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Sample records for tritium removal procedures

  1. 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

  2. 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)

  3. 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

  4. Tritium removal using vanadium hydride

    International Nuclear Information System (INIS)

    Hill, F.B.; Wong, Y.W.; Chan, Y.N.

    1978-01-01

    The results of an initial examination of the feasibility of separation of tritium from gaseous protium-tritium mixtures using vanadium hydride in cyclic processes is reported. Interest was drawn to the vanadium-hydrogen system because of the so-called inverse isotope effect exhibited by this system. Thus the tritide is more stable than the protide, a fact which makes the system attractive for removal of tritium from a mixture in which the light isotope predominates. The initial results of three phases of the research program are reported, dealing with studies of the equilibrium and kinetics properties of isotope exchange, development of an equilibrium theory of isotope separation via heatless adsorption, and experiments on the performance of a single heatless adsorption stage. In the equilibrium and kinetics studies, measurements were made of pressure-composition isotherms, the HT--H 2 separation factors and rates of HT--H 2 exchange. This information was used to evaluate constants in the theory and to understand the performance of the heatless adsorption experiments. A recently developed equilibrium theory of heatless adsorption was applied to the HT--H 2 separation using vanadium hydride. Using the theory it was predicted that no separation would occur by pressure cycling wholly within the β phase but that separation would occur by cycling between the β and γ phases and using high purge-to-feed ratios. Heatless adsorption experiments conducted within the β phase led to inverse separations rather than no separation. A kinetic isotope effect may be responsible. Cycling between the β and γ phases led to separation but not to the predicted complete removal of HT from the product stream, possibly because of finite rates of exchange. Further experimental and theoretical work is suggested which may ultimately make possible assessment of the feasibility and practicability of hydrogen isotope separation by this approach

  5. 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

  6. 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

  7. Mercury and tritium removal from DOE waste oils

    Energy Technology Data Exchange (ETDEWEB)

    Klasson, E.T. [Oak Ridge National Lab., TN (United States)

    1997-10-01

    This work covers the investigation of vacuum extraction as a means to remove tritiated contamination as well as the removal via sorption of dissolved mercury from contaminated oils. The radiation damage in oils from tritium causes production of hydrogen, methane, and low-molecular-weight hydrocarbons. When tritium gas is present in the oil, the tritium atom is incorporated into the formed hydrocarbons. The transformer industry measures gas content/composition of transformer oils as a diagnostic tool for the transformers` condition. The analytical approach (ASTM D3612-90) used for these measurements is vacuum extraction of all gases (H{sub 2}, N{sub 2}, O{sub 2}, CO, CO{sub 2}, etc.) followed by analysis of the evolved gas mixture. This extraction method will be adapted to remove dissolved gases (including tritium) from the SRS vacuum pump oil. It may be necessary to heat (60{degrees}C to 70{degrees}C) the oil during vacuum extraction to remove tritiated water. A method described in the procedures is a stripper column extraction, in which a carrier gas (argon) is used to remove dissolved gases from oil that is dispersed on high surface area beads. This method appears promising for scale-up as a treatment process, and a modified process is also being used as a dewatering technique by SD Myers, Inc. (a transformer consulting company) for transformers in the field by a mobile unit. Although some mercury may be removed during the vacuum extraction, the most common technique for removing mercury from oil is by using sulfur-impregnated activated carbon (SIAC). SIAC is currently being used by the petroleum industry to remove mercury from hydrocarbon mixtures, but the sorbent has not been previously tested on DOE vacuum oil waste. It is anticipated that a final process will be similar to technologies used by the petroleum industry and is comparable to ion exchange operations in large column-type reactors.

  8. Tritium removal by CO2 laser heating

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Mueller, D.

    1997-01-01

    Efficient techniques for rapid tritium removal will be necessary for ITER to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 ms is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

  9. Tritium removal by CO2 laser heating

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Mueller, D.

    1997-10-01

    Efficient techniques for rapid tritium removal will be necessary for ITER (International Thermonuclear Experimental Reactor) to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:YAG laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 msec is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

  10. Tritium removing method and device therefor

    International Nuclear Information System (INIS)

    Kitayama, Hisao.

    1993-01-01

    A part of cleaned gases introduced from the exit of an adsorber is used for regeneration of the adsorbent which removes tritium water by processing gases to be cleaned, and off-gases are dehydrated and joined with the gases to be cleaned before compression. Further, the exits of a plurality of adsorbers for the cleaned gases are in communication with each other by a regenerated gas supply channel having a heater, and a circulation channel is disposed for circulating the regenerated gases on the suction side of the compressor by way of a cooling dehydration device. Then, it is not necessary to prepare exclusive gases for regeneration, and in the regeneration system, only heating to the temperature for regeneration is sufficient for cleaned gases. Further, regenerated gases can be introduced only by switching of adsorption and removing steps and by operating valves for regeneration and, in addition, gases after regeneration are circulated after joining to the gases to be cleaned. Accordingly, it is not necessary to completely remove tritium water upon dehydration treatment and cold trap is also unnecessary. (N.H.)

  11. In-vessel tritium retention and removal in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

    1998-06-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  12. In-vessel tritium retention and removal in ITER

    International Nuclear Information System (INIS)

    Federici, G.; Anderl, R.A.

    1998-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world's fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  13. Measurements of tritium retention and removal on TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Kamperschroer, J.

    1996-05-01

    Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have afforded an opportunity to measure the retention of tritium in a graphite limiter that is subject to erosion, codeposition and high neutron flux. The tritium was injected by both gas puff and neutral beams. The isotopic mix of hydrogenic recycling was measured spectroscopically and the tritium fraction T/(H+D+T) increased to as high as 75%. Some tritium was pumped out during the experimental run and some removed in a subsequent campaign using various clean-up techniques. While the short term retention of tritium was high, various conditioning techniques were successful in removing ∼ 8,000 Ci and restoring the tritium inventory to a level well below the administrative limit

  14. Operating experience and procedures at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Carlson, R.V.; Binning, K.E.; Cole, S.P.; Jenkins, E.M.; Wilhelm, R.C.; Cole, S.P.

    1988-01-01

    Operating procedures are important for the safe and efficient operation of the Tritium Systems Test Assembly (TSTA). TSTA has been operating for four years with tritium in a safe and efficient manner. The inventory of tritium in the process loop is 100 grams and several milestone runs have been completed. This paper describes the methods used to operate TSTA. 3 refs., 1 fig

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

    International Nuclear Information System (INIS)

    Butler, J.P.; Hammerli, M.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Gutowski, H.; Bracha, M.

    1987-01-01

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

  17. 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

  18. Removal of tritium from gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1976-01-01

    Tritium contained in the coolant gas in the primary circuit of a gas cooled nuclear reactor together with further tritium adsorbed on the graphite used as a moderator for the reactor is removed by introducing hydrogen or a hydrogen-containing compound, for example methane or ammonia, into the coolant gas. The addition of the hydrogen or hydrogen-containing compound to the coolant gas causes the adsorbed tritium to be released into the coolant gas and the tritium is then removed from the coolant gas by passing the mixture of coolant gas and hydrogen or hydrogen-containing compound through a gas purification plant before recirculating the coolant gas through the reactor. 14 claims, 1 drawing figure

  19. Tritium isotope fractionation in biological systems and in analytical procedures

    International Nuclear Information System (INIS)

    Kim, M.A.; Baumgaertner, Franz

    1989-01-01

    The organically bound tritium (OBT) is evaluated in biological systems by determining the tritium distribution ratio (R-value), i.e. tritium concentrations in organic substance to cell water. The determination of the R-value always involves isotope fractionation is applied analytical procedures and hence the evaluation of the true OBT -value in a given biological system appears more complicated than hitherto known in the literature. The present work concentrates on the tritium isotope fractionation in the cell water separation and on the resulting effects on the R-value. The analytical procedures examined are vacuum freeze drying under equilibrium and non-equilibrium conditions and azeotropic distillation. The vaporization isotope effects are determined separately in the phase transition of solid or liquid to gas in pure tritium water systems as well as in real biological systems, e.g. corn plant. The results are systematically analyzed and the influence of isotope effects on the R-value is rigorously quantified

  20. Tritium removal: a preliminary evaluation of several getters

    International Nuclear Information System (INIS)

    Schoenfelder, C.W.; West, L.A.

    1975-11-01

    The removal of hydrogen isotopes from flowing gas streams is an important aspect of CTR technology for both decontamination and tritium recovery from plasma exhausts. Several getters have been evaluated for their tritium scrubbing potential at the parts per billion level. Measurements of total capacity and dynamic response have been made for barium, erbium, palladium dispersed on molecular sieve, General Electric H-36 (zirconium alloy), Union Carbide Y-993 (PdMnO 2 ), Societa Apparecchi Electtrici e Scientifici Getters ST101 (Zr--Al), ST171, and ST181, and a Sandia developed organic material, dimerized phenyl propargyl ether (DPPE). Preliminary flow studies were conducted by passing mixtures of either hydrogen or deuterium diluted with argon through packed beds containing the getter and periodically sampling the effluent with a gas chromatograph sensitive to 500 ppB H 2 . The results of this work, similar flow experiments using tritium and total capacity measurements are presented in the text

  1. Development on the technologies for tritium removal processes

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Ki Woong; Kim, Yong Ik; Nah, Jung Won; Koo, Je Hyoo; Kim, Kwang Lak; Chung, Heung Suk; Lee, Han Soo; Cho, Yung Hyun; Paek, Seung Woo; Kang, Heui Suk [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chung, Yong Won [In Hah Univ., Inchun (Korea, Republic of)

    1994-12-01

    While tritium exposure to the site-workers in Wolsung NPP is upto about 40 % of the total personnel exposure, Korea Institute of Nuclear Safety has asked tritium removal facility, as one of the requirements for post reactor construction, after operation of four CANDU reactors in Wolsung site. For the purpose of essential removal of tritium from the heavy water system of the heavy water reactors, an experiment of Ar-N{sub 2} cryogenic distillation tower was carried out as a preliminary study for development of liquid-phase catalytic exchange - cryogenic hydrogen distillation process. The steady-state reached after 50 minutes under 90 K in the Ar-N{sub 2} distillation column (inner diameter 20 mm, height 500 mm) packed with Dixon ring ({phi} 3 mm x H 3 mm), and the ratios of Ar-concentration at the top and at the bottom measured by gas chromatography within {+-}1 % relative error was approximately 93 : 3. This value was distillation performances quite higher than those estimated by computer-simulation, which might be due to good efficiency of the packing materials. Several dynamic characteristics such as height equivalent to theoretical plate or effects of the kind of packing materials for Ar-N{sub 2} distillation column to be produced will be available for design study of cryogenic hydrogen distillation process. 19 figs, 17 tabs, 21 refs. (Author).

  2. Development on the technologies for tritium removal processes

    International Nuclear Information System (INIS)

    Sung, Ki Woong; Kim, Yong Ik; Nah, Jung Won; Koo, Je Hyoo; Kim, Kwang Lak; Chung, Heung Suk; Lee, Han Soo; Cho, Yung Hyun; Paek, Seung Woo; Kang, Heui Suk; Chung, Yong Won

    1994-12-01

    While tritium exposure to the site-workers in Wolsung NPP is upto about 40 % of the total personnel exposure, Korea Institute of Nuclear Safety has asked tritium removal facility, as one of the requirements for post reactor construction, after operation of four CANDU reactors in Wolsung site. For the purpose of essential removal of tritium from the heavy water system of the heavy water reactors, an experiment of Ar-N 2 cryogenic distillation tower was carried out as a preliminary study for development of liquid-phase catalytic exchange - cryogenic hydrogen distillation process. The steady-state reached after 50 minutes under 90 K in the Ar-N 2 distillation column (inner diameter 20 mm, height 500 mm) packed with Dixon ring (φ 3 mm x H 3 mm), and the ratios of Ar-concentration at the top and at the bottom measured by gas chromatography within ±1 % relative error was approximately 93 : 3. This value was distillation performances quite higher than those estimated by computer-simulation, which might be due to good efficiency of the packing materials. Several dynamic characteristics such as height equivalent to theoretical plate or effects of the kind of packing materials for Ar-N 2 distillation column to be produced will be available for design study of cryogenic hydrogen distillation process. 19 figs, 17 tabs, 21 refs. (Author)

  3. Removal mechanism of tritium by variously pretreated silica gel

    International Nuclear Information System (INIS)

    Nakashima, M.; Tachikawa, E.; Saeki, M.; Aratono, Y.

    1981-01-01

    Removal mechanisms of HTO from variously pretreated and non-pretreated silica gel columns were investigated with pulse-loading with tritiated water vapor. With non-pretreated silica gel, the HTO physisorbed on the upper part of the column comes into contact with surface hydroxyl groups while passing downward the column, so that in each equilibration a part of the tritium is incorporated into hydroxyl groups by H/T isotopic exchange reactions. With the silica gel pretreated at a temperature below 400 0 C, most of tritium in the applied HTO is easily incorporated into surface hydroxyl groups in the upper part of the column either by H/T isotopic exchange reactions or by rehydration of the dehydrated surface (siloxyl linkage). In the pretreatment above 400 0 C, essentially all the tritium is trapped by siloxyl groups of various stabilities. The ease of rehydration of siloxyl groups by applied HTO depends on their stabilities, which, in turn, depend on the pretreatment temperature. As a general trend, treatment at higher temperature promotes annealing of the constrained siloxyl groups and thus the rate of rehydration becomes slower. (author)

  4. Tritium fractionation in biological systems and in analytical procedures

    International Nuclear Information System (INIS)

    Kim, M.A.; Baumgaertner, F.

    1991-01-01

    The organically bound tritium (OBT) is evaluated in biological systems by measuring the tritium distribution ratio (R-value), i.e. tritium concentrations in organic substance to tissue water. The determination of the R-value is found to involve always isotope fractionation in applied analytical procedures and hence the evaluation of the true OBT-value in a given biological system appears more complicated than hitherto known in the literature. The present work concentrates on the tritium isotope fraction in the tissue water separation and on the resulting effects on the R-value. The analytical procedures examined are vacuum freeze drying under equilibrium and non-equilibrium conditions and azeotropic distillation. The vaporization isotope effects are determined separately in the phase transition of solid or liquid to gas in pure water systems as well as in real biological systems, e.g. maize plant. The results are systematically analysed and the influence of isotope effects on the R-value is rigorously quantified. (orig.)

  5. Investigations of titamium and zirconium hydrides to determine suitability of recoverable tritium immobilization for the Pickering tritium removal system

    International Nuclear Information System (INIS)

    Noga, J.O.

    1981-11-01

    A tritium removal system will be constructed at Pickering Nuclear Generating station to reduce the adverse effects of this radioactive hydrogen isotope. This report summarizes various properties of titanium and zirconium sponge hydrides which have been selected as suitable candidates for tritium product immobilization. Equilibrium pressure-composition-temperature data indicates that both materials behave suitably to provide a safe, solid form of tritium storage. Titanium tritide is recommended as the best choice due to higher dissociation pressures which can be achieved at equivalent temperatures when compared to zirconium tritide. Higher dissociation pressures would result in faster and more efficient recovery of tritium gas from the immobilized state. It is evident from the stability of these compounds that their utilization as tritides will greatly enhance the integrity of tritium storage

  6. 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

  7. 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.)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  10. Tritium retention in next step devices and the requirements for mitigation and removal techniques

    International Nuclear Information System (INIS)

    Counsell, G; Coad, P; Grisola, C; Hopf, C

    2006-01-01

    Mechanisms underlying the retention of fuel species in tokamaks with carbon plasma-facing components are presented, together with estimates for the corresponding retention of tritium in ITER. The consequential requirement for new and improved schemes to reduce the tritium inventory is highlighted and the results of ongoing studies into a range of techniques are presented, together with estimates of the tritium removal rate in ITER in each case. Finally, an approach involving the integration of many tritium removal techniques into the ITER operational schedule is proposed as a means to extend the period of operations before major intervention is required

  11. Assessment of the cryogenic distillation system in Cernavoda tritium removal facility

    International Nuclear Information System (INIS)

    Pasca, Gheorghe; Draghia, Mirela; Porcariu, Florina; Ana, George

    2010-01-01

    Full text: This paper aims at presenting an assessment of the Cryogenic Distillation system (CD) in the Cernavoda Tritium Removal Facility (CTRF). The cryogenic distillation system is one of the key components of the CTRF which comprises other systems as: the liquid phase catalytic exchange system, designed to transfer tritium from heavy water to a deuterium stream to be fed into the CD system; the atmosphere detritiation system; the tritium recovery system; the tritium/hydrogen monitoring system; the central interlocking system; the tritium extraction and storage system. Thus, the need to build a tritium separation and recovery system results from economic opportunities offered both by heavy water reuse and tritium production, but, at the same time, it offers an alternative for the storage of tritiated heavy water as radioactive waste. (authors)

  12. Current status for applications of hydrophobic platinum catalysts in tritium removal from nuclear effluents

    International Nuclear Information System (INIS)

    Vagner, Irina; Ionita, Gheorghe; Varlam, Carmen

    2008-01-01

    Full text: Based on the long experience of the authors, in the preparation, testing and evaluation of the performances of hydrophobic catalysts, and based on the reviewed references, this paper presents up-to-date R and D results on the preparation methods and applications of the hydrophobic catalysts, in deuterium and tritium separation. The objectives of the paper are: 1. to provide a database for selection of the most appropriate catalyst and catalytic packing for above mentioned processes; 2. to evaluate the potentiality of hydrophobic Pt-catalysts in the deuterium and tritium separation; 3. to assess and find a new procedure for preparation of a new improved hydrophobic catalyst. The merits of the hydrophobic catalysts are shown in comparison to hydrophilic catalysts. As results of the review some general conclusions about the applications of hydrophobic catalysts in environmental field are as follow: 1. the hydrophobic Pt-catalysts packed in the trickle bed reactors showed a high catalytic activity and long stability; 2. the utilization of the hydrophobic Pt-catalysts for tritium removal from liquid and gaseous effluent in nuclear field was entirely confirmed on industrial scale; 3. the extension of the utilization of the hydrophobic Pt-catalysts to other new processes, which take place in presence of liquid water or high humidity, like VOCs oxidation from wastewater or H 2 -O 2 catalytic recombination, are subject to testing

  13. Removal of impurities from environmental water samples for tritium measurement by means of liquid scintillation counter

    International Nuclear Information System (INIS)

    Sakuma, Yoichi; Noda, Mitsuyasu

    2000-01-01

    Tritium concentration in environmental water samples is usually measured by means of liquid scintillation counting. Before the counting distillation operation is necessarily required to remove impurities, which have possibility of bad influence on the measurement, from the samples. But the operation usually takes long time and it is also troublesome. If you could simplify the purification process, you would be much easily able to measure it. Then, we have studied the probability of replacement the process by filtration aiming to simplify the procedure. We prepared several environmental water samples and also several water samples added quenching materials. These samples were purified by means of the distillation and the filtration and the impurities in them were examined. The purified samples were mixed with scintillation cocktail and the tritium concentration was measured. We added small amount of tritium in the same samples and investigated their scintillation spectra and their ESCR values in order to compare the two purification methods. Two kinds of filters were used for the filtration: 0.45 μm and 0.1 μm pore sized membrane filters. The liquid scintillation counter was LB-3 produced by Aloka Co. and Ltd. The scintillation cocktail was Ultima Gold LLT made by Packard Instrument Co and Ltd. The vial was Polyvial 145 LSD made by Zinsser Analytic Co. and Ltd. As the result, there was no significant difference between the two purification methods then the filtration method is feasible instead of the distillation. (author)

  14. Separation of hydrogen isotopes for tritium waste removal

    International Nuclear Information System (INIS)

    Wilkes, W.R.

    1975-01-01

    A distillation cascade for separating hydrogen isotopes was simulated by means of a multicomponent, multistage computer code. A hypothetical test mixture containing equal atomic fractions of protium, deuterium and tritium, equilibrated to high temperature molecular concentrations was used as feed. The results show that a two-column cascade can be used to separate the protium from the tritium. Deuterium appears both in the protium and the tritium product streams. (auth)

  15. 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.)

  16. Tritium removal from contaminated water via infrared laser multiple-photon dissociation

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Magnotta, F.; Herman, I.P.; Aldridge, F.T.; Hsiao, P.

    1983-01-01

    Isotope separation by means of infrared-laser multiple-photon dissociation offers an efficient way to recover tritium from contaminated light or heavy water found in fission and fusion reactors. For tritium recovery from heavy water, chemical exchange of tritium into deuterated chloroform is followed by selective laser dissociation of tritiated chloroform and removal of the tritiated photoproduct, TCl. The single-step separation factor is at least 2700 and is probably greater than 5000. Here we present a description of the tritium recovery process, along with recent accomplishments in photochemical studies and engineering analysis of a recovery system

  17. 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

  18. Method of removing tritium in exhaust water in a nuclear equipment

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, T

    1976-05-12

    A method is claimed to increase the efficiency of removing tritium from waste water through adsorption treatment. Steam is produced by heating waste water containing tritium, and it is passed through a tube filled with an adsorbent such as activated alumina, silica gel or zeolite. When a control limiting value is reached by the concentration of tritium within the steam, the flow of steam is stopped, and the adsorption tube is removed from the path of steam flow. Thereafter, another adsorption tube containing the afore-said adsorbent is provided in the steam flow path, and the steam is then allowed to flow again.

  19. A design assessment of tritium removal systems for the mirror advanced reactor study

    International Nuclear Information System (INIS)

    Sood, S.K.; Kveton, O.K.

    1983-01-01

    This study investigates the available processes for removing tritium from light water, and selects the most appropriate process for recovering tritium from the various tritiated water streams identified in the Mirror Advanced Reactor Study (MARS). A simplified flowsheet is shown for the process and the main process parameters are identified. Previous experience is utilized to predict direct capital costs and power requirement for the Tritiated Water Removal Unit (TWRU). A number of possibilities are discussed for lowering the cost of the TWRU. An estimate is made of the direct capital cost for the Air Detritiation System that has already been selected as the reference design by MARS personnel. The leakage from the MARS coolant loop is estimated, based on the experience obtained with Ontario Hydro's coolant systems. Design targets are identified for tritium levels in the reactor hall atmosphere and in water and air emissions. Tritium levels are predicted for these and are assessed against the previously identified targets

  20. Technical solutions for tritium removal from Cernavoda NPP heavy water systems

    International Nuclear Information System (INIS)

    Barariu, Gheorghe; Panait, Adrian

    2002-01-01

    In CANDU nuclear plants 2400 KCi/GW(e) - year tritium is generated. At a CANDU - 600 reactor similar to Cernavoda NPP Unit 1, 1500 KCi/year of tritium is generated 95% being in the D 2 O moderator, which can achieve a radioactivity level of 80 - 100 Ci/kg. Tritium in heavy water contributes with 30 - 50% to the doses received by operation personnel and with 20% to the radioactivity released to the environment. The extraction of tritium heavy water at CANDU reactors implies the following possibilities: - the radioactivity level reduction in the operation area; - the maintenance and repair cost reduction due to reduction of personnel protection measures and increased labor productivity; - the increase of NPP utilization factor by shutdown time reduction for maintenance and repair; - tritium concentration reduction from technological systems, ensuring thus the possibility of redesigning the systems in order to lower the cost of investment; - profitable use of extracted tritium. Technical measures provided by AECL project for CANDU 600 at Cernavoda make possible to satisfy the current standards concerning tritium concentration in the operation area atmosphere of 5 x 10 -6 Ci/m 3 . The regulations recommend that the radioactivity level should be maintained as low as possible in conformity with ALARA principles. Also, it is possible that norms will become more restrictive in the future, so the tritium removal technology is a good preventive measure which may become very necessary. The methods, which currently reached the industrial or pilot stages, are based on catalyzed chemical exchange, the heavy water electrolysis, and deuterium distillation. They are known as: VPCE - Vapour Phase Catalytic Exchange; LPCE - Liquid Phase Catalytic Exchange; DE - Direct Electrolysis; CD - Cryogenic Distillation. As transfer processes the catalyzed chemical exchange and heavy water electrolysis are used while concentration of tritium gas is done by cryogenic distillation. At present the

  1. Methods of removal of tritium from aqueous effluent: a review of international research and development

    International Nuclear Information System (INIS)

    Segal, M.G.

    1988-01-01

    Tritium is formed in thermal nuclear reactors both by neutron activation of elements such as deuterium and lithium and by ternary fission in the fuel. It is a weak beta-emitter with a short half-life, 12.3 years, and its radiological significance in reactor discharges is very low. In heavy-water-cooled and -moderated reactors, such as the CANDU stations, the tritium concentration in the moderator is sufficiently high to cause a potential hazard to operators, and so a major research and development programme has been carried out on processes to remove the tritium. Detritiation of light water has also been the subject of major R and D effort world-wide, because reprocessing operations can generate significant quantities of tritium in liquid waste, and high concentrations of tritium may arise in some aqueous streams in fusion reactors. This Report presents a review of the methods that have been proposed, studied and developed for removal of tritium from light and heavy water: the principles of individual methods are discussed, and the current status of their development is reviewed. (author)

  2. Pre-Conceptual Design for Northstar ⁹⁹Mo Process Tritium Removal System

    Energy Technology Data Exchange (ETDEWEB)

    Nobile, Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reichert, Heidi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hollis, William Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taylor, Craig Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gordon, John Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dale, Gregory E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-01-12

    In this report we describe a preliminary concept for a Tritium Removal System (TRS) to remove tritium that is generated in the ⁹⁹Mo production process. Preliminary calculations have been performed to evaluate an approximate size for the system. The concept described utilizes well-established detritiation technology based on catalytic oxidation of tritium and tritiated hydrocarbons to water in a high temperature (400 °C) reactor and capture of water in a molecular sieve bed. The TRS concept involves use of a single system that would cycle through each of the seven online target systems and remove tritium that has been accumulated after one week’s run time. The TRS would perform cleanup operations on each target system for a period of approximately 24 hours. This would occur while the system is still online and just prior to target replacement, so tritium levels would at their minimum values for target replacement. In the concept, during normal operation a small fraction (1%) of the helium recirculating in the system would be diverted through the TRS and returned to the flow loop. With this approach sufficient levels of detritiation can be accomplished in a 24 hour period. In the study it was found that because of the need to maintain low oxygen levels in the system (<100 ppm) this increases the size of the catalytic reactor. As a result of this finding, consideration should be given to other methods for removing tritium from the system. Other methods such as catalytic exchange of tritium with an unsaturated organic compound and subsequent trapping on activated carbon or molecular sieve could offer advantages of reducing reactor size and operation at lower reactor temperature. However the most significant advantage of such an approach would be the ability to operate in very low oxygen environments, which would eliminate any concerns for oxidation of the target.

