Measurement of tritium with plastic scintillator surface improvement with plasma treatment

Energy Technology Data Exchange (ETDEWEB)

Yoshihara, Y.; Furuta, E. [Ochanomizu University, Bunkyo-ku, Tokyo (Japan); Ohyama, R.I.; Yokota, S. [Tokai University, Hiratsuka-shi, Kanagawa (Japan); Kato, Y.; Yoshimura, T.; Ogiwara, K. [Hitachi Aloka Medical, Mure, Mitaka-shi, Tokyo (Japan)

2015-03-15

Tritium is usually measured by using a liquid scintillation counter. However, liquid scintillator used for measurement will become radioactive waste fluid. To solve this issue, we have developed a method of measuring tritium samples with plasma-treated plastic scintillator (PS)sheets (Plasma method). The radioactive sample is held between 2 PS sheets and the whole is enclosed in a a low-potassium glass vial. With the Plasma method of 2-min plasma treatment, we have obtained measurement efficiency of 48 ± 2 % for 2 min measurement of tritium except for tritiated water. The plasma treatment makes the PS surface rough and hydrophilic which contributes to improve the contact between tritium and PS. On the other hand, it needed almost 6 hours to obtain constant measurement efficiency. The reason was that the dry-up handling in the vial needed longer time to vaporize H{sub 2}O molecules than in the air. We tried putting silica gel beads into vials to remove H{sub 2}O molecules from PS sheet surface quickly. The silica gel beads worked well and we got constant measurement efficiency within 1-3 hours. Also, we tried using other kinds of PS treated with plasma to obtain higher measurement efficiencies of tritium samples.

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

Directory of Open Access Journals (Sweden)

B.D. Lemeshko

2017-11-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Determination of the Tritium Concentration in Deuterium-Tritium Fusion Plasmas from the Jet TTE Campaign

International Nuclear Information System (INIS)

Gatu Johnson, Maria

2004-01-01

This thesis describes the development and implementation of methods for tritium concentration determination for JET fusion plasmas. The usefulness of MPR data in this context is investigated. It is shown that results from MPR spectral analysis can simplify the calculations for neutral beam heated plasmas and that it is essential for calculations for radio frequency heated plasmas. The described methods are applied to pulses from the Trace Tritium Experiment (TTE), staged at JET in October 2003. Results from simple, time resolved analysis using MPR and other public JET data are presented and the assumptions made in the calculations are discussed. The results agree with expectations but would be even more interesting if spatial variations were taken into account

Determination of the Tritium Concentration in Deuterium-Tritium Fusion Plasmas from the Jet TTE Campaign

Energy Technology Data Exchange (ETDEWEB)

Gatu Johnson, Maria

2004-01-01

This thesis describes the development and implementation of methods for tritium concentration determination for JET fusion plasmas. The usefulness of MPR data in this context is investigated. It is shown that results from MPR spectral analysis can simplify the calculations for neutral beam heated plasmas and that it is essential for calculations for radio frequency heated plasmas. The described methods are applied to pulses from the Trace Tritium Experiment (TTE), staged at JET in October 2003. Results from simple, time resolved analysis using MPR and other public JET data are presented and the assumptions made in the calculations are discussed. The results agree with expectations but would be even more interesting if spatial variations were taken into account.

Isotopic scaling of transport in deuterium-tritium plasmas

International Nuclear Information System (INIS)

Scott, S.D.; Adler, H.; Bell, M.G.; Bell, R.; Budny, R.V.; Bush, C.E.; Chang, Z.; Duong, H.

1995-01-01

Both global and thermal energy confinement improve in high-temperature supershot plasmas in the Tokamak Fusion Test Reactor (TFTR) when deuterium beam heating is partially or wholly replaced by tritium beam heating. For the same heating power, the tritium-rich plasmas obtain up to 22% higher total energy, 30% higher thermal ion energy, and 20-25% higher central ion temperature. Kinetic analysis of the temperature and density profiles indicates a favorable isotopic scaling of ion heat transport and electron particle transport, with τ Ei (a/2) ∝ (A) 0.7-0.8 and τ pe (a) ∝ (A) 0.8

Behavior of water of crystallization in CuSO4·5H2O studied by the tritium tracer method

International Nuclear Information System (INIS)

Sato, Tetsuya; Jiao, Yurong; Imaizumi, Hiroshi; Kano, Naoki

2011-01-01

Tritium (T) is one of hydrogen isotopes, and its chemical behavior is similar to other hydrogen isotopes. Therefore tritium is used as one of tracers in chemical experimental tracer. As one of applications, we tried to apply this method to clarifying the behavior of water of crystallization in an inorganic material. The sample used was copper sulfate pentahydrate. First, this compound was tritiated, then desorbed the water of crystallization from the tritiated compound. Comparing the behavior of amount of substance with the specific activity, the following four matters have been found. (1) There is no relation between each T concentration of HTO water and the mass of the compound within the T concentration used. (2) It can be confirmed that copper sulfate pentahydrate has three kinds of energetically different water of crystallization by T tracer method. (3) Each T concentration of water of crystallization is different at the coordinate position, and the HTO molecule is hard to coordinate at the position having weak binding force. (4) The T tracer method is useful to analyze the behavior of the combined water in materials. (author)

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

Tanabe, T.; Sugiyama, K.; Masaki, K.; Gotoh, Y.; Tobita, K.; Miya, N.

2003-01-01

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

Isotopic scaling of transport in deuterium-tritium plasmas

International Nuclear Information System (INIS)

Scott, S.D.; Murakami, M.; Adler, H.; Chang, Z.; Duong, H.; Grisham, L.R.; Fredrickson, E.D.; Grek, B.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Jassby, D.L.; Johnson, D.W.; Johnson, L.C.; Loughlin, M.J.; Mansfield, D.K.; McGuire, K.M.; Meade, D.M.; Mikkelsen, D.M.; Murphy, J.; Park, H.K.; Ramsey, A.T.; Schivell, J.; Skinner, C.H.; Strachan, J.D.; Synakowski, E.J.; Taylor, G.; Thompson, M.E.; Wieland, R.; Zarnstorff, M.C.

1995-01-01

Both global and thermal energy confinement improve in high-temperature supershot plasmas in the Tokamak Fusion Test Reactor (TFTR) when deuterium beam heating is partially or wholly replaced by tritium beam heating. For the same heating power, the tritium-rich plasmas obtain up to 22% higher total energy, 30% higher thermal ion energy, and 20-25% higher central ion temperature. Kinetic analysis of the temperature and density profiles indicates a favorable isotopic scaling of ion heat transport and electron particle transport, with τ Ei (a/2) ∝ left angle A right angle 0.7-0.8 and τ pe (a) ∝ left angle A right angle 0.8 . (orig.)

Tritium as a tracer for the movement of surface water and groundwater in the Glatt Valley, Switzerland

International Nuclear Information System (INIS)

Santschi, P.H.; Hoehn, E.; Lueck, A.; Farrenkothen, K.

1987-01-01

A pulse of tritiated water (∼ 500 Ci) accidentally discharged by an isotope processing plant in the Glatt River Valley, northern Switzerland, allowed us to observe the migration of a contaminant pulse through a sewage treatment plant, rivers, and various wells of infiltrated groundwater. The accident pointed to various memory effects of the tritium, which acted as a conservative tracer. Tritium concentrations in surface water and groundwater were used to test predictions for the transport of conservative anthropogenic trace contaminants accidentally discharged into the sewer system. Mass balance calculations indicate that about 2-10% of the tritium pulse infiltrated to the groundwater and about 0.5% of the total reached eight major drinking water wells of this densely populated area. In spite of the complex hydrogeology of the lower Glatt River Valley, tritium breakthrough curves could be effectively simulated with modeling approaches developed from an experimental well field

Engineering studies of tritium recovery from CTR blankets and plasma exhaust

International Nuclear Information System (INIS)

Watson, J.S.

1975-01-01

Engineering studies on tritium handling problems in fusion reactors have included conceptual and experimental studies of techniques for recovery of tritium bred in the reactor blanket and conceptual designs for recovery and processing of tritium from plasma exhausts. The process requirements and promising techniques for the blanket system depend upon the materials used for the blanket, coolant, and structure and on the operating temperatures. Process requirements are likely to be set in some systems by allowable loss rates to the steam system or by inventory considerations. Conceptual studies have also been made for tritium handling equipment for fueling, recovery, and processing in plasma recycle systems of fusion reactors, and a specific design has been prepared for ''near-term'' Tokamak experiments. (auth)

Evaluation of natural recharge of Chingshui geothermal reservoir using tritium as a tracer

International Nuclear Information System (INIS)

Cheng, W.; Kuo, T.; Su, C.; Chen, C.; Fan, K.; Liang, H.; Han, Y.

2010-01-01

Naturally existing tritium in groundwater was applied as a tracer to evaluate the natural recharge of the Chingshui geothermal reservoir. The residence time (or, age) of Chingshui geothermal water was first determined with tritium data at 15.2 and 11.3 year using the plug flow and dispersive model, respectively. The annual natural recharge was then estimated by combining the use of the residence time and the fluid-in-place of the Chingshui geothermal reservoir. The natural recharge for Chingshui geothermal reservoir was estimated at 5.0 x 10 5 and 6.7 x 10 5 m 3 year -1 using the plug flow and dispersive model, respectively. Chingshui geothermal water is largely from a fractured zone in the Jentse Member of the Miocene Lushan Formation. The dispersive model more adequately represents the fracture flow system than the simple plug flow model.

Design of a tritium decontamination workstation based on plasma cleaning

International Nuclear Information System (INIS)

Antoniazzi, A.B.; Shmayda, W.T.; Fishbien, B.F.

1993-01-01

A design for a tritium decontamination workstation based on plasma cleaning is presented. The activity of tritiated surfaces are significantly reduced through plasma-surface interactions within the workstation. Such a workstation in a tritium environment can routinely be used to decontaminate tritiated tools and components. The main advantage of such a station is the lack of low level tritiated liquid waste. Gaseous tritiated species are the waste products with can with present technology be separated and contained

Modeling tritium processes in plasma-facing beryllium

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Dolan, T.J.; Mulock, M.J.

1995-01-01

In this paper we present techniques and recommended parameters for modeling tritium implantation, trapping and release, and permeation, in beryllium-clad structures adjacent to the plasma. Among the features that should be considered are the effects of surface films, the mobility of beryllium through those films, damage caused by ion implantation, especially in regions where pitting may be expected, and bubble formation. Tritium transport parameters recommended are based on fits with experimental data and available theory. Estimates of inventories in ITER using these parameters are also given. 31 refs., 2 figs., 1 tab

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-02-01

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

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

Science.gov (United States)

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

1997-02-01

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

Reactivity of hydrogen contained in Raney nickel for ethylene hydrogenation studied by means of a tritium tracer

International Nuclear Information System (INIS)

Miyatani, Daisaku; Takeuchi, Toyosaburo.

1979-01-01

Reactivity of hydrogen contained in Raney nickel with ethylene was studied by using a tritium tracer. Hydrogen in Raney nickel was previously labeled with tritium and distinguished from hydrogen introduced during the hydrogenation reaction. The reactivity of the contained hydrogen was determined by measurement of the radioactivity of ethane produced in the hydrogenation. Ethylene reacted with hydrogen in Raney nickel for no supply of hydrogen during the hydrogenation. However, when ethylene was hydrogenated by both hydrogen in Raney nickel and introduced hydrogen, over 99% of the ethylene reacted with the introduced hydrogen and hardly reacted with the contained hydrogen. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Tritium tracer movement as an analogy for pump and treat remediation

International Nuclear Information System (INIS)

1994-12-01

There has been debate over effectiveness of groundwater pump and treat remediation. The goal of the following discussion is to present evidence from a tracer test that illustrates the difficulty in removing contaminants from fractured shale that is typical of portions of the DOE-Oak Ridge Reservation (ORR). This report provides a brief prelude to more detailed analysis that is in progress. Attempts to remediate groundwater contamination with pump and treat technology have been hampered by difficulties in removing contaminants in slow flow zones. There is interest in using this remediation method on the ORR because it is an existing technology. However, this setting provides a rather extreme contrast between fast flow zones (fractures) and slow flow zones (the matrix surrounding the fractures). Over the past few years, the authors have begun to develop an understanding of how contaminants move in fractures and how contaminant exchange between the fracture and matrix occurs. In particular, they have evidence from a long term tritium tracer test that has direct bearing on potential success or failure of pump and treat remediation in fractured rocks

The tritium confinement and surface chemistry of plasma facing materials in controlled D-T fusion devices

International Nuclear Information System (INIS)

Wu, C.H.

1987-01-01

Tritium permeation through first walls, limiters or divertors subjected to energetic tritium charge exchange neutral bombardment is a potentially serious problem area for advanced D-T reactors operating at elevated temperatures. High concentrations of tritium in the near surface region can be reached by implantation of the charge neutral flux combined with a relatively slow recombination of these atoms into molecules at the plasma/ first wall interface. A concentration gradient is established, causing tritium to diffuse into the bulk and essentially to the outer wall surface where it can enter the first wall coolant. Since tritium separation from cooling water is very costly, release of even a small fraction of tritium to the environment could pose undesirable safety problems. Therefore, it is necessary to reduce the tritium permeation. An analysis of the way of inhibition has been made. The tritium interacts with the solid surface of the plasma facing components, resulting in trapping and material erosion, and posing problems with respect to plasma density control. The erosion of the plasma facing component materials is mainly caused by physical and chemical erosion. A detailed analysis of chemical erosion by tritium has been performed and the results are described. (author)

Physics of high performance deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

McGuire, K.M.; Batha, S.

1996-11-01

During the past two years, deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR) have been used to study fusion power production, isotope effects associated with tritium fueling, and alpha-particle physics in several operational regimes. The peak fusion power has been increased to 10.7 MW in the supershot mode through the use of increased plasma current and toroidal magnetic field and extensive lithium wall conditioning. The high-internal-inductance (high-I i ) regime in TFTR has been extended in plasma current and has achieved 8.7 MW of fusion power. Studies of the effects of tritium on confinement have now been carried out in ohmic, NBI- and ICRF- heated L-mode and reversed-shear plasmas. In general, there is an enhancement in confinement time in D-T plasmas which is most pronounced in supershot and high-I i discharges, weaker in L-mode plasmas with NBI and ICRF heating and smaller still in ohmic plasmas. In reversed-shear discharges with sufficient deuterium-NBI heating power, internal transport barriers have been observed to form, leading to enhanced confinement. Large decreases in the ion heat conductivity and particle transport are inferred within the transport barrier. It appears that higher heating power is required to trigger the formation of a transport barrier with D-T NBI and the isotope effect on energy confinement is nearly absent in these enhanced reverse-shear plasmas. Many alpha-particle physics issues have been studied in the various operating regimes including confinement of the alpha particles, their redistribution by sawteeth, and their loss due to MHD instabilities with low toroidal mode numbers. In weak-shear plasmas, alpha-particle destabilization of a toroidal Alfven eigenmode has been observed

Proceedings of 2nd Internaitonal workshop on tritium effects in plasma facing components

Energy Technology Data Exchange (ETDEWEB)

Morita, Kenji [Nagoya Univ. (Japan). School of Engineering; Noda, Nobuaki [eds.

1994-08-01

This workshop was held at Nagoya University on May 19 and 20, 1994. Approximately 1/3 of the lectures discussed the migration and retention of tritium in graphite and the other forms of carbon. As to this topic, most of the different aspects of the tritium reactions with carbon were generally agreed on. At the temperature lower than 800 K, tritium plasma interacts with graphite by forming a saturated layer on the surface, by forming a codeposited layer of sputtered carbon and tritium, and by allowing tritium diffusion through Pores. At the temperature higher than 800 K, the principal reaction of tritium with carbon is intergranular diffusion with high energy trapping. Because beryllium is the reference plasma-facing material for the ITER, several presentations on the reactions of tritium with beryllium were made. Also the tritium permeation through other metals was the topics. The results of TFTR D-T experiment were reported in the first talk. In this book, the gists of these lectures are collected. (K.I.).

Proceedings of 2nd International workshop on tritium effects in plasma facing components

International Nuclear Information System (INIS)

Morita, Kenji; Noda, Nobuaki

1994-08-01

This workshop was held at Nagoya University on May 19 and 20, 1994. Approximately 1/3 of the lectures discussed the migration and retention of tritium in graphite and the other forms of carbon. As to this topic, most of the different aspects of the tritium reactions with carbon were generally agreed on. At the temperature lower than 800 K, tritium plasma interacts with graphite by forming a saturated layer on the surface, by forming a codeposited layer of sputtered carbon and tritium, and by allowing tritium diffusion through Pores. At the temperature higher than 800 K, the principal reaction of tritium with carbon is intergranular diffusion with high energy trapping. Because beryllium is the reference plasma-facing material for the ITER, several presentations on the reactions of tritium with beryllium were made. Also the tritium permeation through other metals was the topics. The results of TFTR D-T experiment were reported in the first talk. In this book, the gists of these lectures are collected. (K.I.)

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.)

Test determination with tritium as a radioactive tracer of the residence time distribution in the stability pool for Cabrero sewage

International Nuclear Information System (INIS)

Diaz, Francisco; Duran, Oscar; Henriquez, Pedro; Vega, Pedro; Padilla, Liliana; Gonzalez, David; Garcia Agudo, Edmundo

2000-01-01

This work was prepared by the Chilean and International Atomic Energy Agencies and covers the hydrodynamic functioning of sewage stability pools using tracers. The plant selected in the city of Cabrero, 500 km. south of Santiago, and is a rectangular facultative pool with a surface area of 7100 m 2 and a maximum volume of 12,327 m2 that receives an average flow of 20 l/s, serving a population of 7000 individuals. The work aims to characterize the runoff from the flow that enters the pool, using a radioactive tracer test, where the incoming water is marked, and its out-coming passage is determined, to establish the residence time distribution. Tritium was selected in the form of tritiated water as a tracer that is precisely emptied into the water flow from the distribution ravine at the lake entrance. Samples are taken at the outflow to determine the concentration of tritium after distillation, simultaneously measuring the flow, to be analyzed in a liquid flicker counter. An average test time of 5.3 days was obtained and an analysis of the residence time distribution for the tracer shows that it leaves quickly and indicates bad flow distribution in the lake with a major short circuit and probable dead zones

ICRF heating and transport of deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

Murakami, M.; Batchelor, D.B.; Bush, C.E.

1994-01-01

This paper describes results of the first experiments utilizing high-power ion cyclotron range of frequency (ICRF) to heat deuterium-tritium (D-T) plasmas in reactor-relevant regimes on the Tokamak Fusion Test Reactor (TFTR). Results from these experiments have demonstrated efficient core, second harmonic, tritium heating of D-T supershot plasmas with tritium concentrations ranging from 6%--40%. Significant direct ion heating on the order of 60% of the input radio frequency (rf) power has been observed. The measured deposition profiles are in good agreement with two-dimensional modeling code predictions. Confinement in an rf-heated supershot is at least similar to that without rf, and possibly better in the electron channel. Efficient electron heating via mode conversion of fast waves to ion Bernstein waves (IBW) has been demonstrated in ohmic, deuterium-deuterium and DT-neutral beam injection plasmas with high concentrations of minority 3 He (n 3 He /n e > 10%). By changing the 3 He concentration or the toroidal field strength, the location of the mode-conversion radius was varied. The power deposition profile measured with rf power modulation showed that up to 70% of the power can be deposited on electrons at an off-axis position. Preliminary results with up to 4 MW coupled into the plasma by 90-degree phased antennas showed directional propagation of the mode-converted IBW. Heat wave propagation showed no strong inward thermal pinch in off-axis heating of an ohmically-heated (OH) target plasma in TFIR

Silicon Carbide as a tritium permeation barrier in tungsten plasma-facing components

Science.gov (United States)

Wright, G. M.; Durrett, M. G.; Hoover, K. W.; Kesler, L. A.; Whyte, D. G.

2015-03-01

The control of tritium inventory is of great importance in future fusion reactors, not only from a safety standpoint but also to maximize a reactor's efficiency. Due to the high mobility of hydrogenic species in tungsten (W) one concern is the loss of tritium from the system via permeation through the tungsten plasma-facing components (PFC). This can lead to loss of tritium through the cooling channels of the wall thereby mandating tritium monitoring and recovery methods for the cooling system of the first wall. The permeated tritium is then out of the fuel cycle and cannot contribute to energy production until it is recovered and recycled into the system.

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

International Nuclear Information System (INIS)

Wu, C.H.

1986-01-01

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

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

ICRF heating and transport of deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

Rogers, J.H.; Schilling, G.; Stevens, J.E.; Taylor, G.; Wilson, J.R.; Bell, M.G.; Budny, R.V.; Bretz, N.L.; Darrow, D.; Fredrickson, E.

1995-02-01

This paper describes results of the first experiments utilizing high-power ion cyclotron range of frequency (ICRF) to heat deuterium-tritium (D-T) plasmas in reactor-relevant regimes on the Tokamak Fusion Test Reactor (TFTR). Results from these experiments have demonstrated efficient core, second harmonic, tritium beating of D-T supershot plasmas with tritium concentrations ranging from 6%-40%. Significant direct ion heating on the order of 60% of the input radio frequency (rf) power has been observed. The measured deposition profiles are in good agreement with two-dimensional modeling code predictions. Energy confinement in an rf-heated supershot is at least similar to that without rf, and possibly better in the electron channel. Efficient electron heating via mode conversion of fast waves to ion Bernstein waves (IBW) has been demonstrated in ohmic, deuterium-deuterium and DT-neutral beam injection plasmas with high concentrations of minority 3 He (n 3He /n e = 15% - 30%). By changing the 3 He concentration or the toroidal field strength, the location of the mode-conversion radius was varied. The power deposition profile measured with rf power modulation indicated that up to 70% of the power can be deposited on electrons at an off-axis position. Preliminary results with up to 4 MW coupled into the plasma by 90-degree phased antennas showed directional propagation of the mode-converted IBW. Analysis of heat wave propagation showed no strong inward thermal pinch in off-axis heating of an ohmically-heated target plasma in TFTR

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

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

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)

Thermonuclear reaction rates in a deuterium-tritium plasma

International Nuclear Information System (INIS)

Beckman, L.

