001-01-01

    Various long-loved radionuclides, 3 H, 7 Be, 22 Na, 51 Cr, 54 Mn, 56 Co, 57 Co, 60 Co, 134 Cs, 152 Eu and 154 Eu, have been produced in the shielding concrete of high energy proton accelerator facility through both nuclear spallation reactions and thermal neutron capture reactions of concrete elements, during machine operation. Tritium is the most important nuclide from the radiation protection. There were, however, few measurements of tritium concentration induced in the shielding concrete. In this study, the conditions of measurement method of tritium concentration induced in shielding concrete have been investigated using the activated shielding concrete of the 12 GeV proton beam-line tunnel at KEK and the standard rock (JG-1) irradiated of thermal neutron at the reactor. And the depth profiles of tritium induced in the shielding concrete of slow extracted proton beam line at KEK were determined using this method. (author)

  2. Tritium in the environment. Knowledge synthesis

    International Nuclear Information System (INIS)

    2009-01-01

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

  3. Recommended radiological controls for tritium operations

    International Nuclear Information System (INIS)

    Mansfield, G.

    1992-01-01

    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

  4. Tritium in metals: Techniques of preparation

    International Nuclear Information System (INIS)

    Laesser, R.; Klatt, K.H.; Mecking, P.; Wenzl, H.

    1982-08-01

    In order to study the behavior of tritium in metals, an all metal apparatus has been built for the safe handling of 100 mg of tritium. Samples of palladium, vanadium, niobium, and tantalum were loaded with tritium, deuterium or hydrogen. Some details of the phase diagrams could be established by DTA and by measurement of the lattice parameters. The diffusion of tritium in V, Nb, and Ta was studied with the Gorsky-effect. (TWO)

  5. Tritium decontamination of machine components and walls

    International Nuclear Information System (INIS)

    Hircq, B.; Wong, K.Y.; Jalbert, R.A.; Shmayda, W.T.

    1991-01-01

    Tritium decontamination techniques for machine components and their application at tritium handling facilities are reviewed. These include commonly used methods such as vacuuming, purging, thermal desorption and isotopic exchange as well as less common methods such as chemical/electrochemical etching, plasma discharge cleaning, and destructive methods. Problems associated with tritium contamination of walls and use of protective coatings are reviewed. Tritium decontamination considerations at fusion facilities are discussed

  6. Investigating Unsaturated Zone Travel Times with Tritium and Stable Isotopes

    Science.gov (United States)

    Visser, A.; Thaw, M.; Van der Velde, Y.

    2017-12-01

    Travel times in the unsaturated zone are notoriously difficult to assess. Travel time tracers relying on the conservative transport of dissolved (noble) gases (tritium-helium, CFCs or SF6) are not applicable. Large water volume requirements of other cosmogenic radioactive isotopes (sulfur-35, sodium-22) preclude application in the unsaturated zone. Prior investigations have relied on models, introduced tracers, profiles of stable isotopes or tritium, or a combination of these techniques. Significant unsaturated zone travel times (UZTT) complicate the interpretation of stream water travel time tracers by ranked StorAge Selection (rSAS) functions. Close examination of rSAS functions in a sloping soil lysimeter[1] show the effect of the UZTT on the shape of the rSAS cumulative distribution function. We studied the UZTT at the Southern Sierra Critical Zone Observatory (SS-CZO) using profiles of tritium and stable isotopes (18O and 2H) in the unsaturated zone, supported by soil water content data. Tritium analyses require 100-500 mL of soil water and therefore large soil samples (1-5L), and elaborate laboratory procedures (oven drying, degassing and noble gas mass spectrometry). The high seasonal and interannual variability in precipitation of the Mediterranean climate, variable snow pack and high annual ET/P ratios lead to a dynamic hydrology in the deep unsaturated soils and regolith and highly variable travel time distributions. Variability of the tritium concentration in precipitation further complicates direct age estimates. Observed tritium profiles (>3 m deep) are interpreted in terms of advective and dispersive vertical transport of the input variability and radioactive decay of tritium. Significant unsaturated zone travel times corroborate previously observed low activities of short-lived cosmogenic radioactive nuclides in stream water. Under these conditions, incorporating the UZTT is critical to adequately reconstruct stream water travel time distributions. 1

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  8. Weak decays

    International Nuclear Information System (INIS)

    Wojcicki, S.

    1978-11-01

    Lectures are given on weak decays from a phenomenological point of view, emphasizing new results and ideas and the relation of recent results to the new standard theoretical model. The general framework within which the weak decay is viewed and relevant fundamental questions, weak decays of noncharmed hadrons, decays of muons and the tau, and the decays of charmed particles are covered. Limitation is made to the discussion of those topics that either have received recent experimental attention or are relevant to the new physics. (JFP) 178 references

  9. Tritium calorimeter setup and operation

    CERN Document Server

    Rodgers, D E

    2002-01-01

    The LBNL tritium calorimeter is a stable instrument capable of measuring tritium with a sensitivity of 25 Ci. Measurement times range from 8-hr to 7-days depending on the thermal conductivity and mass of the material being measured. The instrument allows accurate tritium measurements without requiring that the sample be opened and subsampled, thus reducing personnel exposure and radioactive waste generation. The sensitivity limit is primarily due to response shifts caused by temperature fluctuation in the water bath. The fluctuations are most likely a combination of insufficient insulation from ambient air and precision limitations in the temperature controller. The sensitivity could probably be reduced to below 5 Ci if the following improvements were made: (1) Extend the external insulation to cover the entire bath and increase the top insulation. (2) Improve the seal between the air space above the bath and the outside air to reduce evaporation. This will limit the response drift as the water level drops. (...

  10. Polymeric media for tritium fixation

    International Nuclear Information System (INIS)

    Franz, J.A.; Burger, L.L.

    1975-01-01

    The synthesis and leach testing of several polymeric media for tritium fixation are presented. Tritiated bakelite, poly(acrylonitrile) and polystyrene successfully fixed tritium. Tritium leach rates at the tracer level appear to be negligible. Advantages and disadvantages of the processes are discussed, and further bench-scale investigations underway are reported. Rough cost estimates are presented for the different media and are compared with alternate approaches such as deep-well injection and long-term tank storage. Polymeric media costs are high compared to deep-well storage and are of the same order of magnitude per liter of water as for isotopic enrichment. With this limitation, polymeric media can be economically feasible only for highly concentrated tritiated wastes. It is recommended that the bakelite and polystyrene processes be examined on a larger scale to permit more accurate cost analysis and process design. (auth)

  11. Tritium processing using metal hydrides

    International Nuclear Information System (INIS)

    Mallett, M.W.

    1986-01-01

    E.I. duPont de Nemours and Company is commissioned by the US Department of Energy to operate the Savannah River Plant and Laboratory. The primary purpose of the plant is to produce radioactive materials for national defense. In keeping with current technology, new processes for the production of tritium are being developed. Three main objectives of this new technology are to ease the processing of, ease the storage of, and to reduce the operating costs of the tritium production facility. Research has indicated that the use of metal hydrides offers a viable solution towards satisfying these objectives. The Hydrogen and Fuels Technology Division has the responsibility to conduct research in support of the tritium production process. Metal hydride technology and its use in the storage and transportation of hydrogen will be reviewed

  12. Implanted-tritium permeation experiments

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Holland, D.F.; Casper, L.A.; Hsu, P.Y.; Miller, L.G.; Schmunk, R.E.; Watts, K.D.; Wilson, C.J.; Kershner, C.J.; Rogers, M.L.

    1982-04-01

    In fusion reactors, charge exchange neutral atoms of tritium coming from the plasma will be implanted into the first wall and other interior structures. EG and G Idaho is conducting two experiments to determine the magnitude of permeation into the coolant streams and the retention of tritium in those structures. One experiment uses an ion gun to implant deuterium. The ion gun will permit measurements to be made for a variety of implantation energies and fluxes. The second experiment utilizes a fission reactor to generate a tritium implantation flux by the 3 He(n,p) 3 H reaction. This experiment will simulate the fusion reactor radiation environment. We also plan to verify a supporting analytical code development program, in progress, by these experiments

  13. Second international comparison on measuring techniques of tritium production rate for fusion neutronics experiments (ICMT-2)

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Maekawa, Hiroshi

    1993-02-01

    An second international comparison on measuring techniques of tritium production rates for fusion neutronics experiments (ICMT-2) has been performed. The purpose is to evaluate the measurement accuracy of tritium production rates in the current measurement techniques. Two 14 MeV neutron source facilities, FNS at JAERI-Japan and LOTUS at EPFL-Switzerland, were used for this purpose. Nine groups out of seven countries participated in this program. A fusion simulated blanket assembly of simple-geometry was served as the test bed at each facility, in which Li-containing samples from the participants were irradiated in an uniform neutron field. The tritium production rates were determined by the participants using their own ways by using the liquid scintillation counting method. Tritiated water sample with unknown but the same concentration was also distributed and its concentration was measured to make a common reference. The standard deviation of measured tritium production rates among participants was about 10 % for both FNS and LOTUS irradiation levels: 4x10 -13 T-atoms/Li-atom and 1.6x10 -12 T-atoms/Li-atom at a sample, respectively. This standard deviation exceeds the expected deviation of 5 % in this program. It is presumed that the deviation of 10 % is caused mainly by the systematic and unknown errors in a process of tritium extraction from the irradiated samples depending on each organization. (author)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  15. Tritium emissions from a detritiation facility