  3. Thermal Removal of Tritium from Concrete and Soil to Reduce Groundwater Impacts - 13197

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, Dennis G. [Savannah River National Laboratory, Building 773-42A, Aiken, South Carolina 29808 (United States); Blount, Gerald C. [Savannah River Nuclear Solutions (United States); Wells, Leslie H.; Cardoso, Joao E.; Kmetz, Thomas F.; Reed, Misty L. [U.S Department of Energy-Savannah River Site (United States)

    2013-07-01

    Legacy heavy-water moderator operations at the Savannah River Site (SRS) have resulted in the contamination of equipment pads, building slabs, and surrounding soil with tritium. At the time of discovery the tritium had impacted the shallow (< 3-m) groundwater at the facility. While tritium was present in the groundwater, characterization efforts determined that a significant source remained in a concrete slab at the surface and within the associated vadose zone soils. To prevent continued long-term impacts to the shallow groundwater a CERCLA non-time critical removal action for these source materials was conducted to reduce the leaching of tritium from the vadose zone soils and concrete slabs. In order to minimize transportation and disposal costs, an on-site thermal treatment process was designed, tested, and implemented. The on-site treatment consisted of thermal detritiation of the concrete rubble and soil. During this process concrete rubble was heated to a temperature of 815 deg. C (1,500 deg. F) resulting in the dehydration and removal of water bound tritium. During heating, tritium contaminated soil was used to provide thermal insulation during which it's temperature exceeded 100 deg. C (212 deg. F), causing drying and removal of tritium. The thermal treatment process volatiles the water bound tritium and releases it to the atmosphere. The released tritium was considered insignificant based upon Clean Air Act Compliance Package (CAP88) analysis and did not exceed exposure thresholds. A treatability study evaluated the effectiveness of this thermal configuration and viability as a decontamination method for tritium in concrete and soil materials. Post treatment sampling confirmed the effectiveness at reducing tritium to acceptable waste site specific levels. With American Recovery and Reinvestment Act (ARRA) funding three additional treatment cells were assembled utilizing commercial heating equipment and common construction materials. This provided a

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  5. Development on the cryogenic hydrogen isotopes distillation process technology for tritium removal (Final report)

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Ki Woung; Kim, Yong Ik; Na, Jeong Won; Ku, Jae Hyu; Kim, Kwang Rak; Jeong, Yong Won; Lee, Han Soo; Cho, Young Hyun; Ahn, Do Hee; Baek, Seung Woo; Kang, Hee Seok; Kim, You Sun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-12-01

    While tritium exposure to the site-workers in Wolsung NPP is up to about 40% of the total personnel exposure, Ministry of Science and Technology has asked tritium removal facility for requirement of post heavy-water reactor construction. For the purpose of essential removal of tritium from the Wolsung heavy-water reactor system, a preliminary study on the cryogenic Ar-N{sub 2} and H{sub 2}-D{sub 2} distillation process for development of liquid-phase catalytic exchange cryogenic hydrogen distillation process technology. The Ar-N{sub 2} distillation column showed good performance with approximately 97% of final Ar concentration, and a computer simulation code was modified using these data. A simulation code developed for cryogenic hydrogen isotopes (H{sub 2}, HD, D{sub 2}, HT, DT, T{sub 2}) distillation column showed good performance after comparison with the result of a JAERI code, and a H{sub 2}-D{sub 2} distillation column was made. Gas chromatography for hydrogen isotopes analysis was established using a vacuum sampling loop, and a schematic diagram of H{sub 2}-D{sub 2} distillation process was suggested. A feasibility on modification of H{sub 2}-D{sub 2} distillation process control system using Laser Raman Spectroscopy was studied, and the consideration points for tritium storage system for Wolsung tritium removal facility was suggested. 31 tabs., 79 figs., 68 refs. (Author).

  6. Development on the cryogenic hydrogen isotopes distillation process technology for tritium removal (Final report)

    International Nuclear Information System (INIS)

    Sung, Ki Woung; Kim, Yong Ik; Na, Jeong Won; Ku, Jae Hyu; Kim, Kwang Rak; Jeong, Yong Won; Lee, Han Soo; Cho, Young Hyun; Ahn, Do Hee; Baek, Seung Woo; Kang, Hee Seok; Kim, You Sun

    1995-12-01

    While tritium exposure to the site-workers in Wolsung NPP is up to about 40% of the total personnel exposure, Ministry of Science and Technology has asked tritium removal facility for requirement of post heavy-water reactor construction. For the purpose of essential removal of tritium from the Wolsung heavy-water reactor system, a preliminary study on the cryogenic Ar-N 2 and H 2 -D 2 distillation process for development of liquid-phase catalytic exchange cryogenic hydrogen distillation process technology. The Ar-N 2 distillation column showed good performance with approximately 97% of final Ar concentration, and a computer simulation code was modified using these data. A simulation code developed for cryogenic hydrogen isotopes (H 2 , HD, D 2 , HT, DT, T 2 ) distillation column showed good performance after comparison with the result of a JAERI code, and a H 2 -D 2 distillation column was made. Gas chromatography for hydrogen isotopes analysis was established using a vacuum sampling loop, and a schematic diagram of H 2 -D 2 distillation process was suggested. A feasibility on modification of H 2 -D 2 distillation process control system using Laser Raman Spectroscopy was studied, and the consideration points for tritium storage system for Wolsung tritium removal facility was suggested. 31 tabs., 79 figs., 68 refs. (Author)

  7. Consideration for a tritium removal facility at the Cernavoda Nuclear Power Station

    International Nuclear Information System (INIS)

    2006-01-01

    Full text: A pre-feasibility study considering process options for a Tritium Removal Facility at the Cernavoda Nuclear Power Station has been completed by ICIT and AECL. Three different process options were considered. These three options differ in the front-end process used to transfer tritium from heavy water to deuterium gas. All three options use cryogenic distillation (CD) as a back end process to extract tritium from the deuterium gas stream and concentrate it into a small volume stream of pure DT or T 2 that can be immobilized on a titanium sponge. The first option for the front-end process is Liquid Phase Catalytic Exchange (LPCE). The LPCE column is used to transfer the tritium from the heavy water to a recirculating stream of deuterium gas. The separation of hydrogen isotopes takes place in the cryogenic distillation column. Tritium-depleted deuterium gas from the CD system is fed back to the LPCE column. The cryogenic distillation system concentrates the tritium into a small volume of elemental tritium for storage. Tritiated heavy water that has been purified to remove catalyst poisons is fed to the top of the LPCE column. The heavy water leaving the column is depleted in deuterium. Both existing detritiation plants built to detritiate CANDU reactors (the Darlington TRF in Canada and the Wolsung TRF in Korea) use variations of the LPCE-CD process. The second option uses electrolysis to convert tritiated heavy water into oxygen and tritiated deuterium gas. The deuterium gas is sent to the Cryogenic Distillation system to extract and concentrate the tritium. The tritium depleted deuterium gas is recombined with the electrolytic oxygen to give a tritium-depleted heavy water product. The third option uses a Combined Electrolysis and Catalytic Exchange (CECE) front end. A CECE process concentrates the tritium in the water and, using water electrolysis, converts the concentrated tritium into deuterium gas. An overhead catalytic recombiner converts the

  8. A study on the primary requirement for the safety of the Wolsong tritium removal facility

    International Nuclear Information System (INIS)

    Hwang, K. H.; Lee, K. J.; Jeong, C. W.

    2001-01-01

    Owing to the using a heavy water as a moderator and a coolant in Heavy water reactor, A large mount of tritium is produced due to a reaction of deuterium with neutron in the reactor and some of tritium is released to the environment. In Wolsong, 4 units (CANDU-600 type) Heavy water reactor is in operation. And the generated amount of tritium is increased with the increase of operational year of the Wolsong nuclear reactor. Decommissioning of the Wolsong unit 1 is expected to start at 2013. Before 2013, to reduce the workers internal radiation doses and environmental release of tritium, Tritium Removal Facility (TRF) is required and should be operated. Wolsong TRF (WTRF) is under developing stage by Korea Electric Power Corporation(KEPCO)and scheduled to start operation about 2006. Once the facility begins operation it can be contributed to the greatly reduction of tritium release to the environment and worker's expose. In this situation, study about the safety assessment method and regulatory requirement is essential for safety insurance of WTRF. And this helps the safety acquirement, successful operation and reliance of WTRF

  9. Development status of the cryogenic distillation system in Cernavoda Tritium Removal Facility

    International Nuclear Information System (INIS)

    Draghia, Mirela; Ana, George; Pasca, Gheorghe; Porcariu, Florina

    2009-01-01

    Full text: The reference design technology for the heavy water detritiation plant of Cernavoda CANDU station is based on combination of Liquid Phase Catalytic Exchange (LPCE) and Cryogenic Distillation (CD) processes. Based on this technology, tritium is transferred from the heavy water to a deuterium stream in the catalyzed isotopic exchange process, LPCE, followed by a final enrichment within the cryogenic distillation cascade. The final step is the tritium storage on metallic hydride. The basic function of the Cryogenic Distillation System (CDS) is the separation of tritium from the tritiated deuterium coming from the LPCE column in the following conditions: - the final product has to be tritium with a concentration of at least 99%; - it must be provided a detritiation factor of at least 100 (the ration between the tritium concentration in the deuterium stream fed to the CD system and the tritium concentration in the returned stream to the LPCE); - the deuterium must be enriched up to 99.995%, by removing the protium; - provisions for safe discharge of the entire inventory of the CD cascade into buffer vessels shall be implemented. To summarize, the present status of the project consists of technical documentation for all the components of CDS, including the P and ID (Pipping and Instrumentation Diagram), preliminary data sheets, technical specifications, drawings for the major components as the buffer vessels, coldbox, etc, and 3D models as well for almost all the components. (authors)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  12. Investigation of tritium removal by means of organic compounds. Catalytic hydrogenation (tritiation) of linoleic acid

    International Nuclear Information System (INIS)

    El-Sharnouby, A.; Weichselgartner, H.

    1984-11-01

    In the presence of noble-metal catalysts unsaturated fatty acids such as eruic acid and linoleic acid capture hydrogen (and tritium) quantitatively. The hydrogenation reaction of eruic acid has already been reported. The experimental results of the reaction of hydrogen (and tritium) with linoleic acid are now discussed in this paper. Obviously, the use of linoleic acid shows some advantages compared with eruic acid: - the hydrogenation reaction is faster, - linoleic acid is liquid, so that the choice of additional solvents is easier, and - linoleic acid is a more or less cheap natural product, which is available from a series of seeds, so that the cost of a technical tritium removal plant is not increased by the basic chemical material. (orig.)

  13. Demonstration tests of tritium removal device under the conditions of nuclear fusion reactor. Cooperation test between Japan and USA

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

    2001-01-01

    Performance of oxidation catalysis in emergency tritium removal device was tested in Los Alamos National Laboratory by cooperation between Japan and USA on November 8, 2000. To reduce the effects of tritium on the environment, a plan of the closed space for trapping tritium was made. A tritium removal device using oxidation catalysis and water vapor adsorption removes the tritium in the closed space. The treatment flow rate of the device is about 2,500 m 3 /h, the same as ITER(3,000 to 4,500 m 3 /h). Catalysis is Pt/ alumina. The closed space is 3,000m 2 . The initial concentration of tritium was about 7 Bq/cm 2 , ten times as large as the concentration limit in atmosphere. The concentration of tritium in the test laboratory decreased linearly with time and attained to the limit value after about 200 min. Residue of tritium on the wall had been removed and the significant quantity was not detected after three days. The results proved to satisfy safety of ITER. (S.Y.)

  14. A Study on Thermal Desorption of Deuterium in D-loaded SS316LN for ITER Tritium Removal System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Myungchul; Kim, Heemoon; Ahn, Sangbok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Jaeyong; Lee, Sanghwa; LanAhn, Nguyen Thi [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    Because Type B radwaste includes tritium on its inside, especially at vicinity of surface, tritium removal from the radwaste is a matter of concern in terms of the radwaste processes. Tritium behavior in materials is related with temperature. Considering a diffusion process, it is expected that tritium removal efficiency is enhanced with increasing baking temperature. However, there is a limitation about temperature due to facility capacity and economic aspect. Therefore, it is necessary to investigate the effect of temperature on the desorption behavior of Tritium in ITER materials. TDS analysis was performed in SS316LN loaded at 120, 240 and 350 °C. D2 concentration and the desorption peak temperature increased with increasing loading temperature. Using peak shift method with three ramp rates of 0.166, 0.332, and 0.5 °C/sec, trap activation energy of D in SS316LN loaded at 350 °C was 56 kJ/mol.

  15. A vacuum disengager for tritium removal from HYLIFE-II Reactor Flibe

    International Nuclear Information System (INIS)

    Dolan, T.J.; Longhurst, G.R.; Garcia-Otero, E.

    1992-01-01

    We have designed a vacuum disengager system to remove tritium from the Flibe (Li 2 BeF 4 ) molten salt coolant of the HYLIFE-II fusion reactor. There is a two-stage vacuum disengager in each of three intermediate heat exchanger (IHX) loops. Each stage consists of a vacuum chamber 4 m in diameter and 7 m tall. As 0.2 mm diameter molten salt droplets fall vertically downward into the vacuum, most of the tritium diffuses out of the droplets and is pumped away. A fraction Φ ∼10 -5 of the 8.6 MCi/day tritium source (from breeding in the Flibe and from unburned fuel) remains in the Flibe as it leaves the vacuum disengagers, and about 21% of that permeates into the intermediate coolant loop, so about 20 Ci/day leak into the steam system. With Flibe primary coolant and a vacuum disengager, it appears that an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate Flibe vacuum disengager operation

  16. Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

    Directory of Open Access Journals (Sweden)

    G. De Temmerman

    2017-08-01

    Full Text Available As a licensed nuclear facility, ITER must limit the in-vessel tritium (T retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

  17. Procedure and technique critique for tritium enrichment by electrolysis at the IAEA Laboratory (effective November 1976)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1976-11-05

    This publication gives a detailed description of the experimental and calculation procedures for tritium enrichment. Most descriptive sections are divided into 2 parts: Section A describes the procedure in the IAEA laboratory; section B discusses the reasons behind the various procedures, and may indicate alternative acceptable, or in some cases even better, procedures. The description of the equipment focuses on electrolysis cells, cooling system and power supply. Routine procedures are discussed including handling and checking of samples after receipt, 'spike' and blank water, initial sample distillation, preparation of cells and samples for electrolysis, electrolysis and completion of electrolysis (weighing of cells, neutralisation and distillation) and precautions against contaminations (prevention, detection and cure). A list of equipment required for electrolytic enrichment of tritium is provided.

  18. Procedure and technique critique for tritium enrichment by electrolysis at the IAEA Laboratory (effective November 1976)

    International Nuclear Information System (INIS)

    1976-01-01

    This publication gives a detailed description of the experimental and calculation procedures for tritium enrichment. Most descriptive sections are divided into 2 parts: Section A describes the procedure in the IAEA laboratory; section B discusses the reasons behind the various procedures, and may indicate alternative acceptable, or in some cases even better, procedures. The description of the equipment focuses on electrolysis cells, cooling system and power supply. Routine procedures are discussed including handling and checking of samples after receipt, 'spike' and blank water, initial sample distillation, preparation of cells and samples for electrolysis, electrolysis and completion of electrolysis (weighing of cells, neutralisation and distillation) and precautions against contaminations (prevention, detection and cure). A list of equipment required for electrolytic enrichment of tritium is provided

  19. New mechanism for enhancing ash removal efficiency and reducing tritium inventory

    International Nuclear Information System (INIS)

    Li Chengyue; Deng Baiquan; Yan Jiancheng

    2007-01-01

    A new mechanism is suggested to suppress ash particle back streams in the divertor region of our fusion experimental breeder (FEB) reactor for enhancing the ash removal efficiency and reducing the tritium inventory by applications of the nonlinear effect of high power rf ponderomotive force potential which reflects the plate-released and re-ionized He + back to the plate. Meanwhile, the potential does not hinder α particles, which are coming from scraping of the layer, flowing to the target plate. However, it does stop tritium ions flowing to the target. Based on the FEB design parameters, our calculations have shown that the ash removal efficiency can be improved by as much as 40% if the parallel component of rf field 150-200 V/cm is applied to the location at a perpendicular distance L=20 cm apart from the plate and the plate-recycling neutral helium atom energy is about 0.75 eV, at the same time, the tritium inventory can be reduced to some extent. (authors)

  20. Simulation and optimisation of the data acquisition system for tritium removal pilot plant

    International Nuclear Information System (INIS)

    Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu

    2004-01-01

    Optimization and simulation of systems especially in science and engineering can help to reduce risk and cost of design and testing processes. A huge number of codes has been developed to support modeling and simulation efforts. All of these software tools support the use of one or more mathematical model classes. Despite all of these efforts, it is hard to find simulation software, which is capable of combining several model classes in a real industry standard environment. The paper presents a simulation software product for controlling and data acquisition system of cryogenic installation process in the tritium removal pilot plant, using an industry standard programming environment widely applied to data acquisition, process control and data visualization, namely LabView. One of the problems in a tritium separation installation is controlling the temperature. To solve this problem it is necessary to develop a simulation system which includes the mathematical model for cryogenic distillation. Also with this simulation system we can approach the safety system which ensures the monitoring of radiations and toxic gases from installation. All elements used in controlling, modeling and simulation of the process, as well as, in the datalogging and supervisory control module from tritium removal installation are new. (authors)

  1. Production of ultrapure D-T gas by removal of molecular tritium by selective adsorption

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Hudson, R.S.; Tsugawa, R.T.; Fearon, E.M.; Souers, P.C.; Collins, G.W.

    1992-01-01

    Production of molecular deuterium-tritium (D-T) with very low molecular tritium (T 2 ) is necessary for application as a nuclear spin polarized fuel. Selective adsorption of hydrogen isotopes on zeolites or alumina can provide the separation needed to produce D-T with very low T 2 . Use of an absorption column at 20-25 K offers low inventory, compact size, and rapid operation, in comparison with conventional separation techniques such as cryogenic distillation or thermal diffusion. In this paper, the authors discuss principles of absorption, and describe a calculational model of the absorption column and operational implications revealed by it. The authors show experimental proof-of-principle data for removal of T 2 from D-T with an adsorption column operated at 23 K

  2. Procedure for the preparation of tritium-labelled insulins

    International Nuclear Information System (INIS)

    Bienert, M.; Haensicke, A.; Beyermann, M.; Kaufmann, K.D.; Oehlke, J.; Klauschenz, E.; Bespalowa, S.; Titov, M.; Pleiss, U.

    1986-01-01

    This invention is concerned with a procedure for the preparation of specific 3 H-labelled insulins with sequences of human, bovine or porcine insulins and without simultaneous chemical modifications of the insulin. On the basis of this procedure a 3 H 2 -Typ (B26)-insulin can be obtained in good yield and purity with a specific radioactivity appropriate to biopharmaceutical and pharmacokinetic purposes in medicine and pharmaceutical industry, resp

  3. 1997 evaluation of tritium removal and mitigation technologies for Hanford Site wastewaters

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Biyani, R.K.; Duncan, J.B.; Flyckt, D.L.; Mohondro, P.C.; Sinton, G.L.

    1997-01-01

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

  4. Use of hydrophobic Pt-catalysts in tritium removal from effluents

    International Nuclear Information System (INIS)

    Gheorghe, Ionita; Popescu, Irina; Stefanescu, Ioan; Steflea, Dumitru; Varlam, Carmen

    2002-01-01

    Based on the long experience of the authors, in the preparation, testing and evaluation of the performances of hydrophobic catalysts, and based on the reviewed references, this paper presents up-to-date R and D activities on the application of the hydrophobic catalysts in tritium removal from nuclear effluents. Tritium removal from the heavy water reactor and nuclear reprocessing plant, the cleanup of atmosphere and gaseous effluents by hydrogen-oxygen recombination, removal of oxygen dissolved in water are presented and discussed. Unlike the conventional hydrophilic catalysts, the hydrophobic catalysts keep a high catalytic activity and stability, even under the direct contact to liquid water or in presence of saturated humidity. A large diversity of catalyst types (over 100 catalysts) was prepared and tested in order to make them feasible for such processes. The objectives of the review are: - to provide a database for selection of the most appropriate catalyst and catalytic packing for above mentioned processes; - the designing and operation of reactor packed with hydrophobic catalysts; - to evaluate the potentiality of hydrophobic Pt-catalysts in the present and future applications. The most important results are the following: - the hydrophobic Pt-catalysts packed in the trickle bed or separated bed reactors, showed a high catalytic activity and long stability; - the utilization of the hydrophobic Pt-catalysts for the hydrogen isotopes (tritium and deuterium) separation and for hydrogen-oxygen recombination in nuclear field was entirely confirmed on industrial scale; - the improvement of the inner geometry of the reactors and of the composition of mixed catalytic packing as well as the evaluation of performances of separation processes constitute a major contribution of the authors; - the extension of the utilization of the hydrophobic Pt-catalysts in the oxidation of volatile organic compounds from wastewater; - the removal of dissolved oxygen, and deuterium

  5. Development of a simplified treatment for measuring tritium concentration in the environmental water. Removal of dissolved ions by reverse osmosis membrane for electrolysis enrichment

    International Nuclear Information System (INIS)

    Koganezawa, Takayuki; Iida, Takao; Ogata, Yoshimune; Tsuji, Naruhito; Kakiuchi, Masahisa; Satake, Hiroshi; Yamanishi, Hirokuni; Sakuma, Yoichi

    2004-01-01

    An apparatus for tritium enrichment by electrolysis using solid polymer electrolyte was recently developed. The apparatus has the advantage that is to be electrolyzed without adding electrolyte to the sample water. The new treatment both being replaced the distillation process with filtration before electrolysis and being omitted the distillation process after electrolysis, was proposed. Impurities eluted by the electrolysis of ultra pure water with the device introduced no influence on tritium measurement. As alternative treatment to distillation before enrichment, micro filtration and reverse osmosis was carried out. When the sample water treated by micro filtration was electrolyzed, ions adhered both to the electrodes and the solid polymer electrolyte of the device since micro filtration cannot remove ions in the sample water. Therefore, the sample water treated by micro filtration caused some troubles in the electrolysis device. On the other hand, the sample water treated by reverse osmosis did not cause any troubles because it could remove ions. Applying the new treatment to measure some environmental waters, such as river water, resulted in an effective measurement without any influence to liquid scintillation counting. The results proved that a period of the pretreatment process of the water sample could be decreased from about 2 days to about 1.5 hours by applying the proposed treatment. A simplified treatment on the procedure of electrolysis enrichment was established for tritium measurements in the environmental water samples via liquid scintillation counting. (author)

  6. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Y. E-mail: nagao@jmtr.oarai.jaeri.go.jp; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H

    2000-11-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of {sup 6}Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high {sup 6}Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10{sup 13} n cm{sup -2} per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2.

  7. Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JMTR

    International Nuclear Information System (INIS)

    Nagao, Y.; Nakamichi, K.; Tsuchiya, M.; Ishitsuka, E.; Kawamura, H.

    2000-01-01

    To evaluate exactly the total amount of tritium production in tritium breeding materials during in-pile test with JMTR, the 'tritium monitor' has been produced and evaluation of total tritium generation was done by using 'tritium monitor' in preliminary in-pile mock-up, and verification of procedure concerning tritium production evaluation was conducted by using Monte Carlo code MCNP and nuclear cross section library of FSXLIBJ3R2. Li-Al alloy (Li 3.4 wt.%, 95.5% enrichment of 6 Li) was selected as tritium monitor material for the evaluation on the total amount of tritium production in high 6 Li enriched materials. From the results of preliminary experiment, calculated amounts of total tritium production at each 'tritium monitor', which was installed in the preliminary in-pile mock-up, were about 50-290% higher than the measured values. Concerning tritium measurement, increase of measurement error in tritium leak form measuring system to measure small amount of tritium (0.2-0.7 mCi in tritium monitor) was found in the results of present experiment. The tendency for overestimation of calculated thermal neutron flux in the range of 1-6x10 13 n cm -2 per s was found in JMTR and the reason may be due to the beryllium cross section data base in JENDL3.2

  8. Production of ultrapure D-T gas by removal of molecular tritium by selective adsorption

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Hudson, R.S.; Tsugawa, R.T.; Fearon, E.M.; Souers, P.C.; Collins, G.W.

    1991-07-01

    The application of selective adsorption to purification of D-T gas by removal of T 2 has been demonstrated for small quantities of gas typical in research applications. This represents a variation on the production of pure spin isomers of deuterium and hydrogen. The use of an adsorption column offers several advantages over conventional separation techniques, such as low tritium inventory, rapid delivery to prevent radiation damage of the accumulated product, compact size, simplicity of design, construction, and operation, and operation without carrier gas. Because a column can have several thousand equilibrium stages, the purity of the product can be very high. The adsorption column has been shown to be an attractive separation tool for small quantities of hydrogen isotopes

  9. Chemical behaviors of tritium formed in a LiF-BeF2 mixture and its removal from a molten mixture

    International Nuclear Information System (INIS)

    Oishi, J.; Moriyama, H.; Maeda, S.; Ohmura, T.; Moritani, K.

    1987-01-01

    Chemical behaviors of tritium formed in a LiF-BeF 2 mixture were studied using a radiometric method. Most of tritium was found to be present in the T + and T - states under no thermal treatment. The distribution of tritium in chemical states was explained by considering hot atom reactions and radiation chemical reactions. Tritium behaviors in a molten LiF-BeF 2 mixture were also studied at 873 K. In the presence of hydrogen, the isotopic exchange reaction which is TF + H 2 → HT + HF was observed to occur probably in the salt phase. The removal of tritium in a molten LiF-BeF 2 mixture was tried by sparging a gas in a melt for tritium purge, and the effects of the composition of purge gas and of the construction material of crucibles containing the melt on the removal rate were observed. (author)

  10. Dynamic informational system for control and monitoring the tritium removal pilot plant with data transfer and process analyses

    International Nuclear Information System (INIS)

    Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu

    2005-01-01

    The dynamic informational system with datalogging and supervisory control module includes a motion control module and is a new conception used in tritium removal installation with isotopic exchange and cryogenic distillation. The control system includes an event-driven engine that maintains a real-time database, logs historical data, processes alarm information, and communicates with I/O devices. Also, it displays the operator interfaces and performs tasks that are defined for advanced control algorithms, supervisory control, analysis, and display with data transfer from data acquisition room to the control room. By using the parameters, we compute the deuterium and tritium concentration, respectively, of the liquid at the inlet of the isotopic exchange column and, consequently, we can compute at the outlet of the column, the tritium concentration in the water vapors. (authors)

  11. 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

  12. Management of Tritium in ITER Waste

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  13. Darlington tritium removal facility and station upgrading plant dynamic process simulation

    International Nuclear Information System (INIS)

    Busigin, A.; Williams, G. I. D.; Wong, T. C. W.; Kulczynski, D.; Reid, A.