1978-12-01

In a deuterium-tritium plasma six thermonuclear reactions take place between the deuterons, tritons and the 3 He-particles formed in about half of the d-d-reactions. The rate constants for these six reactions have been calculated from the latest evaluations of the reaction cross sections which were available. In some cases, notably the reactions t+t, t+ 3 He and 3 He+ 3 He, the number of published cross section measurements is small, and the uncertainty in the calculated rate constants consequently large. Analytical expressions for the rate constants as functions of the plasma temperature have been set up. (author)

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

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

Preliminary assessment of the tritium inventory and permeation in the plasma facing components of ITER

International Nuclear Information System (INIS)

Federici, G.; Holland, D.; Brooks, J.; Causey, R.; Dolan, T.J.; Longhurst, G.

1995-01-01

This paper discusses preliminary quantitative predictions for the tritium inventory in- and permeation through the first-wall and divertor PFC's of ITER. The primary plasma facing material under consideration is beryllium, with possible use of tungsten or carbon fiber composites (CFC's) on high-heat-flux surfaces. They use state-of-the-art tritium transport models, in conjunction with design parameters, and loading conditions anticipated for the first-wall, baffle, limiter and divertor. The analysis includes the synergistic effects of erosion on tritium implantation and trapping, which are expected to play a key role, particularly in the divertor regions where the interaction of the plasma with the surfaces will be most severe. The influence of several key parameters that strongly affect tritium build-up and release is assessed. Finally, they discuss the uncertainties in materials properties under ITER operating conditions and the R and D needed to resolve these uncertainties

High performance deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

Sabbagh, S.A.; Bell, M.G.

1995-03-01

Plasmas composed of nominally equal concentrations of deuterium and tritium (DT) have been created in TFTR with the goals of producing significant levels of fusion power and of examining the effects of DT fusion alpha particles. Conditioning of the limiter by the injection of lithium pellets has led to an approximate doubling of the energy confinement time, τ E , in supershot plasmas at high plasma current (I p ≤ 2.5 MA) and high heating power (P b ≤ 33 MW). Operation with DT typically results in an additional 20% increase in τ E . In the high poloidal beta, advanced tokamak regime in TFTR, confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.5 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasmas, exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Confinement of alpha particles appears to be classical and losses due to collective effects have not been observed. While small fluctuations in fusion product loss were observed during ELMs, no large loss was detected in DT plasmas

Dating of young groundwater using tritium and gaseous tracers (SF6, SF5CF3, CFC-12, H-1301): case study from southern Poland

Science.gov (United States)

Rozanski, Kazimierz; Bartyzel, Jakub; Dulinski, Marek; Kuc, Tadeusz; Sliwka, Ireneusz; Mochalski, Pawel; Kania, Jaroslaw; Witczak, Stanislaw

2013-04-01

Groundwater is an important source of potable water in many countries. While it covers ca. 50% of the global drinking water needs, in Europe this share is even higher, reaching approximately 70%. Nowadays, this strategic resource is at risk due to anthropogenic pollutants of various nature entering shallow aquifers. Proper management of groundwater resources requires thorough understanding of groundwater dynamics on time scales characteristic for the history of pollutant input to groundwater. The bomb-tritium has been used for several decades now as a tracer of choice to detect recent recharge and to quantify groundwater residence times on time scales extending from several years to several decades. The lumped-parameter modeling was the most often employed approach in this context. Since nowadays atmospheric concentrations of tritium are approaching natural levels in most parts of the world, the usage of this tracer has become more problematic. Therefore, there is a growing interest in alternative indicators of groundwater age in shallow aquifers. Anthropogenic trace gases present in the atmosphere, such as freons (CFC-11, CFC-12, CFC-113) and sulfur hexafluoride (SF6), have been applied in numerous case studies as substitutes of tritium. Here we present the results of a comprehensive study aimed at quantifying mean residence time of groundwater in the recharge area of porous sandy aquifer system located in the southern Poland. The principal economic role of the aquifer, consisting of two water-bearing strata, is to provide potable water for public and private users. The yield of the aquifer is insufficient to meet all the needs and, as a consequence, licensing conflicts arise between water supply companies and industry on the amount of water available for safe exploitation. To quantify residence time distribution (RTD) functions of water parcels arriving at the production wells located in the recharge area of the aquifer, tritium along with several gaseous tracers

Deuterium-tritium TFTR plasmas in the high poloidal beta regime

International Nuclear Information System (INIS)

Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.

1995-03-01

Deuterium-tritium plasmas with enhanced energy confinement and stability have been produced in the high poloidal beta, advanced tokamak regime in TFTR. Confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.46 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasma,s exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Fusion power exceeding 6.7 MW with a fusion power gain Q DT = 0.22 has been produced with reduced alpha particle first orbit loss provided by the increased l i

Tritium retention on the surface of stainless steel samples fixed on the plasma-facing wall in LHD

International Nuclear Information System (INIS)

Matsuyama, Masao; Abe, Shinsuke; Nishimura, Kiyohiko; Ashikawa, Naoko; Sagara, Akio; Oya, Yasuhisa; Okuno, Kenji; Yamauchi, Yuji; Nobuta, Yuji

2014-01-01

Effects of pre-heating for retention and distribution of tritium have been studied using samples fixed on the wall of the Large Helical Device during a plasma campaign. The samples were fixed at four different locations. The plasma-facing surface of the samples was covered with deposition layers of different thickness in each sample. Retention behavior in deposition layers was observed using β-ray-induced X-ray spectrometry and imaging plate technique. Pre-heating of the samples in vacuum was changed in a temperature range from 300 to 623 K, and subsequent tritium exposure was carried out at 300 K in every runs. Non-uniformity of tritium distribution clearly appeared even in the as-received samples which was not pre-heated. It is considered, therefore, that non-uniform adsorption sites of tritium have been produced during a formation process of deposition layers. In addition, it was seen that the amount of tritium retention increased with an increase in the pre-heating temperature, indicating that adsorption sites of tritium were newly formed in the deposition layers by heating in vacuum. (author)

Structure, tritium depth profile and desorption from ‘plasma-facing’ beryllium materials of ITER-Like-Wall at JET

Directory of Open Access Journals (Sweden)

E. Pajuste

2017-08-01

Experimental results revealed that > 95% of the tritium was localized in the top 30 – 45µm of the ‘plasma-facing’ surface, however, possible tritium presence up to 100µm cannot be excluded. During temperature programmed desorption at 4.8K/min in the flow of purge gas He+ 0.1% H2 the tritium release started below 475K, the most intense release occurred at 725 – 915K and the degree of detritiation of > 91% can be obtained upon reaching 1075K. The total tritium activity in the samples was in range of 2 – 32kilo Becquerel per square centimetre of the plasma-facing surface area.

Simulation of bomb tritium entry into the Atlantic Ocean

International Nuclear Information System (INIS)

Sarmiento, J.L.

1983-01-01

Tritium is used in a model-calibration study that aimed at developing three-dimensional ocean circulation and mixing models for climate and geochemical simulations. The North Atlantic tritium distribution is modeled using a three-dimensional advective field predicted by a primitive equation ocean circulation model. The effect of wintertime convection is parametrized by homogenizing the tracer to the observed March mixed-layer depth. Mixing is parametrized by horizontal and vertical Fickian diffusivities of 5 x 10 -6 cm 2 s -1 and 0.5 cm 2 s -1 , respectively. The spreading of tritium in the model is dominated by advection in the horizontal, and by wintertime convection and advection in the vertical. The horizontal and vertical mixing provided by the model have negligible effect. A comparison of the model tracer fields with observations shows that most of the basic patterns of the tritium field are repreduced. The model's mean vertical penetration of 543 m in 1972 is comparable to the 592 penetration obtained from the data. The major discrepancy between model and data is an inadequate penetration into deeper portions of the northwestern subtropical gyre main thermoclien. Some of the problem that may contribute to this are identified. A tritium simulation with a smoothed input gives a penetration depth of only 395 m. The smoothing puts a high fraction of the tritium into low-latitude, low-penetration regions such as the equator. This suggests that great care needs to be exercised in using simplified models of tritium observations to predict the behavior of tracers with different input functions, like fossil fuel CO 2

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.

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

Compression of a spherically symmetric deuterium-tritium plasma liner onto a magnetized deuterium-tritium target

International Nuclear Information System (INIS)

Santarius, J. F.

2012-01-01

Converging plasma jets may be able to reach the regime of high energy density plasmas (HEDP). The successful application of plasma jets to magneto-inertial fusion (MIF) would heat the plasma by fusion products and should increase the plasma energy density. This paper reports the results of using the University of Wisconsin’s 1-D Lagrangian, radiation-hydrodynamics, fusion code BUCKY to investigate two MIF converging plasma jet test cases originally analyzed by Samulyak et al.[Physics of Plasmas 17, 092702 (2010)]. In these cases, 15 cm or 5 cm radially thick deuterium-tritium (DT) plasma jets merge at 60 cm from the origin and converge radially onto a DT target magnetized to 2 T and of radius 5 cm. The BUCKY calculations reported here model these cases, starting from the time of initial contact of the jets and target. Compared to the one-temperature Samulyak et al. calculations, the one-temperature BUCKY results show similar behavior, except that the plasma radius remains about twice as long near maximum compression. One-temperature and two-temperature BUCKY results differ, reflecting the sensitivity of the calculations to timing and plasma parameter details, with the two-temperature case giving a more sustained compression.

Tritium in the Physical and Biological Sciences. Vol. II. Proceedings of the Symposium on the Detection and Use of Tritium in the Physical and Biological Sciences

Energy Technology Data Exchange (ETDEWEB)

NONE

1962-02-15

The use of tritium for research in physics, chemistry, biology and hydrology has in recent years become increasingly important. It was for this reason that the first international conference to discuss the progress of new developments was organized by the IAEA in conjunction with the Joint Commission on Applied Radioactivity and held from 3 - 10 May 1961, in Vienna. The first five sessions of the Symposium were devoted to the use of tritium in hydrology, physics and chemistry. Special emphasis was laid on the role of tritium as a tracer in hydrology, especially in the study of water movement. The establishment and improvement of counting and detection techniques to facilitate the application of tritium as a tracer was another aspect discussed in this part of the proceedings. Papers were read on the preparation of tritiated compounds and it was generally agreed that further clarification of the mechanism of various techniques, and of the Wilzbach gas exposure technique in particular, would lead to further developments in the synthesis of a number of tritium compounds important in biology. Other papers were concerned with tritium applications to studies of the mechanism of some chemical reactions together with the effects of tritium isotopes. During the second part of the Symposium the biological applications of tritium and tritiated compounds were discussed. These included general problems connected with the biological uses of tritium and the radiation effects of tritium on living organisms such as viruses, bacteria and cancer cells. The value of tritium in biological studies became apparent because of the ease with which a large number of metabolically active compounds such as hormones, vitamins and other important constituents in the body can be labelled with tritium. Tritium is also a weak beta-emitter and autoradiographs of tissues and single cells containing tritium-labelled compounds allow an excellent localization of the tracer. The Symposium was attended by

Tritium in the Physical and Biological Sciences. Vol. II. Proceedings of the Symposium on the Detection and Use of Tritium in the Physical and Biological Sciences

International Nuclear Information System (INIS)

1962-01-01

The use of tritium for research in physics, chemistry, biology and hydrology has in recent years become increasingly important. It was for this reason that the first international conference to discuss the progress of new developments was organized by the IAEA in conjunction with the Joint Commission on Applied Radioactivity and held from 3 — 10 May 1961, in Vienna. The first five sessions of the Symposium were devoted to the use of tritium in hydrology, physics and chemistry. Special emphasis was laid on the role of tritium as a tracer in hydrology, especially in the study of water movement. The establishment and improvement of counting and detection techniques to facilitate the application of tritium as a tracer was another aspect discussed in this part of the proceedings. Papers were read on the preparation of tritiated compounds and it was generally agreed that further clarification of the mechanism of various techniques, and of the Wilzbach gas exposure technique in particular, would lead to further developments in the synthesis of a number of tritium compounds important in biology. Other papers were concerned with tritium applications to studies of the mechanism of some chemical reactions together with the effects of tritium isotopes. During the second part of the Symposium the biological applications of tritium and tritiated compounds were discussed. These included general problems connected with the biological uses of tritium and the radiation effects of tritium on living organisms such as viruses, bacteria and cancer cells. The value of tritium in biological studies became apparent because of the ease with which a large number of metabolically active compounds such as hormones, vitamins and other important constituents in the body can be labelled with tritium. Tritium is also a weak beta-emitter and autoradiographs of tissues and single cells containing tritium-labelled compounds allow an excellent localization of the tracer. The Symposium was attended

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

Tritium experiments on components for fusion fuel processing at the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

Konishi, S.; Yoshida, H.; Naruse, Y.; Carlson, R.V.; Binning, K.E.; Bartlit, J.R.; Anderson, J.L.

1990-01-01

Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, ''JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing

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

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.

Investigation on the suitability of plasma sprayed Fe-Cr-Al coatings as tritium permeation barrier

International Nuclear Information System (INIS)

Fazio, C.; Serra, E.; Benamati, G.

1999-01-01

Results on the fabrication of a tritium permeation barrier by spraying Fe-Cr-Al powders are described. The sprayed coatings were deposited at temperatures below the A c1 temperature of the ferritic-martensitic steel substrate and no post-deposition heat treatment was applied. The aim of the investigation was the determination of the efficiency of the coatings to act as tritium permeation barrier. Metallurgical investigations as well as hydrogen isotope permeation measurements were carried out onto the produced coatings. The depositions were performed on ferritic-martensitic steels by means of three types of spray techniques: high velocity oxy fuel, air plasma spray and vacuum plasma spray. (orig.)

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

Fusion power production from TFTR plasmas fueled with deuterium and tritium

International Nuclear Information System (INIS)

Strachan, J.D.; Adler, H.; Alling, P.

1994-03-01

Peak fusion power production of 6.2 ± 0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2 x 10 17 m -3 without the appearance of either disruptive MHD events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits

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

Tritium in the Physical and Biological Sciences. V. 1. Proceedings of a Symposium

Energy Technology Data Exchange (ETDEWEB)

NONE

1962-01-15

The use of tritium for research in physics, chemistry, biology and hydrology has in recent years become increasingly important. It was for this reason that the first international conference to discuss the progress of new developments was organized by the IAEA in conjunction with the Joint Commission on Applied Radioactivity and held from 3-10 May 1961, in Vienna. The first five sessions of the Symposium were devoted to the use of tritium in hydrology, physics and chemistry. Special emphasis was laid on the role of tritium as a tracer in hydrology, especially in the study of water movement. The establishment and improvement of counting and detection techniques to facilitate the application of tritium as a tracer was another aspect discussed in this part of the proceedings. Papers were read on the preparation of tritiated compounds and it was generally agreed that further clarification of the mechanism of various techniques, and of the Wilzbach gas exposure technique in particular, would lead to further developments in the synthesis of a number of tritium compounds important in biology. Other papers were concerned with tritium applications to studies of the mechanism of some chemical reactions together with the effects of tritium isotopes. During the second part of the Symposium the biological applications of tritium and tritiated compounds were discussed. These included general problems connected with the biological uses of tritium and the radiation effects of tritium on living organisms such as viruses, bacteria and cancer cells. The value of tritium in biological studies became apparent because of the ease with which a large number of metabolically active compounds such as hormones, vitamins and other important constituents in the body can be labelled with tritium. Tritium is also a weak beta-emitter and autoradiographie s of tissues and single cells containing tritium-labelled compounds allow an excellent localization of the tracer. The Symposium was attended by

Tritium Decay Helium-3 Effects in Tungsten

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-01

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

Improvements in a tracer - encapsulated solid pellet and its injector for more advanced plasma diagnostics

International Nuclear Information System (INIS)

Tamura, Naoki; Suzuki, Chihiro; Funaba, Hisamichi; Hayashi, Hiromi; Maeno, Hiroya; Yokota, Mitsuhiro; Ogawa, Hideki; Sudo, Shigeru; Takagi, Masaru; Satoh, Nakahiro

2015-01-01

Although we are facing the age of the International Thermonuclear Experiment Reactor (ITER), many physics issues related to the confinement of magnetically-confined toroidal plasma still remain to be clarified. For example, under some conditions, impurities inside the magnetically-confined toroidal plasma tend to accumulate into the core region of the plasma. This will cause a dilution of fusion fuel. Moreover, a radiation loss from the core plasma will be enhanced due to the impurity accumulation, and then the temperature in the core region will be decreased dramatically. Consequently, fusion plasma performance will be degraded below the acceptable level. In order to develop strategy for obviating and suppressing the impurity accumulation, it is significantly important to gain a full understanding of the impurity transport in the magnetically-confined toroidal plasma. In consideration of such a situation, we have developed a Tracer-Encapsulated Solid Pellet (TESPEL) for promoting a precise study of the impurity transport. To put it plainly, the TESPEL is a double-layered impurity pellet. This form enables us to produce a both poloidally and toroidally localized 'tracer' impurity source in the plasma, and to specify the total amount of the tracer impurity deposited in the plasma precisely. In this contribution, we introduce new-type TESPELs, which are greatly improved in regard to the above-mentioned features. Owing to this improvement, we have achieved a shallower penetration of the TESPEL into the plasma with sufficient quantities of the tracer particles, which can be measured with the existing diagnostics. In addition, we also introduce a new TESPEL injector, which enables us to inject the TESPEL obliquely into the plasma. This injector can also contribute to a further shallower penetration of the TESPEL into the plasma. Moreover, we will discuss a future strategy of the TESPEL in the research of fusion plasma and plasma application. (author)

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

Improved iodine and tritium control in reprocessing plants

International Nuclear Information System (INIS)

Henrich, E.; Schmieder, H.; Roesch, W.; Weirich, F.

1981-01-01

During spent fuel processing, iodine and tritium are distributed in many aqueous, organic and gaseous process streams, which complicates their control. Small modifications of conventional purex flow sheets, compatible with processing in the headend and the first extraction cycle are necessary to confine the iodine and the tritium to smaller plant areas. The plant area connected to the dissolver off-gas (DOG) system is suited to confine the iodine and the plant area connected to the first aqueous cycle is suited to confine the tritium. A more clear and convenient iodine and tritium control will be achieved. Relevant process steps have been studied on a lab or a pilot plant scale using I-123 and H-3 tracer

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

Investigation of tritium in groundwater at Pickering NGS

International Nuclear Information System (INIS)

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

2001-01-01

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

Studies of tritium dispersion around and close to the buildings on the JET site

International Nuclear Information System (INIS)

Caldwell-Nichols, C.J.

1995-01-01

The dispersion of gases released to the environment at significant distances from the release point can be predicted using propriety computer codes. However during and after the Preliminary Tritium Experiment 1,2 (PTE) at JET in 1991 comparatively high levels of tritium were measured around the buildings and also there was measurable uptake of tritium in the site cooling water. Better assessment of likely tritium concentrations resulting from discharges is required to determine if tritium would tend to concentrate close to the buildings due to the complex air flow patterns around them. Three methods have been considered, namely computational studies, wind tunnel testing and tracer release experiments. A graduated approach has been adopted as each method has its limitations, tracer experiments being particularly expensive. Computational studies indicate that under worst case conditions the maximum ground level concentrations (Bq/m.) per unit stack release rate (Bq/s) is 1.0E-4 but more generally less. The results are presented noting the limitations of this approach. To aid understanding and verify some of the results, wind tunnel tests on a model of the JET site have been undertaken and the results discussed. The need for tracer release studies is considered. 3 refs., 6 figs

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

Determination of cortisol in blood plasma by isotopic dilution

International Nuclear Information System (INIS)

Shimizu, T.

1978-01-01

A method to measure cortisol in blood plasma is developed using tritium a tracer. Thin-layer chromatographer is used for very accurate separation and determination of cortisol free of impurities. Results show a concentration of 14,02+-5,79 μg/100ml for cortisol in blod. This method appears to be especially useful when interfering substances are present [pt

Synergistic effects of surface erosion on tritium inventory and permeation in metallic plasma facing armours

Science.gov (United States)

Federici, G.; Holland, D. F.; Matera, R.

1996-10-01

In the next generation of DT fuelled tokamaks, i.e., the International Thermonuclear Experimental Reactor (ITER) implantation of energetic DT particles on some portions of the plasma facing components (PFCs) will take place along with significant erosion of the armour surfaces. As a result of the simultaneous removal of material from the front surface, the build-up of tritium inventory and the start of permeation originating in the presence of large densities of neutron-induced traps is expected to be influenced considerably and special provisions could be required to minimise the consequences on the design. This paper reports on the results of a tritium transport modelling study based on a new model which describes the migration of implanted tritium across the bulk of metallic plasma facing materials containing neutron-induced traps which can capture it and includes the synergistic effects of surface erosion. The physical basis of the model is summarised, but emphasis is on the discussion of the results of a comparative study performed for beryllium and tungsten armours for ranges of design and operation conditions similar to those anticipated in the divertor of ITER.