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  16. Tritium inventory prediction in a CANDU plant

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  17. Tritium oxidation and exchange: preliminary studies

    International Nuclear Information System (INIS)

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

    1978-05-01

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

  18. Bioassay guideline 2: guidelines for tritium bioassay

    International Nuclear Information System (INIS)

    1983-01-01

    This guideline is one of a series under preparation by the Federal-Provincial Working Group on Bioassay and In Vivo Monitoring Criteria. In this report tritium compounds have been grouped into four categories for the purpose of calculating Annual Limits on Intake and Investigation Levels: tritium gas, tritiated water, tritium-labelled compounds and nucleic acid precursors

  19. 10 CFR 30.55 - Tritium reports.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Tritium reports. 30.55 Section 30.55 Energy NUCLEAR..., Inspections, Tests, and Reports § 30.55 Tritium reports. (a)-(b) [Reserved] (c) Except as specified in paragraph (d) of this section, each licensee who is authorized to possess tritium shall report promptly to...

  20. Toxicity and dosimetry of tritium

    International Nuclear Information System (INIS)

    Myers, D.K.; Johnson, J.R.

    1991-01-01

    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 60 Co 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, 60 Co 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.)

  1. The INEL Tritium Research Facility

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-01-01

    The Tritium Research Facility (TRF) at the Idaho National Engineering Laboratory (INEL) is a small, multi-user facility dedicated to research into processes and phenomena associated with interaction of hydrogen isotopes with other materials. Focusing on bench-scale experiments, the main objectives include resolution of issues related to tritium safety in fusion reactors and the science and technology pertinent to some of those issues. In this report the TRF and many of its capabilities will be described. Work presently or recently underway there will be discussed, and the implications of that work to the development of fusion energy systems will be considered. (orig.)

  2. The INEL Tritium Research Facility

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R. (Idaho National Engineering Lab., Idaho Falls (USA))

    1990-06-01

    The Tritium Research Facility (TRF) at the Idaho National Engineering Laboratory (INEL) is a small, multi-user facility dedicated to research into processes and phenomena associated with interaction of hydrogen isotopes with other materials. Focusing on bench-scale experiments, the main objectives include resolution of issues related to tritium safety in fusion reactors and the science and technology pertinent to some of those issues. In this report the TRF and many of its capabilities will be described. Work presently or recently underway there will be discussed, and the implications of that work to the development of fusion energy systems will be considered. (orig.).

  3. Tritium turnover in succulent plants

    International Nuclear Information System (INIS)

    Krishnamoorthy, T.M.; Gogate, S.S.; Soman, S.D.

    1977-01-01

    Measurements of turnover rates for tissue free water tritium (TFWT) and tissue bound tritium (TBT) were carried out in three succulent plants, Opuntia sp., E. Trigona and E. Mili using tritiated water as tracer. The estimated half-times were 52, 57.5 and 80 days for TFWT and 212, 318 and 132 days for TBT in the stems of the above plants respectively. Opuntia sp. showed significant incorporation of TBT, 10% of TFWT on weight basis, while the other two plants showed lesser incorporation, 2-3% of TFWT. However, the leaves of E. Mili indicated the same level of fixation of TBT as the stem of Opuntia sp. (author)

  4. Limits on neutrino masses from tritium beta decay

    CERN Document Server

    Bonn, J; Bornschein, L; Flatt, B; Kraus, C V; Otten, E W; Schall, J P; Thuemmler, T; Weinheimer, C

    2002-01-01

    The presently lowest limit for the mass of the electron neutrino is m nu < 2.2 eV (95 % C.L.) derived from measurements at Mainz up to 1999. The data taken in 2000 are not fully analyzed yet but limits of possible distortions as reported by the Troitsk group can be given. The next generation neutrino mass experiment KATRIN is briefly discussed.

  5. Present status and program of radiobiological tritium research in Japan

    International Nuclear Information System (INIS)

    Tazima, Y.

    1981-01-01

    Recent projects of the study on the biological effects of tritium in Japan, were reviewed. The investigations were about the behavior of tritium in food chain, i.e., intake to organ or tissue, retention, and distribution, and in genetic stage. As an example RBE studies on the induction of mutations in the silkworm and on transforming were described. Tritiated water was injected into wild type female pupae and radiation dose-rate was calculated from the radioactivity transmitted to the deposited eggs. γ-ray irradiation was carried out, in parallel, at approximately similar dose rates. RBE of THO for mutation fell between 1 and 2. The experiments with transforming principle of B. subtilis showed that the treatment of the bacterial spore with 3 H-glycelin or 3 H-water derived more inactive transformation with the decrease of the concentrations. (Nakanishi, T.)

  6. Quality Control Samples for the Radiological Determination of Tritium in Urine Samples

    International Nuclear Information System (INIS)

    Ost'pezuk, P.; Froning, M.; Laumen, S.; Richert, I.; Hill, P.

    2004-01-01

    The radioactive decay product of tritium is a low energy beta that cannot penetrate the outer dead layer of human skin. Therefore , the main hazard associated with tritium is internal exposure. In addition, due to the relatively long half life and short biological half life, tritium must be ingested in large amounts to pose a significant health risk. On the other hand, the internal exposure should be kept as low as practical. For incorporation monitoring of professional radiation workers the quality control is of utmost importance. In the Research Centre Juelich GmbH (FZJ) a considerable fraction of monitoring by excretion analysis relates to the isotope Tritium. Usually an aliquot of an urine sample is mixed with a liquid scintillator and measured in a liquid scintillation counter. Quality control samples in the form of three kind of internal reference samples (blank, reference samples with low activity and reference sample with elevated activity) were prepared from a mixed, Tritium (free) urine samples. 1 ml of these samples were pipetted into a liquid scintillation vial. In the part of theses vials a known amounts of Tritium were added. All these samples were stored at 20 degrees. Based on long term use of all these reference samples it was possible to construct appropriate control charts with the upper and lower alarm limits. Daily use of these reference samples decrease significantly the risk for false results in original urine with no significant increase of the determination time. (Author) 2 refs

  7. Tritium processing in JT-60U

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Masaki, Kei

    1997-01-01

    Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7 x 10 19 D-D fusion neutrons were produced by D (d, p) T reactions in plasma, which are expected to produce ∼31 GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium Distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ∼ 7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm 3 for radiation work permit requirements. (author)

  8. Hot muonic deuterium and tritium from cold targets

    International Nuclear Information System (INIS)

    Marshall, G.M.; Beveridge, J.L.; Bailey, J.M.; Beer, G.A.; Knowles, P.E.; Mason, G.R.; Olin, A.; Brewer, J.H.; Forster, B.M.; Huber, T.M.; Pippitt, B.; Jacot-Guillarmod, R.; Schellenberg, L.; Martoff, C.J.; Petitjean, C.

    1992-01-01

    Experiments are described which use a solid hydrogen layer to form muonic hydrogen isotopes in vacuum. The method relies on transfer of the muon from protium to either a deuteron or a triton. The resulting muonic deuterium or muonic tritium will not immediately thermalize because of the very low elastic cross sections, and may be emitted from the surface of the layer. Measurements which detect decay electrons, muonic x-rays, and fusion products have been used to study the processes. A target has been constructed which exploits muonic atom emission in order to study the energy dependence of transfer and muon molecular formation

  9. Tritium monitoring in environment at ICIT Tritium Separation Facility

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  10. Automation system for tritium contaminated surface monitoring

    International Nuclear Information System (INIS)

    Culcer, Mihai; Iliescu, Mariana; Curuia, Marian; Raceanu, Mircea; Enache, Adrian; Stefanescu, Ioan; Ducu, Catalin; Malinovschi, Viorel

    2005-01-01

    The low energy of betas makes tritium difficult to detect. However, there are several methods used in tritium detection, such as liquid scintillation and ionization chambers. Tritium on or near a surface can be also detected using proportional counter and, recently, solid state devices. The paper presents our results in the design and achievement of a surface tritium monitor using a PIN photodiode as a solid state charged particle detector to count betas emitted from the surface. That method allows continuous, real-time and non-destructively measuring of tritium. (authors)

  11. Tritium compatibility of alumina and Fosterite

    Energy Technology Data Exchange (ETDEWEB)

    Coffin, D.O.