    2008-01-01

    Ontario Power Generation Nuclear (OPGN) has a 4 x 880 MWe CANDU nuclear station at its Darlington Nuclear Div. located in Bowmanville. The station has been operating a Tritium Removal Facility (TRF) and a D 2 O station Upgrading Plant (SUP) since 1989. Both facilities were designed with a Distributed Control System (DCS) and programmable logic controllers (PLC) for process control. This control system was replaced with a DCS only, in 1998. A dynamic plant simulator was developed for the Darlington TRF (DTRF) and the SUP, as part of the computer control system replacement. The simulator was used to test the new software, required to eliminate the PLCs. The simulator is now used for operator training and testing of process control software changes prior to field installation. Dynamic simulation will be essential for the ITER isotope separation system, where the process is more dynamic than the relatively steady-state DTRF process. This paper describes the development and application of the DTRF and SUP dynamic simulator, its benefits, architecture, and the operational experience with the simulator. (authors)

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

    International Nuclear Information System (INIS)

    Sood, S.K.; Fong, C.; Kalyanam; Woodall, K.B.

    1997-01-01

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

  15. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approx. 15 at. % lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility ( 0 C, the extraction process is not attractive

  16. Study of tritium permeation through Peach Bottom Steam Generator tubes

    International Nuclear Information System (INIS)

    Yang, L.; Baugh, W.A.; Baldwin, N.L.

    1977-06-01

    The report describes the equipment developed, samples tested, procedures used, and results obtained in the tritium permeation tests conducted on steam generator tubing samples which were removed from the Peach Bottom Unit No. 1 reactor

  17. Computer based plant display and digital control system of Wolsong NPP Tritium Removal Facility

    International Nuclear Information System (INIS)

    Jung, C.; Smith, B.; Tosello, G.; Grosbois, J. de; Ahn, J.

    2007-01-01

    The Wolsong Tritium Removal Facility (WTRF) is an AECL-designed, first-of-a-kind facility that removes tritium from the heavy water that is used in systems of the CANDUM reactors in operation at the Wolsong Nuclear Power Plant in South Korea. The Plant Display and Control System (PDCS) provides digital plant monitoring and control for the WTRF and offers the advantages of state-of-the-art digital control system technologies for operations and maintenance. The overall features of the PDCS will be described and some of the specific approaches taken on the project to save construction time and costs, to reduce in-service life-cycle costs and to improve quality will be presented. The PDCS consists of two separate computer sub-systems: the Digital Control System (DCS) and the Plant Display System (PDS). The PDS provides the computer-based Human Machine Interface (HMI) for operators, and permits efficient supervisory or device level monitoring and control. A System Maintenance Console (SMC) is included in the PDS for the purpose of software and hardware configuration and on-line maintenance. A Historical Data System (HDS) is also included in the PDS as a data-server that continuously captures and logs process data and events for long-term storage and on-demand selective retrieval. The PDCS of WTRF has been designed and implemented based on an off-the-self PDS/DCS product combination, the Delta-V System from Emerson. The design includes fully redundant Ethernet network communications, controllers, power supplies and redundancy on selected I/O modules. The DCS provides field bus communications to interface with 3rd party controllers supplied on specialized skids, and supports HART communication with field transmitters. The DCS control logic was configured using a modular and graphical approach. The control strategies are primarily device control modules implemented as autonomous control loops, and implemented using IEC 61131-3 Function Block Diagram (FBD) and Structured

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. Comparison of hand hygiene procedures for removing Bacillus cereus spores.

    Science.gov (United States)

    Sasahara, Teppei; Hayashi, Shunji; Hosoda, Kouichi; Morisawa, Yuji; Hirai, Yoshikazu

    2014-01-01

    Bacillus cereus is a spore-forming bacterium. B. cereus occasionally causes nosocomial infections, in which hand contamination with the spores plays an important role. Therefore, hand hygiene is the most important practice for controlling nosocomial B. cereus infections. This study aimed to determine the appropriate hand hygiene procedure for removing B. cereus spores. Thirty volunteers' hands were experimentally contaminated with B. cereus spores, after which they performed 6 different hand hygiene procedures. We compared the efficacy of the procedures in removing the spores from hands. The alcohol-based hand-rubbing procedures scarcely removed them. The soap washing procedures reduced the number of spores by more than 2 log10. Extending the washing time increased the spore-removing efficacy of the washing procedures. There was no significant difference in efficacy between the use of plain soap and antiseptic soap. Handwashing with soap is appropriate for removing B. cereus spores from hands. Alcohol-based hand-rubbing is not effective.

  20. In-vessel tritium retention and removal in ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Brooks, J.N. [Argonne National Lab., IL (United States); Iseli, M. [ITER Naka Joint Work Site, Naka-gun (Japan); Wu, C.H. [EFDA Close Support Unit, Garching (Germany)

    2001-07-01

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., {proportional_to}350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

  1. In-Vessel Tritium Retention and Removal in ITER-FEAT

    Science.gov (United States)

    Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

  2. In-vessel tritium retention and removal in ITER-FEAT

    International Nuclear Information System (INIS)

    Federici, G.; Brooks, J.N.; Iseli, M.; Wu, C.H.

    2001-01-01

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ∝350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

  3. Gas separation performance of a hollow-filament type polyimide membrane module for a compact tritium removal system

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Yamada, Masayuki; Suzuki, Takumi; Matsuda, Yuji; Okuno, Kenji

    1995-01-01

    A new tritium removal system using gas separation membranes has been studied to develop more compact and cost-effective system for a fusion reactor. To obtain necessary parameters, which are directly scalable to the ITER Atmospheric Detritiation System, the basic tritium recovery performance was investigated with a scaled polyimide membrane module (hollow-filament type : 10 m 3 /hr) loop. The result shows that the H 2 recovery ratio from N 2 or air was more than 99% or about 97%, respectively, at flow rate ratio of permeated/feed = 0.1, feed ampersand permeated side pressures = 2580 ampersand 80 torr, and module temp. = 293 K. Tritium (HT) recovery function was almost the same as H 2 recovery, even though the total hydrogen concentration was a few ppm in the feed of module. H 2 O recovery performance was better than hydrogen recovery. These recovery functions were improved effectively decreasing the pressure ratio of permeated/feed of module. 5 refs., 11 figs

  4. Study of tritium removal from fusion reactor blankets of molten salt and lithium--aluminum

    International Nuclear Information System (INIS)

    Talbot, J.B.

    1976-03-01

    The sorption of tritium by molten lithium--bismuth (Li--Bi, approximately 15 at. percent lithium) and solid equiatomic lithium--aluminum (Li--Al) was investigated experimentally to evaluate the potential applications of both materials in a controlled thermonuclear reactor. The Li--Bi alloy was proposed to countercurrently extract tritium from a molten salt (Li 2 BeF 4 ) blanket. However, because of the low solubility (less than 10 ppb) at temperatures ranging from 500 to 700 0 C, the extraction process is not attractive

  5. Thermal Removal of Tritium from Concrete and Soil to Reduce Groundwater Impacts - 13197

    International Nuclear Information System (INIS)

    Jackson, Dennis G.; Blount, Gerald C.; Wells, Leslie H.; Cardoso, Joao E.; Kmetz, Thomas F.; Reed, Misty L.

    2013-01-01

    Legacy heavy-water moderator operations at the Savannah River Site (SRS) have resulted in the contamination of equipment pads, building slabs, and surrounding soil with tritium. At the time of discovery the tritium had impacted the shallow ( 3 (1,650-yd 3 ) of contaminated concrete and soils were treated with an actual incurred cost of $3,980,000. This represents a unit treatment cost of $3,156/m 3 ($2,412/yd 3 ). In 2011 the project was recognized with an e-Star Sustainability Award by DOE's Office of Environmental Management. (authors)

  6. Experiences with decontaminating tritium-handling apparatus

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

    1992-01-01

    Tritium-handling apparatus has been decontaminated as part of the downsizing of the LLNL Tritium Facility. Two stainless-steel glove boxes that had been used to process lithium deuteride-tritide (LiDT) slat were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. In this paper the details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium, in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given

  7. Development of a wetproofed catalyst recombiner for removal of airborne tritium

    International Nuclear Information System (INIS)

    Chuang, K.T.; Quaiattini, R.J.; Thatcher, D.R.P.; Puissant, L.J.

    1985-01-01

    For cleanup of airborne tritium at tritium handling facilities, it is generally agreed that the most reliable method is to convert the tritium in a recombiner into water vapor followed by adsorption of the vapor in a molecular sieve drier. Decontamination factors of 10 3 to 10 6 have been reported. Wetproofed catalysts developed at Chalk River Nuclear Laboratories have been shown to maintain their activities when exposed to liquid water or air at 100% relative humidity. When a wetproofed catalyst recombiner is used, operation can be carried out at room temperatures thus greatly simplifying the system. Two catalysts, Pt/carbon and Pt/silica, were prepared for this study. The activity of Pt/carbon was measured with hydrogen and found to be comparable to the published results for conventional Pt/alumina catalysts at similar conditions. Experiments were carried out for the following range of operating conditions: flows from 0.3 to 3.0 m/s, pressure from 100 to 500 kPa. Tritium was added to the air stream at 1-5 MBq.m -3 (30-140 μCi.m -3 ). No significant isotope and/or pressure effects were observed. To date lifetime data of greater than four months have been obtained

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

    International Nuclear Information System (INIS)

    Sherwood, A.E.

    1976-06-01

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

  9. Procedure for the preparation of tritium-labelled insulins. Verfahren zur Herstellung von Tritium-markierten Insulinen

    Energy Technology Data Exchange (ETDEWEB)

    Bienert, M; Haensicke, A; Beyermann, M; Kaufmann, K D; Oehlke, J; Klauschenz, E; Bespalowa, S; Titov, M; Pleiss, U

    1986-12-17

    This invention is concerned with a procedure for the preparation of specific /sup 3/H-labelled insulins with sequences of human, bovine or porcine insulins and without simultaneous chemical modifications of the insulin. On the basis of this procedure a /sup 3/H/sub 2/-Typ (B26)-insulin can be obtained in good yield and purity with a specific radioactivity appropriate to biopharmaceutical and pharmacokinetic purposes in medicine and pharmaceutical industry, resp.

  10. 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

  11. 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

  12. 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)

  13. Operability test procedure [Tank] 241-SY-101 equipment removal system

    International Nuclear Information System (INIS)

    Mast, J.C.

    1994-01-01

    The 241-SY-101 equipment removal system (ERS) consists of components, equipment, instrumentation and procedures that will provide the means to disconnect, retrieve, contain, load and transport the Mitigation Pump Assembly (MPA) from waste Tank 241-SY-101 to the Central Waste Complex (CWC). The Operability Test Procedure (OTP) will test the interfaces between ERS components and will rehearse the procedure for MPA removal and transportation to the extent they can be mocked-up at the CTF (Cold Test Facility). At the conclusion of the OTP, the ERS components and equipment will be removed from the CTF, entered into the Component Based Recall System (CBRS), and stored until needed for actual MPA removal and transportation

  14. Possibilities of tritium removal from waste waters of pressurized water reactors and fuel reprocessing plants

    International Nuclear Information System (INIS)

    Ribnikar, S.V.; Pupezin, J.D.

    1975-01-01

    Starting from parameters known for heavy water production processes, a parallel was made with separation of tritium from water. The quantity in common is the total cascade flow. The most efficient processes appear to be hydrogen sulfide, water exchange, hydrogen- and water distillation. Prospects of application of new processes are discussed briefly. Problems concerning detritiation of pressurized water reactors and large fuel reprocessing plants are analyzed. Detritiation of the former should not present problems. With the latter, economical detritiation can be achieved only after some plant flow patterns are changed. (U.S.)

  15. Milestone Report - M4FT-17OR030107025 - Design of a tritium and iodine removal system for use with advanced TPOG

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bruffey, Stephanie H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jordan, Jacob A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jubin, Robert Thomas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    US regulations will require the removal of iodine and tritium, along with other volatile and semi-volatile radionuclides, from the off-gas streams of nuclear fuel reprocessing plants. Advanced tritium pretreatment (TPT) is an additional head-end operation that could be incorporated within nuclear fuel reprocessing plants. It utilizes nitrogen dioxide (NOR2R) as an oxidant to convert UOR2R to UR3ROR8R prior to traditional aqueous dissolution. Advanced TPT can result in the quantitative volatilization of both tritium and iodine. Up-front removal of iodine is of significant advantage because otherwise it distributes to several unit operations and the associated off-gas streams. The off-gas streams will then require treatment to comply with US regulations. Advanced TPT is currently under development at Oak Ridge National Laboratory, and a kilogram-scale hot cell demonstration with used nuclear fuel (UNF) is planned for fiscal year (FY) 2018.

  16. 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

  17. Tritium derivatives of the glycyrrhetinic acid and procedure for its preparation

    International Nuclear Information System (INIS)

    Turner, J.C.

    1977-01-01

    The invention concerns tritium derivatives of glycyrrhetinic acid which is largely used to treat ulcers and inflammations, and it deals with a method for their production. The 3α- 3 H-glycyrrhetinic acid, 3 α- 3 H-carbene oxolone, Na-salt and basic Al salt of this carbene oxolone, as well as the acetyl derivates, piperazine amide derivatives and further derivatives of the glycyrrhetinic acid (e.g. cinnamyl ester) are claimed in nine examples. (HK) [de

  18. Establishing a routine procedure for determination of environmental tritium concentration at ICIT

    International Nuclear Information System (INIS)

    Varlam, C.; Stefanescu, I.; Faurescu, I.; Vagner, I.; Faurescu, D.; Bogdan, D.

    2009-01-01

    Full text: The Cryogenic Pilot is an experimental project within the national nuclear energy research program, which has the aim of developing technologies for tritium and deuterium separation by cryogenic distillation. The process used in this installation is based on a combined method for liquid-phase catalytic exchange (LPCE) and cryogenic distillation. There are two ways in which the Cryogenic Pilot can interact with the environment: by atmospheric release and through the sewage system. 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: spinach, spring wheat, onion, hay, grass, apple, garden lettuce, soil, milk, and meat. For the azeotropic distillation of all types of samples were used two solvents, toluene and cyclohexane, and all measurements for the determination of environmental tritium concentration were carried out using liquid scintillation counting (LSC), with ultra-low liquid scintillation spectrometer Quantulus 1220 specially designed for environmental samples and low radioactivity. Sample scintillation cocktail ratio was 8:12 ml and liquid scintillation cocktail was UltimaGold LLT. The background determined for the prepared blank samples was between 0.926 CPM and 1.002 CPM and the counting efficiency between 25.37% and 26.10%. The counting time was 1000 minutes (50 minutes/20 cycles) for each sample, and the minimum detectable activity according to ISO 9698 was 8.9 TU and 9.05 TU, respectively, at a confidence factor of 3. (authors)

  19. High Heat Flux Interactions and Tritium Removal from Plasma Facing Components by a Scanning Laser

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.A.; Hassanein, A.

    2002-01-01

    A new technique for studying high heat flux interactions with plasma facing components is presented. The beam from a continuous wave 300 W neodymium laser was focused to 80 W/mm2 and scanned at high speed over the surface of carbon tiles. These tiles were previously used in the TFTR [Tokamak Fusion Test Reactor] inner limiter and have a surface layer of amorphous hydrogenated carbon that was codeposited during plasma operations. Laser scanning released up to 84% of the codeposited tritium. The temperature rise of the codeposit on the tiles was significantly higher than that of the manufactured material. In one experiment, the codeposit surface temperature rose to 1,770 C while for the same conditions, the manufactured surface increased to only 1,080 C. The peak temperature did not follow the usual square-root dependence on heat pulse duration. Durations of order 100 ms resulted in brittle destruction and material loss from the surface, while a duration of approximately 10 ms showed minimal change. A digital microscope imaged the codeposit before, during, and after the interaction with the laser and revealed hot spots on a 100-micron scale. These results will be compared to analytic modeling and are relevant to the response of plasma facing components to disruptions and vertical displacement events (VDEs) in next-step magnetic fusion devices

  20. Procedural Sedation for the removal of a rectal foreign body

    Directory of Open Access Journals (Sweden)

    John Costumbrado, MD, MPH

    2018-04-01

    Full Text Available History of present illness: A 40-year-old male with a history of intravenous drug use presented to the emergency department (ED for one week of constant lower abdominal pain associated with bloody stool. He denied fever, nausea, vomiting, urinary symptoms, and testicular pain or swelling. On exam, vital signs were within normal limits. Abdominal exam was non-tender without rebound or guarding. Rectal exam was negative for occult blood but positive for a palpable firm, blunt object. A computed tomography (CT of the abdomen and pelvis was ordered to further investigate. Significant findings: Axial and coronal views on CT showed evidence of a large, tube-shaped foreign body in the rectum (see arrows without evidence of acute gastrointestinal tract disease. Discussion: While rectal foreign bodies (RFB are not uncommon to the ED, accurate epidemiological estimates are not available, due in part to underreporting.1 One study estimated an incidence of one patient per month that needed care for a RFB.2 Generally, patients can remove the object themselves; however, 20% of cases require endoscopic intervention and 1% require surgical intervention.3 RFBs can be removed via the transanal approach manually, instrument-assisted (eg, Kocher clamp, obstetric forceps, or endoscopically. In cases without intestinal perforation, transanal removal of a RFB is generally attempted as a first-line procedure in the ED, with an approximate success rate of 75%.4 However, the limitations of transanal removal depends on the location of the object, level of anal relaxation, and ability to grasp the object, which may be limited by the provider’s hand size and availability of instruments.Ways of facilitating bedside removal of RFBs include the Valsalva maneuver with proper positioning (i.e., lithotomy or prone knee-to-chest position or manual abdominal wall compression to help move the object closer to the anal orifice. Anoscopy may also be used to improve visualization of

  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. 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

  3. Validation of tritium measurements in biological materials

    International Nuclear Information System (INIS)

    Kim, M.A.; Baumgartner, F.

    1988-01-01

    The maximum deviation of experimental R value from its real value, which is defined as the ratio of tissue bound to tissue water tritium, has been calculated and verified experimentally by taking consideration of isotopic fractionation arised in the course of water separation. Experimental procedures examined for the purpose are the azeotropic distillation and lyophilization for the removal of tissue water and the oxidative combustion of organic residue either by thermal process or by low temperature plasma generation. Each procedure optimalized by obviating or correcting isotope effects as well as other sources of error has been tested with mixed standards and biological samples. By washing out the exchangeable tritium and also physically bound tritium, the precision and accuracy of R values are further improved

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-20

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

  6. 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

  7. 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

  8. Monitoring system in Labview data logging and supervisory control module and Labview 8 with process analyses of the Cryogenic Pilot Plant for Tritium Removal

    International Nuclear Information System (INIS)

    Moraru, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Bucur, Ciprian; Stefan, Liviu; Bornea, Anisia; Stefanescu, Ioan

    2008-01-01

    The system responds to the monitoring requirements of the technological processes specific to the nuclear installation that processes radioactive substances, with severe consequences in case of technological failure, as is the case with a tritium processing nuclear plant. The big amount of data that needs to be processed, stored and accessed for real time simulation and optimization demands the achievement of the virtual technologic platform where the data acquiring, control and analysis systems of the technological process can be integrated with an advanced technological monitoring system. Thus, integrated computing and monitoring systems needed for the supervising of the technological process will be effected, to be then continued with optimization of the system, by implementing new and performing methods corresponding to the technological processes within the tritium removal processing nuclear plants. (authors)

  9. 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.

  10. 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.

  11. Procedures for Removal of Pesticides from the Environment

    Directory of Open Access Journals (Sweden)

    Đokić, M.

    2012-07-01

    Full Text Available Pesticides are widely used in food production, and it is believed that more than 1000 types of pesticides are in use. Organochlorines and organophosphorous pesticides are used in large quantities due to their efficacy and low cost. These persistent organic pollutants remain in the soil, silt, and sediment long after application, and enter into watercourses, finding their way directly into the food chain. Today, the following procedures are used to remove pesticides from polluted localities: low temperature thermal desorption, incineration, bioremediation and phytoremediation. Each of these technologies has its advantages and disadvantages. Ultimately, the ideal remediation process would be to completely destroy the pollutant without the production of byproducts. Some of these processes are only capable of moving and stabilizing the contaminant, but do not clean or fully eliminate the pollutant. Low temperature thermal desorption is an ex situ cleaning technology most often used to remove pesticides. The advantage of incineration is the complete elimination of the pollutant; however, this process is very expensive, as it requires transport of the media to the incinerator. The processes of bioremediation are stimulated by natural processes of microbiological degradation of contaminants through the interaction of microorganisms and nutrients from waste. Ex situ procedures include the use of bioreactors. Another procedure is land spreading, in which the contaminated medium is mixed with the soil, in which the native microorganisms degrade the pollutants. Efficacy in the biodegradation of toxic pollutants has been established for white root fungi, particularly those from the genus Phanerochaete. Phytoremediation is a green technology that uses plants, in which plants are not directly included in the process but serve as a catalyser for increasing the growth and activity of root microorga- nisms. In phytoremediation of pesticides, plants of the genus

  12. Experiences with decontaminating tritium-handling apparatus

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

    1991-07-01

    Tritium-handling apparatus has been decontaminated as part of the shutdown of the LLNL Tritium Facility. Two stainless-steel gloveboxes that had been used to process lithium deuteride-tritide (LiDT) salt were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. Further surface decontamination was performed by scrubbing the interior with paper towels and ethyl alcohol or Swish trademark. The surface contamination, as shown by swipe surveys, was reduced from 4x10 4 --10 6 disintegrations per minute (dpm)/cm 2 to 2x10 2 --4x10 4 dpm/cm 2 . Details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given

  13. 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

  14. Tritium in the DIII-D carbon tiles

    International Nuclear Information System (INIS)

    Taylor, P.L.; Kellman, A.G.; Lee, R.L.

    1993-06-01

    The amount of tritium in the carbon tiles used as a first wall in the DIII-D tokamak was measured recently when the tiles were removed and cleaned. The measurements were made as part of the task of developing the appropriate safety procedures for processing of the tiles. The surface tritium concentration on the carbon tiles was surveyed and the total tritium released from tile samples was measured in test bakes. The total tritium in all the carbon tiles at the time the tiles were removed for cleaning is estimated to be 15 mCi and the fraction of tritium retained in the tiles from DIII-D operations has a lower bound of 10%. The tritium was found to be concentrated in a narrow surface layer on the plasma facing side of the tile, was fully released when baked to 1,000 degree C, and was released in the form of tritiated gas (DT) as opposed to tritiated water (DTO) when baked

  15. 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.)

  16. 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.)

  17. 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

  18. 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.

  19. 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.)

  20. 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)

  1. 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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Carlson, R.V.

    1989-01-01

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

  4. 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

  5. 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.

  6. 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)

  7. 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.)

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. Pollutants in drinking water - sources, harmful effects and removal procedures

    International Nuclear Information System (INIS)

    Qadeer, R.

    2005-01-01

    The underground water resources available for human consumption are being continuously contaminated by the natural sources and anthropogenic activities. The pollutants include toxic microorganism, inorganic and organic chemicals and radionuclide etc. This is an acute problem in our country, where free style way of disposal of industrial effluents into the natural water bodies contaminates the surface and ground water. These contaminants make their way into human body through contaminated drinking water, which leads to the malfunctioning of the body organs. Details of some pollutants present in drinking water, their source and harmful effects on human beings are reviewed in this communication Merits and demerits of methods used to remove the pollutants from drinking water are also discussed. (author)

  16. 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

  17. 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)

  18. 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)

  19. The removal of metals from edible oil by a membrane extraction procedure 355

    NARCIS (Netherlands)

    Keurentjes, J.T.F.; Bosklopper, T.G.J.; Dorp, van L.J.; Riet, van 't K.

    1990-01-01

    Edible oils may contain traces of metals. In oil refining procedures these metals have to be removed to guarantee oxidatively stable products. In this study we present a hollow fiber membrane extraction system for the removal of metals from an oil. Several extraction liquids were tested, of which an

  20. 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.))

  1. Tritium monitor with improved gamma-ray discrimination

    Science.gov (United States)

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

    1985-01-01

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

  2. 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)

  3. Design, Fabrication, and Testing of a Laboratory-Scale Voloxidation System for Removal of Tritium and Other Volatile Fission Products from Used Nuclear Fuel

    International Nuclear Information System (INIS)

    Spencer, Barry B; DelCul, Guillermo D; Bradley, Eric Craig; Jubin, Robert Thomas; Hylton, Tom D; Collins, Emory D

    2008-01-01

    Advanced nuclear fuel processing methodologies are being demonstrated at the Oak Ridge National Laboratory (ORNL) as part of the Global Nuclear Energy Partnership (GNEP) program. A coupled end-to-end (CETE) research and development (R and D) capability is being installed to provide all primary processing operations, ranging from spent fuel receipt to production of products and waste forms. This R and D capability is designed for small, laboratory-scale throughput and will permit conduct of experiments in the range of 20 kg of spent fuel per year. The head-end processing segment includes single-pin shearing, voloxidation to remove tritium from the fuel before it enters the aqueous based separations systems, cleanup of the cladding hulls for disposition, and transfer of the fuel powder to the dissolution process. This paper describes the voloxidation system design and presents results from the cold checkout of the hardware. Preliminary results of the initial processing campaign with spent fuel is presented as well

  4. 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)

  5. Comparative efficiency of final endodontic cleansing procedures in removing a radioactive albumin from root canal systems

    International Nuclear Information System (INIS)

    Cecic, P.A.; Peters, D.D.; Grower, M.F.