Synergistic effects of surface erosion on tritium inventory and permeation in metallic plasma facing armours

International Nuclear Information System (INIS)

Federici, G.; Holland, D.F.; Matera, R.

1996-01-01

In the next generation of DT fuelled tokamaks, i.e., the international thermonuclear experimental reactor (ITER) implantation of energetic DT particles on some portions of the plasma facing components (PFCs) will take place along with significant erosion of the armour surfaces. As a result of the simultaneous removal of material from the front surface, the build-up of tritium inventory and the start of permeation originating in the presence of large densities of neutron-induced traps is expected to be influenced considerably and special provisions could be required to minimise the consequences on the design. This paper reports on the results of a tritium transport modelling study based on a new model which describes the migration of implanted tritium across the bulk of metallic plasma facing materials containing neutron-induced traps which can capture it and includes the synergistic effects of surface erosion. The physical basis of the model is summarised, but emphasis is on the discussion of the results of a comparative study performed for beryllium and tungsten armours for ranges of design and operation conditions similar to those anticipated in the divertor of ITER. (orig.)

Tritium decay helium-3 effects in tungsten

Directory of Open Access Journals (Sweden)

M. Shimada

2017-08-01

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

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

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)

Tritium module for ITER/Tiber system code

International Nuclear Information System (INIS)

Finn, P.A.; Willms, S.; Busigin, A.; Kalyanam, K.M.

1988-01-01

A tritium module was developed for the ITER/Tiber system code to provide information on capital costs, tritium inventory, power requirements and building volumes for these systems. In the tritium module, the main tritium subsystems/emdash/plasma processing, atmospheric cleanup, water cleanup, blanket processing/emdash/are each represented by simple scaleable algorithms. 6 refs., 2 tabs

The effective cost of tritium for tokamak fusion power reactors with reduced tritium production systems

International Nuclear Information System (INIS)

Gilligan, J.G.; Evans, K.

1983-01-01

If sufficient tritium cannot be produced and processed in tokamak blankets then at least two alternatives are possible. Tritium can be purchased; or reactors with reduced tritium (RT) content in the plasma can be designed. The latter choice may require development of magnet technology etc., but the authors show that the impact on the cost-of-electricity may be mild. Cost tradeoffs are compared to the market value of tritium. Adequate tritium production in fusion blankets is preferred, but the authors show there is some flexibility in the deployment of fusion if this is not possible

Commissioning of a DT fusion reactor without external supply of tritium

International Nuclear Information System (INIS)

Asaoka, Y.; Konishi, S.; Nishio, S.; Hiwatari, R.; Okano, K.; Yoshida, T.; Tomabechi, K.

2001-01-01

Commissioning of a DT fusion reactor without external supply of tritium is discussed. The DD reactions in a DT-oriented fusion reactor with external power injection by neutral beams produce tritium and neutrons. Tritium produced by the DD reaction together with that produced in the blanket by the 2.45 MeV neutron is re-circulated into the plasma. Then, the DT reaction rate increases gradually, as tritium concentration in plasma builds up towards the level of nominal operation. Time required to reach the nominal operational condition, i.e. 50 % tritium in plasma, is estimated with assumptions based on a model of fusion power plant. As a result, the start-up period of a DT fusion reactor without external supply of tritium is estimated to be approximately 55 days, with the plasma parameters of CREST having a high performance blanket and tritium processing systems. Major factors to determine the start-up period are DD and DT reaction rates, net tritium breeding gain of the plant and dead inventory in/on facing materials. Elimination of a constraint for fusion reactor deployment and operation without any tritium transportation in and out of plant through its entire life may be possible. (author)

A novel method for trace tritium transport studies

International Nuclear Information System (INIS)

Bonheure, Georges; Mlynar, Jan; Murari, A.; Giroud, C.; Popovichev, S.; Belo, P.; Bertalot, L.

2009-01-01

A new method combining a free-form solution for the neutron emissivity and the ratio method (Bonheure et al 2006 Nucl. Fusion 46 725-40) is applied to the investigation of tritium particle transport in JET plasmas. The 2D neutron emissivity is calculated using the minimum Fisher regularization method (MFR) (Anton et al 1996 Plasma Phys. Control. Fusion 38 1849, Mlynar et al 2003 Plasma Phys. Control. Fusion 45 169). This method is being developed and studied alongside other methods at JET. The 2D neutron emissivity was significantly improved compared with the first MFR results by constraining the emissivity along the magnetic flux surfaces. 1D profiles suitable for transport analysis are then obtained by subsequent poloidal integration. In methods on which previous JET publications are based (Stork et al 2005 Nucl. Fusion 45 S181, JET Team (prepared by Zastrow) 1999 Nucl. Fusion 39 1891, Zastrow et al 2004 Plasma Phys. Control. Fusion 46 B255, Adams et al 1993 Nucl. Instrum. Methods A 329 277, Jarvis et al 1997 Fusion Eng. Des. 34-35 59, Jarvis et al 1994 Plasma Phys. Control. Fusion 36 219), the 14.07 MeV D-T neutron line integrals measurements were simulated and the transport coefficients varied until good fits were obtained. In this novel approach, direct knowledge of tritium concentration or the fuel ratio n T /n D is obtained using all available neutron profile information, e.g both 2.45 MeV D-D neutron profiles and 14.07 MeV D-T neutron profiles (Bonheure et al 2006 Nucl.Fusion 46 725-40). Tritium particle transport coefficients are then determined using a linear regression from the dynamic response of the tritium concentration n T /n D profile. The temporal and spatial evolution of tritium particle concentration was studied for a set of JET discharges with tritium gas puffs from the JET trace tritium experiments. Local tritium transport coefficients were derived from the particle flux equation Γ = -D∇n T + Vn T , where D is the particle diffusivity and V

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)

Proceedings of the 4th International Workshop on Tritium Effects in Plasma Facing Components

International Nuclear Information System (INIS)

Causey, R. A.

1999-01-01

The 4th International Workshop on Tritium Effects in Plasma Facing Components was held in Santa Fe, New Mexico on May 14-15, 1998. This workshop occurs every two years, and has previously been held in Livermore/California, Nagoya/Japan, and the JRC-Ispra Site in Italy. The purpose of the workshop is to gather researchers involved in the topic of tritium migration, retention, and recycling in materials used to line magnetic fusion reactor walls and provide a forum for presentation and discussions in this area. This document provides an overall summary of the workshop, the workshop agenda, a summary of the presentations, and a list of attendees

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)

Fusion reactor start-up without an external tritium source

Energy Technology Data Exchange (ETDEWEB)

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

2016-02-15

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

Tritium retention and clean-up in JET

International Nuclear Information System (INIS)

Andrew, P.; Brennan, P.D.; Coad, J.P.

1999-01-01

During 1997 JET operation with D-T plasmas, 35 g of tritium were introduced into the torus, mainly by gas puffing. It was found that during this period, the torus tritium inventory would accumulate at a rate of about 40% of the input. After tritium operation ceased, the experimental program continued with deuterium- and hydrogen-fuelled experiments, during which time the tritium inventory decreased to about 17% of the total input. Techniques aimed at detritiation of the torus included methods using deuterium gas (such as deuterium pulsing) which were used in the middle of the experimental campaign, and methods which could adversely affect the torus vacuum conditions (such as air purges) which were reserved for the period after the experimental campaign. Whilst it was found that the plasma tritium fraction could be reduced to below the 1% level in a few days, the tritium inventory reached a virtually steady level of about 6 g by the end of the campaign. (orig.)

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%

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

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

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

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

Tracer dating and ocean ventilation

International Nuclear Information System (INIS)

Thiele, G.; Sarmiento, J.L.

1990-01-01

The interpretation of transient tracer observations depends on difficult to obtain information on the evolution in time of the tracer boundary conditions and interior distributions. Recent studies have attempted to circumvent this problem by making use of a derived quantity, age, based on the simultaneous distribution of two complementary tracers, such as tritium and its daughter, helium 3. The age is defined with reference to the surface such that the boundary condition takes on a constant value of zero. The authors use a two-dimensional model to explore the circumstances under which such a combination of conservation equations for two complementary tracers can lead to a cancellation of the time derivative terms. An interesting aspect of this approach is that mixing can serve as a source or sink of tracer based age. The authors define an idealized ventilation age tracer that is conservative with respect to mixing, and they explore how its behavior compares with that of the tracer-based ages over a range of advective and diffusive parameters

Elucidation of hydrogen mobility in tetralin under coal liquefaction conditions using a tritium tracer method. Effects of the addition of H2S and H2O; Tritium tracer ho wo mochiita sekitan ekika hanno jokenka deno tetralin no suiso idosei hyoka. Ryuka suiso oyobi mizu no tenka koka

Energy Technology Data Exchange (ETDEWEB)

Kanbe, M.; Saito, M.; Ishihara, A.; Kabe, T. [Tokyo University of Agriculture and Technology, Tokyo (Japan)

1996-10-28

It was previously reported that the tritium tracer method is useful for the quantitative consideration of hydrogen behavior in coal during coal liquefaction reaction. Tetralin is excellent hydrogen donating solvent, and is considered as one of the model compounds of coal. In this study, effects of H2S and H2O on the hydrogen exchange reaction between tetralin and gaseous hydrogen labeled by tritium were investigated. It was suggested that the conversion of tetralin and the hydrogen exchange reaction between gaseous hydrogen and tetralin proceed through the radical reaction mechanism with a tetralyl radical as an intermediate product. When H2S existed in this reaction, the hydrogen exchange yield increased drastically without changing the conversion yield. This suggested that the hydrogen exchange reaction proceeds even in the reaction where radical does not give any effect. In the case of H2O addition, the conversion yield and hydrogen exchange rate decreased into a half or one-third. It was suggested that H2O inhibited the formation process of tetralyl radical. 6 refs., 4 figs.

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

Tritium in groundwater investigation at the Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

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

2001-01-01

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

Neutron Profiles and Fuel Ratio nT /nD Measurements in JET ELMy H-mode Plasmas with Tritium Puff

Czech Academy of Sciences Publication Activity Database

Bonheure, G.; Popovichev, S.; Bertalot, L.; Murari, A.; Conroy, S.; Mlynář, Jan; Voitsekhovitch, I.

2006-01-01

Roč. 46, č. 7 (2006), s. 725-740 ISSN 0029-5515 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * JET * plasma profile * tomography * neutron diagnostics * fuel * tritium transport Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.839, year: 2006

Study on conceptual design system of tritium production fusion reactor

International Nuclear Information System (INIS)

He Kaihui

2004-11-01

Conceptual design of an advanced tritium production reactor based on spherical torus, which is intermediate application of fusion energy, was presented. Different from traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can within vacuum vessel in order to produce 1 kg excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented. Besides systematical analyses; design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (author)

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

Benefits of the self-pumping system are the elimination of vacuum ducts, pumps, and penetration shielding (except for a very small startup system), and the reduction of tritium recycle and refueling. In addition, a self-pumped system may perform better and last longer than alternative systems such as a pumped limiter. The reference case here is a self-cooled lithium/vanadium blanket with a first wall/limiter. This concept combines the functions of first wall and limiter into a single first-wall structure. The wall is shaped in accordance with the outermost plasma flux surface. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. The entire wall area is used for helium trapping. The tritium inventory, tritium permeation rate, and plasma protium concentration for the vanadium wall as a function of the number of years of operation are calculated. The tritium inventory is acceptable, the protium concentration in the plasma is acceptably small, and the tritium permeation rate is moderate. At the start of operation, it is equal to about five times the tritium burnup rate. This tritium will enter the coolant and the cost of the blanket tritium recovery system will be higher

Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

International Nuclear Information System (INIS)

Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

1995-01-01

The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

Long Term Tritium Trapping in TFTR and JET

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Young, K.M.; Coad, J.P.; Hogan, J.T.; Penzhorn, R.-D.; Bekris, N.

2001-01-01

Tritium retention in TFTR [Tokamak Fusion Test Reactor] and JET [Joint European Torus] shows striking similarities and contrasts. In TFTR, 5 g of tritium were injected into circular plasmas over a 3.5 year period, mostly by neutral-beam injection. In JET, 35 g were injected into divertor plasmas over a 6 month campaign, mostly by gas puffing. In TFTR, the bumper limiter provided a large source of eroded carbon and a major part of tritium was co-deposited on the limiter and vessel wall. Only a small area of the co-deposit flaked off. In JET, the wall is a net erosion area, and co-deposition occurs principally in shadowed parts of the inner divertor, with heavy flaking. In both machines, the initial tritium retention, after a change from deuterium [D] to tritium [T] gas puffing, is high and is due to isotope exchange with deuterium on plasma-facing surfaces (dynamic inventory). The contribution of co-deposition is lower but cumulative, and is revealed by including periods of D fueling that reversed the T/D isotope exchange. Ion beam analysis of flakes from TFTR showed an atomic D/C ratio of 0.13 on the plasma facing surface, 0.25 on the back surface and 0.11 in the bulk. Data from a JET divertor tile showed a larger D/C ratio with 46% C, 30% D, 20% H and 4% O. Deuterium, tritium, and beryllium profiles have been measured and show a thin less than 50 micron co-deposited layer. Flakes retrieved from the JET vacuum vessel exhibited a high tritium release rate of 2e10 Bq/month/g. BBQ modeling of the effect of lithium on retention in TFTR showed overlapping lithium and tritium implantation and a 1.3x increase in local T retention

Theory and code development for evaluation of tritium retention and exhaust in fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Ohya, Kaoru; Inai, Kensuke [Univ. of Tokushima, Institute of Technology and Science, Tokushima, Tokushima (Japan); Shimizu, Katsuhiro; Takizuka, Tomonori; Kawashima, Hisato; Hoshino, Kazuo [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki (Japan); Hatayama, Akiyoshi; Toma, Mitsunori [Keio Univ., Faculty of Science and Technology, Yokohama, Kanagawa (Japan); Tomita, Yukihiro; Kawamura, Gakushi; Ashikawa, Naoko; Nakamura, Hiroaki; Ito, Atsushi; Kato, Daiji [National Inst. for Fusion Science, Toki, Gifu (Japan); Tanaka, Yasunori [Kanazawa Univ., College of Science and Engineering, Kanazawa, Ishikawa (Japan); Ono, Tadayoshi; Muramoto, Tetsuya [Okayama Univ. of Science, Faculty of Informatics, Okayama, Okayama (Japan); Kenmotsu, Takahiro [Doshisha Univ., Faculty of Life and Medical Science, Kiyotanabe, Kyoto (Japan)

2009-10-15

As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated three research programs on the theory and code development for evaluation of tritium retention and exhaust in fusion reactor have been conducted by the A02 team. They include: (1) Tritium transport in fusion plasmas and the adsorption and desorption property of tritium in plasma-facing components. (2) Behavior of dusts in fusion plasmas and their adsorption property of tritium. (3) Development of computer codes to simulate tritium retention in and release from plasma-facing materials. In order to study these issues, considerable effort has been paid to the development of computer codes and the database system. (J.P.N.)

Theory and code development for evaluation of tritium retention and exhaust in fusion reactor

International Nuclear Information System (INIS)

Ohya, Kaoru; Inai, Kensuke; Shimizu, Katsuhiro; Takizuka, Tomonori; Kawashima, Hisato; Hoshino, Kazuo; Hatayama, Akiyoshi; Toma, Mitsunori; Tomita, Yukihiro; Kawamura, Gakushi; Ashikawa, Naoko; Nakamura, Hiroaki; Ito, Atsushi; Kato, Daiji; Tanaka, Yasunori; Ono, Tadayoshi; Muramoto, Tetsuya; Kenmotsu, Takahiro

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 three research programs on the theory and code development for evaluation of tritium retention and exhaust in fusion reactor have been conducted by the A02 team. They include: (1) Tritium transport in fusion plasmas and the adsorption and desorption property of tritium in plasma-facing components. (2) Behavior of dusts in fusion plasmas and their adsorption property of tritium. (3) Development of computer codes to simulate tritium retention in and release from plasma-facing materials. In order to study these issues, considerable effort has been paid to the development of computer codes and the database system. (J.P.N.)

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

Influence of radiative processes on the ignition of deuterium–tritium plasma containing inactive impurities

Energy Technology Data Exchange (ETDEWEB)

Gus’kov, S. Yu., E-mail: guskov@sci.lebedev.ru [Russian Academy of Sciences, Lebedev Physical Institute (Russian Federation); Sherman, V. E. [Peter the Great St. Petersburg Polytechnic University (Russian Federation)

2016-08-15

The degree of influence of radiative processes on the ignition of deuterium–tritium (DT) plasma has been theoretically studied as dependent on the content of inactive impurities in plasma. The analytic criterion of plasma ignition in inertial confinement fusion (ICF) targets is modified taking into account the absorption of intrinsic radiation from plasma in the ignition region. The influence of radiative processes on the DT plasma ignition has been analytically and numerically studied for plasma that contains a significant fraction of inactive impurities either as a result of DT fuel mixing with ICF target ablator material or as a result of using light metal DT-hydrides as solid noncryogenic fuel. It has been shown that the effect of the absorption of intrinsic radiation leads to lower impurity-induced increase in the ignition energy as compared to that calculated in the approximation of optically transparent ignition region.

1st IAEA research coordination meeting on tritium retention in fusion reactor plasma facing components. October 5-6, 1995, Vienna, Austria. Summary report

International Nuclear Information System (INIS)

Langley, R.A.

1995-12-01

The proceedings and results of the 1st IAEA research Coordination Meeting on ''Tritium Retention in Fusion Reactor Plasma Facing Components'' held on October 5 and 6, 1995 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of a data survey and needs assessment for the retention, release and removal of tritium from plasma facing components, a summary of data evaluation, and recommendations regarding future work. (author). 4 tabs

Conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production fusion reactor based on spherical torus, which is intermediate application of fusion energy, was presented in this paper. Differing from the traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and maximize tritium breeding ratio with arrangement of tritium production blankets within vacuum vessel as possible in order to produce 1 kg excess tritium except need of self-sufficient plasma core with 40% or more corresponding plant availability. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented, providing the backgrounds and reference for next detailed conceptual design

Temporal variation of tritium in spring water of East Sikkim region

International Nuclear Information System (INIS)

Pant, Diksha; Ansari, Md. Arzoo; Mendhekar, G.N.; Kamble, S.N; Sinha, U.K; Dash, A.; Dhakal, Deepak

2016-01-01

Tritium is produced in the atmosphere by the interaction of cosmic rays with the nuclei of the atmospheric gases (mainly nitrogen, σ = 0.388 barn), principally by neutron induced reactions. It is estimated from the natural abundance of tritium that the rate of production is approximately 0.2 tritium atoms/sec.cm 2 area of the earth's surface. Additionally it is possible that tritium may enter the atmosphere from anthropogenic activities like nuclear bomb testing or nuclear reactor. Tritium (T 1/2 = 4540 days) is a particularly suitable tracer for water since hydrogen is part of the water molecule. Tritium can be used for assessing the recharge characteristics of aquifers, in studying artificial recharge characteristics and in determining the 'age' of water with an upper time limit of about 50 years. The objective is to study the temporal changes of tritium content in spring's water of East Sikkim region. Tritium helps in predicting whether the contribution to spring water in rainwater or some other source

A system dynamics model for tritium cycle of pulsed fusion reactor

International Nuclear Information System (INIS)

Zhu, Zuolong; Nie, Baojie; Chen, Dehong

2017-01-01

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

A system dynamics model for tritium cycle of pulsed fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Zhu, Zuolong; Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Chen, Dehong, E-mail: dehong.chen@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2017-05-15

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

Bromide as a tracer for studying water movement and nitrate displacement in soils: comparison with stable isotope tracers; Bromid als Tracer zur Untersuchung der Wasserbewegung und der Nitratverlagerung in Boeden: Vergleich mit stabilisotopen Tracern

Energy Technology Data Exchange (ETDEWEB)

Russow, R.; Knappe, S. [UFZ - Umweltforschungszentrum Leipzig-Halle GmbH, Bad Lauchstaedt (Germany). Sektion Bodenforschung

1999-02-01

Tracers are an ideal means of studying water movement and associated nitrate displacement. Often bromide is preferred as a tracer because it is considered a representative tracer for water and because, being a conservative tracer (i.e. not involved in chemical and biological soil processes), it can be used for studying anion transport in soils. Moreover, it is less expensive and easier to measure than the stable isotopes deuterium and {sup 15}N. Its great advantage over radioactive tracers (e.g. tritium), which outweighs their extreme sensitivity and ease of measurement and which it has in common with stable isotopes, is that it does not require radiation protection measures. However, there are also constraints on the use of bromide as a tracer in soil/water/plant systems. Our own studies on different soils using D{sub 2}O, bromide and [{sup 15}N]-nitrate in lysimeters suggest that the above assumptions on bromide tracers need not always be valid under conditions as they prevail in biologically active soils. As the present paper shows, these studies permit a good assessment of the possibilities and limits to these tracers. [Deutsch] Fuer die Untersuchung der Wasserbewegung sowie der daran gekoppelten Nitrat-Verlagerung ist der Einsatz von Tracern das Mittel der Wahl. Dabei wird Bromid als Tracer haeufig bevorzugt, da es allgemein als ein repraesentativer Tracer fuer Wasser und als konservativer Tracer (nicht involviert in chemische und biologische Bodenprozesse) zur Untersuchung des Anionentransportes in Boeden angesehen wird und es gegenueber den stabilen Isotopen Deuterium und {sup 15}N billiger und einfacher zu bestimmen ist. Gegenueber den radioaktiven Tracern (z.B. Tritium), die zwar sehr empfindlich und einfach messbar sind, besteht der grosse Vorteil, dass, wie bei den stabilen Isotopen, keine Strahlenschutzmassnahmen ergriffen werden muessen. Es gibt jedoch auch einschraenkende Hinweise fuer die Verwendung von Bromid als Tracer im System Boden

Scoping studies of tritium handling in a tokamak experimental power reactor

International Nuclear Information System (INIS)

Cherdack, R.; Watson, J.S.; Clinton, S.D.; Fisher, P.W.