    1979-09-01

    Many pressure measurements are required to control processing of the fuel gases associated with fusion power reactors. Since most pressure transducers respond to changes in pressure sensitive electrical parameters, insulators will be required to withstand chronic exposures to concentrated tritium. For this investigation samples of alumina and Fosterite were exposed to concentrated tritium gas for 11 weeks. Gas phase impurities were then analyzed for clues that would indicate decomposition of the exposed materials. The only gaseous impurity resulting from these tritium exposures was tritio-methane, which is always produced when tritium is stored in stainless steel containers. There was no evidence that either alumina or Fosterite decomposed in the presence of tritium.

  12. Management of tritium at nuclear facilities

    International Nuclear Information System (INIS)

    1984-01-01

    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

  13. Standard Guide for Simulation of Helium Effects in Irradiated Metals

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1996-01-01

    1.1 This guide provides advice for conducting experiments to investigate the effects of helium on the properties of metals where the technique for introducing the helium differs in some way from the actual mechanism of introduction of helium in service. Simulation techniques considered for introducing helium shall include charged particle implantation, exposure to α-emitting radioisotopes, and tritium decay techniques. Procedures for the analysis of helium content and helium distribution within the specimen are also recommended. 1.2 Two other methods for introducing helium into irradiated materials are not covered in this guide. They are the enhancement of helium production in nickel-bearing alloys by spectral tailoring in mixed-spectrum fission reactors, and isotopic tailoring in both fast and mixed-spectrum fission reactors. These techniques are described in Refs (1-5). Dual ion beam techniques (6) for simultaneously implanting helium and generating displacement damage are also not included here. This lat...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-01

    This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 6}Zr{sub 2}O{sub 7}, Li{sub 8}ZrO{sub 6}, Li{sub 2}O and Li{sub 2}SiO{sub 3} have been irradiated at different temperature levels and up to a maximum lithium burnup of about 3%. Tritium residence times of the various breeding materials have been determined from temperature transients performed during irradiation. After irradiation the tritium inventory has been determined from small samples of the various materials. From the out-of-pile tritium release experiments activation energies were determined. These activities have been performed at ECN within the framework of the European Fusion Technology Programme on Breeding Blankets. (orig.).

  15. Tritium in [18O]water containing [18F]fluoride for [18F]FDG synthesis

    International Nuclear Information System (INIS)

    Ito, Shigeki; Saze, Takuya; Sakane, Hitoshi; Ito, Satoshi; Ito, Shinichi; Nishizawa, Kunihide

    2004-01-01

    The presence of tritium in enriched [ 18 O]water irradiated with 9.6 MeV protons used to produce [ 18 F]fluoride by the 18 O(p, n) 18 F reaction was inferred from the cross sections and threshold energies of the 18 O(p, t) 16 O reaction, and the existence of tritium was confirmed experimentally. Tritium was also detected in both [ 18 O]water recovered for recycling and waste acetonitrile solutions. The purified [ 18 F]FDG was not contaminated with 3 H. The amount of 3 H discharged into the air was far less than the International Basic Safety Standard Level

  16. Mechanical properties of irradiated beryllium

    International Nuclear Information System (INIS)

    Beeston, J.M.; Longhurst, G.R.; Wallace, R.S.

    1992-01-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10 25 n/m 2 (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.)

  17. Mechanical properties of irradiated beryllium

    Science.gov (United States)

    Beeston, J. M.; Longhurst, G. R.; Wallace, R. S.; Abeln, S. P.

    1992-10-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 × 10 25 n/m 2 ( E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.

  18. 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.

  19. Generation of gaseous tritium standards

    International Nuclear Information System (INIS)

    Hohorst, F.A.

    1994-09-01

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

  20. Tritium-labelled abscisic acid

    International Nuclear Information System (INIS)

    Pluciennik, H.; Michalski, L.

    1991-01-01

    A simple method for the preparation of biologically active abscisic acid (growth inhibiting plant hormone) labelled with tritium is described. The product obtained has a specific radioactivity of 1.12 GBq mmol -1 : the yield is about 60% as compared to the initial amount of the substance used. (author) 7 refs.; 2 figs

  1. Tritium Level in Romanian Precipitation

    Energy Technology Data Exchange (ETDEWEB)

    Varlam, C.; Stefanescu, I.; Faurescu, I.; Bogdan, D.; Soare, A. [Institute for Cryogenic and Isotope Technologies, Rm. Valcea (Romania); Duliu, O. G. [Faculty of Physics, University of Bucharest, Magurele (Romania)

    2013-07-15

    Romania is one of the countries that has no station included in GNIP (Global Network of Isotopes in Precipitation) on its territory. This paper presents results regarding the tritium concentration in precipitation for the period 1999-2009. The precipitation fell at the Institute for cryogenic and Isotope technologies (geographical coordinates: altitude 237 m, latitude 45{sup o}02'07' N, longitude 24{sup o}17'03' E) an was collected both individually and as a composite average of each month. It was individually measured and the average was calculated and compared with the tritium concentration measured in the composite sample. tritium concentration levels ranged from 9.9 {+-} 2.1 TU for 2004 and 13.7 {+-} 2.2 TU for 2009. Comparing the arithmetic mean values with the weighted mean for the period of observation, it was noticed that the higher absolute values of the weighted means were constant. It was found that for the calculated monthly average for the period of observation (1999-2009), the months with the maximum tritium concentration are the same as the months with the maximum amount of precipitation. This behaviour is typical for the monitored location. (author)

  2. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  3. Tritium behavior intentionally released in the room

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  4. Tritium migration in nuclear desalination plants

    International Nuclear Information System (INIS)

    Muralev, E.D.

    2003-01-01

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

  5. Tritium recovery from co-deposited layers using 193-nm laser

    Science.gov (United States)

    Shu, W. M.; Kawakubo, Y.; Nishi, M. F.

    Recovery of tritium from co-deposited layers formed in deuterium-tritium plasma operations of the TFTR (Tokamak Fusion Test Reactor) was investigated by the use of an ArF excimer laser operating at the wavelength of 193 nm. At the laser energy density of 0.1 J/cm2, a transient spike of the tritium-release rate was observed at initial irradiation. Hydrogen isotopes were released in the form of hydrogen-isotope molecules during the laser irradiation in vacuum, suggesting that tritium can be recovered readily from the released gases. In a second experiment, hydrogen (tritium) recovery from the co-deposited layers on JT-60 tiles that had experienced hydrogen-plasma operations was investigated by laser ablation with a focused beam of the excimer laser. The removal rate of the co-deposited layers was quite low when the laser energy density was smaller than the ablation threshold (1.0 J/cm2), but reached 1.1 μm/pulse at the laser energy density of 7.6 J/cm2. The effective absorption coefficient in the co-deposited layers at the laser wavelength was determined to be 1.9 μm-1. The temperature of the surface during the irradiation at the laser energy density of 0.5 J/cm2 was measured on the basis of Planck's law of radiation, and the maximum temperature during the irradiation decreased from 3570 K at the initial irradiation to 2550 K at the 1000th pulse of the irradiation.

  6. Storage and Assay of Tritium in STAR

    International Nuclear Information System (INIS)

    Longhurst, Glen R.; Anderl, Robert A.; Pawelko, Robert J.; Stoots, Carl J.

    2005-01-01

    The Safety and Tritium Applied Research (STAR) facility at the Idaho National Engineering and Environmental Laboratory (INEEL) is currently being commissioned to investigate tritium-related safety questions for fusion and other technologies. The tritium inventory for the STAR facility will be maintained below 1.5 g to avoid the need for STAR to be classified as a Category 3 nuclear facility. A key capability in successful operation of the STAR facility is the ability to receive, inventory, and dispense tritium to the various experiments underway there. The system central to that function is the Tritium Storage and Assay System (SAS).The SAS has four major functions: (1) receiving and holding tritium, (2) assaying, (3) dispensing, and (4) purifying hydrogen isotopes from non-hydrogen species.This paper describes the design and operation of the STAR SAS and the procedures used for tritium accountancy in the STAR facility

  7. Tritium pellet injection sequences for TFTR

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Milora, S.L.; Attenberger, S.E.; Singer, C.E.; Schmidt, G.L.

    1983-01-01

    Tritium pellet injection into neutral deuterium, beam heated deuterium plasmas in the Tokamak Fusion Test Reactor (TFTR) is shown to be an attractive means of (1) minimizing tritium use per tritium discharge and over a sequence of tritium discharges; (2) greatly reducing the tritium load in the walls, limiters, getters, and cryopanels; (3) maintaining or improving instantaneous neutron production (Q); (4) reducing or eliminating deuterium-tritium (D-T) neutron production in non-optimized discharges; and (5) generally adding flexibility to the experimental sequences leading to optimal Q operation. Transport analyses of both compression and full-bore TFTR plasmas are used to support the above observations and to provide the basis for a proposed eight-pellet gas gun injector for the 1986 tritium experiments

  8. Imaging of tritium implanted into graphite

    International Nuclear Information System (INIS)

    Malinowski, M.E.; Causey, R.A.