    1984-01-01

    Fifty-six teeth were initially instrumented, with the use of seven irrigants or irrigant combinations, and filled with radioactive albumin. The study then showed the relative ability of three final endodontic procedures (copious reirrigation with saline solution, drying with paper points, and reassuring patency of the canal with the final instrument) to remove the albumin. Even after copious irrigation, each additional procedure removed statistically significant amounts of albumin. Alternating an organic solvent and an inorganic solvent did appear to leave the canal system in the optimal condition for final cleansing procedures. The study then correlated the relative efficiency of irrigation alone versus instrumentation plus irrigation in removing the remaining albumin from the canal systems. Reinstrumentation plus copious irrigation removed significantly more albumin than copious irrigation alone

  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 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)

  8. Design options to minimize tritium inventories at Savannah River

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  9. Design options to minimize tritium inventories at Savannah River

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. Clinical study of treatment of cerebral hemorrhage: remove the intracranial hematoma with a minimal invasive procedure

    International Nuclear Information System (INIS)

    Ke Dongfeng; He Yunguang; Hu Wen; Lin Yang; Yang Danyang; Chen Shaokai; Ma Shaobing

    2004-01-01

    Objective: To study the feasibility and factors of minimal invasive intracranial hematoma removing procedure as a treatment of cerebral hemorrhage. Methods: From May, 2000 to September, 2003, 33 patients with intracerebral hemorrhage underwent minimal invasive intracranial hematoma removing procedure and from May, 1997 to September, 2000, 27 patients with cerebral hemorrhage received conservative treatments. Two groups were compared and analyzed. The quantity of hemorrhage and the indication of procedure were also studied. Results: State of an illness has no significant difference between the two groups (P<0.05). The rate of recovery were higher in the group undergoing the procedure (57.6%) than in the control group (14.8%) (P<0.05). The rate of handicap were lower in the procedure group (24.0%) than in the control group (60.0%) (P<0.05). The mortality were also lower in the procedure group (24.2%) than in the control group (63.0%) (P<0.01). In the control group no patient with a hematoma larger than 70 ml survived. In the procedure group patients with hematoma larger than 70 ml had less chance of survival than the other patients (P<0.01). The mortality rate were respectively 50%, 5.6%, 33.3% when the procedure was done in super early, early, delayed stage. The mortality rate was higher in the super early stage than in early stage (P<0.05). Conclusion: The minimal invasive intracranial hematoma removing procedure has a better clinical outcome than the conservative treatment. The procedure reduces obviously mortality rate and increase the quality of survival. Multiple puncturing and draining or craniotomy are recommended to remove huge hematoma. The earlier treatment brings better clinical effects. This technique is simple, less invasive and provides good clinical outcome, which is worth recommendation

  11. Test procedures and instructions for Hanford tank waste supernatant cesium removal

    Energy Technology Data Exchange (ETDEWEB)

    Hendrickson, D.W., Westinghouse Hanford

    1996-05-31

    This document provides specific test procedures and instructions to implement the test plan for the preparation and conduct of a cesium removal test using Hanford Double-Shell Slurry Feed supernatant liquor from tank 251-AW-101 in a bench-scale column.Cesium sorbents to be tested include resorcinol-formaldehyde resin and crystalline silicotitanate. The test plan for which this provides instructions is WHC-SD-RE-TP-022, Hanford Tank Waste Supernatant Cesium Removal Test Plan.

  12. Test procedures and instructions for Hanford complexant concentrate supernatant cesium removal using CST

    Energy Technology Data Exchange (ETDEWEB)

    Hendrickson, D.W.

    1997-01-08

    This document provides specific test procedures and instructions to implement the test plan for the preparation and conduct of a cesium removal test, using Hanford Complexant Concentrate supernatant liquor from tank 241-AN-107, in a bench-scale column. The cesium sorbent to be tested is crystalline silicotitanate. The test plan for which this provides instructions is WHC-SD-RE-TP-023, Hanford Complexant Concentrate Supernatant Cesium Removal Test Plan.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  14. Computer aided planning of orthopaedic surgeries: the definition of generic planning steps for bone removal procedures.

    Science.gov (United States)

    Putzer, David; Moctezuma, Jose Luis; Nogler, Michael

    2017-11-01

    An increasing number of orthopaedic surgeons are using computer aided planning tools for bone removal applications. The aim of the study was to consolidate a set of generic functions to be used for a 3D computer assisted planning or simulation. A limited subset of 30 surgical procedures was analyzed and verified in 243 surgical procedures of a surgical atlas. Fourteen generic functions to be used in 3D computer assisted planning and simulations were extracted. Our results showed that the average procedure comprises 14 ± 10 (SD) steps with ten different generic planning steps and four generic bone removal steps. In conclusion, the study shows that with a limited number of 14 planning functions it is possible to perform 243 surgical procedures out of Campbell's Operative Orthopedics atlas. The results may be used as a basis for versatile generic intraoperative planning software.

  15. 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)

  16. HYLIFE-II tritium management system

    International Nuclear Information System (INIS)

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

    1993-06-01

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

  17. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  18. 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

  19. Impression Procedures for Metal Frame Removable Partial Dentures as Applied by General Dental Practitioners.

    NARCIS (Netherlands)

    Fokkinga, W.A.; Uchelen, J. van; Witter, D.J.; Mulder, J.; Creugers, N.H.J.

    2016-01-01

    This pilot study analyzed impression procedures for conventional metal frame removable partial dentures (RPDs). Heads of RPD departments of three dental laboratories were asked to record features of all incoming impressions for RPDs during a 2-month period. Records included: (1) impression

  20. Intra-operative removal of chest tube in video-assisted thoracoscopic procedures

    Directory of Open Access Journals (Sweden)

    Moustafa M. El-Badry

    2017-12-01

    Conclusions: Intra-operative removal of chest tube during VATS procedures was a safe technique in well selected patients with an intra-operative successful air-leak test with radiological and clinical follow-up. This technique provided lesser post-operative pain with shorter hospital stay.

  1. Disposal of tritium-exposed metal hydrides

    International Nuclear Information System (INIS)

    Nobile, A.; Motyka, T.

    1991-01-01

    A plan has been established for disposal of tritium-exposed metal hydrides used in Savannah River Site (SRS) tritium production or Materials Test Facility (MTF) R ampersand D operations. The recommended plan assumes that the first tritium-exposed metal hydrides will be disposed of after startup of the Solid Waste Disposal Facility (SWDF) Expansion Project in 1992, and thus the plan is consistent with the new disposal requiremkents that will be in effect for the SWDF Expansion Project. Process beds containing tritium-exposed metal hydride powder will be disposed of without removal of the powder from the bed; however, disposal of tritium-exposed metal hydride powder that has been removed from its process vessel is also addressed

  2. A simplified test procedure for determining the effectiveness of adsorbents for the removal of methyl iodide

    International Nuclear Information System (INIS)

    Underhill, D.W.

    1993-01-01

    ASTM Test Procedure D3803 measures the ability of nuclear-grade carbon to remove methyl iodide from a stream of humidified air. This test, unlike all the other procedures developed by ASTM Committee D28, has evolved to become extremely complex. The intricacy of this test as well as the great difficulty in obtaining inter-laboratory agreement, creates doubt as tot the actual meaning of the results. Here a far simpler test system is described in which thermodynamic principles are used to maintain a constant, reproducible test procedure. This paper describes a system implementing these elements, its cost to build, and the factors affecting its accuracy. 11 refs., 1 fig

  3. 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.)

  4. Tritium Mitigation/Control for Advanced Reactor System

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-03-31

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

  5. Tritium Systems Test Facility. Volume II. Appendixes

    International Nuclear Information System (INIS)

    Anderson, G.W.; Battleson, K.W.; Bauer, W.

    1976-10-01

    This document includes the following appendices: (1) vacuum pumping, (2) tritium migration into the power cycle, (3) separation of hydrogen isotopes, (4) tritium research laboratory, (5) TSTF containment and cleanup, (6) instrumentation and control, (7) gas heating in torus, and (8) TSTF fuel loop operating procedures

  6. 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.

  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. 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

  9. An Efficient Procedure for Removal and Inactivation of Alpha-Synuclein Assemblies from Laboratory Materials.

    Science.gov (United States)

    Bousset, Luc; Brundin, Patrik; Böckmann, Anja; Meier, Beat; Melki, Ronald

    2016-01-01

    Preformed α-synuclein fibrils seed the aggregation of soluble α-synuclein in cultured cells and in vivo. This, and other findings, has kindled the idea that α-synuclein fibrils possess prion-like properties. As α-synuclein fibrils should not be considered as innocuous, there is a need for decontamination and inactivation procedures for laboratory benches and non-disposable laboratory material. We assessed the effectiveness of different procedures designed to disassemble α-synuclein fibrils and reduce their infectivity. We examined different commercially available detergents to remove α-synuclein assemblies adsorbed on materials that are not disposable and that are most found in laboratories (e.g. plastic, glass, aluminum or stainless steel surfaces). We show that methods designed to decrease PrP prion infectivity neither effectively remove α-synuclein assemblies adsorbed to different materials commonly used in the laboratory nor disassemble the fibrillar form of the protein with efficiency. In contrast, both commercial detergents and SDS detached α-synuclein assemblies from contaminated surfaces and disassembled the fibrils. We describe three cleaning procedures that effectively remove and disassemble α-synuclein seeds. The methods rely on the use of detergents that are compatible with most non-disposable tools in a laboratory. The procedures are easy to implement and significantly decrease any potential risks associated to handling α-synuclein assemblies.

  10. Conceptual design of tritium treatment facility

    International Nuclear Information System (INIS)

    Tachikawa, Katsuhiro

    1982-01-01

    In connection with the development of fusion reactors, the development of techniques concerning tritium fuel cycle, such as the refining and circulation of fuel, the recovery of tritium from blanket, waste treatment and safe handling, is necessary. In Japan Atomic Energy Research Institute, the design of the tritium process research laboratory has been performed since fiscal 1977, in which the following research is carried out: 1) development of hydrogen isotope separation techniques by deep cooling distillation method and thermal diffusion method, 2) development of the refining, collection and storage techniques for tritium using metallic getters and palladium-silver alloy films, and 3) development of the safe handling techniques for tritium. The design features of this facility are explained, and the design standard for radiation protection is shown. At present, in the detailed design stage, the containment of tritium and safety analysis are studied. The building is of reinforced concrete, and the size is 48 m x 26 m. Glove boxes and various tritium-removing facilities are installed in two operation rooms. Multiple wall containment system and tritium-removing facilities are explained. (Kako, I.)

  11. Test procedures and instructions for single shell tank saltcake cesium removal with crystalline silicotitanate

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, J.B.

    1997-01-07

    This document provides specific test procedures and instructions to implement the test plan for the preparation and conduct of a cesium removal test, using Hanford Single Shell Tank Saltcake from tanks 24 t -BY- I 10, 24 1 -U- 108, 24 1 -U- 109, 24 1 -A- I 0 1, and 24 t - S-102, in a bench-scale column. The cesium sorbent to be tested is crystalline siticotitanate. The test plan for which this provides instructions is WHC-SD-RE-TP-024, Hanford Single Shell Tank Saltcake Cesium Removal Test Plan.

  12. Procedure to remove a dirt and/or oil film from water

    Energy Technology Data Exchange (ETDEWEB)

    Jager, T; Jager, G P.A.; Jager, K L.E.

    1970-12-11

    A procedure is described to remove dirt and/or oil films from a water surface. A number of rotating wiper scoops moves through the water. The top of the polluted water is brought into motion by the scoops and directed to a gutter system where it is removed. The advantage of the system is that the wiper scoops can be lowered selectively to the depth of the pollutant, thereby avoiding moving large quantities of unnecessary unpolluted liquid which later has to be separated. (12 claims)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  14. Quick management of accidental tritium exposure cases

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  15. Quick management of accidental tritium exposure cases.

    Science.gov (United States)

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

    2012-07-01

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

  16. 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

  17. 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

  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. 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

  20. Developments in data acquisition systems with LabView datalogging and supervisory control module for tritium removal plant, with data base and process analysis

    International Nuclear Information System (INIS)

    Moraru, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu; Bucur, Ciprian; Hartescu, Florin

    2006-01-01

    Full text: The implementation of the new trends for tritium processing nuclear plants, and especially those with an experimental character or of new technology development, shows a very high complexity due to issues raised by the integration of a high diversity of instrumentation and equipment into a unitary control system of the technological process. Keeping the system's flexibility is a demand of the experimental plants for which the change of configuration, process and parameters is something usual. The big amount of data that needs to be monitored, stored and accessed for ulterior analyses demands the achievement of an information network where the data acquiring, control and analysis systems of the technological process can be integrated with a data base system. Thus, integrated computing and control systems needed for the control of the technological process will be executed, to be continued with the execution of failure protection system, by choosing methods corresponding to the technological processes within the tritium processing nuclear plants. (authors)

  1. Impression Procedures for Metal Frame Removable Partial Dentures as Applied by General Dental Practitioners.

    Science.gov (United States)

    Fokkinga, Wietske A; van Uchelen, Judith; Witter, Dick J; Mulder, Jan; Creugers, Nico H J

    2016-01-01

    This pilot study analyzed impression procedures for conventional metal frame removable partial dentures (RPDs). Heads of RPD departments of three dental laboratories were asked to record features of all incoming impressions for RPDs during a 2-month period. Records included: (1) impression procedure, tray type (stock/custom), impression material (elastomer/alginate), use of border-molding material (yes/no); and (2) RPD type requested (distal-extension/tooth-bounded/combination). Of the 132 total RPD impressions, 111 (84%) involved custom trays, of which 73 (55%) were combined with an elastomer. Impression border-molding material was used in 4% of the cases. Associations between impression procedure and RPD type or dentists' year/university of graduation were not found.

  2. Surgical removal of a large vaginal calculus formed after a tension-free vaginal tape procedure.

    Science.gov (United States)

    Zilberlicht, Ariel; Feiner, Benjamin; Haya, Nir; Auslender, Ron; Abramov, Yoram

    2016-11-01

    Vaginal calculus is a rare disorder which has been reported in association with urethral diverticulum, urogenital sinus anomaly, bladder exstrophy and the tension-free vaginal tape (TVT) procedure. We report a 42-year-old woman who presented with persistent, intractable urinary tract infection (UTI) following a TVT procedure. Cystoscopy demonstrated an eroded tape with the formation of a bladder calculus, and the patient underwent laser cystolithotripsy and cystoscopic resection of the tape. Following this procedure, her UTI completely resolved and she remained asymptomatic for several years. Seven years later she presented with a solid vaginal mass. Pelvic examination followed by transvaginal ultrasonography and magnetic resonance imaging demonstrated a large vaginal calculus located at the lower third of the anterior vaginal wall adjacent to the bladder neck. This video presents the transvaginal excision and removal of the vaginal calculus.

  3. Tritium handling experience at Atomic Energy of Canada Limited

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  4. DOE handbook: Tritium handling and safe storage

    International Nuclear Information System (INIS)

    1999-03-01

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

  5. DOE handbook: Tritium handling and safe storage

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

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

  6. Methods of tritium recovery from molten lithium

    International Nuclear Information System (INIS)

    Farookhi, R.; Rogers, J.E.

    1968-01-01

    It is important to keep the tritium inventory in a blanket of a thermonuclear reactor at a low level both to eliminate possible hydriding of structural components and to reduce inventory cost. Removing the tritium from a lithium blanket by fractional distillation, flash vaporization, and fractional crystallization was investigated. No definitive data are available either on the vapor-liquid equilibrium between lithium and tritium at low T 2 concentrations, or on the rate of formation and decomposition of lithium tritide. The final distinction between the recovery systems discussed in this report will depend on such data, but presently distillation appears to be the best alternate to the diffusion scheme proposed by A.P. Fraas. The capital cost of equipment necessary to remove tritium by distillation appears to be greater than 10 million dollars for a 5000 MW system, whereas the capital cost associated with the diffusion process has been estimated to be 4 million dollars

  7. Electrolytic gettering of tritium from air

    International Nuclear Information System (INIS)

    Souers, P.C.; Tsugawa, R.T.; Stevens, C.G.

    1983-01-01

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

  8. 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

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

    International Nuclear Information System (INIS)

    Gentile, C.A.; Langish, S.W.; Skinner, C.H.; Ciebiera, L.P.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Gentile, C.A.; Langish, S.W.; Skinner, C.H.; Ciebiera, L.P.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Muller, M.E.

    1978-01-01

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

  12. 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

  13. 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

  14. 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.)

  15. An improved ring removal procedure for in-line x-ray phase contrast tomography

    Science.gov (United States)

    Massimi, Lorenzo; Brun, Francesco; Fratini, Michela; Bukreeva, Inna; Cedola, Alessia

    2018-02-01

    The suppression of ring artifacts in x-ray computed tomography (CT) is a required step in practical applications; it can be addressed by introducing refined digital low pass filters within the reconstruction process. However, these filters may introduce additional ringing artifacts when simultaneously imaging pure phase objects and elements having a non-negligible absorption coefficient. Ringing originates at sharp interfaces, due to the truncation of spatial high frequencies, and severely affects qualitative and quantitative analysis of the reconstructed slices. In this work, we discuss the causes of ringing artifacts, and present a general compensation procedure to account for it. The proposed procedure has been tested with CT datasets of the mouse central nervous system acquired at different synchrotron radiation facilities. The results demonstrate that the proposed method compensates for ringing artifacts induced by low pass ring removal filters. The effectiveness of the ring suppression filters is not altered; the proposed method can thus be considered as a framework to improve the ring removal step, regardless of the specific filter adopted or the imaged sample.

  16. 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

  17. Preparation of honey sample for tritium monitoring

    International Nuclear Information System (INIS)

    Chen Bingru; Wang Chenlian; Wang Weihua

    1989-01-01

    The method of preparation of honey sample for tritium monitoring was described. The equipments consist of an air and honey supply system, a quartz combustor with CM-type monolithic combustion catalyst and a condensation system. In the equipments, honey sample was converted into cooling water by the distilling, cracking and carbonizing procedures for tritium counting. The recovery ratio is 99.0 ± 4.5 percent for tritiated water and 96.0 ± 2.0 for tritiated organic compounds. It is a feasible preparing method for the total tritium monitoring in honey sample

  18. Polymeric media for tritium fixation. Supplement I

    International Nuclear Information System (INIS)

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

    1976-01-01

    Procedures for the fixation of tritium as TH or THO in two different polymeric media are described. The complete procedure for THO fixation in a polyureylene-polyurethane polumer, including polymer molding procedures and leach tests is presented. The catalytic tritiation of polystyrene under very mild conditions using a rhodium catalyst is also described. Thermal stabilities and cost estimates for the polymers examined under this program are discussed. Organic polymers were found to have attractive features for the fixation and storage of concentrated tritium wastes due to the convenience of fixation procedures and favorable properties of the resulting media

  19. 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.)

  20. 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

  1. Tritium handling facilities at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    1975-08-01

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

  3. 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)

  4. 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.)

  5. 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

  6. Issues Associated with Tritium Legacy Materials

    International Nuclear Information System (INIS)

    Mills, Michael

    2008-01-01

    This paper highlights some of the issues associated with the treatment of legacy materials linked to research into tritium over many years and also of materials used to contain or store tritium. The aim of the work is to recover tritium where practicable, and to leave the residual materials passively safe, either for disposal or for continued storage. A number of materials are currently stored at AWE which either contain tritium or have been used in tritium processing. It is essential that these materials are characterised such that a strategy may be developed for their safe stewardship, and ultimately for their treatment and disposal. Treatment processes for such materials are determined by the application of best practicable means (BPM) studies in accordance with the requirements of the Environment Agency of England and Wales. Clearly, it is necessary to understand the objectives of legacy material treatment / processing and the technical options available before a definitive BPM study is implemented. The majority of tritium legacy materials with which we are concerned originate from the decommissioning of a facility that was operational from the late 1950's through to the late 1990's when, on post-operative clear-out (POCO), the entire removable and transportable tritium inventory was moved to new, purpose built facilities. One of the principle tasks to be undertaken in the new facilities is the treatment of the legacy materials to recover tritium wherever practicable, and render the residual materials passively safe for disposal or continued storage. Where tritium recovery was not reasonably or technically feasible, then a means to assure continued safe storage was to be devised and implemented. The legacy materials are in the following forms: - Uranium beds which may or may not contain adsorbed tritium gas; - Tritium gas stored in containers; - Tritide targets for neutron generation; - Tritides of a broad spectrum of metals manufactured for research / long

  7. 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.))

  8. Investigation of tritium in the aquatic environment

    International Nuclear Information System (INIS)

    Cohen, L.K.

    1977-01-01

    The behavior, cycling and distribution of tritium in an aquatic ecosystem was studied in the field and in the laboratory from 1969 through 1971. Field studies were conducted in the Hudson River Estuary, encompassing a 30 mile region centered about the Indian Point Nuclear Plant. Samples of water, bottom sediment, rooted emergent aquatic plants, fish, and precipitation were collected over a year and a half period from more than 15 locations. Specialized equipment and systems were built to combust and freeze-dry aquatic media to remove and recover the loose water and convert the bound tritium into an aqueous form. An electrolysis system was set up to enrich the tritium concentrations in the aqueous samples to improve the analytical sensitivity. Liquid scintillation techniques were refined to measure the tritium activity in the samples. Over 300 samples were analyzed during the course of the study

  9. 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)

  10. Tritium burning in inertial electrostatic confinement fusion facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  11. Tritium burning in inertial electrostatic confinement fusion facility

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  12. Results of tritium tests performed on Sandia Laboratories decontamination system

    International Nuclear Information System (INIS)

    Gildea, P.D.; Wall, W.R.; Gede, V.P.

    1978-05-01

    The Tritium Research Laboratory (TRL), a facility for performing experiments using gram amounts of tritium, became operational on October 1, 1977. As secondary containment, the TRL employs sealed glove boxes connected on demand to two central decontamination systems, the Gas Purification System and the Vacuum Effluent Recovery System. Performance tests on these systems show the tritium removal systems can achieve concentration reduction factors (ratio of inlet to exhaust concentrations) much in excess of 1000 per pass at inlet concentrations of 1 part per million or less for both tritium and tritiated methane

  13. 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

  14. 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

  15. Tritium Issues in Next Step Devices

    International Nuclear Information System (INIS)

    C.H. Skinner; G. Federici

    2001-01-01

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

  16. Tritium Issues in Next Step Devices

    Energy Technology Data Exchange (ETDEWEB)

    C.H. Skinner; G. Federici

    2001-09-05

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

  17. Development of organic tritium light technology at Ontario Hydro

    International Nuclear Information System (INIS)

    Mullins, D.F.; Krasznai, J.P.; Mueller, D.A.

    1992-01-01

    Tritium is a by-product of CANDU heavy water reactor operations and is the major contributor to internal dose for plant workers. The Darlington Tritium Removal Facility (DTRF) is decontaminating heavy water by removing tritium and storing it as a metal hydride. In view of the large tritium separation capacity, (24 MCi/a, 888 PBq/a). This paper reports that Ontario Hydro is interested in pursuing markets for the peaceful uses of tritium. One of these peaceful uses is in self-luminous lighting. The state of the art at present is a phosphor coated tube filled with tritium gas. However, safety considerations have restricted the use of these lights to outdoor or essential safety applications. Binding the tritium to a solid non-volatile matrix would increase the safety of tritium lights and allow the use of other phosphors, matrices and construction geometries. Solid, organic based tritium lights were produced using two different polymer matrices. While both these materials produced visible light, the intensity was low and radiolytic damage to the polymers was evident

  18. 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.

  19. 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

  20. 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)

  1. 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

  2. Tritium control: October 1982-March 1983

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. A tritium vessel cleanup experiment in TFTR

    International Nuclear Information System (INIS)

    Caorlin, M.; Kamperschroer, J.; Owens, D.K.; Voorhees, D.; Mueller, D.; Ramsey, A.T.; La Marche, P.H.; Loughlin, M.J.

    1995-03-01

    A simple tritium cleanup experiment was carried out in TFTR following the initial high power deuterium-tritium discharges in December 1993. A series of 34 ohmic and deuterium neutral beam fueled shots was used to study the removal of tritium implanted into the wall and limiters. A very large plasma was created in each discharge to ''scrub'' an area as large as possible. Beam-fueled shots at 2.5 to 7.5 MW of injected power were used to monitor tritium concentration levels in the plasma by detection of DT-neutrons. The neutron signal decreased by a factor of 4 during the experiment, remaining well above the expected T-burnup level. The amount of tritium recovered at the end of the cleanup was about 8% of the amount previously injected with high power DT discharges. The experience gained suggests that measurements of tritium inventory in the torus are very difficult to execute and require dedicated systems with overall accuracy of 1%

  4. Accounting strategy of tritium inventory in the heavy water detritiation pilot plant from ICIT Rm. Valcea

    International Nuclear Information System (INIS)

    Bidica, N.; Stefanescu, I.; Cristescu, I.; Bornea, A.; Zamfirache, M.; Lazar, A.; Vasut, F.; Pearsica, C.; Stefan, I.; Prisecaru, I.; Sindilar, G.

    2008-01-01

    In this paper we present a methodology for determination of tritium inventory in a tritium removal facility. The method proposed is based on the developing of computing models for accountancy of the mobile tritium inventory in the separation processes, of the stored tritium and of the trapped tritium inventory in the structure of the process system components. The configuration of the detritiation process is a combination of isotope catalytic exchange between water and hydrogen (LPCE) and the cryogenic distillation of hydrogen isotopes (CD). The computing model for tritium inventory in the LPCE process and the CD process will be developed basing on mass transfer coefficients in catalytic isotope exchange reactions and in dual-phase system (liquid-vapour) of hydrogen isotopes distillation process. Accounting of tritium inventory stored in metallic hydride will be based on in-bed calorimetry. Estimation of the trapped tritium inventory can be made by subtraction of the mobile and stored tritium inventories from the global tritium inventory of the plant area. Determinations of the global tritium inventory of the plant area will be made on a regular basis by measuring any tritium quantity entering or leaving the plant area. This methodology is intended to be applied to the Heavy Water Detritiation Pilot Plant from ICIT Rm. Valcea (Romania) and to the Cernavoda Tritium Removal Facility (which will be built in the next 5-7 years). (authors)

  5. 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.)