1975-01-01

Tritium handling techniques in an experimental fusion power reactor (EPR) are evaluated to determine the requirements of the system and to compare different equipment and techniques for meeting those requirements. Tritium process equipment is needed to (1) evacuate and maintain a vacuum in the plasma vessel and the neutral beam injectors, (2) purify and recycle tritium and deuterium for the plasma fuel cycle, (3) recover tritium from experimental breeding modules, and (4) provide tritium containment and atmospheric cleanup. A development program is outlined to develop and demonstrate the required techniques and equipment and to permit confident design of an EPR for operation by the mid-1980s

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.)

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

Tritium labeling for bio-med research

International Nuclear Information System (INIS)

Lemmon, R.M.

1980-01-01

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

Aggregation effects on tritium-based mean transit times and young water fractions in spatially heterogeneous catchments and groundwater systems

Science.gov (United States)

Stewart, Michael K.; Morgenstern, Uwe; Gusyev, Maksym A.; Małoszewski, Piotr

2017-09-01

Kirchner (2016a) demonstrated that aggregation errors due to spatial heterogeneity, represented by two homogeneous subcatchments, could cause severe underestimation of the mean transit times (MTTs) of water travelling through catchments when simple lumped parameter models were applied to interpret seasonal tracer cycle data. Here we examine the effects of such errors on the MTTs and young water fractions estimated using tritium concentrations in two-part hydrological systems. We find that MTTs derived from tritium concentrations in streamflow are just as susceptible to aggregation bias as those from seasonal tracer cycles. Likewise, groundwater wells or springs fed by two or more water sources with different MTTs will also have aggregation bias. However, the transit times over which the biases are manifested are different because the two methods are applicable over different time ranges, up to 5 years for seasonal tracer cycles and up to 200 years for tritium concentrations. Our virtual experiments with two water components show that the aggregation errors are larger when the MTT differences between the components are larger and the amounts of the components are each close to 50 % of the mixture. We also find that young water fractions derived from tritium (based on a young water threshold of 18 years) are almost immune to aggregation errors as were those derived from seasonal tracer cycles with a threshold of about 2 months.

Simulation of a field scale tritium tracer experiment in a fractured, weathered shale using discrete-fracture/matrix-diffusion and equivalent porous medium models

Energy Technology Data Exchange (ETDEWEB)

Stafford, Paige L. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Geological Sciences

1996-05-01

Simulations of a tritium tracer experiment in fractured shale saprolite, conducted at the Oak Ridge National Laboratory, were performed using 1D and 2D equivalent porous medium (EPM) and discrete-fracture/matrix-diffusion (DFMD) models. The models successfully reproduced the general shape of the breakthrough curves in down-gradient monitoring wells which are characterized by rapid first arrival, a slow-moving center of mass, and a persistent ``tail`` of low concentration. In plan view, the plume shows a large degree of transverse spreading with the width almost as great as the length. EPM models were sensitive to dispersivity coefficient values which had to be large (relative to the 3.7m distance between the injection and monitoring wells) to fit the tail and transverse spreading. For example, to fit the tail a longitudinal dispersivity coefficient, α_L, of 0.8 meters for the 2D simulations was used. To fit the transverse spreading, a transverse dispersivity coefficient, α_T, of 0.8 to 0.08 meters was used indicating an α_L/α_T ratio between 10 and 1. Transverse spreading trends were also simulated using a 2D DFMD model using a few larger aperture fractures superimposed onto an EPM. Of the fracture networks studied, only those with truncated fractures caused transverse spreading. Simulated tritium levels in all of the cases were larger than observed values by a factor of approximately 100. Although this is partly due to input of too much tritium mass by the models it appears that dilution in the wells, which were not purged prior to sampling, is also a significant factor. The 1D and 2D EPM models were fitted to monitoring data from the first five years of the experiment and then used to predict future tritium concentrations.

Simulation of a field scale tritium tracer experiment in a fractured, weathered shale using discrete-fracture/matrix-diffusion and equivalent porous medium models

International Nuclear Information System (INIS)

Stafford, P.L.

1996-05-01

Simulations of a tritium tracer experiment in fractured shale saprolite, conducted at the Oak Ridge National Laboratory, were performed using 1D and 2D equivalent porous medium (EPM) and discrete-fracture/matrix-diffusion (DFMD) models. The models successfully reproduced the general shape of the breakthrough curves in down-gradient monitoring wells which are characterized by rapid first arrival, a slow-moving center of mass, and a persistent ''tail'' of low concentration. In plan view, the plume shows a large degree of transverse spreading with the width almost as great as the length. EPM models were sensitive to dispersivity coefficient values which had to be large (relative to the 3.7m distance between the injection and monitoring wells) to fit the tail and transverse spreading. For example, to fit the tail a longitudinal dispersivity coefficient, α L , of 0.8 meters for the 2D simulations was used. To fit the transverse spreading, a transverse dispersivity coefficient, α T , of 0.8 to 0.08 meters was used indicating an α L /α T ratio between 10 and 1. Transverse spreading trends were also simulated using a 2D DFMD model using a few larger aperture fractures superimposed onto an EPM. Of the fracture networks studied, only those with truncated fractures caused transverse spreading. Simulated tritium levels in all of the cases were larger than observed values by a factor of approximately 100. Although this is partly due to input of too much tritium mass by the models it appears that dilution in the wells, which were not purged prior to sampling, is also a significant factor. The 1D and 2D EPM models were fitted to monitoring data from the first five years of the experiment and then used to predict future tritium concentrations

Use of artificial tracers in hydrology

International Nuclear Information System (INIS)

1991-05-01

The IAEA has convened an Advisory Group Meeting with the following objectives: To define the role of artificial radioactive tracers for water tracing in comparison with other non-radioactive tracers. To evaluate the real needs of artificial radioactive tracers in hydrology. To identify the fields for which artificial radioactive tracers are useful as well as those in which they can be substituted by other tracers. To discuss the strategy to be adopted to overcome the difficulties derived from the restrictions on the use of radioactive tracers in hydrology. The meeting was held at IAEA Headquarters from 19 to 22 March 1990, and was attended by 30 participants from 15 Member States. The conclusions and recommendations are that the use of artificial radioactive tracers should be restricted to cases where other tracers cannot be used or do not provide the same quality of information. Tritium, iodine-131, bromine-82, chromium-51 in the form of Cr-EDTA, technetium-99m obtained from 99 Mo-generators and gold-198 as an adsorbable tracer are, practically, the only radionuclides used for water tracing. The use of other radionuclides for this purpose does not appear to be necessary, possible and/or convenient. Refs, figs and tabs

Bromide as a tracer for studying water movement and nitrate displacement in soils: comparison with stable isotope tracers

International Nuclear Information System (INIS)

Russow, R.; Knappe, S.

1999-01-01

Tracers are an ideal means of studying water movement and associated nitrate displacement. Often bromide is preferred as a tracer because it is considered a representative tracer for water and because, being a conservative tracer (i.e. not involved in chemical and biological soil processes), it can be used for studying anion transport in soils. Moreover, it is less expensive and easier to measure than the stable isotopes deuterium and 15 N. Its great advantage over radioactive tracers (e.g. tritium), which outweighs their extreme sensitivity and ease of measurement and which it has in common with stable isotopes, is that it does not require radiation protection measures. However, there are also constraints on the use of bromide as a tracer in soil/water/plant systems. Our own studies on different soils using D 2 O, bromide and [ 15 N]-nitrate in lysimeters suggest that the above assumptions on bromide tracers need not always be valid under conditions as they prevail in biologically active soils. As the present paper shows, these studies permit a good assessment of the possibilities and limits to these tracers [de

Scoping calculations for groundwater transport of tritium from the Gnome Site, New Mexico

International Nuclear Information System (INIS)

Pohlmann, K.; Andricevic, R.

1994-08-01

Analytic solutions are employed to investigate potential groundwater transport of tritium from a radioactive tracer site near the Project Gnome site in southeastern New Mexico. The tracer test was conducted in 1963 and introduced significant quantities of radionuclides to the transmissive and laterally continuous Culebra dolomite. Groundwater in the Culebra near Gnome travels toward a regional discharge point at the Pecos River, a distance of about 10 to 15 km, depending on flow path. Groundwater transport of radionuclides from the Gnome site is therefore of interest due to the proximity of the accessible environment and the 31-year time period during which migration is likely to have occurred. The analytical stochastic solutions used incorporate the heterogeneity observed in the Culebra by treating transmissivity as a spatially correlated random field. The results indicate that significant spreading of tritium will occur in the Culebra dolomite as a result of the combination of relatively high transmissivity, high spatial variability, and high spatial correlation of transmissivity. Longitudinal spreading may cause a very small fraction of tritium mass to arrive at the Pecos River within the 31 years since the tracer test. However, dilution and transverse dispersion will act to distribute this mass over a very large volume, thereby reducing groundwater concentrations. Despite the high degree of spreading, the calculations indicate that most of the tritium remains near the source. At present, the center of mass is estimated to have moved approximately 260 m downgradient of the test location and about 95 percent of the mass is estimated to have remained within about 1 km downgradient

Tritium contamination experience in an operational D-T fusion reactor

International Nuclear Information System (INIS)

Gentile, C.A.; Ascione, G.

1994-01-01

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

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

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

Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties

OpenAIRE

中村 博文; 西 正孝

2003-01-01

Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium transport properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authors' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evalua...

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

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

Future fusion devices will use tritium and deuterium fuel. Because tritium is both radioactive and expensive, it is absolutely necessary that there be an understanding of the tritium retention characteristics of the materials used in these devices as well as how to recover the tritium. There are three materials that are strong candidates for plasma-facing-material use in next-step fusion devices. These are beryllium, tungsten, and carbon. While beryllium has the disadvantage of high sputtering and low melting point (which limits its power handling capabilities in divertor areas), it has the advantages of being a low-Z material with a good thermal conductivity and the ability to get oxygen from the plasma. Due to beryllium's very low solubility for hydrogen, implantation of beryllium with deuterium and tritium results in a saturated layer in the very near-surface with limited inventory (J. Nucl. Mater. 273 (1999) 1). Unfortunately, there are nuclear reactions generated by neutrons that will breed tritium and helium in the material bulk (J. Nucl. Mater. 179 (1991) 329). This process will lead to a substantial tritium inventory in the bulk of the beryllium after long-term neutron exposure (i.e. well beyond the operation life time of a next-step reactor like ITER). Tungsten is a high-Z material that will be used in the divertor region of next-step devices (e.g. ITER) and possibly as a first wall material in later devices. The divertor is the preferred location for tungsten use because net erosion is very low there due to low sputtering and high redeposition. While experiments are still continuing on tritium retention in tungsten, present data suggest that relatively low tritium inventories will result with this material (J. Nucl. Mater. 290-293 (2001) 505). For tritium inventories, carbon is the problem material. Neutron damage to the graphite can result in substantial bulk tritium retention (J. Nucl. Mater. 191-194 (1992) 368), and codeposition of the sputtered carbon

Chemical Tracer Methods: Chapter 7

Science.gov (United States)

Healy, Richard W.

2017-01-01

Tracers have a wide variety of uses in hydrologic studies: providing quantitative or qualitative estimates of recharge, identifying sources of recharge, providing information on velocities and travel times of water movement, assessing the importance of preferential flow paths, providing information on hydrodynamic dispersion, and providing data for calibration of water flow and solute-transport models (Walker, 1998; Cook and Herczeg, 2000; Scanlon et al., 2002b). Tracers generally are ions, isotopes, or gases that move with water and that can be detected in the atmosphere, in surface waters, and in the subsurface. Heat also is transported by water; therefore, temperatures can be used to trace water movement. This chapter focuses on the use of chemical and isotopic tracers in the subsurface to estimate recharge. Tracer use in surface-water studies to determine groundwater discharge to streams is addressed in Chapter 4; the use of temperature as a tracer is described in Chapter 8.Following the nomenclature of Scanlon et al. (2002b), tracers are grouped into three categories: natural environmental tracers, historical tracers, and applied tracers. Natural environmental tracers are those that are transported to or created within the atmosphere under natural processes; these tracers are carried to the Earth’s surface as wet or dry atmospheric deposition. The most commonly used natural environmental tracer is chloride (Cl) (Allison and Hughes, 1978). Ocean water, through the process of evaporation, is the primary source of atmospheric Cl. Other tracers in this category include chlorine-36 (36Cl) and tritium (3H); these two isotopes are produced naturally in the Earth’s atmosphere; however, there are additional anthropogenic sources of them.

Influence of neutron irradiation on the tritium retention in beryllium

Energy Technology Data Exchange (ETDEWEB)

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

1998-01-01

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

Aggregation effects on tritium-based mean transit times and young water fractions in spatially heterogeneous catchments and groundwater systems

Directory of Open Access Journals (Sweden)

M. K. Stewart

2017-09-01

Full Text Available Kirchner (2016a demonstrated that aggregation errors due to spatial heterogeneity, represented by two homogeneous subcatchments, could cause severe underestimation of the mean transit times (MTTs of water travelling through catchments when simple lumped parameter models were applied to interpret seasonal tracer cycle data. Here we examine the effects of such errors on the MTTs and young water fractions estimated using tritium concentrations in two-part hydrological systems. We find that MTTs derived from tritium concentrations in streamflow are just as susceptible to aggregation bias as those from seasonal tracer cycles. Likewise, groundwater wells or springs fed by two or more water sources with different MTTs will also have aggregation bias. However, the transit times over which the biases are manifested are different because the two methods are applicable over different time ranges, up to 5 years for seasonal tracer cycles and up to 200 years for tritium concentrations. Our virtual experiments with two water components show that the aggregation errors are larger when the MTT differences between the components are larger and the amounts of the components are each close to 50 % of the mixture. We also find that young water fractions derived from tritium (based on a young water threshold of 18 years are almost immune to aggregation errors as were those derived from seasonal tracer cycles with a threshold of about 2 months.

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

Tritium contamination experience in an operational D-T fusion reactor

Energy Technology Data Exchange (ETDEWEB)

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

1994-09-01

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

Specific features of X-ray generation by plasma focus chambers with deuterium and deuterium–tritium fillings

Energy Technology Data Exchange (ETDEWEB)

Dulatov, A. K., E-mail: bogolubov@vniia.ru; Krapiva, P. S.; Lemeshko, B. D.; Mikhailov, Yu. V.; Moskalenko, I. N.; Prokuratov, I. A.; Selifanov, A. N. [All-Russia Research Institute of Automatics (Russian Federation)

2016-01-15

The process of hard X-ray (HXR) generation in plasma focus (PF) chambers was studied experimentally. The radiation was recorded using scintillation detectors with a high time resolution and thermoluminescent detectors in combination with the method of absorbing filters. Time-resolved analysis of the processes of neutron and X-ray generation in PFs is performed. The spectra of HXR emission from PF chambers with deuterium and deuterium–tritium fillings are determined. In experiments with PF chambers filled with a deuterium–tritium mixture, in addition to the HXR pulse with photon energies of up to 200–300 keV, a γ-ray pulse with photon energies of up to 2.5–3.0 MeV is recorded, and a mechanism of its generation is proposed.

Tritium systems for the TITAN reversed-field pinch fusion reactor design

International Nuclear Information System (INIS)

Martin, R.C.; Sze, D.K.; Bartlit, J.R.; Gierszewski, P.J.

1987-01-01

Tritium systems for the TITAN reversed-field pinch (RFP) fusion reactor study have been designed for two blanket concepts. The TITAN-1 design uses a self-cooled liquid-lithium blanket. The TITAN-2 design uses a self-cooled aqueous-solution blanket, with lithium nitrate dissolved in the water for tritium breeding. Tritium inventory, release, and safety margins are within regulatory limits, at acceptable costs. Major issues for TITAN-1 are plasma-driven permeation, the need for a secondary coolant loop, tritium storage requirements, redundancy in the plasma exhaust system, and minimal isotopic distillation of the exhaust. TITAN-1 fuel cleanup, reprocessing, and air detritiation systems are described in detail

Results from deuterium-tritium tokamak confinement experiments

International Nuclear Information System (INIS)

Hawryluk, R.J.

1997-02-01

Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which enabled the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation which have both good confinement and are MHD stable have enabled a broad study of burning plasma physics issues. A review of the technical and scientific results from the deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) is given with particular emphasis on alpha-particle physics issues

Technology and component development for a closed tritium cycle

International Nuclear Information System (INIS)

Penzhorn, R.D.; Haange, R.; Hircq, B.; Meikle, A.; Naruse, Y.

1991-01-01

A brief summary on recent advances in the field of tritium technology concerning the most important subsystems of the fuel cycle of a fusion reactor, i.e. the plasma exhaust pumping system, the exhaust gas clean up system, the isotope separation, the tritium storage and the tritium extraction from a blanket is provided. Experimental results, single component developments, and technical tests including those with relevant amounts of tritium that constitute the basis of proposed integral process concepts are described. 48 refs., 2 tabs

Technology and component development for a closed tritium cycle

International Nuclear Information System (INIS)

Hircq, B.; Penzhorn, R.D.; Haange, R.; Naruse, Y.

1991-01-01

A brief summary on recent advances in the field of tritium technology concerning the most important subsystems of the fuel cycle of a fusion reactor, i.e. the plasma exhaust pumping system, the exhaust gas clean up system, the isotope separation, the tritium storage and the tritium extraction from a blanket is provided. Experimental results, single component developments, and technical tests including those with relevants amounts of tritium that constitute the basis of proposed integral process concepts are described. 48 refs

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

Deuterium-tritium fuel self-sufficiency in fusion reactors

International Nuclear Information System (INIS)

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

1986-01-01

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

Tritium production, management and its impact on safety for a D-3He fusion reactor

International Nuclear Information System (INIS)

Sze, D.K.; Herring, S.; Sawan, M.

1991-11-01

About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

TPE upgrade for enhancing operational safety and improving in-vessel tritium inventory assessment in fusion nuclear environment

Energy Technology Data Exchange (ETDEWEB)

Shimada, M., E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Taylor, C.N.; Moore-McAteer, L.; Pawelko, R.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Kolasinski, R.D.; Buchenauer, D.A. [Sandia National Laboratories, Hydrogen and Materials Science Department, Livermore, CA 94550 (United States); Cadwallader, L.C.; Merrill, B.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2016-11-01

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γ{sub i}{sup max} of >1.0 × 10{sup 23} m{sup −2} s{sup −1} and q{sub heat} of >1 MW m{sup −2} with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

Tritium breeding blanket device of D-T reactors

International Nuclear Information System (INIS)

Chevereau, G.

1984-01-01

This blanket device uses solid tritium breeding materials as those which include, in a known manner, near a neutron breeding plasma, a neutron multiplier medium and a tritium breeding medium, cooled by a cooling fluid circulation. This device is characterized by the fact that the association of the multiplier media and the tritium breeding media is realized by pellet alternated piling up of each of those both media, help in close contact on all their lateral surfaces [fr

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)

Tritium pellet injector for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

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) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

Detection of atmospheric tritium by scintillation. Variations in its concentration in France

International Nuclear Information System (INIS)

Bibron, R.

1964-06-01

The tritium released into the atmosphere as a result of the explosion of thermonuclear devices is a useful radioactive tracer for the study of certain geophysical problems. The low concentrations found however call for the use of extremely sensitive detectors. Two detection methods using liquid scintillators are described. In the first method, the sample is introduced into the scintillator in liquid form, after prior concentration of the tritium by electrolysis. In the second method the tritium is incorporated into the scintillator solvent molecule by chemical synthesis. In the last part of the report are examined the variations in the tritium concentration in rain-water and of the free hydrogen in the air in France. A discussion is then made of the seasonal variations in the case of rain-water and these are compared to the variations in the strontium-90 concentrations. (author) [fr

Treatment of tritiated exhaust gases at the Tritium Laboratory Karlsruhe

Energy Technology Data Exchange (ETDEWEB)

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

1995-02-01

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

Application of proton-conducting ceramics and polymer permeable membranes for gaseous tritium recovery

International Nuclear Information System (INIS)

Asakura, Yamato; Sugiyama, Takahiko; Kawano, Takao; Uda, Tatsuhiko; Tanaka, Masahiro; Tsuji, Naruhito; Katahira, Koji; Iwahara, Hiroyasu

2004-01-01

In order to carry out deuterium plasma experiments on the Large Helical Device (LHD), the National Institute for Fusion Science (NIFS) is planning to install a system for the recovery of tritium from exhaust gas and effluent liquid. As well as adopting proven conventional tritium recovery systems, NIFS is planning to apply the latest technologies such as proton-conducting ceramics and membrane-type dehumidifiers in an overall strategy to ensure minimal risk in the tritium recovery process. Application of these new technologies to the tritium recovery system for the LHD deuterium plasma experiment is evaluated quantitatively using recent experimental data. (author)

A study on conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production reactor based on spherical torus (ST), which is an intermediate application of fusion energy, is presented. Different from traditional Tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST are used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can do within vacuum vessel in order to produce certain amount of excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR is presented. Based on systematical analysis, design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (authors)

A system dynamics model for stock and flow of tritium in fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kwon, Saerom [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Sakamoto, Yoshiteru; Yamanishi, Toshihiko; Tobita, Kenji [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori-ken 039-3212 (Japan)

2015-10-15

Highlights: • System dynamics model of tritium fuel cycle was developed for analyzing stock and flow of tritium in fusion power plants. • Sensitivity of tritium build-up to breeding ratio parameters has been assessed to two plant concepts having 3 GW and 1.5 GW fusion power. • D-D start-up absolutely without initial loading of tritium is possible for both of the 3 GW and 1.5 GW fusion power plant concepts. • Excess stock of tritium is generated by the steady state operation with the value of tritium breeding ratio over unity. - Abstract: In order to analyze self-efficiency of tritium fuel cycle (TFC) and share the systems thinking of TFC among researchers and engineers in the vast area of fusion reactor technology, we develop a system dynamics (SD) TFC model using a commercial software STELLA. The SD-TFC model is illustrated as a pipe diagram which consists of tritium stocks, such as plasma, fuel clean up, isotope separation, fueling with storage and blanket, and pipes connecting among them. By using this model, we survey a possibility of D-D start-up without initial loading of tritium on two kinds of fusion plant having different plasma parameters. The D-D start-up scenario can reduce the necessity of initial loading of tritium through the production in plasma by D-D reaction and in breeding blanket by D-D neutron. The model is also used for considering operation scenario to avoid excess stock of tritium which must be produced at tritium breeding ratio over unity.