    1988-01-01

    The extensive use of graphite in plasma-facing surfaces of tokamaks such as the Tokamak Fusion Test Reactor, which has planned tritium discharges, makes two-dimensional tritium detection techniques important in helping to determine torus tritium inventories. We have performed experiments in which highly oriented pyrolytic graphite (HOPG) samples were first tritium implanted with fluences of ∼10 16 T/cm 2 at energies approx. 0 C resulted in no discernible motion of tritium along the basal plane, but did show that significant desorption of the implanted tritium occurred. The current results indicate that tritium in quantities of 10 12 T/cm 2 in tritiated components could be readily detected by imaging at lower magnifications

  9. Calculation of tritium release from reactor's stack

    International Nuclear Information System (INIS)

    Akhadi, M.

    1996-01-01

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

  10. Simultaneous measurement of tritium and radiocarbon by ultra-low-background proportional counting.

    Science.gov (United States)

    Mace, Emily; Aalseth, Craig; Alexander, Tom; Back, Henning; Day, Anthony; Hoppe, Eric; Keillor, Martin; Moran, Jim; Overman, Cory; Panisko, Mark; Seifert, Allen

    2017-08-01

    Use of ultra-low-background capabilities at Pacific Northwest National Laboratory provide enhanced sensitivity for measurement of low-activity sources of tritium and radiocarbon using proportional counters. Tritium levels are nearly back to pre-nuclear test backgrounds (~2-8 TU in rainwater), which can complicate their dual measurement with radiocarbon due to overlap in the beta decay spectra. We present results of single-isotope proportional counter measurements used to analyze a dual-isotope methane sample synthesized from ~120mg of H 2 O and present sensitivity results. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Application of NaI(Tl) scintillation detector to measurement of tritium concentration

    International Nuclear Information System (INIS)

    Nishizawa, Kazushige; Endo, Yoshio; Shinagawa, Mutsuaki

    1977-01-01

    Measurements of tritium concentration in various media have been successfully performed by detecting with an ordinary NaI(Tl) scintillation counter the bremsstrahlung emitted from tritium decay. To distinguish meaningful signals from noises emanating mainly from thermoelectrons present in the phototube, differences in rise time were discriminated by means of an electronic circuit, instead of separation by pulse height analysis. This pulse-shape discriminator successfully reduced noise counts below 18 keV in a multichannel analyzer from 100 cpm to 4 cpm, which permitted direct counting of tritiated water as dilute as 1 μCi/ml, without requiring any complicated preparation. (auth.)

  12. A basic study on the ITER tritium storage vessel design and components

    International Nuclear Information System (INIS)

    Chung, H. S.; Ahn, D. H.; Kim, K. R.; Yim, S. P.; Paek, S. W.; Lee, M. S.; Lee, S. H.; Shim, M. H.

    2006-01-01

    The ZrCo getter beds are built of a primary vessel which contains the ZrCo powder mixed with Cu spheres of less than one mm diameter and of a secondary outer vessel. The purpose of the secondary outer vessel is to capture permeated or leaked tritium and to present a good thermal insulation when properly evacuated. A third volume, a helium filled loop, is installed in the primary volume to remove the decay heat and is used to perform tritium accountancy measurements

  13. Tritium systems test assembly stabilization

    International Nuclear Information System (INIS)

    Jasen, William G.; Michelotti, Roy A.; Anast, Kurt R.; Tesch, Charles

    2004-01-01

    The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium technology Research and Development (R and D) primarily for future fusion power reactors. The facility was conceived in mid 1970's, operations commenced in early 1980's, stabilization and deactivation began in 2000 and were completed in 2003. The facility will remain in a Surveillance and Maintenance (S and M) mode until the Department of Energy (DOE) funds demolition of the facility, tentatively in 2009. A safe and stable end state was achieved by the TSTA Facility Stabilization Project (TFSP) in anticipation of long term S and M. At the start of the stabilization project, with an inventory of approximately 140 grams of tritium, the facility was designated a Hazard Category (HC) 2 Non-Reactor Nuclear facility as defined by US Department of Energy standard DOE-STD-1027-92 (1997). The TSTA facility comprises a laboratory area, supporting rooms, offices and associated laboratory space that included more than 20 major tritium handling systems. The project's focus was to reduce the tritium inventory by removing bulk tritium, tritiated water wastes, and tritium-contaminated high-inventory components. Any equipment that remained in the facility was stabilized in place. All of the gloveboxes and piping were rendered inoperative and vented to atmosphere. All equipment, and inventoried tritium contamination, remaining in the facility was left in a safe-and-stable state. The project used the End Points process as defined by the DOE Office of Environmental Management (web page http://www.em.doe.- gov/deact/epman.htmtlo) document and define the end state required for the stabilization of TSTA Facility. The End Points process added structure that was beneficial through virtually all phases of the project. At completion of the facility stabilization project the residual tritium inventory was approximately 3,000 curies, considerably less than the 1.6-gram threshold for a HC 3 facility. TSTA is now

  14. Influence of impurities in Beryllium on tritium breeding ratio

    International Nuclear Information System (INIS)

    Yamauchi, M.; Ochiai, K.; Verzilov, Y.; Ito, M.; Wada, M.; Nishitani, T.

    2004-01-01

    Several neutronics experiments simulating fusion blankets have been conducted with 14 MeV neutron source to assess the reliability of nuclear analysis codes. However, the analyses have not always presented good agreements so far between calculated and measured tritium production rates. One of the reasons was considered as impurities in beryllium which has negligibly small neutron absorption cross section in low energy range. Chemical compositions of beryllium were analyzed by Inductively Coupled Plasma (ICP) method, and a pulsed neutron decay experiment discovered that the macroscopic neutron absorption cross section for beryllium medium may be about 30% larger than the value calculated by the data specified by manufacturing company. The influence of the impurities on the calculations was studied on the basis of the fusion DEMO-reactor blanket design. As a result of the study, it was made clear that the impurities affect the local tritium production rates when the size of beryllium medium is more than 20-30 mean free paths (30-40 cm) in thickness. In case of some blanket designs that meet the above condition, the effect on tritium breeding ratio may become as large as about 4%. (author)

  15. Influence of impurities in Beryllium on tritium breeding ratio

    Energy Technology Data Exchange (ETDEWEB)

    Yamauchi, M; Ochiai, K; Verzilov, Y; Ito, M; Wada, M; Nishitani, T [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2004-03-01

    Several neutronics experiments simulating fusion blankets have been conducted with 14 MeV neutron source to assess the reliability of nuclear analysis codes. However, the analyses have not always presented good agreements so far between calculated and measured tritium production rates. One of the reasons was considered as impurities in beryllium which has negligibly small neutron absorption cross section in low energy range. Chemical compositions of beryllium were analyzed by Inductively Coupled Plasma (ICP) method, and a pulsed neutron decay experiment discovered that the macroscopic neutron absorption cross section for beryllium medium may be about 30% larger than the value calculated by the data specified by manufacturing company. The influence of the impurities on the calculations was studied on the basis of the fusion DEMO-reactor blanket design. As a result of the study, it was made clear that the impurities affect the local tritium production rates when the size of beryllium medium is more than 20-30 mean free paths (30-40 cm) in thickness. In case of some blanket designs that meet the above condition, the effect on tritium breeding ratio may become as large as about 4%. (author)

  16. Tritium issues in plasma wall interactions

    International Nuclear Information System (INIS)

    Tanabe, T.

    2009-01-01

    In order to establish a D-T fusion reactor as an energy source, it is not enough to have a DT burning plasma, and economical conversion of fusion energy to electricity and/or heat, a large enough margin of tritium breeding and tritium safety must be simultaneously achieved. In particular, handling of huge amount of tritium needs a significant effort to ensure that the radiation dose of radiological workers and of the public is below the limits specified by the International Commission on Radiological Protection. For the safety reasons, tritium in a reactor will be limited to only a few kg orders in weight, with radioactivity up to 10 17 Bq. Since public exposure to tritium is regulated at a level as tiny as a few Bq/cm 2 , tritium must be strictly confined in a reactor system with accountancy of an order of pg (pico-gram). Generally qualitative analysis with the accuracy of more than 3 orders of magnitude is hardly possible. We are facing to lots of safety concerns in the handling of huge amounts of radioactive tritium as a fuel and to be bred in a blanket. In addition, tritium resources are very limited. Not only for the safety reason but also for the saving of tritium resources, tritium retention in a reactor must be kept as small as possible. In the present tokamaks, however, hydrogen retention is significantly large, i.e. more than 20% of fueled hydrogen is continuously piled up in the vacuum vessel, which must not be allowed in a reactor. After the introduction of tritium as a hydrogen radioisotope, this lecture will 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. This document is made of the slides of the presentation. (author)

  17. Neutron irradiation effects on plasma facing materials

    Science.gov (United States)

    Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

    2000-12-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

  18. Neutron irradiation effects on plasma facing materials

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

    2000-01-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

  19. Analysis on tritium permeation in tritium storage bed with gas flowing calorimetry

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hirofumi; Hayashi, Takumi; Suzuki, Takumi; Nishi, Masataka [Japan Atomic Energy Research Inst., Naka Fusion Research Establishment, Department of Fusion Engineering Research, Naka, Ibaraki (Japan); Yoshida, Hiroshi [Japan Atomic Energy Research Inst., Naka Fusion Research Establishment, ITER-Joint Centeral Team, Naka, Ibaraki (Japan)

    2000-10-01

    Tritium permeation amount in a tritium storage bed with gas flowing calorimetric was evaluated under a condition of new operation mode for International Thermonuclear Experimental Reactor (ITER). As a result, tritium permeation under the new operation mode was estimated to be about twice of that under the practical operation mode. This result show that it would be regardless in a view point of material control of tritium, however, it was suggested to be required additional tritium removal or evacuate system in a view points of safety control or performance of accountability or thermal insulating of the tritium storage bed. (author)

  20. Tritium transport studies with use of the ISEP NPA during tritium trace experimental campaign on JET

    International Nuclear Information System (INIS)

    Mironov, M I; Afanasyev, V I; Murari, A; Santala, M; Beaumont, P

    2010-01-01

    The neutral particle analyzer (NPA) known as ISEP (Ion SEParator) was applied to measure the tritium neutral flux during the tritium trace experiment (TTE) on JET. The energy dependence (in the 5-28 keV energy range) of the tritium neutral flux rise time after a short ∼100 ms tritium gas puff into deuterium plasmas has been observed for the first time. The dependence has been interpreted as being due to the penetration of the tritium ions from the plasma boundary into the core and has been used for the calculation of the tritium diffusion coefficient and convective velocity values.

  1. Tritium confinement in a new tritium processing facility at the Savannah River Site

    International Nuclear Information System (INIS)

    Heung, L.K.; Owen, J.H.; Hsu, R.H.; Hashinger, R.F.; Ward, D.E.; Bandola, P.E.

    1991-01-01

    A new tritium processing facility, named the Replacement Tritium Facility (RTF), has been completed and is being prepared for startup at the Savannah River Site (SRS). The RTF has the capability to recover, purify and separate hydrogen isotopes from recycled gas containers. A multilayered confinement system is designed to reduce tritium losses to the environment. This confinement system is expected to confine and recover any tritium that might escape the process equipment, and to maintain the tritium concentration in the nitrogen glovebox atmosphere to less than 10 -2 μCi/cc tritium

  2. Drinking water standard for tritium-what's the risk?

    Science.gov (United States)

    Kocher, D C; Hoffman, F O

    2011-09-01

    This paper presents an assessment of lifetime risks of cancer incidence associated with the drinking water standard for tritium established by the U.S. Environmental Protection Agency (USEPA); this standard is an annual-average maximum contaminant level (MCL) of 740 Bq L(-1). This risk assessment has several defining characteristics: (1) an accounting of uncertainty in all parameters that relate a given concentration of tritium in drinking water to lifetime risk (except the number of days of consumption of drinking water in a year and the number of years of consumption) and an accounting of correlations of uncertain parameters to obtain probability distributions that represent uncertainty in estimated lifetime risks of cancer incidence; (2) inclusion of a radiation effectiveness factor (REF) to represent an increased biological effectiveness of low-energy electrons emitted in decay of tritium compared with high-energy photons; (3) use of recent estimates of risks of cancer incidence from exposure to high-energy photons, including the dependence of risks on an individual's gender and age, in the BEIR VII report; and (4) inclusion of risks of incidence of skin cancer, principally basal cell carcinoma. By assuming ingestion of tritium in drinking water at the MCL over an average life expectancy of 80 y in females and 75 y in males, 95% credibility intervals of lifetime risks of cancer incidence obtained in this assessment are (0.35, 12) × 10(-4) in females and (0.30, 15) × 10(-4) in males. Mean risks, which are considered to provide the best single measure of expected risks, are about 3 × 10(-4) in both genders. In comparison, USEPA's point estimate of the lifetime risk of cancer incidence, assuming a daily consumption of drinking water of 2 L over an average life expectancy of 75.2 y and excluding an REF for tritium and incidence of skin cancer, is 5.6 × 10(-5). Probability distributions of annual equivalent doses to the whole body associated with the drinking

  3. Reactions of recoil tritium generated by the 3He(n,p)3H reaction with aromatic compounds -intramolecular tritium distribution and radiochemical yield

    International Nuclear Information System (INIS)

    Nogawa, N.; Morikawa, Naotake; Oohashi, Kunio; Matuoka, H.; Moki, T.; Moriya, T.

    1986-01-01

    Reactions of recoil tritium with benzoic acid, acetanilide and β-phenethyl alcohol were studied using the 3 He(=n,p) 3 H reaction. The tritium distribution in the aromatic ring is approximately uniform in all the irradiated compounds and the tritium activity per C-H bond in the methyl and ethylene groups is 7 to 8 relative to that of the corresponding ring as standard. These findings are substantially the same as those obtained previously by the 6 Li(n,α) 3 H reactions, suggesting the same mechanism of tritiation for both recoil reactions. The tritiated parent compounds were obtained in high radiochemical yields: 45% for benzoic acid, 30% for acetanilide, 12% for β-phenethyl alcohol. (author)

  4. Tau decays

    International Nuclear Information System (INIS)

    Golutvin, A.

    1994-09-01

    The most recent experimental results of τ physics are reviewed. The covered topics include precision measurements of semihadronic τ decay and their impact on tau branching ratio budget, the current status of the tau consistency test, a determination of Michel parameters and τ neutrino helicity, and upper limits on lepton-number violating τ decays. (orig.)

  5. 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.

  6. Decay tank

    International Nuclear Information System (INIS)

    Matsumura, Seiichi; Tagishi, Akinori; Sakata, Yuji; Kontani, Koji; Sudo, Yukio; Kaminaga, Masanori; Kameyama, Iwao; Ando, Koei; Ishiki, Masahiko.

    1990-01-01

    The present invention concerns an decay tank for decaying a radioactivity concentration of a fluid containing radioactive material. The inside of an decay tank body is partitioned by partitioning plates to form a flow channel. A porous plate is attached at the portion above the end of the partitioning plate, that is, a portion where the flow is just turned. A part of the porous plate has a slit-like opening on the side close to the partitioning plate, that is, the inner side of the flow at the turning portion thereof. Accordingly, the primary coolants passed through the pool type nuclear reactor and flown into the decay tank are flow caused to uniformly over the entire part of the tank without causing swirling. Since a distribution in a staying time is thus decreased, the effect of decaying 16 N as radioactive nuclides in the primary coolants is increased even in a limited volume of the tank. (I.N.)

  7. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

    Ono, F.; Yamawaki, M.

    1995-01-01

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

  8. The LLNL portable tritium processing system

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The end of the Cold War significantly reduced the need for facilities to handle radioactive materials for the US nuclear weapons program. The LLNL Tritium Facility was among those slated for decommissioning. The plans for the facility have since been reversed, and it remains open. Nevertheless, in the early 1990s, the cleanup (the Tritium Inventory Removal Project) was undertaken. However, removing the inventory of tritium within the facility and cleaning up any pockets of high-level residual contamination required that we design a system adequate to the task and meeting today's stringent standards of worker and environmental protection. In collaboration with Sandia National Laboratory and EG ampersand G Mound Applied Technologies, we fabricated a three-module Portable Tritium Processing System (PTPS) that meets current glovebox standards, is operated from a portable console, and is movable from laboratory to laboratory for performing the basic tritium processing operations: pumping and gas transfer, gas analysis, and gas-phase tritium scrubbing. The Tritium Inventory Removal Project is now in its final year, and the portable system continues to be the workhorse. To meet a strong demand for tritium services, the LLNL Tritium Facility will be reconfigured to provide state-of-the-art tritium and radioactive decontamination research and development. The PTPS will play a key role in this new facility

  9. The Chalk River Tritium Extraction Plant

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Harrison, T.E.; Spagnolo, D.A.

    1990-01-01

    The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1-35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T 2 . The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: it produces pure tritium for use in fusion research and development, and it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors (author)

  10. The Chalk River Tritium Extraction Plant

    Energy Technology Data Exchange (ETDEWEB)

    Holtslander, W J; Harrison, T E; Spagnolo, D A

    1990-07-01

    The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1-35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T{sub 2}. The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: it produces pure tritium for use in fusion research and development, and it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors (author)

  11. Estimation of Biological Effects of Tritium.

    Science.gov (United States)

    Umata, Toshiyuki

    2017-01-01

    Nuclear fusion technology is expected to create new energy in the future. However, nuclear fusion requires a large amount of tritium as a fuel, leading to concern about the exposure of radiation workers to tritium beta radiation. Furthermore, countermeasures for tritium-polluted water produced in decommissioning of the reactor at Fukushima Daiichi Nuclear Power Station may potentially cause health problems in radiation workers. Although, internal exposure to tritium at a low dose/low dose rate can be assumed, biological effect of tritium exposure is not negligible, because tritiated water (HTO) intake to the body via the mouth/inhalation/skin would lead to homogeneous distribution throughout the whole body. Furthermore, organically-bound tritium (OBT) stays in the body as parts of the molecules that comprise living organisms resulting in long-term exposure, and the chemical form of tritium should be considered. To evaluate the biological effect of tritium, the effect should be compared with that of other radiation types. Many studies have examined the relative biological effectiveness (RBE) of tritium. Hence, we report the RBE, which was obtained with radiation carcinogenesis classified as a stochastic effect, and serves as a reference for cancer risk. We also introduce the outline of the tritium experiment and the principle of a recently developed animal experimental system using transgenic mouse to detect the biological influence of radiation exposure at a low dose/low dose rate.