  6. 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)

  7. 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

  8. Internal dose from tritium at Wolsung nuclear power plant

    International Nuclear Information System (INIS)

    Hee Geun Kim; Jeong Yull Dho; Myung Jae Song

    1995-01-01

    Tritium is produced in large quantities at heavy water nuclear power reactors via the neutron activation reaction 2 H(n,γ) 3 H. At Wolsung nuclear power plant which has a CANDU reactor, the tritium concentrations in coolant and in moderator systems are 1.5 Ci/Kg-D 2 O and 35 Ci/kg-D 2 O, respectively, after 12 years of operation. The airborne tritium concentration in main access area is normally less than 5 MPCa except short-term peaks. The average tritium concentrations in main access controlled areas are normally less than 100 MPCa. Tritium is mainly present in the air of workplace of CANDU reactors as a tritiated water vapour. Airborne tritiated water vapour enters the workers body via inhalation and absorption through skin and can result in a significant dose. The occupational doses from tritium at Wolsung NPP have been maintained below 1 man-Sv per year so far. The tritium contribution to the total plant man-Sv changes between 30 percent and 50 percent. For the mitigation of tritium inhalation, various protective equipment are being used at Wolsung NPP. The respirator system was devised at Wolsung NPP in order to remove tritiated water vapours from the inhaled air. A respirator is connected to a small plastic bottle filled with ice cubes. The system devised shows a good tritium removal efficiency. The air pressure drop through the ice cubes is minimal. The operation cost of the system is also very cheap. Further mitigation of tritium inhalation is heavily dependant on the source term reduction. One of the ultimate solutions is to introduce a tritium removal facility. (author). 7 figs., 3 tabs

  9. Analysis and speciation of the tritium in environmental matrices

    International Nuclear Information System (INIS)

    Bacchetta, Audrey

    2014-01-01

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

  10. Initial experience of tritium exposure control at JET

    International Nuclear Information System (INIS)

    Patel, B.; Campling, D.C.; Schofield, P.A.; Macheta, P.; Sandland, K.

    1998-01-01

    Some of the safety procedures and controls in place for work with tritium are described, and initial operational experience of handling tritium is discussed. A description is given of work to rectify a water leak in a JET neutral beam heating component, which involved man-access to a confined volume to perform repairs, at tritium levels about 100 DAC (80 MBq/m 3 . HTO). Control measures involving use of purge and extract ventilation, and of personal protection using air-fed pressurized suits are described. Results are given of the internal doses to project staff and of atmospheric discharges of tritium during the repair outage. (P.A.)

  11. Tritium waste control: April-September 1982

    International Nuclear Information System (INIS)

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

    1983-01-01

    The pilot-scale, water feed cleanup system was used to successfully remove organic and inorganic impurities from Effluent Removal System (ERS) water. Tests with activated carbon traps removed organic impurities to as low as 2.5 ppM total carbon. Traps containing Amberlite resins for removing organic impurities were not successful and actually contaminated the water with higher levels (>2000 ppM) of organics. Gas generation rates caused by radiolysis of tritiated waste materials were determined for polymer and nonpolymer-impregnated tritiated concrete and fixated and nonfixated tritiated waste vacuum pump oil. In addition, the pressure change of hydrogen cover gas over tritiated water on cement-plaster was determined. The test program to measure and compare the release of tritium from tritiated concrete with and without styrene impregnation continued. Tritium permeation data from small test blocks are given. The drum study monitoring the release of tritium from actual burial packages continued. The maximum fractional release rate for the three types of high activity, tritiated liquid waste generated is 2.97 x 10 -5 , and the maximum total permeation is 158 mCi after 8 yr. These two values represent a 13% increase for the past 6 months. Tritium release from the polymer-impregnated, tritiated concrete (PITC) and from the control (non-PITC) remains very low

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. 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

  17. 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.

  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. 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.

  20. Tritium activity balance in hairless rats following skin-contact exposure to tritium-gas-contaminated stainless-steel surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A

    1994-06-01

    Studies using animals and human volunteers have demonstrated that the dosimetry for skin-contact exposure to contaminated metal surfaces differs from that for the intake of tritiated water or tritium gas. However, despite the availability of some information on the dosimetry for skin-contact with tritium-gas-contaminated metal surfaces, uncertainties in estimating skin doses remain, because of poor accounting for the applied tritium activity in the body (Eakins et al., 1975; Trivedi, 1993). Experiments on hairless rats were performed to account for the tritium activity applied onto the skin. Hairless rats were contaminated through skin-contact exposure to tritium-gas-contaminated stainless-steel planchets. The activity in the first smear was about 35% of the total removable activity (measured by summing ten consecutive swipes). The amount of tritium applied onto the skin can be approximated by estimating the tritium activity in the first smear removed form the contaminated surfaces. 87 {+-} 9% of the transferred tritium was retained in the exposed skin 30 min post-exposure. 30 min post exposure, the unexposed skin and the carcass retained 8 {+-} 6% and 3 {+-} 2% of the total applied tritium activity, respectively. The percentage of tritium evolved from the body or breathed out was estimated to be 2 {+-} 1% of the total applied activity 30 min post-exposure. It is recommended that to evaluate accurately the amount of tritium transferred to the skin, alternative measurement approaches are required that can directly account for the transferred activity onto the skin. 15 refs., 13 tabs., 7 figs.

  1. Tritium and radon risks for humans

    International Nuclear Information System (INIS)

    Mauna, Traian; Mauna, Andriesica

    2008-01-01

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

  2. 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

  3. Tritium isotopic exchange in air detritiation dryers

    International Nuclear Information System (INIS)

    Everatt, A.E.; Johnson, R.E.; Senohrabek, J.A.; Shultz, C.M.

    1989-02-01

    Isotopic exchange between tritiated and non-tritiated water species in a molecular sieve bed has been demonstrated. At high humidities (+6 degrees Celsius dew point) the rate of tritium isotopic exchange in a 2.4 L molecular sieve bed has been demonstrated to be at least 50% of published exchange rates. In an industrial-sized air detritiation dryer, utilizing the pretreatment technique of H 2 O steam washing to elute the residual tritium, a DF of 12 600 has been demonstrated when operating at an inlet vapor tritium concentration of 14 Ci/kg and at inlet and outlet dew points of 4.8 and -54 degrees Celsius, respectively. In the NPD dryer bed studied, which was not optimally designed for full benefit from isotopic exchange, at least one order of magnitude in additional detritiation is attributed to isotopic exchange in the unsaturated zone. The technique of eluting the residual tritium from an industrial sized bed by H 2 O washing at high temperature, high humidity and low bed loading has been demonstrated to be a fast and effective way of removing tritium from a molecular sieve bed during regeneration. The isotopic exchange model accurately predicted the exchange between tritiated and non-tritiated water species in a molecular sieve bed where there is no net adsorption or desorption. The model's prediction of the tritium breakthrough trend observed in the NPD tests was poor; however, a forced fit can be achieved if the exchange rates in the MTZ and the unsaturated zone are manipulated. More experiments are needed to determine the relative rates of tritium exchange in the saturated, mass transfer, and unsaturated zones of a dryer bed

  4. Incorporation of tritium from wrist watches

    International Nuclear Information System (INIS)

    Schoenhofer, F.; Pock, K.

    1995-01-01

    Watches are consumer products and are subject to the regulations that control food and consumer products. Elevated concentrations of tritium were found in the urine of persons who wore wrist watches with luminous dials and plastic cases. High emission of tritium from these watches were observed. In an experiment, a volunteer wore a watch with high emissions and the build-up of the tritium concentration in urine was monitored, as well as the decline after removing the watch. Possible pathways for the incorporation and its mechanism are considered. In spite of the relatively high activity concentrations observed, the dose is negligible. On the other hand, the principle 'ALARA' can be achieved without any costs by simply choosing other types of watches. (author). 12 refs., 2 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    Muller, M.E.

    1978-01-01

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

  6. Experimental determination of reaction rates of water. Hydrogen exchange of tritium with hydrophobic catalysts

    International Nuclear Information System (INIS)

    Bixel, J.C.; Hartzell, B.W.; Park, W.K.

    1976-01-01

    This study was undertaken to obtain data needed for further development of a process for the enrichment and removal of tritium from the water associated with light-water reactors, fuel-reprocessing plants, and tritium-handling laboratories. The approach is based on the use of antiwetting, hydrophobic catalysts which permit the chemical exchange reactions between liquid water and gaseous hydrogen in direct contact, thus eliminating problems of catalyst deactivation and the complexity of reactor design normally associated with current catalytic-detritiation techniques involving gas-phase catalysis. An apparatus and procedure were developed for measuring reaction rates of water-hydrogen chemical exchange with hydrophobic catalysts. Preliminary economic evaluations of the process were made as it might apply to the AGNS fuel reprocessing plant

  7. The advanced CECE process for enriching tritium by the chemical exchange method with a hydrophobic catalyst

    International Nuclear Information System (INIS)

    Kitamoto, Asashi; Shimizu, Masami; Masui, Takashi.

    1992-01-01

    The monothermal chemical exchange process with electrolysis, i.e., CECE process, was an effective method for enriching and removing tritium from tritiated water with low to middle level activity. The purpose of this study is to propose the theoretical background of the two-parameter evaluation method, which is based on a two-step isotope exchange reaction between hydrogen gas and liquid water, for improvement of the performance of a hydrophobic catalyst by a trickle bed-type column. Finally, a two-parameter method could attain the highest performance of isotope separation and the lowest liquid holdup for a trickle bed-type column. Therefore, this method will present some effective and practical procedures in scaling up a tritium enrichment process. The main aspect of the CECE process in engineering design and system evaluation was to develop the isotope exchange column with a high performance catalyst. (author)

  8. 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.)

  9. 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)

  10. 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)

  11. 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

  12. Selection of fluids for tritium pumping systems

    International Nuclear Information System (INIS)

    Chastagner, P.

    1984-02-01

    The degradation characteristics of three types of vacuum pump fluids, polyphenyl ethers, perfluoropolyethers and hydrocarbon oils were reviewed. Fluid selection proved to be a critical factor in the long-term performance of tritium pumping systems and subsequent tritium recovery operations. Thermal degradation and tritium radiolysis of pump fluids produce contaminants which can damage equipment and interfere with tritium recovery operations. General characteristics of these fluids are as follows: polyphenyl ether has outstanding radiation resistance, is very stable under normal diffusion pump conditions, but breaks down in the presence of oxygen at anticipated operating temperatures. Perfluoropolyether fluids are very stable and do not react chemically with most gases. Thermal and mechanical degradation products are inert, but the radiolysis products are very corrosive. Most of the degradation products of hydrogen oils are volatile and the principal radiolysis product is methane. Our studies show that polyphenyl ethers and hydrocarbon oils are the preferred fluids for use in tritium pumping systems. No corrosive materials are formed and most of the degradation products can be removed with suitable filter systems

  13. Design and operational experience with a portable tritium cleanup system

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Wilson, S.W.; Garcia, F.

    1991-06-01

    We built a portable tritium cleanup system to scavenge tritium from contaminated gases in any tritium-containing system in the LLNL Tritium Facility. The cleanup system uses standard catalytic oxidation of tritium to water followed by water removal with a molecular sieve dryer. The cleanup unit, complete with instrumentation, is contained in a portable cart that is rolled into place and connected to the apparatus to be cleaned. The cleanup systems is effective, low-tech, simple, and reliable. The nominal flow rate of the system is 30 liters/minute, and the decontamination factor is > 1000. In this paper we will show design information on our portable cleanup system, and will discuss our operational experience with it over the past several years

  14. 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

  15. 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.

  16. Bacteriological study and structural composition of staghorn stones removed by the anatrophic nephrolithotomic procedure

    Directory of Open Access Journals (Sweden)

    Hamid Shafi

    2013-01-01

    Full Text Available This study was conducted to determine the composition of staghorn stones and to assess the proportion of infected stones as well as the correlation between infection in the stones and bacteria grown in urine. Samples of 45 consecutive stones removed through anatrophic nephrolithotomic procedures were taken from the operation site and samples of urine were obtained by simultaneous bladder catheterization. The frequency of infection in the stones and correlation between infection of stone and urine samples were determined with respect to the composition of the stones. Twenty-two males and 23 females, with respective mean ages of 48.3 ± 15.6 years and 51 ± 7.4 years, were studied. The stone and urine cultures yielded positive results in ten and 16 patients, respectively, of a total of 45 patients (22.2% and 35.5%, respectively. Calcium oxalate was the main constituent of staghorn stones, seen in 31 patients (68.8%, uric acid in 12 patients (26.6% and struvite and/or calcium phosphate in 11 patients (24.4%. In seven of ten stones with bacterial growth, bacteria were isolated from urine cultures as well, which accounted for a concordance rate of 70%. The bacteria grown in the stone were the cause of urinary tract infection (UTI in 43.5% of the cases. Stone infection was significantly associated with UTI (OR = 6.47; 95% CI 1.43-31.7, P = 0.021 and presence of phosphate in the stones (OR = 18, 95% CI 3.28-99.6, P = 0.0006. E. coli was the most common bacteria grown from the stones, and was isolated in 50% of the cases; Ureaplasma urealyticum was the most common organism causing UTI, grown in 62.5% of the urine samples. There was a high concordance rate between bacteria in the stones and urine. These findings indicate that the urine culture can provide information for selection of an appropriate anti-microbial agent for stone sterilization. In addition, preventing re-growth or recurrence of stones and treatment of post-surgical infections would be

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

    International Nuclear Information System (INIS)

    Mendoza, Norman; Sucgang, Raymond; Castaneda, Soledad

    2009-01-01

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

  18. 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.)

  19. 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

  20. 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.)

  1. 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

  2. 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.

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

    International Nuclear Information System (INIS)

    Kim, K.; Lee, S.K.; Chung, E.S.; Kim, K.S.; Kim, W.S.; Nam, G.J.

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  5. 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)

  6. 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)

  7. Release of gaseous tritium during reprocessing

    International Nuclear Information System (INIS)

    Bruecher, H.; Hartmann, K.

    1983-01-01

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

  8. Tritium inventory and recovery in next-step fusion devices

    International Nuclear Information System (INIS)

    Causey, R.A.; Brooks, J.N.; Federici, G.

    2002-01-01

    with the tritium from the plasma will produce a layer of carbonaceous material potentially containing kilograms of tritium in the cooler areas of the tokamak (J. Vac. Sci. Technol. A5 (1987) 2286). This paper reviews the tritium retention mechanisms for the three materials discussed above. Tritium removal techniques, including those used in situ to minimize in-vessel inventories as well as those used to reduce contamination prior to waste disposal, are discussed

  9. 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

  10. Development of tritium cleanup system for LHD

    International Nuclear Information System (INIS)

    Sakuma, Yoichi; Kawano, Takao; Shibuya, Mamoru; Kabutomori, Toshiki

    2000-01-01

    Energy is vital for humans and we have been consuming a large amount of fossil fuel especially from the beginning of the industrial revolution. Nowadays its huge consumption has however come to threaten our life and we have to prepare nonfossil fuels, for instance solar energy, biomass energy, nuclear energy and so on. Fusion energy is an unlimited resource and one of the strongest candidates of the future energy source. At the National Institute for Fusion Science (referred to as 'NIFS' hereafter), we have constructed a new fusion experimental device called large helical device (referred to as 'LHD' hereafter) in 1998. The device will generate a small amount of tritium, as a fusion product. In order to remove it from the exhaust gas, we have designed a tritium cleanup system based on a new concept. This system is mainly composed of a palladium permeater, a decomposer and hydrogen absorbing alloys. It may perfectly recover the tritium from exhaust gas without oxidizing it. This system is applicable for the future needs at fusion power plants. In order to remove tritium discharged from fusion experimental facilities, it is usual to employ a system by which tritiated constituents, in various chemical forms, are entirely converted to a form of water vapor by catalytic oxidation. The water vapor containing tritiated form is then absorbed by molecular sieve (referred to as 'wet system' hereafter). However, in the case of LHD, it is not rational to deliberately convert the discharged tritium into the water vapor, because the tritium discharged from LHD is almost in a form of hydrogen molecules. Moreover, the tritium in the form of water vapor affects the human body 18000 times stronger than that of hydrogen molecules. In accordance with these view points, we have developed another type of tritium cleanup system based on a new concept, in which hydrogen molecules including tritiated ones (HT, DT and T 2 ) found in the exhaust gas of LHD are directly fixed to hydrogen

  11. Determination of tritium in wine yeast samples

    International Nuclear Information System (INIS)

    Cotarlea, Monica-Ionela; Paunescu Niculina; Galeriu, D; Mocanu, N.; Margineanu, R.; Marin, G.

    1998-01-01

    Analytical procedures were developed to determine tritium in wine and wine yeast samples. The content of organic compounds affecting the LSC measurement is reduced by fractioning distillation for wine samples and azeotropic distillation/fractional distillation for wine yeast samples. Finally, the water samples were normally distilled with K MO 4 . The established procedures were successfully applied for wine and wine samples from Murfatlar harvests of the years 1995 and 1996. (authors)

  12. Tritium concentration in the heavy water upgrading plants

    International Nuclear Information System (INIS)

    Croitoru, C.; Pop, F.; Titescu, Gh.; Dumitrescu, M.; Ciortea, C.; Stefanescu, I.; Peculea, M.; Pitigoi, Gh.; Trancota, D. . E-mail of corresponding author: croitoru@icsi.ro; Croitoru, C.)

    2005-01-01

    In the course of time heavy water used in CANDU nuclear power plants, as moderator or coolant, degrades, as a result of its impurification with light water and tritium. Concentration diminution below 99.8% mol for moderator and 99.75% mol for coolant causes an inefficient functioning of CANDU reactor. By isotopic distillation, light water is removed. Simultaneously tritium concentration takes place. The heavy water upgrading plant from Cernavoda is an isotopic separation cascade with two stages. The paper presents, for this plant, a theoretical study of the tritium concentration. (author)

  13. 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)

  14. A methodology for determination of tritium inventory to the heavy water detritiation pilot plant from ICIT Rm. Valcea

    International Nuclear Information System (INIS)

    Bidica, N.; Stefanescu, I.; Bornea, A.; Zamfirache, M.; Lazar, A.; Vasut, F.; Pearsica, C.; Stefan, I.; Cristescu, I.; Prisecaru, I.; Sindilar, G.

    2007-01-01

    Full text: In this paper we present a methodology for determination of tritium inventory in a tritium removal facility. The method proposed is based on the developing of computing models for accountancy of the mobile tritium inventory in the separation processes of the stored tritium and of the trapped tritium inventory in the structure of the process system components. The configuration of the detritiation process is a combination of isotope catalytic exchange between water and hydrogen (LPCE) and the cryogenic distillation of hydrogen isotopes (CD). The computing model for tritium inventory in the LPCE process and the CD process will be developed based on mass transfer coefficients in catalytic isotope exchange reactions and in dual-phase system (liquid-vapour) of hydrogen isotopes distillation process. Accounting of tritium inventory stored in metallic hydride will be based on in-bed calorimetry. Estimation of the trapped tritium inventory can be made by subtraction of the mobile and stored tritium inventories from the global tritium inventory of the plant area. Determinations of the global tritium inventory of the plant area will be made on a regular basis by measuring any tritium amount entering or leaving the plant area. This methodology is intended to be applied to the Heavy Water Detritiation Pilot Plant from ICIT Rm. Valcea (Romania) and at the Cernavoda Tritium Removal Facility (which will be built in the next 5-7 years). (authors)

  15. 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

  16. Design procedure for sizing a submerged-bed scrubber for airborne particulate removal

    International Nuclear Information System (INIS)

    Ruecker, C.M.; Scott, P.A.

    1987-04-01

    Performance correlations to design and operate the submerged bed scrubber were developed for various applications. Structural design procedure outlined in this report focuses on off-gas scrubbing for HLW vitrification applications; however, the method is appropriate for other applications

  17. 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

  18. 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

  19. Maxillary Swing Approach for Removal of Palatal Carcinoma: A Modified Procedure

    Directory of Open Access Journals (Sweden)

    Tsutomu Nomura

    2018-01-01

    Full Text Available Introduction. We report a modification of the maxillary swing approach to remove a palatal tumor while preserving the anterior alveolar area. Methods. Case report using clinical records. Results. The patient was a 54-year-old male. TNM grade was T4bN0M0, and invasion to the base of the pterygoid process was seen. Two courses of induction chemotherapy were administered prior to the operation. Because there was no evidence of anterior maxillary invasion, the maxillary swing approach was chosen. The left anterior maxilla was cut and swung laterally, preserving the blood supply. After removal of the palatal tumor, the maxilla was repositioned and the defect was restored with an anterior lateral thigh flap. Postoperative course was typical, and facial appearance, speech, and masticatory function were satisfactory. Conclusions. This technique is particularly useful for preserving appearance as well as speech and mastication.

  20. Oil adsorbing package, also procedure to remove oil from a water surface

    Energy Technology Data Exchange (ETDEWEB)

    1971-05-01

    A method is given to remove oil from water to prevent water pollution. Use is made of an oil-adsorbing packet having a specific gravity which is lower than the specific gravity of water. The hull is manufactured from any material which is not a water-insoluble nonpolar material. The hull is partly permeable to water and encloses a solid oil-adsorbing compound having a large adsorbing surface. (10 claims)

  1. Procedure to remove dissolved nickel and/or radium compounds from water and facility therefor

    International Nuclear Information System (INIS)

    Moravec, J.

    2004-01-01

    Dissolved nickel and/or radium compounds are removed from water on a granular material such as quartz sand, crushed coal or granulated MnO 2 whose surface contains oxides of manganese MnO x . The compounds to be removed are adsorbed into the MnO x layer. Subsequently the adsorbed compounds are desorbed with a reductant, such as sodium sulfite, which is present in a concentration forming a redox potential of -5 to -120 mV, and with a solution of sodium polyphosphate, such as sodium hexametaphosphate (NaPO 3 ) n . Two variants are possible: either MnO x is first acted upon with the reductant and subsequently with the polyphosphate, or a mixed solution of the two agents is used. The excess of the agents is removed with water or with a KMnO 4 solution at 0.001 to 25 g/L. The granular material as well as the agent solutions (after concentration) are reusable. (P.A.)

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

    International Nuclear Information System (INIS)

    Sherman, R.H.; Bartlit, J.R.

    1988-01-01

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

  3. 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

  4. New procedure for the control of the treatment of industrial effluents to remove volatile organosulfur compounds.

    Science.gov (United States)

    Boczkaj, Grzegorz; Makoś, Patrycja; Fernandes, André; Przyjazny, Andrzej

    2016-10-01

    We present a new procedure for the determination of volatile organosulfur compounds in samples of industrial effluents using dispersive liquid-liquid microextraction and gas chromatography with flame photometric detection. Initially, the extraction parameters were optimized. These included: type and volume of extraction solvent, volume of disperser solvent, salting out effect, pH, time and speed of centrifugation as well as extraction time. The procedure was validated for 30 compounds. The developed procedure has low detection limits of 0.0071-0.49 μg/L and a good precision (relative standard deviation values of 1.2-5.0 and 0.6-4.1% at concentrations of 1 and 10 μg/L, respectively). The procedure was used to determine the content of volatile organosulfur compounds in samples of effluents from the production of bitumens before and after chemical treatment, in which six compounds were identified, including 2-mercaptoethanol, thiophenol, thioanisole, dipropyl disulfide, 1-decanethiol, and phenyl isothiocyanate at concentrations ranging from 0.47 to 8.89 μg/L. Problems in the determination of organosulfur compounds related to considerable changes in composition of the effluents, increase in concentration of individual compounds and appearance of secondary pollutants during effluent treatment processes are also discussed. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  5. 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

  6. Handling of tritium at TFTR

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  7. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.; Masson, M.; Briec, M.

    1986-09-01

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 10 13 Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 10 10 Bq (0.5 Ci) per day per ton of fuel

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

    International Nuclear Information System (INIS)

    Ehrenkaufer, R.L.E.; Hembree, W.C.; Wolf, A.P.

    1988-01-01

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

  9. 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

  10. Release of tritium from fuel and collection for storage

    International Nuclear Information System (INIS)

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

    1976-04-01

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

  11. 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.

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

    International Nuclear Information System (INIS)

    Klein, J.E.; Wermer, J.R.

    1994-01-01

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

  13. Successful Insular Glioma Removal in a Deaf Signer Patient During an Awake Craniotomy Procedure.

    Science.gov (United States)

    Metellus, Philippe; Boussen, Salah; Guye, Maxime; Trebuchon, Agnes

    2017-02-01

    Resection of tumors located within the insula of the dominant hemisphere represents a technical challenge because of the complex anatomy, including the surrounding vasculature, and the relationship to functional (motor and language) structures. We report here the case of a successful resection of a left insular glioma in a native deaf signer during an awake craniotomy. The patient, a congenitally deaf right-handed patient who is a native user of sign language, presented with a seizure 1 week before he was referred to our department. Magnetic resonance imaging revealed a left heterogeneous insular tumor enhanced after intravenous gadolinium infusion. Because of its deep and dominant hemisphere location, an awake craniotomy was decided. The patient was evaluated intraoperatively using object naming, text reading, and sign repetition tasks. An isolated inferior frontal gyrus site evoked repeated object naming errors. A transopercular parietal approach was performed and allowed the successful removal of the tumor under direct electric stimulation and electrocorticography. To our knowledge, this is the first report of successful removal of a left insular tumor without any functional sequelae in a native deaf signer using intraoperative direct cerebral stimulation during an awake craniotomy. The methodology used also provides the first evidence of the actual anatomo-functional organization of language in deaf signers. Copyright © 2016 Elsevier Inc. All rights reserved.

  14. 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)

  15. 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

  16. Pollutants in drinking water: their sources, harmful effects and removal procedures

    International Nuclear Information System (INIS)

    Qadeer, R.

    2004-01-01

    The underground water resources available for human consumption are being continuously contaminated by the natural sources and anthropogenic activities. The pollutants include toxic microorganism, inorganic and organic chemical and radionuclide etc. this is an acute problem in our country, where free style way of disposal of industrial effluents into the natural water bodies contaminates the surface and ground water. These contaminants make their way into human body through contaminated drinking water, which leads to the malfunctioning of the body organs. Details of some pollutants present in drinking water, their source and harmful effects on human beings are reviewed in this communication. Merits and demerits of methods used to remove the pollutants from drinking water are also discussed. (author)

  17. Rapid assessment of soil and groundwater tritium by vegetation sampling

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.

    1995-01-01

    A rapid and relatively inexpensive technique for defining the extent of groundwater contamination by tritium has been investigated. The technique uses existing vegetation to sample the groundwater. Water taken up by deep rooted trees is collected by enclosing tree branches in clear plastic bags. Water evaporated from the leaves condenses on the inner surface of the bag. The water is removed from the bag with a syringe. The bags can be sampled many times. Tritium in the water is detected by liquid scintillation counting. The water collected in the bags has no color and counts as well as distilled water reference samples. The technique was used in an area of known tritium contamination and proved to be useful in defining the extent of tritium contamination

  18. Tritium permeation losses in HYLIFE-II heat exchanger tubes

    International Nuclear Information System (INIS)

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

    1990-01-01

    Tritium permeation through the intermediate heat exchanger of the HYLIFE-II inertial fusion design concept is evaluated for routine operating conditions. The permeation process is modelled using the Lewis analogy combined with surface recombination. It is demonstrated that at very low driving potentials, permeation becomes proportional to the first power of the driving potential. The model predicts that under anticipated conditions the primary cooling loop will pass about 6% of the tritium entering it to the intermediate coolant. Possible approached to reducing tritium permeation are explored. Permeation is limited by turbulent diffusion transport through the molten salt. Hence, surface barriers with impendance factors typical of present technology can do very little to reduce permeation. Low Flibe viscosity is desirable. An efficient tritium removal system operating on the Flibe before it gets to the intermediate heat exchanger is required. Needs for further research are highlighted. 9 refs., 2 figs., 1 tab

  19. 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

  20. 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

  1. 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

  2. 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.