Overview of tritium systems for the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Bartlit, J.R.; Gruetzmacher, K.M.; Fleming, R.B.

1987-01-01

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

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

Halon-1301, a new Groundwater Age Tracer

Science.gov (United States)

Beyer, Monique; van der Raaij, Rob; Morgenstern, Uwe; Jackson, Bethanna

2015-04-01

Groundwater dating is an important tool to assess groundwater resources in regards to direction and time scale of groundwater flow and recharge and to assess contamination risks and manage remediation. To infer groundwater age information, a combination of different environmental tracers, such as tritium and SF6, are commonly used. However ambiguous age interpretations are often faced, due to a limited set of available tracers and limitations of each tracer method when applied alone. There is a need for additional, complementary groundwater age tracers. We recently discovered that Halon-1301, a water soluble and entirely anthropogenic gaseous substance, may be a promising candidate [Beyer et al, 2014]. Halon-1301 can be determined along with SF6, SF5CF3 and CFC-12 in groundwater using a gas chromatography setup with attached electron capture detector developed by Busenberg and Plummer [2008]. Halon-1301 has not been assessed in groundwater. This study assesses the behaviour of Halon-1301 in water and its suitability as a groundwater age tracer. We determined Halon-1301 in 17 groundwater and various modern (river) waters sites located in 3 different groundwater systems in the Wellington Region, New Zealand. These waters have been previously dated with tritium, CFC-12, CFC-11 and SF6 with mean residence times ranging from 0.5 to over 100 years. The waters range from oxic to anoxic and some show evidence of CFC contamination or degradation. This allows us to assess the different properties affecting the suitability of Halon-1301 as groundwater age tracer, such as its conservativeness in water and local contamination potential. The samples are analysed for Halon-1301 and SF6simultaneously, which allows identification of issues commonly faced when using gaseous tracers such as contamination with modern air during sampling. Overall we found in the assessed groundwater samples Halon-1301 is a feasible new groundwater tracer. No sample indicated significantly elevated

Study on hydrogen transfer in coal liquefaction by tritium and carbon-14 tracers

International Nuclear Information System (INIS)

Nitoh, Osamu; Kabe, Toshiaki; Kabe, Yaeko.

1985-01-01

For the analysis of mechanism of hydrogenation and cracking of coal, the liquefaction of Taiheiyo coal using tritium labeled gaseous hydrogen and tritium labeled tetralin with small amounts of carbon-14 labeled naphthalene has been studied. Taiheiyo coal(25g) was thermally decomposed in tetralin or naphthalene solvent(75g) at 400--440 0 C under the initial hydrogen pressure of 5.9MPa for 30min with Ni-Mo-Al 2 O 3 catalyst(0--5g). The reaction mixture in an autoclave was separated by filtration, distillation and solvent extraction. Produced gas, oils and the solvent were analyzed by gas chromatography. The tritium and carbon-14 contents of separated reaction products were measured with a liquid scintilation counter to study the hydrogen transfer mechanism. The distribution of reaction products and the amount of hydrogen transfer from gas or solvent to the products were also determined. In hydrogen donor solvent such as tetralin, the coal liquefaction yield was independent from the catalyst, but the catalyst was effective in hydrocracking of preasphaltene and asphaltene. In naphthalene solvent, the coal liquefaction reaction hardly occured in the absence of the catalyst, because hydrogen transfer from both the solvent and gaseous hydrogen was scarce. Tritium distribution in the reaction products showed that complicated hydrogen exchange reactions between gaseous hydrogen, coal liquids and solvent came out by the presence of coal liquids and catalyst. The very small amounts of carbon-14 transferred to the liquefaction products showed that carbon exchange or transfer between solvent and coal did not take place. (author)

Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Bell, M.G.; Beer, M.

1997-02-01

Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l i ). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q a ∼ 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l i plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q 0 > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions

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.)

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

Tritium resources available for fusion reactors

Science.gov (United States)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

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

Theoretical studies on the stopping power of deuterium-tritium mixed with uranium plasmas for α particles

International Nuclear Information System (INIS)

Wang, Zhigang; Fu, Zhen-Guo; Zhang, Ping

2014-01-01

The stopping power of a compressed and highly ionized deuterium-tritium (DT) and uranium (U) plasma for α particles at very high temperatures (T = 5 keV) is examined theoretically with the dimensional continuation method. We show that with increasing density of U, both the magnitude and width of the resonance peak in the stopping power (as a function of the α particle energy), increases because of the ions, while the penetration distance of the α particles decreases. A simple relation of decreasing penetration distance as a function of plasma density is observed, which may be useful for inertial confinement fusion experiments. Moreover, by comparing the results with the case of a DT plasma mixed with beryllium, we find that the effect of a higher Z plasma is stronger, with regard to energy loss as well as the penetration distance of α particles, than that of a lower Z plasma

The Safety and Tritium Applied Research (STAR) Facility: Status-2004

International Nuclear Information System (INIS)

Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Sharpe, J.P.; Schuetz, S.T.; Petti, D.A.

2005-01-01

The Safety and Tritium Applied Research (STAR) Facility, a US DOE National User Facility at the Idaho National Engineering and Environmental Laboratory (INEEL), comprises capabilities and infrastructure to support both tritium and non-tritium research activities important to the development of safe and environmentally friendly fusion energy. Research thrusts include (1) interactions of tritium and deuterium with plasma-facing-component (PFC) materials, (2) fusion safety issues [PFC material chemical reactivity and dust/debris generation, activation product mobilization, tritium behavior in fusion systems], and (3) molten salts and fusion liquids for tritium breeder and coolant applications. This paper updates the status of STAR and the capabilities for ongoing research activities, with an emphasis on the development, testing and integration of the infrastructure to support tritium research activities. Key elements of this infrastructure include a tritium storage and assay system, a tritium cleanup system to process glovebox and experiment tritiated effluent gases, and facility tritium monitoring systems

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

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

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

Data for Erosion and Tritium Retention in Beryllium Plasma-Facing Materials. Summary Report of the First Research Coordination Meeting

International Nuclear Information System (INIS)

Braams, B.J.

2013-04-01

Nine experts in the field of plasma-wall interaction on beryllium surfaces together with IAEA staff met at IAEA Headquarters 26-28 September 2012 for the First Research Coordination Meeting of an IAEA Coordinated Research Project on data for erosion and tritium retention in beryllium plasma-facing materials. They described their on-going research, reviewed the main data needs and made plans for coordinated research during the remaining years of the project. The proceedings of the meeting are summarized in this report. (author)

Tritium evolution from various morphologies of palladium

International Nuclear Information System (INIS)

Tuggle, D.G.; Claytor, T.N.; Taylor, S.F.

1994-01-01

The authors have been able to extend the tritium production techniques to various novel morphologies of palladium. These include small solid wires of various diameters and a type of pressed powder wire and a plasma cell. In most successful experiments, the amount of palladium required, for an equivalent tritium output, has been reduced by a factor of 100 over the older powder methods. In addition, they have observed rates of tritium production (>5 nCi/h) that far exceed most of the previous results. Unfortunately, the methods that they currently use to obtain the tritium are poorly understood and consequently there are numerous variables that need to be investigated before the new methods are as reliable and repeatable as the previous techniques. For instance, it seems that surface and/or bulk impurities play a major role in the successful generation of any tritium. In those samples with total impurity concentrations of >400 ppM essentially no tritium has been generated by the gas loading and electrical simulation methods

New technique of in-situ soil-moisture sampling for environmental isotope analysis applied at Pilat sand dune near Bordeaux. HETP modelling of bomb tritium propagation in the unsaturated and saturated zones

International Nuclear Information System (INIS)

Thoma, G.; Esser, N.; Sonntag, C.; Weiss, W.; Rudolph, J.; Leveque, P.

1979-01-01

A new soil-air suction method with soil-water vapour adsorption by a 4-A molecular sieve provides soil-moisture samples from various depths for environmental isotope analysis and yields soil temperature profiles. A field tritium tracer experiment shows that this in-situ sampling method has an isotope profile resolution of about 5-10cm only. Application of this method in the Pilat sand dune (Bordeaux/France) yielded deuterium and tritium profiles down to 25m depth. Bomb tritium measurements of monthly lysimeter percolate samples available since 1961 show that the tritium response has a mean delay of five months in the case of a sand lysimeter and of 2.5 years for a loess loam lysimeter. A simple HETP model simulates the layered downward movement of soil water and the longitudinal dispersion in the lysimeters. Field capacity and evapotranspiration taken as open parameters yield tritium concentration values of the lysimeters' percolate which agree well with the experimental results. Based on local meteorological data the HETP model applied to tritium tracer experiments in the unsaturated zone yields in addition an individual prediction of the momentary tracer position and of the soil-moisture distribution. This prediction can be checked experimentally at selected intervals by coring. (author)

Tritium Records to Trace Stratospheric Moisture Inputs in Antarctica

Science.gov (United States)

Fourré, E.; Landais, A.; Cauquoin, A.; Jean-Baptiste, P.; Lipenkov, V.; Petit, J.-R.

2018-03-01

Better assessing the dynamic of stratosphere-troposphere exchange is a key point to improve our understanding of the climate dynamic in the East Antarctica Plateau, a region where stratospheric inputs are expected to be important. Although tritium (3H or T), a nuclide naturally produced mainly in the stratosphere and rapidly entering the water cycle as HTO, seems a first-rate tracer to study these processes, tritium data are very sparse in this region. We present the first high-resolution measurements of tritium concentration over the last 50 years in three snow pits drilled at the Vostok station. Natural variability of the tritium records reveals two prominent frequencies, one at about 10 years (to be related to the solar Schwabe cycles) and the other one at a shorter periodicity: despite dating uncertainty at this short scale, a good correlation is observed between 3H and Na+ and an anticorrelation between 3H and δ18O measured on an individual pit. The outputs from the LMDZ Atmospheric General Circulation Model including stable water isotopes and tritium show the same 3H-δ18O anticorrelation and allow further investigation on the associated mechanism. At the interannual scale, the modeled 3H variability matches well with the Southern Annular Mode index. At the seasonal scale, we show that modeled stratospheric tritium inputs in the troposphere are favored in winter cold and dry conditions.

Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

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 was the impurity control system for the reactor in the Tokamak Power Systems Study (TPSS). The use of a self-pumped limiter had a major impact on the design of the tritium systems of the TPSS fusion reactor. The self-pumped limiter functions by depositing the helium ash under a layer of metal (vanadium). The majority of the hydrogen species are recycled at the plasma edge; a small fraction permeates to the blanket/coolant which was lithium in TPSS. Use of the self-pumped limiter results in the elimination of the plasma processing system. Thus, the blanket tritium processing system becomes the major tritium system. The main advantages achieved for the tritium systems with a self-pumped limiter are a reduction in the capital cost of tritium processing equipment as well as a reduction in building space, a reduced tritium inventory for processing and for reserve storage, an increase in the inherent safety of the fusion plant and an improvement in economics for a fusion world economy

Regeneration and tritium recovery from the large JET neutral injection cryopump system after the FTE

International Nuclear Information System (INIS)

Obert, W.; Bell, A.; Davies, J.; Mayaux, C.; Perinic, G.; Saibene, G.; Sartori, R.; Thompson, E.; Anderson, J.; Jenkins, E.; Walthers, C.

1992-01-01

Neutral Beam Injection (NBI) was used to introduce tritium into the plasma for the First Tritium Experiment In addition to the decisive advantage of depositing the tritium into the centre of the plasma, the use of NBI also minimized the total quantity of tritium introduced into the Torus and the contamination of the vacuum vessel. However, because of the relatively low gas efficiency of the positive ion injection system approximately 95% of the total quantity of tritium introduced was pumped by the large condensation cryopumps which form an integral part of the injector. Several hardware and associated software changes were implemented in order to making provision for possible fault scenarios during operation with tritium and to ensure complete regeneration of the tritium from the cryopumps. The tritium released after all subsequent regeneration's has been monitored carefully in order to determine the amount of tritium retained by the black anodized liquid nitrogen panel surfaces of the cryopump and to compare it with experiments at TSTA on JET samples before the FTE

Tritium handling and vacuum considerations for the STARFIRE commercial tokamak reactor

International Nuclear Information System (INIS)

Finn, P.A.; Clemmer, R.G.; Maroni, V.A.; Dillow, C.

1979-01-01

Tritium processing and vacuum pumping requirements were analyzed for the STARFIRE commercial fusion reactor design. It was found that vacuum pumps having a helium capture probability of 0.5 (total helium pump speed 1.2 x 10 4 m 3 /s) in combination with the proposed STARFIRE limiter-vacuum concept is sufficient to achieve plasma impurity control and, simultaneously, high fractional burnup (11%). The high fractional burnup and minimum fuel recycle time result in a very low fuel cycle tritium inventory, approx. 1300 g. A Lean-T burn method that can further reduce the fuel cycle inventory by 30 to 50% is discussed. D 2 O is proposed as a first wall coolant from considerations of plasma contamination (due to hydrogen isotope permeation through coolant tubes) and enrichment of recycled tritium from the coolant circuit. Tritium recovery from solid breeders, under realistic structural and breeder materials constraints, appears to represent a formidable task. The tritium inventory in the solid breeder is estimated to be as high as 10 kg, which would make the blanket the largest single hold-up point for tritium in the plant

Radioimmunoassay of human. beta. -lipotropin in unextracted plasma. [/sup 125/I tracer technique

Energy Technology Data Exchange (ETDEWEB)

Wiedemann, E. (Univ. of California, Berkeley); Saito, T.; Linfoot, J.A.; Li, C.H.

1977-11-01

A sensitive and specific radioimmunoassay for human ..beta..-lipotropin (..beta../sub h/-LPH) in unextracted plasma was developed using pure ..beta../sub h/-LPH as tracer and standard and an antiserum not cross-reacting with human ..beta..-MSH and hACTH. In healthy volunteers plasma ..beta../sub h/-LPH ranged from <20 to 150 pg/ml at 8:00 a.m. and rose after metyrapone administration. ..beta../sub h/-LPH was very low in panhypopituitarism, normal in most patients with untreated Cushing's disease, elevated in acromegaly and extremely high in Nelson's syndrome.

Tritium tracing in hydrogeochemical studies using model-lysimeters

International Nuclear Information System (INIS)

Matthess, G.; Pekdeger, A.; Schulz, H.D.; Rast, H.; Rauert, W.

1978-01-01

Tritium was used as a reference tracer for hydrogeochemical studies in the unsaturated zone. The investigators used different lysimeter types (25, 50, 100 cm), with and without suction plates filled with undisturbed soil monoliths of sandy podsol and loamy lessive. The tritium loss was greater than the evaporation amount determined. Water logging takes place in lysimeter bottoms increasing the evaporation in up to 100 cm lysimeters filled with loamy lessive and 25 cm with sandy podsol. After a 20 mm rain event seepage characteristics indicate 'by-passing' water besides intergranular seepage. Dispersion coefficients (8.5 x 10 -5 cm 2 s -1 ) are higher than molecular diffusion coefficient. Dispersion takes place mainly in top soil with wide ranging pore size distribution. Distribution coefficients of tritium in soil are rather low. Concentrations of anions and dissolved organic substance are different depending on residence time of seepage water in soil. Even a short residence time of seepage water in unsaturated soil is enough for cation exchange reactions to take place. (orig.) [de

Competition of bulk trapping and surface erosion in the kinetics of tritium inventory and permeation in plasma protection metals

International Nuclear Information System (INIS)

Federici, G.; Holland, D.F.; Esser, B.

1996-01-01

A simplified transient model is presented to describe the migration of implanted tritium in the presence of trap sites across the bulk of metallic substrates whose thickness is decreasing with time due to erosion. The subject is relevant for quantifying the tritium inventory in - and permeation through -plasma facing armours in the next generation of D-T fuelled tokamak devices (i.e., the International Thermonuclear Experimental Reactor). This paper describes the equations of the physical model and the main assumptions used to simplify the complex analysis, and surveys the influence of several parameters such as the implantation flux, the erosion rate, the armour temperature, the armour thickness, the density and trapping energy of neutron-induced traps, etc., which are all expected to play a key role in the phenomena investigated. The examples presented to show the applicability of the model include the results of a study performed for beryllium armours exposed to heat and particle loads similar to those expected on the ITER divertor plasma facing components and comparison is made with cases where erosion does not play any role. (orig.)

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.

An injected gamma-tracer method for soil-moisture movement investigations in arid zones

International Nuclear Information System (INIS)

Nair, A.R.; Navada, S.V.; Rao, S.M.

1980-01-01

A method for the in-situ determination of soil-moisture transport rates using K 3 60 Co(CN) 6 is discussed. The tracer compares well with tritiated water in laboratory investigations and the results obtained in limited field studies are very encouraging. The method promises to be of specific interest in arid-zone investigations where the soil-moisture fluxes in liquid and vapour phases could cause complications for tritium tracer data interpretation. (author)

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

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

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.)

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.)

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.

The impact of tritium solubility and diffusivity on inventory and permeation in liquid breeder blankets

International Nuclear Information System (INIS)

Caorlin, M.; Gervasini, G.; Reiter, F.

1988-01-01

The authors reviewed hydrogen solubility and diffusivity data for liquid lithium-based compounds which are potential breeding blanket materials in NET-type fusion devices. These data have been used to assess tritium permeation and inventory in separately cooled NET blankets and in self cooled blankets with a vanadium first wall. The results for the separately cooled NET-liquid breeder show that tritium permeation is negligible for lithium, a serious problem for Pb-17Li and a critical one for Flibe. The total tritium inventory is lowest in lithium, high in Pb-17Li and very high in Flibe. The high tritium partial pressure for Flibe or Pb-17Li can be reduced in a self cooled blanket with a vanadium first wall. Permeation into the plasma reduces the blanket tritium inventory and permeation. Tritium recovery can be combined with the plasma exhaust

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)

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.

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties

International Nuclear Information System (INIS)

Nakamura, Hirofumi; Nishi, Masataka

2003-11-01

Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authours' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evaluation was also carried out for comparison (previous data). The permeation analysis was carried out individually by classifying into the armor region (Carbon Fiber Composites and tungsten) and the slit region without armor (3% of armor surface area) assuming the incident flux and temperature for each region. As the results of the permeation analysis, estimated permeation amount with the authors' data was one order less than that with the previous data at the end of lifetime of the divertor due to authors' small diffusion coefficient of tritium in tungsten. It also indicated the possibility that permeation through the slit region of the armor tiles could dominate total permeation through the vertical target, since tritium permeation amount through tungsten armor with the authors' data was estimated to be reduced drastically smaller than that with the previous evaluation data. The result of a little tritium permeation amount through the vertical target with the authors' data ensured the conservatism of the current evaluation of tritium concentration in the primary cooling water in ITER divertor, as it indicated the possibility of direct drainage of the divertor primary cooling water. (author)

Measurements of tritium recycling and isotope exchange in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Kamperschroer, J.; Mueller, D.; Nagy, A.; Stotler, D.P.