  12. Tritium inventories and tritium safety design principles for the fuel cycle of ITER

    International Nuclear Information System (INIS)

    Cristescu, I.R.; Cristescu, I.; Doerr, L.; Glugla, M.; Murdoch, D.

    2007-01-01

    Within the tritium plant of ITER a total inventory of about 2-3 kg will be necessary to operate the machine in the DT phase. During plasma operation, tritium will be distributed in the different sub-systems of the fuel cycle. A tool for tritium inventory evaluation within each sub-system of the fuel cycle is important with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems; however, tritium accounting may be achieved by modelling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the sub-systems. To get reliable results, an accurate dynamic modelling of the tritium content in each sub-system is necessary. A dynamic model (TRIMO) for tritium inventory calculation reflecting the design of each fuel cycle sub-systems was developed. The amount of tritium needed for ITER operation has a direct impact on the tritium inventories within the fuel cycle sub-systems. As ITER will function in pulses, the main characteristics that influence the rapid tritium recovery from the fuel cycle as necessary for refuelling are discussed. The confinement of tritium within the respective sub-systems of the fuel cycle is one of the most important safety objectives. The design of the deuterium/tritium fuel cycle of ITER includes a multiple barrier concept for the confinement of tritium. The buildings are equipped with a vent detritiation system and re-circulation type room atmosphere detritiation systems, required for tritium confinement barrier during possible tritium spillage events. Complementarily to the atmosphere detritiation systems, in ITER a water detritiation system for tritium recovery from various sources will also be operated

  13. The tritium content of precipitation and groundwater at Yola, Nigeria ...

    African Journals Online (AJOL)

    Tritium is a radioactive isotope of hydrogen which occurs in precipitation. In groundwater studies tritium measurements give information on the time of recharge to the system; the tritium content of precipitation being used to estimate the input of tritium to the groundwater system. At Yola, the tritium ontents in precipitation and ...

  14. Tritium in Exit Signs | RadTown USA | US EPA

    Science.gov (United States)

    2017-08-07

    Many exit signs contain tritium to light the sign without batteries or electricity. Using tritium in exit signs allows the sign to remain lit if the power goes out. Tritium is most dangerous when it is inhaled or swallowed. Never tamper with a tritium exit sign. If a tritium exit sign is broken, leave the area immediately and notify the building maintenance staff.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  16. Nuclear irradiation parameters of beryllium under fusion, fission and IFMIF irradiation conditions

    International Nuclear Information System (INIS)

    Fischer, U.; Chen, Y.; Leichtle, D.; Simakov, S.; Moeslang, A.; Vladimirov, P.

    2004-01-01

    A computational analysis is presented of the nuclear irradiation parameters for Beryllium under irradiation in typical neutron environments of fission and fusion reactors, and of the presently designed intense fusion neutron source IFMIF. The analysis shows that dpa and Tritium production rates at fusion relevant levels can be achieved with existing high flux fission reactors while the achievable Helium production is too low. The resulting He-Tritium and He/dpa ratios do not meet typical fusion irradiation conditions. Irradiation simulations in the medium flux test modules of the IFMIF neutron source facility were shown to be more suitable to match fusion typical irradiation conditions. To achieve sufficiently high production rates it is suggested to remove the creep-fatigue testing machine together with the W spectra shifter plate and move the tritium release module upstream towards the high flux test module. (author)

  17. 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

  18. Dynamic investigation of electron trapping and charge decay in electron-irradiated Al sub 2 O sub 3 in a scanning electron microscope: Methodology and mechanisms

    CERN Document Server

    Fakhfakh, S; Belhaj, M; Fakhfakh, Z; Kallel, A; Rau, E I

    2002-01-01

    The charging and discharging of polycrystalline Al sub 2 O sub 3 submitted to electron-irradiation in a scanning electron microscope (SEM) are investigated by means of the displacement current method. To circumvent experimental shortcomings inherent to the use of the basic sample holder, a redesign of the latter is proposed and tests are carried out to verify its operation. The effects of the primary beam accelerating voltage on charging, flashover and discharging phenomena during and after electron-irradiation are studied. The experimental results are then analyzed. In particular, the divergence between the experimental data and those predicted by the total electron emission yield approach (TEEYA) is discussed. A partial discharge was observed immediately after the end of the electron-irradiation exposure. The experimental data suggests, that the discharge is due to the evacuation to the ground, along the insulator surface, of released electrons from shallow traps at (or in the close vicinity of) the insulat...

  19. Behaviour of tritium in the environment

    International Nuclear Information System (INIS)

    1979-01-01

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

  20. An experiment to measure the electron neutrino mass using a cryogenic tritium source

    International Nuclear Information System (INIS)

    Fackler, O.; Jeziorski, B.; Kolos, W.; Monkhorst, H.; Mugge, M.; Sticker, H.; Szalewicz, K.; White, R.M.; Woerner, R.

    1985-01-01

    An experiment has been performed to determine the electron neutrino mass with the precision of a few eV by measuring the tritium beta decay energy distribution near the endpoint. Key features of the experiment are a 2 eV resolution electrostatic spectrometer and a high-activity frozen tritium source. It is important that the source have electronic wavefunctions which can be accurately calculated. These calculations have been made for tritium and the HeT + daughter ion and allow determination of branching fractions to 0.1% and energy of the excited states to 0.1 eV. The excited final molecular state calculations and the experimental apparatus are discussed. 4 refs., 5 figs

  1. Stability of tritium permeation prevention barrier with TiC and TiN + TiC coating

    International Nuclear Information System (INIS)

    Shan Changqi; Chen Qingwang; Dai Shaoxia; Jiang Weisheng

    1999-01-01

    The stability of tritium permeation prevention barrier of 316L stainless steel with coating TiC and TiN + TiC under the conditions of very large thermal gradient, thermal cycling and plasma irradiation is researched. The research includes two aspects: one is the study on the stability resisting H + plasma irradiation; another is on the ability of two coating materials when they are used in long term under the condition of very large thermal gradient and cycling. The results show that TiC and TiN + TiC composite coating materials, after chemical heat treatment and forming tritium permeation prevention barrier, can resist H + ion irradiation, and also can resist very large thermal gradient and thermal cycling. The long time experiments show that tritium permeation prevention barrier of those coating materials is stable when they are used in long term

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

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

    International Nuclear Information System (INIS)

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

    1986-10-01

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

  4. Tritium labelled steroids, preparation process and application to synthesis of tritium labelled estrane derivatives

    International Nuclear Information System (INIS)

    1978-01-01

    Process for preparing new steroids labelled with tritium in 6.7 and comprising in 3 a blocked ketonic group as ketal, thioketal or derivatives. Application of these products to the synthesis of tritium labelled estrane derivatives [fr

  5. Radiological training for tritium facilities

    International Nuclear Information System (INIS)

    1996-12-01

    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

  6. Radiological training for tritium facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    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.

  7. Tritium environmental transport studies at TFTR

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  8. Tritium environmental transport studies at TFTR

    Science.gov (United States)

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

    1993-06-01

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

  9. Purification of tritium-free water

    International Nuclear Information System (INIS)

    Hussain, S.D.

    1982-10-01

    Ground water which has been out of contact with the atmosphere for a long time as compared to the half life of tritium (12.43 years) does not contain any measureable amount of tritium. Such water is called tritium-free water. It may contain dissolved and suspended impurities and has to be purified before it can be used for the preparation of blanks and standards required in the routine measurement of low level tritium in water samples. The purification of tritium-free water by distillation in a closed system has been described. The quality of processed tritium-free water was precisely checked at International Atomic Energy Agency (IAEA) Vienna and found satisfactory. (authors)

  10. Modeling tritium transport in the environment

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.