  3. 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.)

  4. Development of method of tritium labeling of pharmacological preparate of drotaverine hydrochloride (NOSPA)

    International Nuclear Information System (INIS)

    Kim, A.A.; Djuraeva, G.T.; Shukurov, B.V.

    2004-01-01

    Full text: The method for tritium labeling of pharmacological preparate of drotaverine hydrochloride (no spa) was developed. Drotaverine hydrochloride was labeled by thermally activated tritium in apparatus for tritium labeling. The optimum regime of labeling was selected. The system of purification of tritium labeled drotaverine hydrochloride by thin layer chromatography (TLC) has been developed. The TLC system of purification of tritium labeled drotaverine hydrochloride was developed. Tritium labeled preparation of drotaverine hydrochloride was purified by TLC on silicagel in system isopropanol: ammonia: water (8:1:1). We found appearance of additional fractions in tritium labeled preparation of drotaverine hydrochloride that testifies to partial transformation of drotaverine hydrochloride during procedure of labeling. Application of TLC for purification of tritium labeled preparation allows to purify completely drotaverine hydrochloride of by-products. The output of purified tritium labeled preparation of drotaverine hydrochloride was about 25 %. The received preparation had specific radioactivity - 3,2 MBq/mg, radiochemical purity of a preparation was 95 %. TLC purification seems inexpensive, fast and suitable for purification of tritium-labeled drotaverine hydrochloride. Thus developed method allows obtain tritium labeled preparation of drotaverine hydrochloride (no - spa), suitable for medical and biologic researches

  5. 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

  6. 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

  7. Separation of tritium from reprocessing effluents

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  8. A new methodology for assessment of pectus excavatum correction after bar removal in Nuss procedure: Preliminary study.

    Science.gov (United States)

    Gomes-Fonseca, João; Vilaça, João L; Henriques-Coelho, Tiago; Direito-Santos, Bruno; Pinho, António C M; Fonseca, Jaime C; Correia-Pinto, Jorge

    2017-07-01

    The objective is to present a new methodology to assess quantitatively the impact of bar removal on the anterior chest wall, among patients with pectus excavatum who have undergone the Nuss procedure, and present a preliminary study using this methodology. We propose to acquire, for each patient, the surface of the anterior chest wall using a three-dimensional laser scanner at subsequent time points (short term: before and after surgery; long term: follow-up visit, 6months, and 12months after surgery). After surfaces postprocessing, the changes are assessed by overlapping and measuring the distances between surfaces. In this preliminary study, three time points were acquired and two assessments were performed: before vs after bar removal (early) and before vs 2-8weeks after bar removal (interim). In 21 patients, the signed distances and volumes between surfaces were computed and the data analysis was performed. This methodology revealed useful for monitoring changes in the anterior chest wall. On average, the mean, maximum, and volume variations, in the early assessment, were -0.1±0.1cm, -0.6±0.2cm, and 47.8±22.2cm 3 , respectively; and, in the interim assessment, were -0.5±0.2cm, -1.3±0.4cm, and 122.1±47.3cm 3 , respectively (pbar was in situ was inversely and significantly correlated with postretraction and was a relevant predictor of its decrease following surgery (pbar was in situ may be the main determinant of the anterior chest wall retraction following bar removal. Further studies should continue to corroborate and reinforce the preliminary findings, by increasing the sample size and performing long-term assessments. III. Copyright © 2017 Elsevier Inc. All rights reserved.

  9. 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

  10. 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.)

  11. 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.)

  12. 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

  13. 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

  14. Two-stage electrolysis to enrich tritium in environmental water

    International Nuclear Information System (INIS)

    Shima, Nagayoshi; Muranaka, Takeshi

    2007-01-01

    We present a two-stage electrolyzing procedure to enrich tritium in environmental waters. Tritium is first enriched rapidly through a commercially-available electrolyser with a large 50A current, and then through a newly-designed electrolyser that avoids the memory effect, with a 6A current. Tritium recovery factor obtained by such a two-stage electrolysis was greater than that obtained when using the commercially-available device solely. Water samples collected in 2006 in lakes and along the Pacific coast of Aomori prefecture, Japan, were electrolyzed using the two-stage method. Tritium concentrations in these samples ranged from 0.2 to 0.9 Bq/L and were half or less, that in samples collected at the same sites in 1992. (author)

  15. Binder-free Na-mordenite pellets for tritium processing

    International Nuclear Information System (INIS)

    Toci, F.; Viola, A.; Edwards, R.A.H.; Mencarelli, T.; Brossa, P.

    1995-01-01

    Gas separation systems based on adsorption on zeolites are used in various applications involving tritium: air and inert gas detritiation, purification of Q 2 and Q 2 O, and isotope separation. Differential adsorption processes are attractive because efficient separation can be combined with small plant dimensions, low energy consumption and a small tritium inventory. Zeolites are the usual choice for the adsorbate because they combine high adsorption capacity with high selectivity and stability. However, commercial pellets show appreciable tritium retention due to inappropriate activation procedures or the presence of a binder. In this paper we report a research study aimed at producing a pelletized zeolite without binder (self-bound) with low tritium retention. (orig.)

  16. Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-14

    Under the Tritium Facility Modernization {ampersand} Consolidation (TFM{ampersand}C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM{ampersand}C Project also provides for a new replacement R&D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H.

  17. Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

    International Nuclear Information System (INIS)

    Hsu, R.H.; Oji, L.N.

    1997-01-01

    Under the Tritium Facility Modernization ampersand Consolidation (TFM ampersand C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM ampersand C Project also provides for a new replacement R ampersand D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H

  18. In-bed accountability of tritium in production scale metal hydride storage beds

    International Nuclear Information System (INIS)

    Klein, J.E.

    1995-01-01

    An ''in-bed accountability'' (IBA) flowing gas calorimetric measurement method has been developed and implemented to eliminate the need to remove tritium from production scale metal hydride storage beds for inventory measurement purposes. Six-point tritium IBA calibration curves have been completed for two, 390 gram tritium metal hydride storage beds. The calibration curves for the two tritium beds are similar to those obtained from the ''cold'' test program. Tritium inventory errors at the 95 percent confidence level ranged from ± 7.3 to 8.6 grams for the cold test results compared to ± 4.2 to 7.5 grams obtained for the two tritium calibrated beds

  19. Preparation of tritium-labelled dextran and inulin

    International Nuclear Information System (INIS)

    Akulov, G.P.; Kaminski, Ju.L.; Korsakova, N.A.; Kudelin, B.K.

    1992-01-01

    Tritiated dextran and inulin were prepared by both a catalytic solid state and a liquid phase isotropic exchange with gaseous tritium. The liquid phase procedure is convenient for preparation of the polysaccharides with specific activities up to 5 mCi/g, while the solid state procedure allows specific activities up to 700 mCi/g. (Author)

  20. 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)

  1. 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)

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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)

  7. Tritium volume activity in natural waters of NPP Temelin region

    Energy Technology Data Exchange (ETDEWEB)

    Tomasek, M; Wilhelmova, L [Academy of Sciences of the Czech Rep., Prague (Czech Republic). Nuclear Physics Inst., Dept. of Radiation Dosimetry

    1996-12-31

    This paper presents the results of tritium measurement in selected rivers of NPP Temelin before its operation obtained during the period 1991-1994. Particular attention is paid to Vltava river into which liquid effluents will be discharged and which is also utilized as a drinking water supply for the capital Prague. Samples from the Vltava river were collected near the mouth of NPP waste canal (point Hladna)and in front of the intake into Prague water works (point Podoli). Tritium content was analysed also in surface waters of Paleckuv, Temelinsky and Strouha streams which can be affected by gaseous effluents due to atmospheric removal processes. Tritium activity was measured with Tric-Carb 1050 TR/LL liquid scintillation counter. The mean annual tritium activities of investigated river waters varied within 1.9-3.0 Bq/l during the period 1991-1994 and that their trend has been slowly decreasing. This fact, as well as seasonal variability, suggests, that tritium level in the surface waters of studied region is largely governed by this radionuclide global atmospheric fallout. The results of this work indicate the trend of background tritium in examined natural waters and make possible the evaluation of their potential future contamination. (J.K.) 1 tab., 2 figs., 4 refs.

  8. Separation of tritium from gaseous and aqueous effluent systems

    International Nuclear Information System (INIS)

    Kobisk, E.H.

    1977-01-01

    Removal or reduction of tritium content in a wide variety of effluent streams has been extensively studied in the United States. This paper specifically reviews three processes involving tritium separation in the gaseous phase and the aqueous phase. Diffusion through a selective Pd-25Ag alloy membrane at temperatures up to 600 0 C and at pressures up to 700 kg/cm 2 has resulted in successful separation of hydrogen-deuterium mixtures with an associated separation factor of 1.65 (and gives a calculated separation factor for hydrogen-tritium mixtures of 2.0). Use of a single palladium bipolar membrane in an electrolysis system has been found to yield a hydrogen-deuterium separation factor of 4 and a hydrogen-tritium factor of 6 to 11 without the production of gaseous hydrogen. Finally, countercurrent catalytic exchange between tritium-containing hydrogen gas and water has yielded a separation factor of 6.3. The specific advantages of each of these systems will be discussed in terms of their potential applications. In all cases, further investigations are necessary to scale the systems to handle large quantities of feed material in a continuous mode and to minimize energy requirements. Such separative systems must necessarily be cascaded to yield gaseous or aqueous product streams suitable for recycling to the tritium producing systems, for storage or for discharge to the environment. (orig./HP) [de

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. Absolute measurement of a tritium standard

    International Nuclear Information System (INIS)

    Hadzisehovic, M.; Mocilnik, I.; Buraei, K.; Pongrac, S.; Milojevic, A.

    1978-01-01

    For the determination of a tritium absolute activity standard, a method of internal gas counting has been used. The procedure involves water reduction by uranium and zinc further the measurement of the absolute disintegration rate of tritium per unit of the effective volume of the counter by a compensation method. Criteria for the choice of methods and procedures concerning the determination and measurement of gaseous 3 H yield, parameters of gaseous hydrogen, sample mass of HTO and the absolute disintegration rate of tritium are discussed. In order to obtain gaseous sources of 3 H (and 2 H), the same reversible chemical reaction was used, namely, the water - uranium hydride - hydrogen system. This reaction was proved to be quantitative above 500 deg C by measuring the yield of the gas obtained and the absolute activity of an HTO standard. A brief description of the measuring apparatus is given, as well as a critical discussion of the brass counter quality and the possibility of obtaining equal working conditions at the counter ends. (T.G.)

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

    International Nuclear Information System (INIS)

    Kobayashi, K.; Miya, N.; Ikeda, Y.; Torikai, Y.; Saito, M.; Alimov, V.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  19. Effect of bleaching and repolishing procedures on coffee and tea stain removal from three anterior composite veneering materials.

    Science.gov (United States)

    Türkün, L Sebnem; Türkün, Murat

    2004-01-01

    Discolored teeth can be treated with resin veneers, but their color changes when confronted with staining solutions. Polishing procedures can provide a remedy for highly stained composites, but they tend to remove some materials as well. However, bleaching procedures are an effective, nondestructive method for solving the problem. The aim of this study was to compare the color change of three veneer composites exposed to staining solutions and to evaluate the effectiveness of a 15% hydrogen peroxide bleaching agent and three polishing systems to remove the stain. Forty-five disks (12 x 2 mm) each of Clearfil ST (Kuraray Co. Ltd., Osaka, Japan), Esthet-X (Dentsply/Caulk, Milford DE, USA), and Filtek A110 (3M ESPE, St. Paul, MN, USA) were prepared. The specimens were polished with Sof-Lex (3M ESPE), Enhance (Dentsply/Caulk), or PoGo (Dentsply/Caulk). Five specimens for each material-polishing system combination were immersed in coffee (Nescafe Classic, Nestle SA, Vevey, Switzerland) or tea (Earl Grey, Lipton, Blackfriars-London, England) for 7 days. The remaining disks were stored in water. Color measurements were made with a spectrophotometer (X-Rite Seroice SP78, Loaner, Köln, Germany) at baseline; after 1, 3, 5, and 7 days; and after bleaching and repolishing. After 1 week, one side of the specimens was bleached with Illuminé-office (Dentsply De Trey GmbH, Konstanz, Germany) for 1 hour, and the other side was repolished for 30 seconds. All comparisons of color change for the polishing systems, times, and staining solutions were subjected to repeated measurements of analysis of variance. Paired t-test was used to examine whether significant color differences (deltaE*) occurred during immersion at the specified time intervals (p < or = .05). Filtek A110 was the least stained resin composite. Its color remained under a deltaE* value of 2 during the study. Clearfil ST exhibited the most color change after 1 week. All specimens polished with Enhance showed less

  20. 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

  1. 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

  2. Automatic isotope gas analysis of tritium labelled organic materials Pt. 1

    International Nuclear Information System (INIS)

    Gacs, I.; Mlinko, S.

    1978-01-01

    A new automatic procedure developed to convert tritium in HTO hydrogen for subsequent on-line gas counting is described. The water containing tritium is introduced into a column prepared from molecular sieve-5A and heated to 550 deg C. The tritium is transferred by isotopic exchange into hydrogen flowing through the column. The radioactive gas is led into an internal detector for radioactivity measurement. The procedure is free of memory effects, provides quantitative recovery with analytical reproducibility better than 0.5% rel. at a preset number of counts. The experimental and analytical results indicate that isotopic exchange between HTO and hydrogen over a column prepared from alumina or molecular sieve-5A can be successfully applied for the quantitative transfer of tritium from HTO into hydrogen for on-line gas countinq. This provides an analytical procedure for the automatic determination of tritium in water with an analytical reproducibility better than 0.5% rel. The exchange process will also be suitable for rapid tritium transfer from water formed during the decomposition of tritium-labelled organic compounds or biological materials. The application of the procedure in automatic isotope gas analysis of organic materials labelled with tritium will be described in subsequent papers (Parts II and III). (T.G.)

  3. Preparation of tritium labelled synthanecine A and its bis-N-ethylcarbamate

    International Nuclear Information System (INIS)

    Mattocks, A.R.

    1982-01-01

    A procedure is described for incorporating tritium into the 3-CH 2 side chain of synthanecine A, and preparing the carbamate, 2,3-bis-N-ethylcarbamoyloxymethyl-1-methyl-3-pyrroline, a hepatotoxic pyrrolizidine alkaloid analogue. The pyrrolizidine amino alcohol, retronecine, can be tritium labelled in a similar way. (author)

  4. Tritium operating safety seminar, Los Alamos, New Mexico, July 30, 1975

    International Nuclear Information System (INIS)

    1976-03-01

    A seminar for the exchange of information on tritium operating and safety problems was held at the Los Alamos Scientific Laboratory. The topics discussed are: (1) material use (tubing, lubricants, valves, seals, etc.); (2) hardware selection (valves, fittings, pumps, etc.); (3) biological effects; (4) high pressure; (5) operating procedures (high pressure tritium experiment at LLL); (6) incidents; and (7) emergency planning

  5. Preparation of tritium labelled synthanecine A and its bis-N-ethylcarbamate

    Energy Technology Data Exchange (ETDEWEB)

    Mattocks, A.R. (Medical Research Council, Carshalton (UK))

    1982-04-01

    A procedure is described for incorporating tritium into the 3-CH/sub 2/ side chain of synthanecine A, and preparing the carbamate, 2,3-bis-N-ethylcarbamoyloxymethyl-1-methyl-3-pyrroline, a hepatotoxic pyrrolizidine alkaloid analogue. The pyrrolizidine amino alcohol, retronecine, can be tritium labelled in a similar way.

  6. 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

  7. Effective tritium processing using polyimide films

    International Nuclear Information System (INIS)

    Hayashi, T.; Okuno, K.; Ishida, T.; Yamada, M.; Suzuki, T.

    1998-01-01

    Applying a gas separation membrane module of polyimide hollow fiber films, a new tritium removal system has been studied and designed to develop a more compact and cost-effective system than the conventional type of catalytic reactors and molecular sieves dryers. The recent investigations are focused on the development of a more effective membrane module, specifically, an increase in the processing capacity for a unit module. One idea is to purge the permeated side of the module by using a small part of the bleed flow as a counter-current flow. Another idea is to apply a new polyimide membrane module (Φ 0.1 x 1.8 m) with 5 times larger permeability of N 2 (0.24 std. m 3 h -1 atm -1 ) than the original one, though the selectivity (permeability ratio of H 2 /N 2 : 80) is reduced by about a half. The results show that the purging effect improves the module capacity to be 3 times larger and the new membrane has almost 5 times larger capacity under reasonable operation conditions with the same tritium decontamination ability. The total capacity of a unit module is being improved by more than 10 times. Using the recent results, a case design of the membrane detritiation system is discussed for an application to the ITER scale tritium facility. (orig.)

  8. Existence and lifetime of laser fusion pellets containing tritium

    International Nuclear Information System (INIS)

    Devaney, J.J.

    1979-05-01

    Cryogenic pellets containing significant amounts of solid tritium cannot be maintained in a pure vacuum for longer than (typically) some tens of seconds because radiative cooling at low temperatures is inefficient. The steady state temperatures in typical one- and two-shell pellet designs both in vacuum and with external cooling, as well as the lifetimes of pellets following cooling removal, are calculated

  9. Shipment and Storage Containers for Tritium Production Transportation Casks

    International Nuclear Information System (INIS)

    Massey, W.M.

    1998-04-01

    The need for a shipping and storage container for the Tritium production transportation casks is addressed in this report. It is concluded that a shipping and storage container is not required. A recommendation is made to eliminate the requirement for this container because structural support and inerting requirements can be satisfied completely by the cask with a removable basket

  10. Laser separation of hydrogen isotopes: Tritium-from-deuterium recovery

    International Nuclear Information System (INIS)

    Magnotta, F.; Herman, I.P.; Aldridge, F.T.; Maienschein, J.L.

    1984-01-01

    Single-step enrichment factors exceeding 15,000 have been observed in the removal of tritium-from-deuterium by 12 μm laser multiple-photon dissociation of chloroform. The photochemistry and photophysics of this process is discussed along with prospects for implementation of this method in practical heavy water reactor detritiation. 7 refs., 7 figs., 1 tab

  11. 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

  12. TSTA loop operation with 100 grams-level of tritium

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Hirata, Shingo; Naito, Taisei

    1988-10-01

    The first loop operation tests of Tritium Systems Test Assembly (TSTA) with 100 grams-level of tritium were carried out at Los Alamos National Laboratory(LANL) on June and July, 1987. The tests were one of the milestones for TSTA goal scheduled in June, 1987 through June, 1988. The objectives were (i) to operate TSTA process loop composed of tritium supply system, fuel gas purification system, hydrogen isotope separation system, etc, (ii) to demonstrate TSTA safety subsystems such as secondary containment system, tritium waste treatment system and tritium monitoring system, and (iii) to accumulate handling experience of a large amount of tritium. This report describes the plan and procedures of the milestone run done in June and the summary results especially on the safety aspects. Analysis of the emergency shutdown of the process loop, which happened in the June run, is also reported. A brief description of the process and safety subsystems as well as the summary of the TSTA safety analysis report is included. (author)

  13. Recent progress of China HCCB TBM tritium system

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Deli, E-mail: luodeli2005@hotmail.com; Huang, Guoqiang; Huang, Zhiyong; Qin, Cheng; Song, Jiangfeng; He, Kanghao; Chen, Chang’an; Zhang, Guikai; Fu, Jun; Yao, Yong; An, Yongtao

    2016-11-01

    Highlights: • Comparing with our previous design, improvements have been made according to the up-to-date experiments and simulations: (1) The palladium alloy tube in the previous design is now removed in the upgraded one and the cryogenic molecular sieve bed is replaced by the getter bed to reduce tritium inventory; (2) Hot metal reduction bed is relocated from T-Plant to Port Cell; (3) TAS is now integrated into TES. • The proposed coolant purification is based on catalytic oxidation and molecular sieve bed adsorption for tritium removal, as well as hot metal adsorption for the elimination of non-tritium gaseous impurities. Some operation parameters and functional components are improved. The interface with the high pressure HCS and other plant systems was incorporated taking into account of the requirement from the ITER port management group meetings. - Abstract: China tritium system including Tritium Extraction System (TES) with Tritium Accountancy System (TAS) integrated in and Coolant Purification System (CPS), which is subordinate to Helium Coolant System (HCS), is of great importance for China Helium Cooled Ceramic Breeder Test Blanket Module (CN HCCB TBM). The purge gas (99.9% He + 0.1% H{sub 2}) carrying Q{sub 2}O (Q = H, D, T) and Q{sub 2} from Li{sub 4}SiO{sub 4} ceramic breeder flows through the reduction bed where Q{sub 2}O is reduced into Q{sub 2} and then absorbed by the getter bed. The HT/HTO ratio and the total tritium are determined by TAS. Catalytic oxidation combines with molecular sieve absorption and hot metal purification are applied to remove tritium and other impurities in helium coolant. A loop including depressurization, helium-sweeping assisted thermal desorption, and cold trapping for the regeneration of saturated molecular sieve bed until the concentration of the desorbed Q{sub 2}O is reduced to an acceptable level. This paper introduces the recent progress of China tritium system including updated conceptual designs of TES and

  14. Tritium handling and processing experience at TSTA

    International Nuclear Information System (INIS)

    Anderson, J.L.; Okuno, K.

    1994-01-01

    In 1987, the Japan Atomic Energy Research Institute (JAERI) and the US Department of Energy (DOE) signed a collaborative agreement (Annex IV) for the joint funding and operation of the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory (LANL) for a five year period ending June, 1992. After this initial five year collaboration, the Annex IV agreement was extended for another two year period ending June, 1994. During the first five years, a number of the integrated process loop tests of TSTA were conducted, as well as off-line testing of TSTA subsystems. During integrated loop testing the vacuum system, fuel cleanup systems, isotope separation system, transfer pumping system and gas analysis system, are interconnected and tested using 100 g-inventories of tritium to demonstrate steady-state operation of a tritium fuel processing cycle for a fusion reactor. These tests have resulted in a number of significant accomplishments and an experience data base on research, development and operation of the fuel processing system. One of the most significant accomplishments during the initial five year period was the continuous operation of the fuel processing loop for 25 days. During this 25-day extended operation, both the JAERI fuel cleanup system (J-FCU) and the original TSTA fuel cleanup system (FCU) were operated under similar conditions of flow, pressure, and impurity content of the DT gas. Both fuel cleanup systems were demonstrated to provide adequate impurity removal for plasma exhaust gas processing. The isotope separation system was operated continuously, producing pure tritium while rejecting protium as an impurity

  15. 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

  16. 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.)

  17. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-15

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

  19. Radiation doses to lungs and whole body from use of tritium in luminous paint industry

    International Nuclear Information System (INIS)

    Rudran, K.

    1988-01-01

    The radiation dose to persons exposed to tritium in the luminous paint industry is reported. The biological half-life of labile tritium is observed to be 7 to 10 days. There is evidence of exposure of lung tissue from tritium labelled polystyrene deposited in the pulmonary region and of soft tissue from organically bound tritium. Delayed excretion of labile tritium in urine following removal of the individuals from tritium handling, presence of tritium in organic constituents of blood and urine, and presence of non-volatile tritium in faecal excretion have been verified. From in vitro studies using fresh bovine serum, solubilisation half-life of tritium from the labelled paint is estimated to be 35 to 70 days after the initial fast clearance. Probable annual doses to the whole body, soft tissue and lungs under the prevailing working conditions have been estimated from the urinary and faecal excretion data. It is revealed that the actual values thus estimated are likely to exceed the values estimated by the conventional technique based on urine analysis for tritiated water. (author)

  20. 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.)

  1. 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

  2. Development of tritium plant system for fusion reactors. Achievements in the 14-year US-Japan collaboration

    International Nuclear Information System (INIS)

    Nishi, Masataka; Yamanishi, Toshihiko; Shu, Wataru

    2003-01-01

    Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safe-operation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore, distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work. (author)

  3. Development of Tritium Plant System for Fusion Reactors - Achievements in the 14-year US-Japan Collaboration -

    Science.gov (United States)

    Nishi, Masataka; Yamanishi, Toshihiko; Shu, Wataru

    Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safeoperation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore,distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work.

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

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

    International Nuclear Information System (INIS)

    Forrester, R.C. III; Watson, J.S.

    1975-01-01

    Recent experimental successes have generated increased interest in the development of thermonuclear reactors as power sources for the future. This paper examines tritium containment problems posed by an operating CTR and sets forth some processing schemes currently being evaluated at the Oak Ridge National Laboratory. An appreciation of the CTR tritium management problem can best be realized by recalling that tritium production rates for various fission reactors range from 2 x 10 4 to 9 x 10 5 Ci/yr per 1000 MW(e). Present estimates of tritium production in a CTR blanket exceed 10 9 Ci/yr for the same level of power generation, and tritium process systems may handle 10 to 20 times that amount. Tritium's high permeability through most materials of construction at high temperatures makes secondary containment mandatory for most piping. Processing of these containment atmospheres will probably involve conversion of the tritium to a nonpermeating form (T 2 O) followed by trapping on conventional beds of desiccant material. In a similar fashion, all purge streams and process fluid vent gases will be subjected to tritium recovery prior to atmospheric release. Two tritium process systems will be required, one to recover tritium produced by breeding in the blanket and another to recover unburned tritium in the plasma exhaust. Plasma exhaust processing will be unconventional since the exhaust gas pressure will lie between 10 -3 and 10 -6 torr. Treatment of this gas stream will entail the removal of small quantities of protium and helium from a much larger deuterium-tritium mixture which will be recycled. (U.S.)

  10. Tritium control in reprocessing plants

    International Nuclear Information System (INIS)

    Goumondy, J.P.; Miquel, P.