1996-05-01

Tritium Balmer-alpha (T α ) emission, along with H α and D α is observed in the current D-T experimental campaign in TFTR. The data are a measure of the fueling of the plasma by tritium accumulated in the TFTR limiter and the spectral profile maps neutral hydrogenic velocities. T α is relatively slow to appear in tritium neutral beam heated discharges, (T α /(H α + D α + T α ) = 11% after 8 tritium-only neutral beam discharges). In contrast, the T α fraction in a sequence of six discharges fueled with tritium puff,s increased to 44%. Larger transient increases (up to 75% T α ) were observed during subsequent tritium gas puffs. Analysis of the Doppler broadened spectral profiles revealed overall agreement with the dissociation, charge exchange, sputtering and reflection velocities predicted by the neutral Monte-Carlo code DEGAS with some deficiency in the treatment of dissociation products in the 10--100 eV range

Detection of atmospheric tritium by scintillation. Variations in its concentration in France; Detection du tritium atmospherique par scintillation. Evolution de sa concentration en France

Energy Technology Data Exchange (ETDEWEB)

Bibron, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-06-01

The tritium released into the atmosphere as a result of the explosion of thermonuclear devices is a useful radioactive tracer for the study of certain geophysical problems. The low concentrations found however call for the use of extremely sensitive detectors. Two detection methods using liquid scintillators are described. In the first method, the sample is introduced into the scintillator in liquid form, after prior concentration of the tritium by electrolysis. In the second method the tritium is incorporated into the scintillator solvent molecule by chemical synthesis. In the last part of the report are examined the variations in the tritium concentration in rain-water and of the free hydrogen in the air in France. A discussion is then made of the seasonal variations in the case of rain-water and these are compared to the variations in the strontium-90 concentrations. (author) [French] Le tritium introduit dans l'atmosphere par les explosions d'armes thermonucleaires est un traceur radioactif Interessant pour l'etude de certains problemes de geophysique. Les faibles concentrations rencontrees obligent toutefois a utiliser des detecteurs extremement sensibles. On decrit deux methodes de detection utilisant des scintillateurs liquides. Dans la premiere methode, l'echantillon est introduit dans le scintillateur, sous forme aqueuse, apres une concentration prealable du tritium par electrolyse. Dans la seconde methode, le tritium est incorpore a la molecule du solvant du scintillateur par synthese chimique. Dans la derniere partie du rapport, on examine l'evolution de la concentration du tritium dans les eaux de precipitation et l'hydrogene libre de l'air en France. On discute ensuite les variations saisonnieres dans le cas des eaux de precipitation et on les compare aux variations du strontium 90. (auteur)

Detection of atmospheric tritium by scintillation. Variations in its concentration in France; Detection du tritium atmospherique par scintillation. Evolution de sa concentration en France

Energy Technology Data Exchange (ETDEWEB)

Bibron, R. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-06-01

The tritium released into the atmosphere as a result of the explosion of thermonuclear devices is a useful radioactive tracer for the study of certain geophysical problems. The low concentrations found however call for the use of extremely sensitive detectors. Two detection methods using liquid scintillators are described. In the first method, the sample is introduced into the scintillator in liquid form, after prior concentration of the tritium by electrolysis. In the second method the tritium is incorporated into the scintillator solvent molecule by chemical synthesis. In the last part of the report are examined the variations in the tritium concentration in rain-water and of the free hydrogen in the air in France. A discussion is then made of the seasonal variations in the case of rain-water and these are compared to the variations in the strontium-90 concentrations. (author) [French] Le tritium introduit dans l'atmosphere par les explosions d'armes thermonucleaires est un traceur radioactif Interessant pour l'etude de certains problemes de geophysique. Les faibles concentrations rencontrees obligent toutefois a utiliser des detecteurs extremement sensibles. On decrit deux methodes de detection utilisant des scintillateurs liquides. Dans la premiere methode, l'echantillon est introduit dans le scintillateur, sous forme aqueuse, apres une concentration prealable du tritium par electrolyse. Dans la seconde methode, le tritium est incorpore a la molecule du solvant du scintillateur par synthese chimique. Dans la derniere partie du rapport, on examine l'evolution de la concentration du tritium dans les eaux de precipitation et l'hydrogene libre de l'air en France. On discute ensuite les variations saisonnieres dans le cas des eaux de precipitation et on les compare aux variations du strontium 90. (auteur)

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

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.

Earth mechanisms (fluid and solid), life mechanisms and stable isotope tracers. Isotopes and biology, a great project

International Nuclear Information System (INIS)

Fromageot, P.

1997-01-01

Historical and recent review of the development and use of radioactive isotopes for biological studies in France: study of the intermediate metabolism with 14 C tracers in organic molecules; study and biosynthesis of macromolecules (DNA, RNA and polynucleotides) through the use of marked nucleotides; tracer proteins for use in NMR and protein engineering, use of tritium for the study of hormonal regulation

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.)

Current CTR-related tritium handling studies at ORNL

International Nuclear Information System (INIS)

Watson, J.S.; Bell, J.T.; Clinton, S.D.; Fisher, P.W.; Redman, J.D.; Smith, F.J.; Talbot, J.B.; Tung, C.P.

1976-01-01

The Oak Ridge National Laboratory has a comprehensive program concerned with plasma fuel recycle, tritium recovery from blankets, and tritium containment in fusion reactors. Two studies of most current interest are investigations of cryosorption pumping of hydrogen isotopes and measurements of tritium permeation rates through steam generator materials. Cryosorption pumping speeds have been measured for hydrogen, deuterium, and helium at pressures from 10 -8 torr to 3 x 10 -3 torr. Permeation rates through Incoloy 800 have been shown to be drastically reduced when the low pressure side of permeation tubes are exposed to steam. These results will be important considerations in the design of fusion reactor steam generators

Tracer water transport and subgrid precipitation variation within atmospheric general circulation models

Science.gov (United States)

Koster, Randal D.; Eagleson, Peter S.; Broecker, Wallace S.

1988-03-01

A capability is developed for monitoring tracer water movement in the three-dimensional Goddard Institute for Space Science Atmospheric General Circulation Model (GCM). A typical experiment with the tracer water model follows water evaporating from selected grid squares and determines where this water first returns to the Earth's surface as precipitation or condensate, thereby providing information on the lateral scales of hydrological transport in the GCM. Through a comparison of model results with observations in nature, inferences can be drawn concerning real world water transport. Tests of the tracer water model include a comparison of simulated and observed vertically-integrated vapor flux fields and simulations of atomic tritium transport from the stratosphere to the oceans. The inter-annual variability of the tracer water model results is also examined.

Tracer water transport and subgrid precipitation variation within atmospheric general circulation models

Science.gov (United States)

Koster, Randal D.; Eagleson, Peter S.; Broecker, Wallace S.

1988-01-01

A capability is developed for monitoring tracer water movement in the three-dimensional Goddard Institute for Space Science Atmospheric General Circulation Model (GCM). A typical experiment with the tracer water model follows water evaporating from selected grid squares and determines where this water first returns to the Earth's surface as precipitation or condensate, thereby providing information on the lateral scales of hydrological transport in the GCM. Through a comparison of model results with observations in nature, inferences can be drawn concerning real world water transport. Tests of the tracer water model include a comparison of simulated and observed vertically-integrated vapor flux fields and simulations of atomic tritium transport from the stratosphere to the oceans. The inter-annual variability of the tracer water model results is also examined.

Indian contribution to applications of artificially injected tritium in hydrological investigations

International Nuclear Information System (INIS)

Datta, P.S.

1982-01-01

The paper gives a brief description on significance of groundwater hydrology and sets it in the context of radioisotopic investigations. The topics described pertain to potential applications of artificially injected tritium in local or regional scale to determine water movement in the unsaturated zone, rate of infiltration, groundwater recharge, direction and velocity of groundwater, interconnection of groundwater bodies, dispersion of pollutants, etc. The Indian contribution on these topics is incorporated giving informations on techniques adopted and the major findings and conclusions of the experiments conducted. Merits and demerits of each technique have also been described. Some aspects deserving urgent consideration are outlined to gain maximum benefits from the applications of artificially injected tritium tracer techniques in hydrology. (author)

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

Wet scrubber technology for tritium confinement at ITER

Energy Technology Data Exchange (ETDEWEB)

Perevezentsev, A.N., E-mail: alexander.perevezentsev@iter.org [ITER Organization, CS 90 046, 13067 St Paul lez Durance Cedex (France); Andreev, B.M.; Rozenkevich, M.B.; Pak, Yu.S.; Ovcharov, A.V.; Marunich, S.A. [Mendeleev University of Chemical Technology, 125047 Miusskaya Sq. 9, Moscow (Russian Federation)

2010-12-15

Operation of the ITER machine with tritium plasma requires tritium confinement systems to protect workers and the environment. Tritium confinement at ITER is based on multistage approach. The final stage provides tritium confinement in building sectors and consists of building's walls as physical barriers and control of sub-atmospheric pressure in those volumes as a dynamic barrier. The dynamic part of the confinement function shall be provided by safety important components that are available all the time when required. Detritiation of air prior to its release to the environment is based on catalytic conversion of tritium containing gaseous species to water vapour followed by their isotopic exchange with liquid water in scrubber column of packed bed type. Wet scrubber technology has been selected because of its advantages over conventional air detritiation technique based on gas drying by water adsorption. The most important design target of system availability was very difficult to meet with conventional water adsorption driers. This paper presents results of experimental trial for validation of wet scrubber technology application in the ITER tritium confinement system and process evaluation using developed simulation computer code.

Tritium systems concepts for the next European torus (NET)

International Nuclear Information System (INIS)

Sood, S.K.; Bagli, K.S.; Busigin, A.; Kveton, O.K.; Dombra, A.H.; Miller, A.I.

1986-09-01

The study deals with the design of the various tritium processing facilities that will be required for the Next European Torus (NET) design. The reference data for the design of the NET Tritium Systems was provided by the NET team. Significant achievements of this study were: (a) Identification of new ways of handling some problems for example: 1) Recovery of tritium from the helium purge of the lithium-ceramic blanket using a novel Adsoprtion and Catalytic Exchange Process, 2) A new way of combining fuel component separation and coolant water detritiation using cryogenic distillation, 3) The use of parasitic refrigeration for the cryogenic isotope separation, 4) Tritium extraction from effluent gas streams at their respective sources, 5) Attempt to eliminate the need for Air Cleanup Systems. (b) Identification of uncertainties, for example: composition of plasma exhaust, required helium purge rate of Li-Pb for tritium recovery, uncertainty in requirements for decontaminating blanket sectors, etc. (c) Review of ways to limit tritium permeation into steam by swamping with hydrogen and to provide quantitative estimates for this permeation

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)

Fusion Energy-Production from a Deuterium-Tritium Plasma in the Jet Tokamak

NARCIS (Netherlands)

Rebut, P. H.; Gibson, A.; Huguet, M.; Adams, J. M.; Alper, B.; Altmann, H.; Andersen, A.; Andrew, P.; Angelone, M.; Aliarshad, S.; Baigger, P.; Bailey, W.; Balet, B.; Barabaschi, P.; Barker, P.; Barnsley, R.; Baronian, M.; Bartlett, D. V.; Baylor, L.; Bell, A. C.; Benali, G.; Bertoldi, P.; Bertolini, E.; Bhatnagar, V.; Bickley, A. J.; Binder, D.; Bindslev, H.; Bonicelli, T.; Booth, S. J.; Bosia, G.; Botman, M.; Boucher, D.; Boucquey, P.; Breger, P.; Brelen, H.; Brinkschulte, H.; Brooks, D.; Brown, A.; Brown, T.; Brusati, M.; Bryan, S.; Brzozowski, J.; Buchse, R.; Budd, T.; Bures, M.; Businaro, T.; Butcher, P.; Buttgereit, H.; Caldwellnichols, C.; Campbell, D. J.; Card, P.; Celentano, G.; Challis, C. D.; Chankin, A. V.; Cherubini, A.; Chiron, D.; Christiansen, J.; Chuilon, P.; Claesen, R.; Clement, S.; Clipsham, E.; Coad, J. P.; Coffey, I. H.; Colton, A.; Comiskey, M.; Conroy, S.; Cooke, M.; Cooper, D.; Cooper, S.; Cordey, J. G.; Core, W.; Corrigan, G.; Corti, S.; Costley, A. E.; Cottrell, G.; Cox, M.; Cripwell, P.; Dacosta, O.; Davies, J.; Davies, N.; de Blank, H.; De Esch, H.; Dekock, L.; Deksnis, E.; Delvart, F.; Dennehinnov, G. B.; Deschamps, G.; Dickson, W. J.; Dietz, K. J.; Dmitrenko, S. L.; Dmitrieva, M.; Dobbing, J.; Doglio, A.; Dolgetta, N.; Dorling, S. E.; Doyle, P. G.; Duchs, D. F.; Duquenoy, H.; Edwards, A.; Ehrenberg, J.; Ekedahl, A.; Elevant, T.; Erents, S.K.; Eriksson, L. G.; Fajemirokun, H.; Falter, H.; Freiling, J.; Freville, F.; Froger, C.; Froissard, P.; Fullard, K.; Gadeberg, M.; Galetsas, A.; Gallagher, T.; Gambier, D.; Garribba, M.; Gaze, P.; Giannella, R.; Gill, R. D.; Girard, A.; Gondhalekar, A.; Goodall, D.; Gormezano, C.; Gottardi, N. A.; Gowers, C.; Green, B. J.; Grievson, B.; Haange, R.; Haigh, A.; Hancock, C. J.; Harbour, P. J.; Hartrampf, T.; Hawkes, N. C.; Haynes, P.; Hemmerich, J. L.; Hender, T.; Hoekzema, J.; Holland, D.; Hone, M.; Horton, L.; How, J.; Huart, M.; Hughes, I.; Hughes, T. P.; Hugon, M.; Huo, Y.; Ida, K.; Ingram, B.; Irving, M.; Jacquinot, J.; Jaeckel, H.; Jaeger, J. F.; Janeschitz, G.; Jankovicz, Z.; Jarvis, O. N.; Jensen, F.; Jones, E. M.; Jones, H. D.; Jones, Lpdf; Jones, S.; Jones, T. T. C.; Junger, J. F.; Junique, F.; Kaye, A.; Keen, B. E.; Keilhacker, M.; Kelly, G. J.; Kerner, W.; Khudoleev, A.; Konig, R.; Konstantellos, A.; Kovanen, M.; Kramer, G.; Kupschus, P.; Lasser, R.; Last, J. R.; Laundy, B.; Laurotaroni, L.; Laveyry, M.; Lawson, K.; Lennholm, M.; Lingertat, J.; Litunovski, R. N.; Loarte, A.; Lobel, R.; Lomas, P.; Loughlin, M.; Lowry, C.; Lupo, J.; Maas, A. C.; Machuzak, J.; Macklin, B.; Maddison, G.; Maggi, C. F.; Magyar, G.; Mandl, W.; Marchese, V.; Marcon, G.; Marcus, F.; Mart, J.; Martin, D.; Martin, E.; Martinsolis, R.; Massmann, P.; Matthews, G.; McBryan, H.; McCracken, G.; McKivitt, J.; Meriguet, P.; Miele, P.; Miller, A.; Mills, J.; Mills, S. F.; Millward, P.; Milverton, P.; Minardi, E.; Mohanti, R.; Mondino, P. L.; Montgomery, D.; Montvai, A.; Morgan, P.; Morsi, H.; Muir, D.; Murphy, G.; Myrnas, R.; Nave, F.; Newbert, G.; Newman, M.; Nielsen, P.; Noll, P.; Obert, W.; Obrien, D.; Orchard, J.; Orourke, J.; Ostrom, R.; Ottaviani, M.; Pain, M.; Paoletti, F.; Papastergiou, S.; Parsons, W.; Pasini, D.; Patel, D.; Peacock, A.; Peacock, N.; Pearce, R. J. M.; Pearson, D.; Peng, J. F.; Desilva, R. P.; Perinic, G.; Perry, C.; Petrov, M.; Pick, M. A.; Plancoulaine, J.; Poffe, J. P.; Pohlchen, R.; Porcelli, F.; Porte, L.; Prentice, R.; Puppin, S.; Putvinskii, S.; Radford, G.; Raimondi, T.; Deandrade, M. C. R.; Reichle, R.; Reid, J.; Richards, S.; Righi, E.; Rimini, F.; Robinson, D.; Rolfe, A.; Ross, R. T.; Rossi, L.; Russ, R.; Rutter, P.; Sack, H. C.; Sadler, G.; Saibene, G.; Salanave, J. L.; Sanazzaro, G.; Santagiustina, A.; Sartori, R.; Sborchia, C.; Schild, P.; Schmid, M.; Schmidt, G.; Schunke, B.; Scott, S. M.; Serio, L.; Sibley, A.; Simonini, R.; Sips, A.C.C.; Smeulders, P.; Smith, R.; Stagg, R.; Stamp, M.; Stangeby, P.; Stankiewicz, R.; Start, D. F.; Steed, C. A.; Stork, D.; Stott, P.E.; Stubberfield, P.; Summers, D.; Summers, H.; Svensson, L.; Tagle, J. A.; Talbot, M.; Tanga, A.; Taroni, A.; Terella, C.; Terrington, A.; Tesini, A.; Thomas, P. R.; Thompson, E.; Thomsen, K.; Tibone, F.; Tiscornia, A.; Trevalion, P.; Tubbing, B.; Vanbelle, P.; Vanderbeken, H.; Vlases, G.; von Hellermann, M.; Wade, T.; Walker, C.; Walton, R.; Ward, D.; Watkins, M. L.; Watkins, N.; Watson, M. J.; Weber, S.; Wesson, J.; Wijnands, T. J.; Wilks, J.; Wilson, D.; Winkel, T.; Wolf, R.; Wong, D.; Woodward, C.; Wu, Y.; Wykes, M.; Young, D.; Young, I. D.; Zannelli, L.; Zolfaghari, A.; Zwingmann, W.

1992-01-01

The paper describes a series of experiments in the Joint European Torus (JET), culminating in the first tokamak discharges in deuterium-tritium fuelled mixtures. The experiments were undertaken within limits imposed by restrictions on vessel activation and tritium usage. The objectives were: (i) to

Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G.

1994-04-01

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinet trademark system, modification of the vacuum system to handle tritium, preparation and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of ∼ 10 megawatts (MW) of fusion power, evaluation of confinement and heating in deuterium-tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α-particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed

Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Jaeger, E.F.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Laughlin, M.J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D.E.; Marchlik, M.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M.P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Milora, S.L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Paul, S.F.; Pavlov, Y.I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Pitcher, S.; Popovichev, S.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.

1994-01-01

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert TM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of ∼10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed

Tritium processing and management during D-T experiments on TFTR

International Nuclear Information System (INIS)

La Marche, P.H.; Anderson, J.L.; Gentile, C.A.; Hawryluk, R.J.; Hosea, J.; Kalish, M.; Kozub, T.; Murray, H.; Nagy, A.; Raftopoulos, S.

1994-11-01

TFTR performance has surpassed many of the previous tokamak records. This has been made possible by the use of tritium as fuel for DT plasma discharges. Stable operations of tritium systems provide for safe, routine DT operation of TFTR. In the preparation for DT operation, in the commissioning of the tritium systems and in the operation of the Nuclear Facility several key lessons have been learned. They include: the facility must take the lead in interpreting the applicable regulations and orders and then seek regulator approval; the use of ultra high vacuum technology in tritium system design and construction simplifies and enhances operations and maintenance; and central facility control under a single supervisory position is crucial to safely orchestrate operational and maintenance activities

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

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.

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

Studies of deep water formation and circulation in the Weddell Sea using natural and anthropogenic tracers

International Nuclear Information System (INIS)

Schlosser, Peter; Bayer, Reinhold

1991-01-01

The application of natural and anthropogenic trace substances in oceanographic studies of the Weddell Sea is reviewed. The potential of some steady-state and transient tracers (tritium, CFC-11 and CFC-12, 18 O, and helium isotopes) for studies of deep water formation and circulation is discussed on the basis of data sets collected mainly on cruises of R/V 'Polastern' to the Weddell Sea during the 1980s. CFC/ tritium ratio dating of young water masses is applied to estimate mean age and transit times of water involved in Weddell Sea Bottom Water formation. The history of the CFC-11/tritium ratio through time is derived for Weddell Sea shelf waters. (author). 36 refs.; 18 figs

Modeling PWR systems for monitoring primary-to-secondary leakage using tritium tracer

International Nuclear Information System (INIS)

Peiffer, D.G.

1992-01-01

This paper discusses several techniques available for monitoring primary to secondary leakage, focusing on the advantages and disadvantages of each. A mathematical model of Millstone 2 describes the behavior of tritium activity in the secondary plant water when a leak exists. Real data from Millstone 2 illustrate the accuracy and reliability of the model and use of the model to measure the mass of water in the secondary system

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

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

Tritium Decontamination of TFTR D-T Graphite Tiles Employing Ultra Violet Light and a Nd:YAG Laser

International Nuclear Information System (INIS)

Gentile, C.A.; Skinner, C.H.; Young, K.M.; Ciebiera, L.

1999-01-01

The use of an ultra violet (UV) light source (wavelength = 172 nm) and a Nd:YAG Laser for the decontamination of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles will be investigated at the Princeton Plasma Physics Laboratory (PPPL). The development of this form of tritium decontamination may be useful for future D-T burning fusion devices which employ carbon plasma-facing components on the first wall. Carbon tiles retain hydrogen isotopes, and the in-situ tritium decontamination of carbon can be extremely important in maintaining resident in-vessel tritium inventory to a minimum. A test chamber has been designed and fabricated at PPPL. The chamber has the ability to be maintained under vacuum, be baked to 200 *C, and provides sample ports for gas analyses. Tiles from TFTR that have been exposed to D-T plasmas will be placed within the chamber and exposed to either an UV light source or the ND:YAG Laser. The experiment will determine the effectiveness of these two techniques for the removal of tritium. In addition, exposure rates and scan times for the UV light source and/or Nd:YAG Laser will be determined for tritium removal optimization from D-T tiles

Tritium processing and containment technology for fusion reactors: perspective and status

International Nuclear Information System (INIS)

Maroni, V.A.

1976-01-01

This paper reviews the status of selected tritium processing and containment technologies that will be required to support the development of the fusion energy program. Considered in order are the fuel conditioning and recycle systems, the containment and cleanup systems, the blanket processing systems, and two unique problems relating to tritium interactions in neutral beam injectors and first wall coolant circuits. The major technical problem areas appear to lie in the development of (1) high-capacity, rapid recycle plasma chamber evacuation systems; (2) large-capacity (greater than or equal to 100,000 cfm) air handling and processing systems for atmospheric detritiation; (3) tritium recovery technology for liquid lithium blanket concepts; (4) tritium compatible neutral injector systems; and (5) an overall approach to tritium handling and containment that guarantees near zero release to the environment at a bearable cost

Studies applications through tracers techniques and effluent contaminants dispersing in Montevideo coastal waters and east beaches

International Nuclear Information System (INIS)

Suarez, R.; Dellepere, A.; Pintos, A.; Barreiro, M.; Odino, R.; Souto, B.; Badano, A.; Crosignani, L.; Moreno, S.