    1986-01-01

    A model of tritium transport in the environment near an atmospheric source of tritium is presented in the general context of modeling material cycling in ecosystems. The model was developed to test hypotheses about the process involved in tritium cycling. The temporal and spatial scales of the model were picked to allow comparison to environmental monitoring data collected in the vicinity of the Savannah River Plant. Initial simulations with the model showed good agreement with monitoring data, including atmospheric and vegetation tritium concentrations. The model can also simulate values of tritium in vegetation organic matter if the key parameter distributing the source of organic hydrogen is varied to fit the data. However, because of the lack of independent conformation of the distribution parameter, there is still uncertainty about the role of organic movement of tritium in the food chain, and its effect on the dose to man

  11. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  12. Comparison of Tritium Component Failure Rate Data

    International Nuclear Information System (INIS)

    Lee C. Cadwallader

    2004-01-01

    Published failure rate values from the US Tritium Systems Test Assembly, the Japanese Tritium Process Laboratory, the German Tritium Laboratory Karlsruhe, and the Joint European Torus Active Gas Handling System have been compared. This comparison is on a limited set of components, but there is a good variety of data sets in the comparison. The data compared reasonably well. The most reasonable failure rate values are recommended for use on next generation tritium handling system components, such as those in the tritium plant systems for the International Thermonuclear Experimental Reactor and the tritium fuel systems of inertial fusion facilities, such as the US National Ignition Facility. These data and the comparison results are also shared with the International Energy Agency cooperative task on fusion component failure rate data

  13. History of 232-F, tritium extraction processing

    International Nuclear Information System (INIS)

    Blackburn, G.W.

    1994-08-01

    In 1950 the Atomic Energy Commission authorized the Savannah River Project principally for the production of tritium and plutonium-239 for use in thermonuclear weapons. 232-F was built as an interim facility in 1953--1954, at a cost of $3.9M. Tritium extraction operations began in October, 1955, after the reactor and separations startups. In July, 1957 a larger tritium facility began operation in 232-H. In 1958 the capacity of 232-H was doubled. Also, in 1957 a new task was assigned to Savannah River, the loading of tritium into reservoirs that would be actual components of thermonuclear weapons. This report describes the history of 232-F, the process for tritium extraction, and the lessons learned over the years that were eventually incorporated into the new Replacement Tritium Facility

  14. Effects of interfering constituents on tritium smears

    International Nuclear Information System (INIS)

    Levi, G.D. Jr.; Cheeks, K.E.

    1993-01-01

    Tritium smears are performed by Health Protection Operations (HPO) to assess transferable contamination on work place surfaces, materials for movement outside Radiologically Controlled Areas (RCA), and product containers being shipped between facilities. Historically, gas proportional counters were used to detect transferable tritium contamination collected by smearing. Because tritium is a low-energy beta emitter, gas proportional counters do not provide the sensitivity or the counting efficiency to accurately measure the tritium activity on the smear. Liquid Scintillation Counters (LSC) provide greater counting efficiency for the low-energy beta particles along with greater reliability and reproducibility compared to gas flow proportional counters. The purpose of this technical evaluation was to determine the effects of interfering constituents such as filters, dirt and oil on the counting efficiency and tritium recoveries of tritium smears by LSC

  15. Movement of a tritium plume in shallow groundwater at a legacy low-level radioactive waste disposal site in eastern Australia.

    Science.gov (United States)

    Hughes, C E; Cendón, D I; Harrison, J J; Hankin, S I; Johansen, M P; Payne, T E; Vine, M; Collins, R N; Hoffmann, E L; Loosz, T

    2011-10-01

    Between 1960 and 1968 low-level radioactive waste was buried in a series of shallow trenches near the Lucas Heights facility, south of Sydney, Australia. Groundwater monitoring carried out since the mid 1970s indicates that with the exception of tritium, no radioactivity above typical background levels has been detected outside the immediate vicinity of the trenches. The maximum tritium level detected in ground water was 390 kBq/L and the median value was 5400 Bq/L, decay corrected to the time of disposal. Since 1968, a plume of tritiated water has migrated from the disposal trenches and extends at least 100 m from the source area. Tritium in rainfall is negligible, however leachate from an adjacent and fill represents a significant additional tritium source. Study data indicate variation in concentration levels and plume distribution in response to wet and dry climatic periods and have been used to determine pathways for tritium migration through the subsurface.

  16. Regulating tritium in drinking water

    International Nuclear Information System (INIS)

    Fluke, R.

    1994-01-01

    This article incorporates an article by E. Koehl from an internal Ontario Hydro publication, and a letter from the Joint Committee of Health and Safety of the Royal Society of Canada and the Canadian Academy of Engineering, submitted to the Ontario Minister of the Environment and Energy. The Advisory Committee on Environmental Standards had recommended that the limit for tritium in Ontario drinking water be reduced from 40,000 to 100 Bq/L, with a further reduction to 20 in five years. Some facts and figures are adduced to show that the effect of tritium in drinking water in Ontario is negligible compared to the effect of background radiation. The risk from tritium to the people of Ontario is undetectably small, and the attempt to estimate this risk by linear extrapolation is extremely dubious. Regulation entails social and economic costs, and the government ought to ensure that the benefits exceed the costs. The costs translate into nothing less than wasted opportunity to save lives in other ways. 3 refs

  17. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

  18. Refurbishing tritium contaminated ion sources

    International Nuclear Information System (INIS)

    Wright, K.E.; Carnevale, R.H.; McCormack, B.E.; Stevenson, T.; Halle, A. von

    1995-01-01

    Extended tritium experimentation on TFTR has necessitated refurbishing Neutral Beam Long Pulse Ion Sources (LPIS) which developed operational difficulties, both in the TFTR Test Cell and later, in the NB Source Refurbishment Shop. Shipping contaminated sources off-site for repair was not permissible from a transport and safety perspective. Therefore, the NB source repair facility was upgraded by relocating fixtures, tooling, test apparatus, and three-axis coordinate measuring equipment; purchasing and fabricating fume hoods; installing exhaust vents; and providing a controlled negative pressure environment in the source degreaser/decon area. Appropriate air flow monitors, pressure indicators, tritium detectors and safety alarms were also included. The effectiveness of various decontamination methods was explored while the activation was monitored. Procedures and methods were developed to permit complete disassembly and rebuild of an ion source while continuously exhausting the internal volume to the TFTR Stack to avoid concentrations of tritium from outgassing and minimize personnel exposure. This paper presents upgrades made to the LPIS repair facility, various repair tasks performed, and discusses the effectiveness of the decontamination processes utilized

  19. Tritium production in fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.

    1981-08-01

    The present analyses on the possibilities of extracting tritium from the liquid and solid fusion reactor blankets show up many problems. A consistent ensemble of materials and devices for extracting the heat and the tritium has not yet been integrated in a fusion reactor blanket project. The dimensioning of the many pipes required for shifting the tritium can only be done very approximately and the volume taken up by the blanket is difficult to evaluate, etc. The utilization of present data leads to over-dimensioning the installations by prudence and perhaps rejecting the best solutions. In order to measure the parameters of the most promising materials, work must be carried out on well defined samples and not only determine the base physical-chemical coefficients, such as thermal conductivity, scattering coefficients, Sievert parameters, but also the kinetic parameters conventional in chemical engineering, such as the hourly space rates of degassing. It is also necessary to perform long duration experiments under radiation and at operating temperatures, or above, in order to study the ageing of the bodies employed [fr

  20. Tritium calorimeter setup and operation

    International Nuclear Information System (INIS)

    Rodgers, David E.

    2002-01-01

    The LBNL tritium calorimeter is a stable instrument capable of measuring tritium with a sensitivity of 25 Ci. Measurement times range from 8-hr to 7-days depending on the thermal conductivity and mass of the material being measured. The instrument allows accurate tritium measurements without requiring that the sample be opened and subsampled, thus reducing personnel exposure and radioactive waste generation. The sensitivity limit is primarily due to response shifts caused by temperature fluctuation in the water bath. The fluctuations are most likely a combination of insufficient insulation from ambient air and precision limitations in the temperature controller. The sensitivity could probably be reduced to below 5 Ci if the following improvements were made: (1) Extend the external insulation to cover the entire bath and increase the top insulation. (2) Improve the seal between the air space above the bath and the outside air to reduce evaporation. This will limit the response drift as the water level drops. (3) Install an improved temperature controller, preferably with a built in chiller, capable of temperature control to ±0.001 C

  1. The oxygen enhancement ratio for single- and double-strand breaks induced by tritium incorporated in DNA of cultured human T1 cells. Impact of the transmutation effect.

    Science.gov (United States)

    Tisljar-Lentulis, G; Henneberg, P; Feinendegen, L E; Commerford, S L

    1983-04-01

    The effect of oxygen, expressed as the oxygen enhancement ratio (OER), on the number of single-strand breaks (SSB) and double-strand breaks (DSB) induced in DNA by the radioactive decay of tritium was measured in human T1 cells whose DNA had been labeled with tritium at carbon atom number 6 of thymidine. Decays were accumulated in vivo under aerobic conditions at 0-1 degrees C and at -196 degrees C and in a nitrogen atmosphere at 0-1 degrees C. The number of SSB and DSB produced was analyzed by sucrose gradient centrifugation. For each tritium decay there were 0.25 DSB in cells exposed to air at 0-1 degrees C and 0.07 in cells kept under nitrogen, indicating an OER of 3.6, a value expected for such low-LET radiation. However, for each tritium decay there were 1.25 SSB in cells exposed to air at 0-1 degrees C and 0.76 in cells kept under nitrogen indicating an OER of only 1.7. The corresponding values for 60Co gamma radiation, expressed as SSB per 100 eV absorbed energy, were 4.5 and 1.0, giving an OER of 4.5. The low OER value found for SSB induced by tritium decay can be explained if 31% of the total SSB produced in air result from transmutation by a mechanism which does not produce DSB and is unaffected by oxygen.