    1977-01-01

    There is a danger that the T which is formed in water reactors will prove detrimental to the environment over the next few years, and studies have been undertaken to develop techniques to contain and process it where possible. In order to retain T, which is present largely in the fuel and on the possible to adapt for use in the conventional design of reprocessing plant. In this process T is maintained in the form of an aqueous solution in the high-active area of the plant. Control is achieved by restricting as far as possible the ingress of non-tritiated water into this area, and by setting up a tritiated water barrier at the first U and Pu extraction stage, stripping the tritium-containing solvent at that point with ordinary water. In this way the T can be extracted in a small volume of water with a view to intermediate storage, disposal at sea additional processing to remove the T from the water. Experiments carried out so far have demonstrated the effectiveness of the T barrier and have shown what equipment would be required for the application of the process in new reprocessing plants. (orig.) [de

  11. 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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  13. 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

  14. 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

  15. Process and device for stage by stage enrichment of deuterium and/or tritium in a material suitable for isotope exchange of deuterium and tritium with hydrogen

    International Nuclear Information System (INIS)

    Iniotakis, N.; Decken, C.B. von der.

    1983-01-01

    Water containing deuterium and/or tritium is first introduced into a carrier gas flow and reduced for the stage by stage enrichment of deuterium and/or tritium. A hydrogen partial pressure of a maximum of 100 millibar is set in the carrier gas flow. The carrier gas flow is taken along the primary side of an exchange wall suitable for the permeation of hydrogen, and a further carrier gas flow flows on its secondary side, which contains water or hydrogen. Reaction products formed after isotope exchange of deuterium and/or tritium with hydrogen are removed by the secondary carrier gas flow. (orig./HP) [de

  16. Some recent changes in tritium handling and control at Mound Laboratory

    International Nuclear Information System (INIS)

    Rhinehammer, T.B.

    1976-01-01

    Significant reductions in tritium effluents and personnel exposures at Mound Laboratory have been made during the past 5 yr. Yearly effluents are less than 3 percent of former levels and personnel exposures have been reduced by a factor of 300. Several recent changes which have contributed to these reductions include lowered tritium levels in gloveboxes, and the efficiency and capacity of Mound's new effluent removal system. Personnel exposures have been reduced dramatically by changing to precious metal catalytic converters or oxidizers for use with the glovebox gas purification system. Unlike some former systems using hot copper or proprietary reactants for oxygen removal, a catalyst provides very effective removal of both oxygen and tritium. Both oxygen and tritium can be monitored and, if necessary, increments of hydrogen in argon can be added until the oxygen level is brought down to the desired value

  17. 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

  18. 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)

  19. 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)

  20. A programmable autosampler for a field deployable tritium analysis system

    International Nuclear Information System (INIS)

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

    1996-01-01

    Researchers in the Environmental Technology Section of the Savannah River Technology Center, in cooperation with Sampling Systems, Inc. are developing a fully programmable, remotely operated, fixed volume, automatic sampler for use with the field deployable tritium analysis system currently under development at U. of GA's Center for Applied Isotope Studies. The sampler will collect a limited-volume sample and perform on-line sample purification for tritium analyses from multiple collection sites. Pneumatically operated stainless steel samplers operate satisfactorily upon remote activation. The one-step purification system removes all impurities with interfere with tritium analysis by liquid scintillation. Field testing has confirmed system operation. The autosampler may act as a stand-alone device and is enclosed in a rugged, field-portable case with wheels. The system weighs about 40 lbs

  1. Concept of a tritium extraction facility for a reprocessing plant

    International Nuclear Information System (INIS)

    Tunaboylu, K.; Paulovic, M.; Ulrich, D.

    1991-01-01

    There are several alternatives for reducing the release of tritium to the environment originating from the wastewater of a reprocessing plant. Such alternatives, which are applicable for sites not located by the sea or by large rivers, are limited to either injection of tritiated wastewater into suitable deep geological formations, or final disposal into a deep underground repository after adequate treatment similar to other low and intermediate active waste. Removal of tritium from the wastewater by enrichment represents a further feasible option of the second alternative, which allows reduction of the huge volume of tritiated water to be treated before disposal. A significant volume reduction increases the safety of the subsequent steps such as transport, interim storage and final disposal of tritiated waste, furthermore, decreases the corresponding overall waste management cost. The projected Wackersdorf reprocessing plant has been considered as a reference for assessing the permitted tritium releases and other site characteristics. (orig.)

  2. 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

  3. Procedures for the characterization of the detritiation of steel, Inconel and graphite

    International Nuclear Information System (INIS)

    Poletiko, C.; Trabuc, P.; Durand, J.; Tormos, B.; Pignoly, L.

    2006-01-01

    Due to its high diffusivity and different trapping phenomena, tritium is present in materials, such as steel or Inconel that are in use in different parts of a nuclear power reactor, or even in graphite which is present in fusion reactor or in future HTR. From waste management point of view, it is necessary to know as accurately as possible the tritium inventory in such materials before disposal. Moreover the knowledge of tritium species (HTO or HT, etc) is also a significant information in case of detritiation prior to storage, since countries regulation already limit tritium contents and releases. Three different strategies for tritiated waste management are foreseen: the first one is based upon a storage with confined packages, the second one is waiting for radioactive decay while the third one consists in the application of detritiation processes. Studies have been performed to determine different processes that could be used for tritium removal. The aim of this paper was, to study, at laboratory scale, different detritiation procedures which may be used for stainless steel, Inconel and carbon materials. Thermal detritiation kinetics till 1300 K has been studied under various atmospheres; full chemical dissolution of samples has also been performed for steel, Inconel and graphite, this to perfectly know the tritium content in such matrices. A particular attention must be applied to Inconel, the main reason is linked to the presence of titanium which is supposed to be a tritium trap. Finally, a study of tritium content in steel and Inconel layers has also been made, to learn about the tritium behaviour. All results are given, allowing the possibility to take a decision either for detritiation procedure or storage conditions. The main result is that thermal out-gassing for steel and graphite enables higher than 95 % tritium extraction from the bulk at temperature in the range of 600 K, without any material destruction under hi-tech gas (Ar + 5% volume H 2 ), on

  4. Comparison of procedures for immediate reconstruction of large osseous defects resulting from removal of a single tooth to prepare for insertion of an endosseous implant after healing

    NARCIS (Netherlands)

    Raghoebar, G. M.; Slater, J. J. H.; den Hartog, L.; Meijer, H. J. A.; Vissink, A.

    This study evaluated the treatment outcome of immediate reconstruction of 45 large osseous defects resulting from removal of a single tooth with a 1:2 mixture of Bio-Oss(R) and autologous tuberosity bone, and three different procedures for soft tissue closing (Bio-Gide(R) membrane, connective tissue

  5. Tritium systems test assembly quality assurance program

    International Nuclear Information System (INIS)

    Kerstiens, F.L.; Wilhelm, R.C.

    1986-07-01

    A quality assurance program should establish the planned and systematic actions necessary to provide adequate confidence that fusion facilities and their subsystems will perform satisfactorily in service. The Tritium Systems Test Assembly (TSTA) Quality Assurance Program has been designed to assure that the designs, tests, data, and interpretive reports developed at TSTA are valid, accurate, and consistent with formally specified procedures and reviews. The quality consideration in all TSTA activities is directed toward the early detection of quality problems, coupled with timely and positive disposition and corrective action

  6. Methane formation in tritium gas exposed to stainless steel

    International Nuclear Information System (INIS)

    Morris, G.A.

    1977-01-01

    Tests were performed to determine the effect cleanliness of a surface exposed to tritium gas had on methane formation. These tests performed on 304 stainless steel vessels, cleaned in various ways, showed that the methane formation was reduced by the use of various cleaning procedures

  7. Experimental investigation of buried tritium in plant and animal tissues

    International Nuclear Information System (INIS)

    Kim, S. B.; Workman, W. J. G.; Davis, P. A.

    2008-01-01

    Buried exchangeable tritium appears as part of organically bound tritium (OBT) in the traditional experimental determination of OBT. Since buried tritium quickly exchanges with hydrogen atoms in the body following ingestion, assuming that it is part of OBT rather than part of tritiated water (HTO) could result in a significant overestimate of the ingestion dose. This paper documents an experimental investigation into the existence, amount and significance of buried tritium in plant and fish samples. OBT concentrations in the samples were determined in the traditional way and also following denaturing with five chemical solutions that break down large molecules and expose buried tritium to exchange with free hydrogen atoms. A comparison of the OBT concentrations before and after denaturing, together with the concentration of HTO in the supernatant obtained after denaturing, suggests that buried OBT may exist but makes up less than 5% of the OBT concentration in plants and at most 20% of the OBT concentration in fish. The effects of rinse time and rinse water volumes were investigated to optimize the removal of exchangeable OBT from the samples. (authors)

  8. How to Surgically Remove the Permanent Mesh Ring after the Onstep Procedure for Alleviation of Chronic Pain following Inguinal Hernia Repair

    Directory of Open Access Journals (Sweden)

    Stina Öberg

    2016-01-01

    Full Text Available A promising open inguinal hernia operation called Onstep was developed in 2005. The technique is without sutures to the surrounding tissue, causing minimal tension. A specific mesh is used with a memory recoil ring in the border, which may cause pain superficial to the lateral part of the mesh for slender patients. The aim of this study was to illustrate an easy procedure that alleviates/removes the pain. A male patient had persistent pain six months after the Onstep operation and therefore had a ring removal operation. The procedure is presented as a video and a protocol. At the eleven-month follow-up, the patient was free of pain, without a recurrence. It is advised to wait some months after the initial hernia repair before removing the ring, since the mesh needs time to become well integrated into the surrounding tissue. The operation is safe and easy to perform, which is demonstrated in a video.

  9. Catalyst study for the decontamination of glove-box atmospheres containing tritium at MPC levels

    International Nuclear Information System (INIS)

    Chobot, J.; Montel, J.; Sannier, J.

    1988-01-01

    The BEATRICE loop was designed for studying the conversion of tritium at very low activity levels using catalytic oxidation followed by water trapping. The purpose is to study kinetic parameters required for the design of the NET tritium cleanup system with the two main objectives to operate without isotopic swamping and to determine the ability of efficient conversion at room temperature. From experiments carried out between 20 and 250 0 C it is concluded that two palladium/alumina and platinum/alumina catalysts are very efficient in removing tritium from contaminated gas mixtures down to a few MPC levels without isotopic swamping and even at room temperature. However at room temperature, in relation to tritium species trapped on the catalyst surface a progressive deactivation with time occurs. This phenomenon may be a concern for process efficiency and tritium inventory and regeneration conditions have to be determined in order to demonstrate industrial feasibility of operating at room temperature

  10. Catalyst study for the decontamination of glove-boxe atmospheres containing tritium at MPC levels

    International Nuclear Information System (INIS)

    Chabot, J.; Montel, J.; Sannier, J.

    1988-01-01

    The BEATRICE loop was designed for studying the conversion of tritium at very low activity levels using catalytic oxidation followed by water trapping. The purpose is to study kinetic parameters required for the design of the NET tritium clean-up system with the two main objectives to operate without isotopic swamping and to determine the ability of efficient conversion at room temperature. From experiments carried out between 20 and 250 0 C it is concluded that two palladium/alumina and platinum/alumina catalysts are very efficient in removing tritium from contaminated gas mixtures down to a few MPC levels without isotopic swamping and even at room temperature. However at room temperature, in relation to tritium species trapped on the catalyst surface a progressive deactivation with time occurs. This phenomenon may be a concern for process efficiency and tritium inventory and best regeneration conditions have to be determined in order to demonstrate industrial feasibility of operating at room temperature

  11. System for deuterium-tritium mixture filling the working chamber of a dense plasma focus device

    International Nuclear Information System (INIS)

    Bondar', A.I.; Vyskubov, V.P.; Gerasimov, S.A.

    1981-01-01

    A gas-vacuum system designed for filling the gas-discharge chamber of a plasma focus device with equal-coaponent deuterium-tritium mixture is described. The system consists of a unit for gaseous mixture prepa ration and a unit for mixture absorption and device evacuation. The system provides the gaseous mixture purification of O 2 and N 2 impurities. Final tritium content in the gas-discharge chamber after tritium removal is not greater than 2x10 8 Bq/l. Tritium content in a sealed box in which the device is placed does not exceed 30 Bq/l that is less than limiting safe value. The conclusion is made that the described system design gives an opportunity to begin experimental studies at plasma focus devices with deuterium-tritium mixture [ru

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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.)

  19. Health physics manual of good practices for tritium facilities

    International Nuclear Information System (INIS)

    Blauvelt, R.K.; Deaton, M.R.; Gill, J.T.

    1991-12-01

    The purpose of this document is to provide written guidance defining the generally accepted good practices in use at Department of Energy (DOE) tritium facilities. A open-quotes good practiceclose quotes is an action, policy, or procedure that enhances the radiation protection program at a DOE site. The information selected for inclusion in this document should help readers achieve an understanding of the key radiation protection issues at tritium facilities and provide guidance as to what characterizes excellence from a radiation protection point of view. The ALARA (As Low as Reasonable Achievable) program at DOE sites should be based, in part, on following the good practices that apply to their operations

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. 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

  6. Quantification of tritium ''heels'' and isotope exchange mechanisms in La-Ni-Al tritides

    International Nuclear Information System (INIS)

    Wermer, J.R.

    1992-01-01

    Formation of tritium heels in LANA (LaNi 5-x Al x ) 0.30 (x=0.30) and 0.75 tritides was quantified; size of the heel is dependent on storage and processing conditions. Absorption-desorption cycling of the tritide beds mitigates formation of the tritium heel and can reduce its size. The higher pressure material LANA 0.30 showed slower heel formation than LANA 0.75; this allows more tritium to be removed at the maximum processing temperature. In plant application, LANA 0.30 beds are used as compressors; except during compressor operation, their aging will be very slow. Tritium heel removal by D exchange was demonstrated. Absorption-desorption cycling during an exchange cycle does not improve the exchange efficiency. Residual tritium can be removed to very low levels. For a tritide bed scheduled for removal from the process, a final tritium level can be estimated based on the number of D exchange cycles. 13 refs, 8 figs, 6 tabs

  7. 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

  8. 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

  9. 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)

  10. Measurement of organically bound tritium in urine and feces

    International Nuclear Information System (INIS)

    Trivedi, A.; Duong, T.; Leon, J.W.; Linauskas, S.H.

    1993-11-01

    A bioassay method was developed for directly measuring organically bound tritium (OBT) in urine and feces. Samples first undergo low-temperature distillation and vacuum separation to isolate tritiated water (HTO) and exchangeable tritium. This is followed by converting the non-exchangeable tritium (i.e., OBT) into HTO through oxygen combustion. The method was investigated to: optimise the sample preparation procedures; establish OBT recovery (64% ± 7% for urine and 71% ± 8% for feces); and, determine the detection limit for OBT in urine (0.3 Bq · g -1 ) and feces (5 Bq · g -1 ). The method was evaluated for error sources that are associated with the exchange between HTO and OBT. It is concluded that this bioassay method can reliably measure OBT in urine and feces within the range of ± 10%

  11. Computer control of the TFTR tritium storage and delivery system

    International Nuclear Information System (INIS)

    Youssef, N.; Phillips, H.; Yemin, L.; Dong, J.; Pierce, C.

    1980-01-01

    The Tritium Storage and Delivery System (TSDS) will deliver to the torus the required tritium gas in precisely controlled injection profiles. This system will utilize advanced Central Instrumentation, Control and Data Acquisition (CICADA) computer-control techniques, in normal and malfunction-recovery modes of operation. The control scheme of the TSDS is built of three main control scenarios. An operating mode defines the permissives, sequence and path of a process during each scenario. The computerized control of the TSDS has four distinct advantages: (1) versatile control with fast response times both for tritium gas generation and for gas injection into the torus; (2) ease of selecting the proper operating modes of a control scenario, (3) ease of operation without disturbing the multiple levels of containment, and (4) simple fast trouble shooting of system malfunction utilizing programmed procedures and on-line diagnosis. The TSDS has both remote nd local control capability

  12. 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.

  13. Conceptual design of Tritium Extraction System for the European HCPB Test Blanket Module

    International Nuclear Information System (INIS)

    Ciampichetti, A.; Nitti, F.S.; Aiello, A.; Ricapito, I.; Liger, K.; Demange, D.; Sedano, L.; Moreno, C.; Succi, M.

    2012-01-01

    Highlights: ► HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM) to be tested in ITER. ► Tritium extraction by gas purging, removal and transfer to the Tritium Plant. ► Conceptual design of TES and revision of the previous configuration. ► Main components: adsorption column, ZrCo getter beds and PERMCAT reactor. - Abstract: The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80 bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported.

  14. 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.)

  15. Preparation of paraherquamide labeled with deueterium or tritium

    Energy Technology Data Exchange (ETDEWEB)

    Blizzard, T.A.; Rosegay, A.; Mrozik, H.; Fisher, M.H. (Merck Sharp and Dohme Research Labs., Rahway, NJ (United States))

    1990-04-01

    Deprotonation of paraherquamide (1a) at C-24 and subsequent deuteration (or tritiation) is described. The procedure afforded 24-[sup 2]H-paraherquamide (1b) with 66% deuterium incorporation at C-24. Modification of the deuteration procedure to allow the introduction of tritium resulted in the preparation of 24-[sup 3]H-paraherquamide (1c) with specific activity 3.7Ci/mmol. (author).

  16. Fuel cleanup system for the tritium systems test assembly: design and experiments

    International Nuclear Information System (INIS)

    Kerr, E.C.; Bartlit, J.R.; Sherman, R.H.

    1980-01-01

    A major subsystem of the Tritium Systems Test Assembly is the Fuel Cleanup System (FCU) whose functons are to: (1) remove impurities in the form of argon and tritiated methane, water, and ammonia from the reactor exhaust stream and (2) recover tritium for reuse from the tritiated impurities. To do this, a hybrid cleanup system has been designed which utilizes and will test concurrently two differing technologies - one based on disposable, hot metal (U and Ti) getter beds and a second based on regenerable cryogenic asdorption beds followed by catalytic oxidation of impurities to DTO and stackable gases and freezout of the resultant DTO to recover essentially all tritium for reuse

  17. Organically bound tritium analysis in environmental samples

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  18. Tritium waste control: July--September 1978

    International Nuclear Information System (INIS)

    1978-01-01

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

  19. 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)

  20. 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)

  1. Assessment of a chemical getter for scavenging tritium from an inert gas

    International Nuclear Information System (INIS)

    Maienschein, J.L.

    1976-01-01

    Results are presented of a study aimed at determining the feasibility of using chemical getter beds to scavenge tritium from inert gases. Two types of getter bed, fixed and fluidized, were considered, using cerium as the getter material. Mathematical-modeling results and capital-cost estimates indicate that not only is the gettering approach technically feasible, it could lead to considerable cost savings over catalytic oxidation, the tritium-removal method traditionally used

  2. An analysis of the tritium content in fish from Upper Three Runs Creek

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.

    1991-01-01

    In November of 1988 the F/H-area effluent treatment facility (ETF) began releasing treated waste water to Upper Three Runs Creek. Previous to that time, there has been minimal discharge of plant waste water to this tributary of the Savannah River. The ETF is designed to remove the toxic and radioactive waste materials from the effluent stream and to meet the discharge limits of the South Carolina Department of Health and Environmental Control (SCDHEC). The only radioactive nuclide not removed by the process is tritium. Tritium, an isotope of hydrogen, is chemically associated with the water molecules in the waste stream and can not be economically removed at this time. The objective of this study was to determine the relationship between the concentration of tritium in the stream water and the concentration of tritium in the fish. Fish collections were made at two locations. The most upstream location was 50 meters downstream from the SRS Road C bridge. This is immediately downstream of the effluent discharge pipe from the ETF. The other location was at the bridge of SRS Road A (SC Highway 125). The water is removed from the fish by freeze drying under vacuum. This study suggests that, on the average, the tritium concentration of fish in Upper Three Runs Creek will be in equilibrium with the tritium in the water of the creek. The water in the fish comes into equilibrium with the water in the stream quite rapidly and it is quite likely that any single fish sampled will be higher or lower in tritium content of an integrated water sample, such as those collected by the Environmental Monitoring samplers. Both the time of sampling and the sampling of a sufficient number of fish is important in obtaining an accurate estimate of the average tritium concentration in the tissue water of the fish

  3. Simplified fuel cycle tritium inventory model for systems studies -- An illustrative example with an optimized cryopump exhaust system

    International Nuclear Information System (INIS)

    Kuan, W.; Ho, S.K.

    1995-01-01

    It is desirable to incorporate safety constraints due to fuel cycle tritium inventories into tokamak reactor design optimization. An optimal scenario to minimize tritium inventories without much degradation of plasma performance can be defined for each tritium processing component. In this work, the computer code TRUFFLES is used exclusively to obtain numerical data for a simplified model to be used for systems studies. As an illustration, the cryopump plasma exhaust subsystem is examined in detail for optimization purposes. This optimization procedure will then be used to further reduce its window of operation and provide constraints on the data used for the simplified tritium inventory model

  4. 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.)

  5. Mutagenic effect of tritium on DNA of Drosophila melanogaster. Comprehensive performance report, December 15, 1985-June 1, 1988

    International Nuclear Information System (INIS)

    Lee, W.R.

    1988-01-01

    The results of the RBE determination of tritium to cobalt-69 gamma radiation along with a description of methods of treatment and dose determination are given. Using the described procedures for exposing Drosophila to tritiated water, the authors induced mutations by tritium beta radiation and recovered them at the Adh locus

  6. 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

  7. 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)

  8. 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

  9. Radioecological studies of tritium movement in a tropical rain forest

    Energy Technology Data Exchange (ETDEWEB)

    Martin, J R; Jordan, C F; Koranda, J J; Kline, J R [Bio-Medical Division, Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-01

    injected input pulse due to the continuous root uptake of tritium as the diffuse peak moved down into the soil past the root zone. Tritium was removed from the plot by transpiration and by interflow. Using transpiration rates from the previous experiment, rainfall records, tree density data and other measurements, average transpiration for the Puerto Rico rainforest was computed to be 3.64 kg/m{sup 2}/day. The effective capacity of the soil compartment was calculated to be 280 {+-} 12 kg/m{sup 2}. In the final experiment, tritiated water was injected directly into several species of successional trees in a cleared plot. After several weeks, the trees were harvested and aliquots selected for bound tritium assay. The amount of tritium incorporated into the tissue was about 0.1 percent of the total amount applied to the tree. Based on all experimental data, the distribution of tritium from a simulated rainout following a one megaton thermonuclear detonation is presented for a climax tropical rainforest and for successional vegetation. The fraction of input tritium remaining in each compartment as a function of time is tabulated. The residence time for each of the compartments determines the persistence of tritium deposited in a tropical ecosystem. (author)

  10. Radioecological studies of tritium movement in a tropical rain forest

    International Nuclear Information System (INIS)

    Martin, J.R.; Jordan, C.F.; Koranda, J.J.; Kline, J.R.

    1970-01-01

    pulse due to the continuous root uptake of tritium as the diffuse peak moved down into the soil past the root zone. Tritium was removed from the plot by transpiration and by interflow. Using transpiration rates from the previous experiment, rainfall records, tree density data and other measurements, average transpiration for the Puerto Rico rainforest was computed to be 3.64 kg/m 2 /day. The effective capacity of the soil compartment was calculated to be 280 ± 12 kg/m 2 . In the final experiment, tritiated water was injected directly into several species of successional trees in a cleared plot. After several weeks, the trees were harvested and aliquots selected for bound tritium assay. The amount of tritium incorporated into the tissue was about 0.1 percent of the total amount applied to the tree. Based on all experimental data, the distribution of tritium from a simulated rainout following a one megaton thermonuclear detonation is presented for a climax tropical rainforest and for successional vegetation. The fraction of input tritium remaining in each compartment as a function of time is tabulated. The residence time for each of the compartments determines the persistence of tritium deposited in a tropical ecosystem. (author)

  11. Tentative reference method for measurement of tritium in environmental waters. Environmental monitoring series

    International Nuclear Information System (INIS)

    1975-12-01

    A tentative reference method for the measurement of tritium in potable and nonpotable environmental water is described. Water samples are treated with sodium hydroxide and potassium permanganate and then a water fraction is separated from interferences by distillation. Two distillation procedures are described, a simple aqueous distillation for samples from potable water sources, and an aqueous-azeotropic-benzene distillation for nonpotable water sources. Alliquots of a designated distillate fraction are measured for tritium activity by liquid scintillation detection. Distillation recovery and counting efficiency factors are determined with tritium standards. Results are reported in picocuries per milliliter

  12. 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. (...

  13. 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)

  14. 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

  15. 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

  16. Automatic isotope gas analysis of tritium labelled organic materials Pt. 3

    International Nuclear Information System (INIS)

    Gacs, I.; Mlinko, S.; Payer, K.; Otvos, L.; Banfi, D.; Palagyi, T.

    1978-01-01

    An isotope analytical procedure and an automatic instrument developed for the determination of tritium in organic compounds and biological materials by internal gas counting are described. The sample is burnt in a stream of oxygen and the combustion products including water vapour carrying the tritium are led onto a column of molecular sieve-5A heated to 550 deg C. Tritium is retained temporarily on the column, then transferred into a stream of hydrogen by isotope exchange. After addition of butane, the tritiated hydrogen is led into an internal detector and enclosed there for radioactivity measurement. The procedure, providing quantitative recovery, is completed in five minutes. It is free of memory effect and suitable for the determination of tritium in a wide range of organic compounds and samples of biological origin. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

    A series of tests involving 10 h baking under the current ITER design conditions (240 deg. C with 933 Pa O{sub 2}) was performed using a cube of a carbon fiber composite tile that had been used in Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium burning operation. The removal rate of the codeposits was about 3 {mu}m/h near the surface and 0.9 {mu}m/h in the deeper region. Total amount of tritium released from the cube during 10 h baking was 202 MBq, while remaining tritium in the cube after baking was 403 MBq. Thus 10 h baking at 240 deg. C with 933 Pa O{sub 2} removed 1/3 of tritium from the cube. After 10 h baking, the tritium concentration on the cube surface also dropped by about 1/3. In addition, some tritium was released from another cube of the tile during baking at 240 deg. C in pure Ar, and a rapid increase of tritium release was observed when the purging gas was shifted from pure Ar to Ar-1%O{sub 2}. When a whole TFTR tile was baked in air at 350 deg. C for 1 h and then at 500 deg. C for 1 h, the ratios of tritium released were 53 and 47%, respectively. Oxygen reacted with carbon to produce carbon monoxide during baking in air.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

    A series of tests involving 10 h baking under the current ITER design conditions (240 deg. C with 933 Pa O 2 ) was performed using a cube of a carbon fiber composite tile that had been used in Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium burning operation. The removal rate of the codeposits was about 3 μm/h near the surface and 0.9 μm/h in the deeper region. Total amount of tritium released from the cube during 10 h baking was 202 MBq, while remaining tritium in the cube after baking was 403 MBq. Thus 10 h baking at 240 deg. C with 933 Pa O 2 removed 1/3 of tritium from the cube. After 10 h baking, the tritium concentration on the cube surface also dropped by about 1/3. In addition, some tritium was released from another cube of the tile during baking at 240 deg. C in pure Ar, and a rapid increase of tritium release was observed when the purging gas was shifted from pure Ar to Ar-1%O 2 . When a whole TFTR tile was baked in air at 350 deg. C for 1 h and then at 500 deg. C for 1 h, the ratios of tritium released were 53 and 47%, respectively. Oxygen reacted with carbon to produce carbon monoxide during baking in air

  2. Tritium separation from light and heavy water by bipolar electrolysis

    International Nuclear Information System (INIS)

    Ramey, D.W.; Petek, M.; Taylor, R.D.; Kobisk, E.H.; Ramey, J.; Sampson, C.A.