1995-01-01

With the purpose to define or not the contamination influence in Montevideo coastal waters, uranine and tritium tracers were injected in outlet river. A higher grade of contamination was found in the Montevideo Bay, and several recommendations were given for the future

Tritium recovery from co-deposited layers using 193-nm laser

Science.gov (United States)

Shu, W. M.; Kawakubo, Y.; Nishi, M. F.

Recovery of tritium from co-deposited layers formed in deuterium-tritium plasma operations of the TFTR (Tokamak Fusion Test Reactor) was investigated by the use of an ArF excimer laser operating at the wavelength of 193 nm. At the laser energy density of 0.1 J/cm2, a transient spike of the tritium-release rate was observed at initial irradiation. Hydrogen isotopes were released in the form of hydrogen-isotope molecules during the laser irradiation in vacuum, suggesting that tritium can be recovered readily from the released gases. In a second experiment, hydrogen (tritium) recovery from the co-deposited layers on JT-60 tiles that had experienced hydrogen-plasma operations was investigated by laser ablation with a focused beam of the excimer laser. The removal rate of the co-deposited layers was quite low when the laser energy density was smaller than the ablation threshold (1.0 J/cm2), but reached 1.1 μm/pulse at the laser energy density of 7.6 J/cm2. The effective absorption coefficient in the co-deposited layers at the laser wavelength was determined to be 1.9 μm-1. The temperature of the surface during the irradiation at the laser energy density of 0.5 J/cm2 was measured on the basis of Planck's law of radiation, and the maximum temperature during the irradiation decreased from 3570 K at the initial irradiation to 2550 K at the 1000th pulse of the irradiation.

Upgrade to the Tritium Remote Control and Monitoring System for TFTR D and D

International Nuclear Information System (INIS)

Sichta, P.; Oliaro, G.; Sengupta, S.

2002-01-01

Since 1988, the Tritium Remote Control and Monitoring System (TRECAMS) has performed crucial functions in support of D-T [deuterium-tritium] operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). Although plasma operations on TFTR were completed in 1997, the need for TRECAMS continued. During this period TRECAMS supported the TFTR tritium systems, the TFTR's Shutdown and Safing phase, and the TFTR Decontamination and Decommissioning (D and D) project. The most critical function of the TRECAMS in the post-TFTR era has been to provide a real-time indication of the airborne tritium levels in the tritium areas and the (HVAC) stacks. TRECAMS is a critical tool in conducting safe TFTR D and D tritium-line breaks and other tritium-related work activities. Beginning in 1998, the failure rate of the system's hardware sharply increased. Furthermore, the specialized knowledge required to maintain the original software and hardware was diminishing. It soon became apparent that a failure of the TRECAMS could significantly impact the TFTR D and D project's cost and schedule. To preclude this, the TRECAMS hardware and software was upgraded in the year 2000 to use modern components. This paper will describe that successful upgrade, including a review of the engineering processes and our operating experiences with the upgraded system

Laser system for measuring small changes in plasma tracer concentrations.

Science.gov (United States)

Klaesner, J W; Pou, N A; Parker, R E; Galloway, R L; Roselli, R J

1996-01-01

The authors developed a laser-diode system that can be used for on-line optical concentration measurements in physiologic systems. Previous optical systems applied to whole blood have been hampered by artifacts introduced by red blood cells (RBCs). The system introduced here uses a commercially available filter cartridge to separate RBCs from plasma before plasma concentration measurements are made at a single wavelength. The filtering characteristics of the Cellco filter cartridge (#4007-10, German-town, MD) were adequate for use in the on-line measurement system. The response time of the filter cartridge was less than 40 seconds, and the sieving characteristics of the filter for macromolecules were excellent, with filtrate-to-plasma albumin ratios of 0.98 +/- 0.11 for studies in sheep and 0.94 +/- 0.15 for studies in dogs. The 635-nm laser diode system developed was shown to be more sensitive than the spectrophotometer used in previous studies (Klaesner et al., Annals of Biomedical Engineering, 1994; 22, 660-73). The new system was used to measure the product of filtration coefficient (Kfc) and reflection coefficient for albumin (delta f) in an isolated canine lung preparation. The delta fKfc values [mL/(cmH2O.min.100 g dry lung weight)] measured with the laser diode system (0.33 +/- 0.22) compared favorably with the delta fKfc obtained using a spectrophotometer (0.27 +/- 0.20) and with the Kfc obtained using the blood-corrected gravimetric method (0.32 +/- 0.23). Thus, this new optical system was shown to accurately measure plasma concentration changes in whole blood for physiologic levels of Kfc. The same system can be used with different optical tracers and different source wavelengths to make optical plasma concentration measurements for other physiologic applications.

Confined trapped-alpha behavior in TFTR deuterium-tritium plasmas

International Nuclear Information System (INIS)

Medley, S.S.; Budny, R.V.; Redi, M.H.; Roquemore, A.L.; White, R.B.; Petrov, M.P.; Gorelenkov, N.N.

1997-10-01

Confined trapped-alpha energy spectra and differential radial density profiles in TFTR D-T plasmas are obtained with the Pellet Charge-eXchange (PCX) diagnostic which measures high energy (E α = 0.5--3.5 MeV), trapped alphas (v parallel /v = - 0.048) at a single time slice (Δt ∼ 1 msec) with a spatial resolution of Δr ∼ 5 cm. Tritons produced in D-D plasmas and RF-driven ion tails (H, 3 He or T) were also observed and energetic tritium ion tail measurements will be discussed. PCX alpha and triton energy spectra extending up to their birth energies were measured in the core of MHD-quiescent discharges where the expected classical slowing down and pitch angle scattering effects are not complicated by stochastic ripple diffusion and sawtooth activity. Both the shape of the measured alpha and triton energy distributions and their density ratios are in good agreement with TRANSP predictions, indicating that the PCX measurements are consistent with classical thermalization of the fusion-generated alphas and tritons. From calculations, these results set an upper limit on possible anomalous radial diffusion for trapped alphas of D α ≤ 0.01 m 2 s -1 . Outside the core, where the trapped alphas are influenced by stochastic ripple diffusion effects, the PCX measurements are consistent with the functional dependence of the Goldston-White-Boozer stochastic ripple threshold on the alpha energy and the q-profile. In the presence of strong sawtooth activity, the PCX diagnostic observes significant redistribution of the alpha signal radial profile wherein alphas are depleted in the core and redistributed to well outside the q = 1 radius, but apparently not beyond the energy-dependent stochastic ripple loss boundary

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)

Radioactive or natural tracer techniques for leak determining of dam abutment

International Nuclear Information System (INIS)

Chen Jiansheng; Du Guoping; Zheng Zheng; Sun Jing

1995-01-01

Infiltration and localization of preferential infiltration zones at the dam abutment are measured using radioactive tracer tests of flow in boreholes, meanwhile interconnection between boreholes and the observing water points is analysed. The theory and practice of radioactive tracer synthetic detective method are described to give methods and calculation formulae used under the condition of stable flow in single well to measure permeability coefficient and hydrostatic heads. Major single hole techniques including measurement for seepage line, velocity, rate of seepage flow and relationship of recharge of groundwater in aquifers are introduced briefly. The possibilities offered by natural tracers are analysed, including electric-conduct, ph-value and temperature of water as well as stable isotopes (D, 18 O) and tritium. Furthermore, the sensibilities of this theory and methods were confirmed by detecting seepage flow field of Xinanjiang Dam

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

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

Standardized 15N tracer methods for the evaluation of the plasma protein turnover in clinical practice. 1

International Nuclear Information System (INIS)

Bornhak, H.

1984-01-01

Methods for quantitative isolation of plasma proteins or groups of proteins (total plasma or serum proteins, fibrin, total globulines, α, β, γ-globolines, albumin) are described based on combination of chromatography with precipitation and extraction techniques. These methods are adapted to the special requirements of 15 N analysis. They can be performed in clinic-chemical standard laboratories without special apparatuses or devices. The described procedures are the biochemico-analytical basis for the quantitative evaluation of tracer kinetics data by means of mathematic modelling. (author)

Water renewal in Montevideo's bay: a two compartments model for tritium kinetics

International Nuclear Information System (INIS)

Suarez-Antola, Roberto

2013-01-01

During field work about dynamics and renewal of water in Montevideo's Bay, 100 Ci of tritiated water were evenly distributed in the north-east region of the bay, by a continuous injection of a solution, during 5 hours, from a 200 litres tank, using a peristaltic pump. The whole bay was divided in 20 concentration cells, taking into account available bathymetric charts and corrections from field data obtained in situ. Tritium concentrations (activities per unit volume) and other relevant parameters (temperature, electrical conductivity, etc.) were measured in vertical profiles during three weeks, in the mid-point of each cell, first twice a day and the on a daily basis. Remnant total tritium activity was estimated from cells volumes and midpoint cells activity concentrations. Consistency checks were done. A one compartment model was used to estimate a global renewal time of circa 29 hours. However, the details of the measured tritium kinetics, a careful consideration of bathymetric data, water movements in a tidal environment (measured with drogues, fluorescent tracers and current meters), as well as the results of computer fluid dynamics modelling (in depth averaged) suggests that the bay can be meaningfully divided in two main compartments: a North-East and a South-West compartment. The purpose of this paper is threefold: (1) to describe the construction of a two compartments model for water renewal in Montevideo's Bay, (2) to apply experimental data of tritium kinetics to estimate the parameters of the model, and (3) to discuss the validity of the model and its practical applicability. The meaning of the renewal time of each compartment and its relation with the measured tritium kinetics in each cell is discussed. The perturbations in water circulation and renewal produced by civil works already done or the perturbations that could be expected due to civil works to be done, in relation with Montevideo's harbour, is discussed. The tracer model, jointly with other

Tritium Aspects of Fueling and Exhaust Pumping in Magnetic Fusion Energy

Energy Technology Data Exchange (ETDEWEB)

Baylor, Larry R. [ORNL; Meitner, Steven J. [ORNL

2017-04-01

Magnetically confined fusion plasmas generate energy from deuterium-tritium (DT) fusion reactions that produce energetic 3.5 MeV alpha particles and 14 MeV neutrons. Since the DT fusion reaction rate is a strong function of plasma density, an efficient fueling source is needed to maintain high plasma density in such systems. Energetic ions in fusion plasmas are able to escape the confining magnetic fields at a much higher rate than the fusion reactions occur, thus dictating the fueling rate needed. These lost ions become neutralized and need to be pumped away as exhaust gas to be reinjected into the plasma as fuel atoms.The technology to fuel and pump fusion plasmas has to be inherently compatible with the tritium fuel. An ideal holistic solution would couple the pumping and fueling such that the pump exhaust is directly fed back into pellet formation without including impurity gases. This would greatly reduce the processing needs for the exhaust. Concepts to accomplish this are discussed along with the fueling and pumping needs for a DT fusion reactor.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

(18)F-FDG dynamic PET/CT in patients with multiple myeloma: patterns of tracer uptake and correlation with bone marrow plasma cell infiltration rate.

Science.gov (United States)

Sachpekidis, Christos; Mai, Elias K; Goldschmidt, Hartmut; Hillengass, Jens; Hose, Dirk; Pan, Leyun; Haberkorn, Uwe; Dimitrakopoulou-Strauss, Antonia

2015-06-01

The value of F-FDG PET in the diagnostic approach of multiple myeloma (MM) remains incompletely elicited. Little is known about the kinetics of F-FDG in the bone marrow and extramedullary sites in MM. This study aimed to evaluate quantitative data on kinetics and distribution patterns of F-FDG in MM patients with regard to pelvic bone marrow plasma cell infiltration. The study included 40 patients with primary MM. Dynamic PET/CT scanning of the lower lumbar spine and pelvis was performed after the administration of F-FDG. Whole-body PET/CT studies were performed. Sites of focal increased tracer uptake were considered as highly suggestive of myelomatous involvement after taking into account the patient history and CT findings. Bone marrow of the os ilium without pathologic tracer accumulation served as reference. The evaluation of dynamic PET/CT studies was based in addition to the conventional visual (qualitative) assessment, on semiquantitative (SUV) calculations, as well as on absolute quantitative estimations after application of a 2-tissue compartment model and a noncompartmental approach. F-FDG quantitative information and corresponding distribution patterns were correlated with pelvic bone marrow plasma cell infiltration. Fifty-two myelomatous lesions were detected in the pelvis. All parameters in suspected MM lesions ranged in significantly higher levels than in reference tissue (P PET/CT imaging demonstrated 4 patterns of tracer uptake; these are as follows: negative, focal, diffuse, and mixed (focal/diffuse) tracer uptake. Patients with a mixed pattern of radiotracer uptake had the highest mean plasma cell infiltration rate in their bone marrow, whereas those with negative PET/CT scans demonstrated the lowest bone marrow plasma cell infiltration. In total, 265 focal myeloma-indicative F-FDG-avid lesions were detected, 129 of which correlated with low-dose CT osteolytic findings. No significant correlation between the number of focal lesions detected in PET

Surface erosion and tritium inventory analysis for CIT [Compact Ignition Tokamak

International Nuclear Information System (INIS)

Brooks, J.N.; Dylla, H.F.

1990-09-01

The expected buildup of co-deposited tritium on the CIT carbon divertor and first wall surfaces and operational methods of minimizing the inventory have been examined. The analysis uses impurity transport computer codes, and associated plasma and tritium retention models, to compute the thickness of redeposited sputtered carbon and the resulting co-deposited tritium inventory on the divertor plates and first wall. Predicted erosion/growth rates are dominated by the effect of gaps between carbon tiles. The overall results appear favorable, showing stable operation (finite self-sputtering) and acceptably low (∼25 Ci/pulse) co-deposited tritium rates, at high surface temperature (1700 degree C) design conditions. These results, however, are highly speculative due to serious model inadequacies at the high sputtering rates predicted. If stable operation is obtainable, the prospects appear good for adequate tritium inventory control via helium-oxygen glow discharge cleaning. 25 refs

Estimation of the contribution of gaps to tritium retention in the divertor of ITER

International Nuclear Information System (INIS)

Matveev, D; Kirschner, A; Litnovsky, A; Borodin, D; Samm, U; Schmid, K; Komm, M; Van Oost, G

2014-01-01

An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s −1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas. (paper)

Tracer element studies on deep water formation and circulation in the European Artic Sea

International Nuclear Information System (INIS)

Boenisch, G.

1991-01-01

Tracer element investigations (tritium, helium 3, carbon 14, argon 39, krypton 85 and fluorohydrocarbons) were carried out in the European Arctic Sea. The findings are discussed with a view to their validity in the case of deep water formation and circulation. The data cover the period of 1972 through 1989. (BBR) [de

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

Discharge measurements of the River Rufiji (Tanzania) using artificial tritium

International Nuclear Information System (INIS)

Dincer, T.; Florkowski, T.; Salamba, S.

1984-01-01

The use of chemical or radioactive tracers for measuring stream flow is now the established method for discharges up to about 200 m 3 /s. For larger flows and higher suspended load the chemical tracers and also gamma-emitting radioactive tracers become cumbersome if not impossible to use when good accuracy is required. Tritiated water proved to be a good and safe tracer, provided care is taken in handling (no contamination of samples) and the experiments are adequately planned (good estimation of mixing lengths, water velocity and sampling duration). The paper describes discharge measurements performed in 1982 and 1983 in the river Rufiji (Tanzania). Flow rates up to 2000 m 3 /s have been measured, with estimated errors varying between 2 and 4%. Because of high river turbulence in the measurement section, good mixing has been observed over a distance of 7 km (this is much shorter than the distance recommended by various formulae for calculating the mixing length). The problem of selecting the mixing length is discussed and recommendations are given for planning future experiments. Sample contamination as experienced during the first phase of measurements in the river Rufiji is also treated. It is concluded that, technically and economically, the tritium tracer method is feasible for calibrating rating curves (water stage/flow relationship) in turbulent large rivers, also in remote areas. (author)

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

Analysis of air mass trajectories to explain observed variability of tritium in precipitation at the Southern Sierra Critical Zone Observatory, California, USA.

Science.gov (United States)

Visser, Ate; Thaw, Melissa; Esser, Brad

2018-01-01

Understanding the behavior of tritium, a radioactive isotope of hydrogen, in the environment is important to evaluate the exposure risk of anthropogenic releases, and for its application as a tracer in hydrology and oceanography. To understand and predict the variability of tritium in precipitation, HYSPLIT air mass trajectories were analyzed for 16 aggregate precipitation samples collected over a 2 year period at irregular intervals at a research site located at 2000 m elevation in the southern Sierra Nevada (California, USA). Attributing the variation in tritium to specific source areas confirms the hypothesis that higher latitude or inland sources bring higher tritium levels in precipitation than precipitation originating in the lower latitude Pacific Ocean. In this case, the source of precipitation accounts for 79% of the variation observed in tritium concentrations. Air mass trajectory analysis is a promising tool to improve the predictions of tritium in precipitation at unmonitored locations and thoroughly understand the processes controlling transport of tritium in the environment. Copyright © 2017 Elsevier Ltd. All rights reserved.

Hydrogen, deuterium, and tritium isotope exchange experiments in JET

Energy Technology Data Exchange (ETDEWEB)

Horton, L.D.; Andrew, P.; Bracco, G.; Conroy, S.; Corti, S.; Ehrenberg, J.; Goodall, D.H.J.; Jarvis, O.N.; Lomas, P.; Loughlin, M.; Peacock, A.T.; Saibene, G.; Sadler, G.; Sartori, R.; Stamp, M.F.; Thomas, P.R.; Belle, P. van (JET Joint Untertaking, Abingdon, Oxfordshire (United Kingdom))

1992-12-01

Isotope exchange experiments have been performed in JET using hydrogen, deuterium, and, in the recent preliminary tritium experiment (PTE), tritium. The rate of change-over from one isotope to another involves two quite different time constants. We have modelled this behaviour using a multireservoir model which splits the accessible hydrogenic particles into two groups, each having a different rate of exchange of particles with the plasma. By applying this model to the sequence of discharges during and after the PTE, we can determine the parameters in the model. The resulting fit also gives a good representation of hydrogen/deuterium change-over experiments, indicating that the tritium behaves in the same manner as other hydrogen isotopes, at least as far as recycling is concerned. Discrepancies between the model and the actual measurements of tritium recovery after the PTE lead us to conclude that isotope exchange processes resulting from collisions of molecules with the vessel walls play a significant role in spreading tritrium around the machine. (orig.).

Thermal effect of periodical bakeout on tritium inventory in first wall and permeation to coolant in reactor life

International Nuclear Information System (INIS)

Nakahara, Katsuhiko

1989-01-01

In view of safety, it is very important to control the tritium inventory in first walls and permeation to the coolant. A time-dependent diffusion and temperature calculation code, TPERM, was developed. Using this code, a numerical study on the long term effects of the bakeout temperature on tritium inventory and tritium permeation to the coolant was made. In this study, an FER type first wall (stainless steel) was considered and a cyclic operation (one cycle includes a plasma burn phase and a bakeout phase) was assumed. The results are as follows: (i) There is almost no difference in the tritium inventory in the first wall between the operation with 150 0 C-bakeout and the continuous burning operation (without bakeout). In both cases there is not tritium permeation to the coolant at 5 years' integrated burn time. The 150 0 C-bakeout is effective to release tritium in the surface (to 0.1 mm depth) region on the plasma side, but it is not effective to decrease the tritium inventory over the reactor life. (ii) To decrease the tritium inventory, a bakeout at a temperature higher than 150 0 C is necessary. But a high temperature bakeout causes earlier tritium permeation to the coolant. (iii) From these results it is suggested that the decrease the tritium inventory over the reactor life by bakeout, some form of protection against tritium permeation or a decontamination device in the cooling (or bakeout) system becomes necessary. (orig.)

JAERI Fuel Cleanup System (J-FCU) stand-alone tritium test at the TSTA

International Nuclear Information System (INIS)

Konishi, Satoshi; Hayashi, Takumi; Inoue, Masahiko

1993-03-01

JAERI designed, fabricated, and installed the JAERI Fuel Cleanup System (J-FCU) as a subsystem of simulated fusion fuel loop at the TSTA. The main function of the J-FCU is to purify and to recover hydrogen isotopes from simulated plasma exhaust while exhausting tritium free impurities. After a lot of deuterium tests, a first tritium test of the J-FCU was performed with one gram of tritium at the TSTA on June 1991. Main purpose of this test was to evaluate the total integrity and function of the J-FCU system with a DT mixture. Through this test, the J-FCU was operated well and its function with tritium was demonstrated. This report describes the detail test results of the J-FCU first tritium test and discuss its functions by stand-alone mode. Residual tritium inventory of the J-FCU system was also discussed. (author)

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.)

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)

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)

Use of tracer tests to evaluate hydraulic properties of constructed wetlands

International Nuclear Information System (INIS)

Wachniew, P.; Czuprynski, P.; Maloszewski, P.

2004-01-01

Knowledge of hydraulic properties is a perquisite for studies of constructed wetlands functioning. Bromide ions and tritium were used as a tracers to derive RTDs for two constructed wetlands: a reed bed with subsurface flow and a Lemna pond. Quantitative hydraulic characteristics (mean travel time of water, dispersion number) of the wetlands were evaluated from RTDs (Residence Time Distributions) by means of a mathematical model of waste water flow. (author)

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)

Tritium assay of Li2O pellets in the LBM/LOTUS experiments

International Nuclear Information System (INIS)

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

1986-01-01

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

Heating and transport in TFTR D-T plasmas

International Nuclear Information System (INIS)

Zarnstorff, M.C.; Scott, S.D.