  2. 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.

  3. Overview of light sources powered by tritium

    International Nuclear Information System (INIS)

    Wu Jian; Lei Jiarong; Liu Wenke

    2012-01-01

    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)

  4. Degradation of elastomers by tritium beta radiation

    International Nuclear Information System (INIS)

    Zapp, P.E.; Tuer, G.L. Jr.

    1984-01-01

    Based on its tritium radiation resistance, ethylene propylene rubber has been selected as a candidate for replacement of nitrile rubber in the SRP tritium facilities. A specification for flange gasket material has been developed for ethylene propylene such that its mechanical properties are similar to those of nitrile rubber. In-process testing of ethylene propylene and nitrile gaskets will be conducted in the tritium facilities under identical exposure conditions

  5. Tritium radioluminescent devices, Health and Safety Manual

    International Nuclear Information System (INIS)

    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

  6. Effluent Treatment Facility tritium emissions monitoring

    International Nuclear Information System (INIS)

    Dunn, D.L.

    1991-01-01

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

  7. The movement of tritium in ecological systems

    International Nuclear Information System (INIS)

    Polevoy, Y; Laichter, Y.

    1988-11-01

    This literature survey summarizes the interaction of tritium gas and tritiated water with various components of the ecological system. The intake of tritium gas and tritiated water in plants and soil is described as well as the location of the highest measurable concentration. This information may serve as a basis for risk assessment from tritium to man through the food chain and enables effective tracing of its concentration in the environment. (author)

  8. Tritium monitoring equipments for animal experiment facilities

    International Nuclear Information System (INIS)

    Sato, Hiroo

    1980-01-01

    Animal experiment facilities using tritium are described with reference to laws and regulations concerning radiological safety. Usual breeding facilities and surrounding conditions at non-radioactive animal experiments are summarized on feasible and effective designs of tritium monitors. Characteristics and desirable arrangements of various kinds of tritium monitors such as ionization chambers, proportional counters and liquid scintillation detectors are discussed from the standpoint of monitoring for room, glove-box, stack, liquid waste and personnel. (J.P.N.)

  9. Tritium in the environment. NCRP Report No. 62

    International Nuclear Information System (INIS)

    Eisenbud, M.

    1979-01-01

    The NCRP (National Council on Radiation Protection and Measurements) Report No. 62 on tritium is described. Tritium production from various sources, distribution and environmental kinetics, biological behaviour and the dosimetry of tritium are discussed. (author)

  10. The introduction of tritium in lactose and saccharose by isotope exchange with gaseous tritium

    International Nuclear Information System (INIS)

    Akulov, G.P.; Snetkova, E.V.; Kaminskij, Yu.L.; Kudelin, B.K.; Efimova, V.L.

    1991-01-01

    Methods for conducting reactions of catalytic protium-tritium isotopic exchange with gaseous tritium were developed in order to synthesize tritium labelled lactose and saccharose. These methods enabled to prepare these labelled disaccharides with high molar activity. The yield was equal to 50-60%, radiochemical purity ∼ 95%

  11. Tritium in precipitation of Vostok (Antarctica): conclusions on the tritium latitude effect.

    Science.gov (United States)

    Hebert, Detlef

    2011-09-01

    During the Antarctic summer of 1985 near the Soviet Antarctic station Vostok, firn samples for tritium measurements were obtained down to a depth of 2.40 m. The results of the tritium measurements are presented and discussed. Based on this and other data, conclusions regarding the tritium latitude effect are derived.

  12. Cytogenetic effects of tritium incorporated into DNA of human lymphocytes

    Energy Technology Data Exchange (ETDEWEB)

    Beno, M [Inst. of Preventive and Clinical Medicine, 83301 Bratislava (Slovakia)

    1996-12-31

    In the reported in vitro experiments the numbers of chromosomal aberrations (CA) in correlation to the physical dose as assessed by determining the specific radioactivity of DNA have been followed in vitro human lymphocytes from adult donors. Lymphocytes from healthy adult donors of age from 20 to 59 of both sexes (24 males and 20 females) were isolated from blood by centrifugation. After washing the cells were irradiated from tritium incorporated during in vitro incubation in phytohemagglutinin containing medium with tritium labelled thymidine. Slides for standard CA counting have been done from every sample 48 hours after the begin cultivation. The CA were counted in at least 200 metaphases on each slide. Parallel samples of lymphocytes served for preparation smears for autoradiography to determine the labeling index. Other parallel samples were used for the determination of tritium concentration in DNA by the diphenylamine method, as well as determination of the specific radioactivity in lymphocyte DNA by scintillation counting. The dose absorbed in DNA was estimated using the conversion factor implicating that 37 kBq of tritium uniformly distributed per gram of tissue of unit density delivers a dose rate of 121.4 {sup m}i{sup G}y/hour. The contamination of cells by precursors of nucleic acids - like tritiated thymidine - causes an uneven distribution of doses in the cell population. A proportion of the population of cells remains unlabelled. The dose-response curve is flat showing signs of loss of heavily damaged cells and signs of repair of damage. Both these signs are based on the nature of biological processes which lead to internal contamination of cells and to expression of effects in terms of numbers of CA. (J.K.) 5 figs., 4 refs.

  13. Tritium loss in molten flibe systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  14. Process and system for removing tritium

    International Nuclear Information System (INIS)

    Ridgely, J.N.

    1976-01-01

    A process and system for removing tritium, particularly from high temperature gas cooled atomic reactors (HTGR), is disclosed. Portions of the reactor coolant, which is permeated with the pervasive tritium atom, are processed to remove the tritium. Under conditions of elevated temperature and pressure, the reactor coolant is combined with gaseous oxygen, resulting in the formation of tritiated water vapor from the tritium in the reactor coolant and the gaseous oxygen. The tritiated water vapor and the remaining gaseous oxygen are then successively removed by fractional liquefaction steps. The reactor coolant is then recirculated to the reactor

  15. Tritium stripping by a catalytic exchange stripper

    International Nuclear Information System (INIS)

    Heung, L.K.; Gibson, G.W.; Ortman, M.S.

    1991-01-01

    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

  16. Tritium immobilization and packaging using metal hydrides

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Yaraskavitch, J.M.

    1981-04-01

    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, T 2 . 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)

  17. Development of tritium handing technology(II)

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. S.; Ahn, D. H.; Kim, K. R. [KAERI, Taejon (Korea, Republic of); Yook, D. S.; Song, K. M.; Son, S. H. [KEPRI, Taejon (Korea, Republic of); Lee, K. J.; Jung, H. Y.; Song, M. C. [KAIST, Taejon (Korea, Republic of)

    2004-02-01

    The buildup rate of tritium in heavy water moderator and coolant of pressurized heavy water reactors in Wolsong Nuclear Power Plant is about 4MCi/a. The control of tritium is of increasing concern to the power reactor industry and general public in Korea. Metal tritides have the advantage of significantly decreasing the volume required to store tritium without increasing the pressure of storage vessel. Titanium hydride was safely used for the long-term storage of tritium. The experimental thermodynamic P-C-T data show that titanium soaks up hydrogen isotope gas at ambient temperature and modest pressures.

  18. Tritium loss in molten flibe systems

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  19. Tritium resources available for fusion reactors

    Science.gov (United States)

    Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

    2018-02-01

    The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

  20. Tritium proof-of-principle pellet injector

    International Nuclear Information System (INIS)

    Fisher, P.W.

    1991-07-01

    The tritium proof-of-principle (TPOP) experiment was designed and built by Oak Ridge National Laboratory (ORNL) to demonstrate the formation and acceleration of the world's first tritium pellets for fueling of future fusion reactors. The experiment was first used to produce hydrogen and deuterium pellets at ORNL. It was then moved to the Tritium Systems Test Assembly at Los Alamos National Laboratory for the production of tritium pellets. The injector used in situ condensation to produce cylindrical pellets in a 1-m-long, 4-mm-ID barrel. A cryogenic 3 He separator, which was an integral part of the gun assembly, was capable of lowering 3 He levels in the feed gas to <0.005%. The experiment was housed to a glovebox for tritium containment. Nearly 1500 pellets were produced during the course of the experiment, and about a third of these were pure tritium or mixtures of deuterium and tritium. Over 100 kCi of tritium was processed through the experiment without incident. Tritium pellet velocities of 1400 m/s were achieved with high-pressure hydrogen propellant. The design, operation, and results of this experiment are summarized. 34 refs., 44 figs., 3 tabs