    1979-10-01

    Use of bipolar electrolysis with countercurrent electrolyte flow to separate hydrogen isotopes was investigated for the removal of tritium from light water effluents or from heavy water moderator. Deuterium-tritium and protium-tritium separation factors occurring on a Pd-25% Ag bipolar electrode were measured to be 2.05 to 2.16 and 11.6 to 12.4 respectively, at current densities between 0.21 and 0.50 A cm -2 , and at 35 to 90 0 C. Current densities up to 0.3 A cm -2 have been achieved in continuous operation, at 80 to 90 0 C, without significant gas formation on the bipolar electrodes. From the measured overvoltage at the bipolar electrodes and the electrolyte conductivity the power consumption per stage was calculated to be 3.0 kwh/kg H 2 O at 0.2 A cm -2 and 5.0 kwh/kg H 2 O at 0.5 A cm -2 current density, compared to 6.4 and 8.0 kwh/kg H 2 O for normal electrolysis. A mathematical model derived for hydrogen isotope separation by bipolar electrolysis, i.e., for a square cascade, accurately describes the results for protium-tritium separation in two laboratory scale, multistage experiments with countercurrent electrolyte flow; the measured tiritum concentration gradient through the cascade agreed with the calculated values

  3. Evaluation and mitigation of tritium memory in detritiation dryers

    International Nuclear Information System (INIS)

    Malara, C.; Ricapito, I.; Edwards, R.A.H.; Toci, F.

    1999-01-01

    In atmospheric detritiation, and other tritium processes, tritium is adsorbed on zeolites (molecular sieves) in the form of tritiated water. Regeneration removes almost all the physically adsorbed water, but a proportion remains permanently in the zeolite and binder structure as chemically bound water or hydroxyl groups. Exchange between adsorbed water and bound water means that tritiated water is retained in the structure after regeneration. At the end of its life, the zeolite therefore constitutes a tritiated waste. Furthermore, if an atmosphere detritiation dryer (ADD) gets highly contaminated from a tritium spill, retained tritium contaminates both the small amount of vapour leaving the bed during the next drying cycle, and the water produced in the subsequent regeneration. This report first describes experiments to measure the tritiated water retained in a 5A zeolite bed after standard regeneration treatments, and then investigates strategies to mitigate the effect: more thorough regeneration and isotope swamping or elution. The effect of zeolite ageing after thermal cycling is also seen. (orig.)

  4. 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)

  5. 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

  6. Tritium Concentrations in Environmental Samples and Transpiration Rates from the Vicinity of Mary's Branch Creek and Background Areas, Barnwell, South Carolina, 2007-2009

    Science.gov (United States)

    Vroblesky, Don A.; Canova, Judy L.; Bradley, Paul M.; Landmeyer, James E.

    2009-01-01

    Tritium in groundwater from a low-level radioactive waste disposal facility near Barnwell, South Carolina, is discharging to Mary's Branch Creek. The U.S. Geological Survey conducted an investigation from 2007 to 2009 to examine the tritium concentration in trees and air samples near the creek and in background areas, in groundwater near the creek, and in surface water from the creek. Tritium was found in trees near the creek, but not in trees from background areas or from sites unlikely to be in direct root contact with tritium-contaminated groundwater. Tritium was found in groundwater near the creek and in the surface water of the creek. Analysis of tree material has the potential to be a useful tool in locating shallow tritium-contaminated groundwater. A tritium concentration of 1.4 million picocuries per liter was measured in shallow groundwater collected near a tulip poplar located in an area of tritium-contaminated groundwater discharge. Evapotranspiration rates from the tree and tritium concentrations in water extracted from tree cores indicate that during the summer, this tulip poplar may remove more than 17.1 million picocuries of tritium per day from the groundwater that otherwise would discharge to Mary's Branch Creek. Analysis of air samples near the tree showed no evidence that the transpirative release of tritium to the air created a vapor hazard in the forest.

  7. 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.)

  8. 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.).

  9. 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)

  10. Distribution and behavior of tritium in the Coolant-Salt Technology Facility

    International Nuclear Information System (INIS)

    Mays, G.T.; Smith, A.N.; Engel, J.R.

    1977-04-01

    A 1000-MW(e) Molten-Salt Breeder Reactor (MSBR) is expected to produce 2420 Ci/day of tritium. As much as 60 percent of the tritium produced may be transported to the reactor steam system (assuming no retention by the secondary coolant salt), where it would be released to the environment. Such a release rate would be unacceptable. Experiments were conducted in an engineering-scale facility--the Coolant-Salt Technology Facility (CSTF)--to examine the potential of sodium fluoroborate, the proposed coolant salt for an MSBR, for sequestering tritium. The salt was believed to contain chemical species capable of trapping tritium. A series of 5 experiments--3 transient and 2 steady-state experiments--was conducted from July of 1975 through June of 1976 where tritium was added to the CSTF. The CSTF circulated sodium fluoroborate at temperatures and pressures typical of MSBR operating conditions. Results from the experiments indicated that over 90 percent of tritium added at steady-state conditions was trapped by sodium fluoroborate and appeared in the off-gas system in a chemically combined (water-soluble) form and that a total of approximately 98 percent of the tritium added at steady-state conditions was removed through the off-gas system overall

  11. Synthesis of high specific activity tritium labelled [2-3H]-adenosine-5'-triphosphate

    International Nuclear Information System (INIS)

    Jaiswal, D.K.; Morimoto, H.; Trump, E.L.; Williams, P.G.; Wemmer, D.E.

    1996-01-01

    A procedure for high level tritium labelling at the C2-H position of adenosine 5'-triphosphate ([2- 3 H]-ATP, 1), based on the tritiodehalogenation reaction of 2-bromoadenosine 5'-triphosphate (2) has been elaborated. This precursor was prepared in a six-step synthesis from guanosine. The tritiodehalogenation of (2) for three hours over palladium oxide in phosphate buffer yielded tritium labelled ATP with high specific activity, in good chemical yield. (author)

  12. Radiation protection data sheets for the use of Tritium in unsealed sources

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    This radiation protection data sheet is intended for supervisors and staff in the different medical, hospital, pharmaceutical, university and industrial laboratories and departments where Tritium is handled, and also for all those involved in risk prevention in this field. It provides essential data on radiation protection measures during the use of Tritium in unsealed sources: physical characteristics, risk assessment, administrative procedures, recommendations, regulations and bibliography

  13. Predicted fate of tritium residuum from groundwater tracer experiments in the Amargosa Desert, southern Nevada

    International Nuclear Information System (INIS)

    Brikowski, T.

    1993-07-01

    Analytic solutions are used in this study to evaluate potential groundwater transport of tritium used in goundwater tracer tests southwest of the Nevada Test Site. Possible transport from this site is of interest because initial radionuclide concentrations were high and the site is close to goundwater discharge points (12 km). Anecdotal evidence indicates that 90 percent of these tracers were removed by pumping at the completion of the tests; this study examines the probable transport of the tracers with and without the removal. Classical dispersive transport analytic solutions are used, treating the tracer test as a point slug injection. Input parameters for the solutions were measured at the site, and consideration of parameter uncertainty is incorporated in the results. With removal of the tracer, the maximum expected region with above-Safe Drinking Water Act (40 CFR 121) concentrations of tritium extends 5 km from the injection point, and does not reach any sites of public access. Detectable tritium from the tests is likely to have reached the Ash Meadows fault zone, but flow along the fault probably diluted the tracer to below detection limits before arrival at springs along the fault. Arrival at the springs would have occurred 20 to 25 years after the tests. Without removal of the tracer, the solutions indicate that tritium concentrations just above Safe Drinking Water Act standards would have reached the Ash Meadows fault zone. In this case, detectable tritium might have been found in Devil's Hole or Longstreet Spring, the nearest points of possible public exposure

  14. 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)

  15. Enhancing radiolytic stability upon concentration of tritium-labeled pharmaceuticals utilizing centrifugal evaporation.

    Science.gov (United States)

    Marques, Rosemary; Helmy, Roy; Waterhouse, David

    2015-05-30

    Tritium radiopharmaceuticals are often used in drug development because of their desirable specific activity. The inherent instability of these radioactive tracers often leads to a requirement to purify prior to use. Purification methodologies such as preparative chromatography and solid/liquid extractions often utilize water as a solvent, which is not suitable for long-term storage and necessitates removal. Rotary evaporation has traditionally been utilized for the removal of this unwanted solvent, however, this method has been shown to lead to decomposition of the tritium species in some cases. Centrifugal evaporation is a milder concentration method which has been demonstrated to effectively remove solvents. In this study, we show that centrifugal evaporation leads to effective concentration of tritium samples without the decomposition typically observed by rotary evaporation. Copyright © 2015 John Wiley & Sons, Ltd.

  16. 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)

  17. 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.

  18. 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)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-10-01

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

  1. R and D of tritium technology as SHI (Sumitomo Heavy Industries)

    International Nuclear Information System (INIS)

    Yokogawa, N.

    1997-01-01

    Sumitomo Heavy Industries (SHI) participated in an R and D programme on tritium processing for the first time in 1967 by joining the advanced thermal reactor project. (The thermal reactor is cooled by light water and moderated by heavy water.) From that time SHI has developed various kind of tritium handling technologies. On the basis of cooperation with Sulzer (Sulzer Chemtech Ltd. Switzerland), SHI developed a system for removing waste water for fuel reprocessing plants by water distillation technology. In the field of fusion technology, SHI has developed a hydrogen isotope separation system by cryogenic distillation and thermal diffusion methods, and a tritium storage bed. Fundamental data required for the system design were obtained through the production and operation of the above prototype systems. Recently, SHI has also been taking part in the design and planning of ITER. In the future, along with ITER design, SHI will aim at developing tritium measuring technology. (author)

  2. Catalyst study for the decontamination of atmospheres containing few traces of tritium

    International Nuclear Information System (INIS)

    Chabot, J.; Montel, J.; Sannier, J.

    1988-01-01

    The conversion of tritium at very low activity level using catalytic oxidation followed by water trapping is studied in the loop BEATRICE in order to measure kinetic parameters required for the design of the NET tritium clean-up system. Two precious-metal catalysts (Pd/alumina and Pt/alumina) are very efficient in removing tritium from contaminated gas mixtures down to a few MPC level at low temperatures, without need of isotopic swamping. However at room temperature, the trapping of tritium species on the catalyst surface gives rise to a progressive deactivation with time. Best regeneration conditions have to be determined in order to demonstrate industrial feasibility of operating at low temperatures

  3. Tritium system for a tokamak reactor with a self-pumped limiter

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Sze, D.K.

    1986-01-01

    The self-pumping concept was proposed as a means of simplifying the impurity control system in a fusion reactor. The idea is to remove helium in-situ by trapping in freshly deposited metal surface layers of a limiter or divertor. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. Some of the key issues for this concept are the tritium inventory in the trapping material and the permeation of protium and recycling of tritium. These quantities are shown to be acceptable for the reference design. The tritium issues for a helium-cooled solid breeder reactor design with vanadium alloy as a structural material are also examined. Models are presented for tritium permeation and inventory calculation for structure materials with the effect of a thin layer of coating material

  4. Determination of tritium in wine and wine yeast samples

    International Nuclear Information System (INIS)

    Cotarlea, Monica-Ionela; Paunescu, Niculina; Galeriu, D.; Mocanu, N.; Margineanu, R.; Marin, G.

    1997-01-01

    A sensitive method for evaluating the tritium content in wine and wine yeast was applied to estimate tritium impact on the environment in the surrounding area of nuclear power plant Cernavoda, where the vineyards are part of representative agricultural ecosystem. Analytical procedures were developed to determine HTO in wine and wine yeast samples. The content of organic compounds affecting the LSC measurement is reduced by fractionating distillation for wine samples and azeotropic distillation followed by fractional distillation for wine yeast samples. Finally, the water samples obtained after fractional distillation were normally distilled with KMO 4 . The established procedures were successfully applied for wine and wine yeast samples from Mulfatlar harvests of the years 1995 and 1996. (authors)

  5. 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.

  6. 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

  7. 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)

  8. 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.

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. Advancement in tritium transport simulations for solid breeding blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Ying, Alice, E-mail: ying@fusion.ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Zhang, Hongjie [Mechanical and Aerospace Engineering Department, UCLA, Los Angeles, CA 90095 (United States); Merrill, Brad J. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2016-11-01

    In this paper, advancement on tritium transport simulations was demonstrated for a solid breeder blanket HCCR TBS, where multi-physics and detailed engineering descriptions are considered using a commercial simulation code. The physics involved includes compressible purge gas fluid flow, heat transfer, chemical reaction, isotope swamping effect, and tritium isotopes mass transport. The strategy adopted here is to develop numerical procedures and techniques that allow critical details of material, geometric and operational heterogeneity in a most complete engineering description of the TBS being incorporated into the simulation. Our application focuses on the transient assessment in view of ITER being pulsed operations. An immediate advantage is a more realistic predictive and design analysis tool accounting pulsed operations induced temperature variations which impact helium purge gas flow as well as Q{sub 2} composition concentration time and space evolutions in the breeding regions. This affords a more accurate prediction of tritium permeation into the He coolant by accounting correct temperature and partial pressure effects and realistic diffusion paths. The analysis also shows that by introducing by-pass line to accommodate ITER pulsed operations in the TES loop allows tritium extraction design being more cost effective.

  15. Canadian inter-laboratory organically bound tritium (OBT) analysis exercise.

    Science.gov (United States)

    Kim, S B; Olfert, J; Baglan, N; St-Amant, N; Carter, B; Clark, I; Bucur, C

    2015-12-01

    Tritium emissions are one of the main concerns with regard to CANDU reactors and Canadian nuclear facilities. After the Fukushima accident, the Canadian Nuclear Regulatory Commission suggested that models used in risk assessment of Canadian nuclear facilities be firmly based on measured data. Procedures for measurement of tritium as HTO (tritiated water) are well established, but there are no standard methods and certified reference materials for measurement of organically bound tritium (OBT) in environmental samples. This paper describes and discusses an inter-laboratory comparison study in which OBT in three different dried environmental samples (fish, Swiss chard and potato) was measured to evaluate OBT analysis methods currently used by CANDU Owners Group (COG) members. The variations in the measured OBT activity concentrations between all laboratories were less than approximately 20%, with a total uncertainty between 11 and 17%. Based on the results using the dried samples, the current OBT analysis methods for combustion, distillation and counting are generally acceptable. However, a complete consensus OBT analysis methodology with respect to freeze-drying, rinsing, combustion, distillation and counting is required. Also, an exercise using low-level tritium samples (less than 100 Bq/L or 20 Bq/kg-fresh) would be useful in the near future to more fully evaluate the current OBT analysis methods. Crown Copyright © 2015. Published by Elsevier Ltd. All rights reserved.

  16. Experimental study of permeation and selectivity of zeolite membranes for tritium processes

    Energy Technology Data Exchange (ETDEWEB)

    Borisevich, Olga; Antunes, Rodrigo; Demange, David, E-mail: david.demange@kit.edu

    2015-10-15

    Highlights: • We report about new experimental results on advanced membranes for tritium processing especially for the DEMO breeding blanket. • High permeances are measured on different zeolite MFI membranes made by film deposition or pore plugging. • Selectivity for H{sub 2}/He is limited requiring a multi-stage membrane process. • Selectivity of H{sub 2}O/He seems high enough to operate one single module. - Abstract: Zeolites are known as tritium compatible inorganic materials widely used in packed beds as driers in detritiation systems and are also suggested for tritium removal from helium at cryogenic temperature. The Tritium Laboratory Karlsruhe (TLK) proposed a new fully continuous approach for tritium extraction from the solid breeding blanket of fusion machines that improves the overall tritium management and minimizes both the tritium inventory and processing time. It is based on membrane permeation as a pre-concentration stage upstream of a final tritium recovery stage using a catalytic Pd-based membrane reactor. Zeolite membranes were identified as the most promising candidates for the pre-concentration stage. In the present work the tubular zeolite MFI membrane provided by the Institute for Ceramic Technologies and Systems (IKTS, Hermsdorf, Germany) is studied to consolidate the proposed approach. The permeation measurements for single gases hydrogen (replacing radioactive tritium) and helium, for binary mixtures H{sub 2}/He and H{sub 2}O/He at different concentrations and temperatures are presented. The tested membrane demonstrates a high performance, almost independent from the inlet composition in the case of a gaseous mixture, while the transport in the presence of water vapour is strongly related to the temperature of the mixture and component concentrations.

  17. Depth of cervical cone removed by loop electrosurgical excision procedure and subsequent risk of spontaneous preterm delivery

    DEFF Research Database (Denmark)

    Noehr, Bugge; Jensen, Allan; Frederiksen, Kirsten

    2009-01-01

    OBJECTIVE: To investigate the association between cone depth of the loop electrosurgical excision procedure (LEEP) of the cervix and subsequent risk of spontaneous preterm delivery. METHODS: The study included all deliveries in Denmark over a 9-year period, 1997-2005, with information obtained fr...

  18. Investigation of internal contamination by tritium in A-1 nuclear power plant personnel in 1974

    International Nuclear Information System (INIS)

    Ondris, D.; Herchl, M.; Homolova, E.

    1977-01-01

    The results are presented of the 1974 personnel monitoring of the Bohunice A-1 nuclear power plant staff for internal contamination with tritium. Totally, 650 urine samples taken from 103 workers were analyzed using the recommended ICRP procedure. In routine examinations, the highest dose equivalent value of tritium incorporated within two weeks did not exceed 10 mrem, i.e., the maximum annual dose equivalent did not exceed 260 mrem. 8.5 μCi tritium per 1 litre urine was considered to be an alarm value. In a selected group of 21 high-risk persons analyses were conducted before and after each operation associated with tritium hazards. The limit dose was set to 5.8 μCi.l -1 , i.e., the tritium concentration equivalent to 10% of the maximum permissible annual intake. In 18 workers where tritium risk was of a more serious nature the biological half-life was followed up, with the average biological half-life being 8.5 days, with 5 days for the minimum and 12 days for the maximum values. The results show that in 1974 the tritium burden did not exceed 1/10 of the maximum permissible dose for any of the A-1 nuclear power plant workers. (L.O.)

  19. Seasonal variations of bomb tritium in rain

    Energy Technology Data Exchange (ETDEWEB)

    Israel, G.; Roether, W.; Schumann, G.

    1962-08-15

    The tritium activity of precipitation at Sindorf, near Cologne, Germany, has been monitored since 1960. Experimental results are presented and discussed briefly with the aid of a comparison to Sr/sup 90/ as another artificial nuclide characterized by similar tropospheric behavior. (auth) A series of analytical schemes for the measurement of Ba/sup 140/, Sr/sup 89/, Sr/sup 90/, Cs/sup 137/, I/sup 131/, and Pu/sup 238/ in various environmental materials is described. These procedures are based on radiochemical methods in use at Chalk River and have been thoroughly tested in practice. The manual is divided into distinct sections in the belief that such an arrangement will be a convenience where sample preparation, gamma -spectrometry, radiochemistry, and counting are performed by different people working in separate areas of a laboratory. (auth)

  20. 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)

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

    International Nuclear Information System (INIS)

    Varghese, C.; Singh, I.; Agarwal, R.P.; Ramani, M.P.S.; Khan, A.A.

    1983-01-01

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

  2. Effect of the self-pumped limiter concept on the tritium fuel cycle

    International Nuclear Information System (INIS)

    Finn, P.A.; Sze, D.K.; Hassanein, A.

    1988-01-01

    The self-pumped limiter concept for impurity control of the plasma of a fusion reactor has a major impact on the design of the tritium systems. To achieve a sustained burn, conventional limiters and divertors remove large quantities of unburnt tritium and deuterium from the plasma which must be then recycled using a plasma processing system. The self-pumped limiter which does not remove the hydrogen species, does not require any plasma processing equipment. The blanket system and the coolant processing systems acquire greater importance with the use of this unconventional impurity control system. 3 refs., 2 figs

  3. 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

  4. New tritium monitor for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Jalbert, R.A.

    1985-01-01

    At DT-fueled fusion reactors, there will be a need for tritium monitors that can simultaneously measure in real time the concentrations of HTO, HT and the activated air produced by fusion neutrons. Such a monitor has been developed, tested and delivered to the Princeton Plasma Physics Laboratory for use at the Tokamak Fusion Test Reactor (TFTR). It uses semipermeable membranes to achieve the removal of HTO from the sampled air for monitoring and a catalyst to convert the HT to HTO, also for removal and monitoring. The remaining air, devoid of tritium, is routed to a third detector for monitoring the activated air. The sensitivities are those that would be expected from tritium instruments employing conventional flow-through ionization chambers: 1 to 3 μCi/m 3 . Its discriminating ability is approximately 10 -3 for any of the three components (HTO, HT and activated air) in any of the other two channels. For instance, the concentration of HT in the HTO channel is 10 -3 times its original concentration in the sampled air. This will meet the needs of TFTR

  5. In-vessel dust and tritium control strategy in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M., E-mail: michiya.shimada@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France); Pitts, R.A.; Ciattaglia, S.; Carpentier, S.; Choi, C.H.; Dell Orco, G.; Hirai, T.; Kukushkin, A.; Lisgo, S.; Palmer, J.; Shu, W.; Veshchev, E. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2013-07-15

    A baseline strategy for dust and tritium-inventory control and recovery in ITER has been established and preparations are underway for its implementation. Limits on dust and tritium-inventory are an integral part of the ITER safety case and are fixed at 1 kg for tritium, 1000 kg for mobilisable dust and 11 kg (beryllium)/76 kg (tungsten) for dust on hot surfaces. Maximum average T-retention rates of ∼1 g/shot are estimated for baseline inductive operation at Q{sub DT} = 10, suggesting that the in-vessel T-retention could reach the administrative limit of 640 g in as little as ∼2 months of operation. Baking is expected to remove a significant fraction of the T co-deposited on the divertor targets. Despite large uncertainties, dust quantities are expected to remain well below safety limits over the divertor cassette lifetime. In situ aspiration during divertor cassette exchange is foreseen as the main dust removal technique.

  6. Method of separating tritium contained in gaseous wastes

    International Nuclear Information System (INIS)

    Hashimoto, Yasuo; Oozono, Hideaki.

    1981-01-01

    Purpose: To decrease tritium concentration in gaseous wastes to less than the allowable level by removing tritium in gaseous wastes generated upon combustion of radioactive wastes by using a plurality of heat exchangers. Method: Gaseous wastes at high temperature generated upon combustion of radioactive wastes in an incinerator are removed with radioactive solid substances through filters, cooled down to a temperature below 10 0 C by the passage through first and second heat exchangers and decreased with tritium content to less than the allowable concentration by the gaseous wastes at low temperature from the second heat exhcanger. The gaseous wastes at low temperature are used as the cooling medium for the first heat exchanger. They are heat exchanged at the upstream of the second heat exchanger with the cooling water from the third heat exchanger and cooled at the downstream by the cooling water cooled by the cooling medium. The gaseous wastes at low temperature thus cooled below 10 0 C are heated to about 350 0 C in the first heat exchanger and discharged. (Moriyama, K.)

  7. 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.

  8. 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

  9. Applications of hydrophobic Pt catalysts in separation of tritium from liquid effluents

    International Nuclear Information System (INIS)

    Ionita, Gheorghe; Popescu, Irina; Stefanescu, Ioan; Varlam, Carmen

    2003-01-01

    Hydrophobic Pt catalysts were first prepared and used in deuterium or tritium separation while after their application was extended to chemical reactions occurring in liquid water or saturated humidity environments. Capillary condensing produced at the contact with liquid water or vapors engenders in classical hydrophilic catalysts a decrease in activity what makes them inefficient. Consequently, liquid water 'repealing' catalysts are to be used allowing, at the same time gaseous reactants and reaction products to diffuse to and fro the catalytic active centers. These catalysts were successfully applied in deuterium enrichment and tritium separation based on hydrogen- liquid water isotopic exchange at both pilot and industrial scale. High activity and a prolonged stability were demonstrated and checked in: - detritiation of the heavy water used as both moderator and coolant in CANDU type reactors; removing of tritium from light water recirculated in nuclear fuel reprocessing facilities; removal and recovery of tritium from atmosphere and tritium processing installations. Due to their incontestable advantages the use of these catalysts was recently extended to other chemical processes occurring in the presence of liquid water or in high humidity environment or else when water occurs as a reaction product, such as catalytic hydrogen - oxygen recombination at room temperature or removal of stable organic pollutants from waste waters

  10. 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.)

  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. Repair procedure used in removing corroded pits in the distillation towers of the Getulio Vargas Refinery Unit 2100

    Energy Technology Data Exchange (ETDEWEB)

    Lordelos, H.M.; Santin, J.L.

    1977-07-01

    A description is given of the corrosion pits on ASTM A240, Type 405 steel cladded to carbon steel plates used in Petroleo Brasileiro S.A.'s Getulio Vargas Refinery Unit 2100 distillation towers; the repair procedure used, including sand blasting of the corroded surfaces, grinding of the pits, and welding of those pits whose depth was above a maximum limit, and the use of liquid penetrant to check the repairs made; and hydrostatic testing of the T2201 catalytic cracking unit, which used also cladded metals and on which the pits were much smaller in size and number than those on the distillation units.

  16. 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

  17. 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

  18. 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.

  19. 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)

  20. 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