1994-01-01

The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in TFTR. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. The radial profiles of the deuterium and tritium densities are determined from the DT fusion neutron emission profile

A study of the effectiveness and application of environmental tritium as a groundwater tracer in a semi-arid region of Botswana

International Nuclear Information System (INIS)

Sellschop, J.

1974-12-01

To assess the potential use of environmental tritium in groundwater studies, an extensive sampling programme was carried out from a network of wells in groundwater basins of Lobatse and Serowe in South Africa. Statistical evaluations were performed on collected data to assess the significance of low-level tritium concentrations observed in groundwaters. Tritium depth profiles were studied at selected sites to estimate the rate of direct recharge to the aquifers. The tritium data were used to study mixing patterns and mechanism of recharge, and quantitative estimates of storage capacity were made for different aquifer units in the areas studied. Supplementary, carbon-14, stable isotope and hydrochemical data were discussed to achieve a full understanding of the behaviour of the aquifers. The results of the study indicate that environmental tritium, even at the low concentrations found in the southern hemisphere, is a powerful tool in hydrogeological studies and, together with simultaneous measurements of other environmental isotopes, could prove very useful for studying the behaviour of aquifers and for quantitative estimates of certain parameters of groundwater systems

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)

Estimation of the tritium retention in ITER tungsten divertor target using macroscopic rate equations simulations

Science.gov (United States)

Hodille, E. A.; Bernard, E.; Markelj, S.; Mougenot, J.; Becquart, C. S.; Bisson, R.; Grisolia, C.

2017-12-01

Based on macroscopic rate equation simulations of tritium migration in an actively cooled tungsten (W) plasma facing component (PFC) using the code MHIMS (migration of hydrogen isotopes in metals), an estimation has been made of the tritium retention in ITER W divertor target during a non-uniform exponential distribution of particle fluxes. Two grades of materials are considered to be exposed to tritium ions: an undamaged W and a damaged W exposed to fast fusion neutrons. Due to strong temperature gradient in the PFC, Soret effect’s impacts on tritium retention is also evaluated for both cases. Thanks to the simulation, the evolutions of the tritium retention and the tritium migration depth are obtained as a function of the implanted flux and the number of cycles. From these evolutions, extrapolation laws are built to estimate the number of cycles needed for tritium to permeate from the implantation zone to the cooled surface and to quantify the corresponding retention of tritium throughout the W PFC.

Estimation of transit times in a Karst Aquifer system using environmental tracers: Application on the Jeita Aquifer system-Lebanon.

Science.gov (United States)

Doummar, Joanna; Hamdan, Ahmad

2016-04-01

Estimating transit times is essential for the assessment of aquifer vulnerability to contaminants. Groundwater in karst aquifer is assumed to be relatively young due to fast preferential pathways; slow flow components are present in water stored in the fissured matrix. Furthermore, transit times are site specific as they depend on recharge rates, temperatures, elevation, and flow media; saturated and unsaturated zones. These differences create significant variation in the groundwater age in karst systems as the water sampled will be a mix of different water that has been transported through different flow pathways (fissured matrix and conduits). Several methods can be applied to estimate water transit time of an aquifer such as artificial tracers, which provide an estimate for fast flow velocities. In this study, groundwater residence times in the Jeita spring aquifer (Lebanon) were estimated using several environmental tracers such as Chlorofluorocarbons (CFCs), Sulfur Hexafluoride (SF6), Helium-Tritium (3H, 3H- 3He). Additional stable isotope and major ion analysis was performed to characterize water types. Groundwater samples were collected from six different wells in the Jeita catchment area (Jurassic Kesrouane aquifer) as well as from the spring and cave itself. The results are reproducible for the Tritium-Helium method, unlike for the CFC/SF6 methods that yielded poor results due to sampling problems. Tritium concentrations in all groundwater samples show nearly the same concentration (~2.73 TU) except for one sample with relatively lower tritium concentration (~2.26 TU). Ages ranging from 0.07 ± 0.07 years to 23.59 ± 0.00 years were obtained. The youngest age is attributed to the spring/ cave while the oldest ages were obtained in wells tapping the fissured matrix. Neon in these samples showed considerable variations and high delta Ne in some samples indicating high excess air. Four (4) samples showed extreme excess air (Delta-Ne is greater than 70 %) and

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

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.)

Transport of deuterium, tritium and helium in a tokamak

International Nuclear Information System (INIS)

Potters, J.H.H.M.

1984-02-01

A one-dimensional numerical model for determining steady-state radial profiles of the densities of the particles, including neutrals, in a multispecies toroidal plasma is described. For prescribed temperature profiles, the coupled momentum and particle balances of the ions are solved numerically with a newly developed compact finite difference scheme for a non-equidistant mesh. Neutral densities are obtained by solving the Boltzmann equations, using a collocation method. The model is applied to deuterium-tritium plasmas without and with a helium admixture. For the charged particles, Pfirsch-Schlueter transport, including the highly collisional extension, and either of two anomalous transport models are adopted. For equal densities of deuterons and tritons in the plasma centre, the neutral tritium density in front of the wall is found to be 1.3 to 1.6 times higher than that of deuterium, depending on the plasma density, the temperature profile and the transport model. Secondly, it is found that pumping neutral helium, originating from fusion alpha particles, out of a cold plasma/gas blanket surrounding the hot plasma is not feasible, as the helium gas density, corresponding to a relative abundance of alpha-particles in the plasma core below 10%, is very low. Although depending strongly on the ion transport model and being increased by elastic collisions between neutral helium and charged hydrogen isotopes, the neutral helium enrichment ratio is always much less than unity. (Auth.)

In-Situ Imaging and Quantification of Tritium Surface Contamination via Coherent Fiber Bundle

International Nuclear Information System (INIS)

Gentile, Charles A.; Parker, John J.; Zweben, Stewart J.

2001-01-01

Princeton Plasma Physics Laboratory (PPPL) has developed a method of imaging tritium on in-situ surfaces for the purpose of real-time data collection. This method expands upon a previous tritium imaging concept, also developed at PPPL. Enhancements include an objective lens coupled to the entry aperture of a coherent fiber optic (CFO) bundle, and a relay lens connecting the exit aperture of the fiber bundle to an intensifier tube and a charge-coupled device (CCD) camera. The system has been specifically fabricated for use in determining tritium concentrations on first wall materials. One potential complication associated with the development of D-T [deuterium-tritium] fueled fusion reactors is the deposition of tritium (i.e., co-deposited layer) on the surface of the primary wall of the vacuum vessel. It would be advantageous to implement a process to accurately determine tritium distribution on these inner surfaces. This fiber optic imaging device provides a highly practical method for determining the location, concentration, and activity of surface tritium deposition. In addition, it can be employed for detection of tritium ''hot-spots'' and ''hide-out'' regions present on the surfaces being imaged

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)

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)

Development of a plasma driven permeation experiment for TPE

Energy Technology Data Exchange (ETDEWEB)

Buchenauer, Dean, E-mail: dabuche@sandia.gov [Sandia National Laboratories, Livermore, CA (United States); Kolasinski, Robert [Sandia National Laboratories, Livermore, CA (United States); Shimada, Masa [Idaho National Laboratory, Idaho Falls, ID (United States); Donovan, David [Sandia National Laboratories, Livermore, CA (United States); Youchison, Dennis [Sandia National Laboratories, Albuquerque, NM (United States); Merrill, Brad [Idaho National Laboratory, Idaho Falls, ID (United States)

2014-10-15

Highlights: • We have designed and fabricated a novel tritium permeation membrane holder for use in the Tritium Plasma Experiment (TPE). • The membrane temperature is controlled by varying the cooling flow rate and proximity of a spiral cooling channel. • Sealing tests have demonstrated adequate helium leak rates up to temperatures of 1000 °C. • Flow modeling indicates a minimal helium pressure drop across the membrane holder (<700 Pa). • Thermal modeling shows good heat removal and minimal membrane temperature variation (±2%) even up to peak TPE ion fluxes. - Abstract: Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.

Development of a plasma driven permeation experiment for TPE

International Nuclear Information System (INIS)

Buchenauer, Dean; Kolasinski, Robert; Shimada, Masa; Donovan, David; Youchison, Dennis; Merrill, Brad

2014-01-01

Highlights: • We have designed and fabricated a novel tritium permeation membrane holder for use in the Tritium Plasma Experiment (TPE). • The membrane temperature is controlled by varying the cooling flow rate and proximity of a spiral cooling channel. • Sealing tests have demonstrated adequate helium leak rates up to temperatures of 1000 °C. • Flow modeling indicates a minimal helium pressure drop across the membrane holder (<700 Pa). • Thermal modeling shows good heat removal and minimal membrane temperature variation (±2%) even up to peak TPE ion fluxes. - Abstract: Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes

Isotopes as tracers of the oceanic circulation: Results from the World Ocean Circulation Experiment

International Nuclear Information System (INIS)

Schlosser, P.; Jenkins, W.J.; Key, R.; Lupton, J.

2002-01-01

During the past decades, natural and anthropogenic isotopes such as tritium ( 3 H), radiocarbon ( 14 C), 3 He, or the stable isotopes of water have been used in studies of the dynamics of natural systems. Early applications of tracers to studies of the ocean were directed at determination of circulation patterns and mean residence times of specific water masses, as well as estimates of mixing coefficients. These exploratory studies suggested that tracers can add significantly to our understanding of the oceanic circulation. In order to fully exploit this potential, the first global tracer study, the GEochemical Ocean SECtions Study (GEOSECS), was launched. From the GEOSECS results it was immediately apparent that very close coordination of tracer programs with physical oceanography studies is required for full utilization of tracer data. During the 1980s plans for the World OCean Experiment (WOCE) were developed. As part of its Hydrographic Program (WHP), especially during the one-time survey, a set of tracers were measured on a global scale with unprecedented spatial resolution (both lateral and vertical). The original plan included a larger number of tracers (CFCs, 3 H/ 3 He, 14 C, 39 Ar, stable isotopes of water, helium isotopes, 228 Ra, 90 Sr, 137 Cs, 85 Kr) than could actually be measured systematically (CFCs, 3 H/ 3 He, 14 C, H 2 18 O/H 2 16 O, helium isotopes). Nevertheless, the resulting data set, which presently is under evaluation, exceeds those obtained from pre-WOCE tracer studies by a wide margin. In this contribution, we describe the existing WOCE data set and demonstrate the type of results that can be expected from its interpretation on the basis of a few selected examples. These examples include: (1) the application of tritium and 3 He to studies of the ventilation of the upper waters in the Pacific Ocean, (2) the spreading of intermediate water in the Pacific and Indian oceans as derived from the distribution of 3 He, and (3) the evaluation of

Tritium inventory control--the experience with DT tokamaks and its relevance for future machines

International Nuclear Information System (INIS)

Bell, A.C.; Gentile, C.A.; Laesser, R.L.K.; Coad, J.P.

2003-01-01

At present, the commercial use of tritium is relatively small scale. The main source of supply is as a by-product of heavy water moderated fission reactors and the products are mainly discrete sources or tracers with activity typically in the GBq range. There are in general no restrictions on the use of tritium other than those, which would normally apply to the use of radioactive material. The future use of tritium as intermediate fuel for a fusion power plant series will involve an increase by several orders of magnitude in the industrial use of tritium and may increase concerns relating to safety, transport and waste disposal. In addition, the use of tritium in fusion power will be unable to be satisfied by current sources of supply and tritium production in future fusion power plants will be essential for the operation of the plants as well as for the start of new ones. Power plant studies have, however, shown that these issues can be satisfactorily addressed. In addition the values for clearance of tritiated materials in a number of countries are consistent with the low environmental impact of disposal of tritiated waste. There are, however, many practical operational and regulatory problems, which will need to be solved in the context of the experimental programmes. The current regulations for control and accountancy of tritium inventory, as applied internationally and in specific countries, are reviewed and their influence on the DT fuel cycle considered. The effect of safety case limits on the need for control of tritium inventory in TFTR, JET and ITER is analysed. The sensitivity of the fuel cycle to tritium inventory is considered. The experience of controlling tritium inventory in TFTR and JET is reviewed and the latest results from JET presented. This takes into account the limits and constraints, the differing requirements for tritium processing, in-vessel retention, the needs for waste management and decommissioning including detritiation, and

Tissue factor-dependent activation of tritium-labeled factor IX and factor X in human plasma

International Nuclear Information System (INIS)

Morrison, S.A.; Jesty, J.

1984-01-01

A comparism was made of the tissue factor-dependent activation of tritium-labeled factor IX and factor X in a human plasma system and a study was made of the role of proteases known to stimulate factor VII activity. Plasma was defibrinated by heating and depleted of its factors IX and X by passing it through antibody columns. Addition of human brain thromboplastin, Ca2+, and purified 3H-labeled factor X to the plasma resulted, after a short lag, in burst-like activation of the factor X, measured as the release of radiolabeled activation peptide. The progress of activation was slowed by both heparin and a specific inhibitor of factor Xa but factor X activation could not be completely abolished by such inhibitors. In the case of 3H-factor IX activation, the rate also increased for approximately 3 min after addition of thromboplastin, but was not subsequently curtailed. A survey of proteases implicated as activators of factor VII in other settings showed that both factor Xa and factor IXa could accelerate the activation of factor IX. However, factor Xa was unique in obliterating activation when present at concentrations greater than approximately 1 nM. Heparin inhibited the tissue factor-dependent activation of factor IX almost completely, apparently through the effect of antithrombin on the feedback reactions of factors Xa and IXa on factor VII. These results suggest that a very tight, biphasic control of factor VII activity exists in human plasma, which is modulated mainly by factor Xa. At saturation of factor VIIa/tissue factor, factor IX activation was significantly more rapid than was previously found in bovine plasma under similar conditions. The activation of factor X at saturation was slightly more rapid than in bovine plasma, despite the presence of heparin

Tritium recovery from fusion blankets using solid lithium compounds. I. Design and minimization of tritium inventory

International Nuclear Information System (INIS)

Powell, J.R.

1975-01-01

Tritium blanket inventories of 100 curies/MW(e) are readily achievable, and inventories as low as 10 curies/MW(e) are possible for blankets with small lithium compound particulates (less than or equal to 50μ) at T greater than or equal to 800 0 C. Of the three release modes [A - to the main coolant (e.g., He) stream; B - to a separate processing circuit; and C - to the plasma region], mode A appears optimum for blankets using gas-cooled metallic structures (e.g., Al, stainless), while mode C appears optimum for high temperature refractory (e.g., C, SiC) structures. The greater structural complexity of mode B makes it less attractive than modes A and C. No recovery method is required for mode C release. With mode A release, tritium inventory in the coolant circuits ranges from 1 to 10 curies/MW(e), depending on processing parameters. Tritium leak rates to the environment during normal operation can be kept to less than or equal to 10 -3 curies/MW(e) per day with low permeability barriers. In general, a mixture of T 2 and T 2 O is present in the coolant stream. Three methods of tritium recovery are examined: (1) Conversion to T 2 followed by absorption in a metal hydride bed. (2) Conversion to T 2 followed by condensation at approximately 6 0 K. (3) Conversion to T 2 O followed by condensation at approximately 100 0 K

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)

Plasma heating by non-linear wave-Plasma interaction | Echi ...

African Journals Online (AJOL)

We simulate the non-linear interaction of waves with magnetized tritium plasma with the aim of determining the parameter values that characterize the response of the plasma. The wave-plasma interaction has a non-conservative Hamiltonian description. The resulting system of Hamilton's equations is integrated numerically ...

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

Advances in the use of tritium as a radiotracer for oil consumption measurement

International Nuclear Information System (INIS)

Shore, P.R.

1988-01-01

The oil consumption of a turbocharged, aftercooled direct-injection truck diesel engine was measured using a tritium-tracer technique. The advantages of the method over other chemical and radioactive tracers are described, and supplemented with data from radioanalysis of tritiated oils. As a proportion of fuel consumption, the oil consumption was shown to range from 0.4% depending upon the engine's load and speed, with the highest consumption at idle and at full load conditions. The mass consumption rate ranged from 6 g/h at light load, low speed to 230 g/h at full load, rated speed. The contribution of consumed oil to another truck engine's particulate-bound hydrocarbon emission was shown to be greatest at light and intermediate loads and negligible at high loads

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.))

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

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

Reacting plasma project at IPP Japan

International Nuclear Information System (INIS)

Miyahara, A.; Momota, H.; Hamada, Y.; Kawamura, K.; Akimune, H.

1981-01-01

Contributed papers of the seminar on burning plasma held at UCLA are collected. Paper on ''overview of reacting plasma project'' described aim and philosophy of R-Project in Japan. Paper on ''Burning plasma and requirements for design'' gave theoretical aspect of reacting plasma physics while paper on ''plasma container, heating and diagnostics'' treated experimental aspect. Tritium handling is essential for the next step experiment; therefore, paper on ''Tritium problems in burning plasma experiments'' took an important part of this seminar. As appendix, paper on ''a new type of D - ion source using Si-semiconductor'' was added because such an advanced R and D work is essential for R-Project. (author)

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.)

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

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

International Nuclear Information System (INIS)

Strack, S.; Koenig, L.A.

1981-12-01

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

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.)

The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

International Nuclear Information System (INIS)

Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse.

1988-03-01

We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

Stability of tritium permeation prevention barrier with TiC and TiN + TiC coating

International Nuclear Information System (INIS)

Shan Changqi; Chen Qingwang; Dai Shaoxia; Jiang Weisheng

1999-01-01

The stability of tritium permeation prevention barrier of 316L stainless steel with coating TiC and TiN + TiC under the conditions of very large thermal gradient, thermal cycling and plasma irradiation is researched. The research includes two aspects: one is the study on the stability resisting H + plasma irradiation; another is on the ability of two coating materials when they are used in long term under the condition of very large thermal gradient and cycling. The results show that TiC and TiN + TiC composite coating materials, after chemical heat treatment and forming tritium permeation prevention barrier, can resist H + ion irradiation, and also can resist very large thermal gradient and thermal cycling. The long time experiments show that tritium permeation prevention barrier of those coating materials is stable when they are used in long term

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

Development of a compact tritium activity monitor and first tritium measurements

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-15

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

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

In depth fusion flame spreading with a deuterium—tritium plane fuel density profile for plasma block ignition

International Nuclear Information System (INIS)

Malekynia, B.; Razavipour, S. S.

2012-01-01

Solid-state fuel ignition was given by Chu and Bobin according to the hydrodynamic theory at x = 0 qualitatively. A high threshold energy flux density, i.e., E* = 4.3 × 10 12 J/m 2 , has been reached. Recently, fast ignition by employing clean petawatt—picosecond laser pulses was performed. The anomalous phenomena were observed to be based on suppression of prepulses. The accelerated plasma block was used to ignite deuterium—tritium fuel at solid-state density. The detailed analysis of the thermonuclear wave propagation was investigated. Also the fusion conditions at x ≠ 0 layers were clarified by exactly solving hydrodynamic equations for plasma block ignition. In this paper, the applied physical mechanisms are determined for nonlinear force laser driven plasma blocks, thermonuclear reaction, heat transfer, electron—ion equilibration, stopping power of alpha particles, bremsstrahlung, expansion, density dependence, and fluid dynamics. New ignition conditions may be obtained by using temperature equations, including the density profile that is obtained by the continuity equation and expansion velocity. The density is only a function of x and independent of time. The ignition energy flux density, E* t , for the x ≠ 0 layers is 1.95 × 10 12 J/m 2 . Thus threshold ignition energy in comparison with that at x = 0 layers would be reduced to less than 50 percent. (physics of gases, plasmas, and electric discharges)

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

International Nuclear Information System (INIS)

Malara, C.; Casini, G.; Viola, A.

1995-01-01

The question of the tritium recovery in water-cooled Pb17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pb17Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit. (orig.)

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

A Visual Detection System for Determining Tritium Surface Deposition Employing Phosphor Coated Materials

International Nuclear Information System (INIS)

Gentile, C.A.; Skinner, C.H.; Young, K.M.; Zweben, S.J.

1999-01-01

A method for visually observing tritium deposition on the surface of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles is being investigated at the Princeton Plasma Physics Laboratory. A green phosphor (P31, zinc sulfide: copper) similar to that used in oscilloscope screens with a wavelength peak of 530 nm was positioned on the surface of a TFTR D-T tile. The approximately 600 gram tile, which contains approximately 1.5 Ci of tritium located on the top approximately 1-50 microns of the surface, was placed in a two liter lexan chamber at Standard Temperature and Pressure (STP). The phosphor plates and phosphor powder were placed on the surface of the tile which resulted in visible light being observed, the consequence of tritium betas interacting with the phosphor. This technique provides a method of visually observing varying concentrations of tritium on the surface of D-T carbon tiles, and may be employed (in a calibrated system) to obtain quantitative data

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

Li2O-pebble type tritium breeding blanket for fusion experimental reactor, 1

International Nuclear Information System (INIS)

Tone, Tatsuzo; Iida, Hiromasa; Tanaka, Yoshihisa

1984-01-01

The fusion experimental reactor is the next stage device in Japan, which is planned to be constructed following the critical plasma experimental device JT-60 being constructed at present. The breeding blanket installed in nuclear fusion reactors is one of most important structures, and it is required to satisfy the fundamental performance of producing and continuously recovering tritium as the nuclear fusion fuel, and other requirement in good coordination. The Li 2 O pebble type breeding blanket that Kawasaki Heavy Industries Ltd. has examined is the concept for resolving the problems of the mass transfer and thermal stress cracking of Li 2 O, which are important in blanket design. In this paper, the concept and characteristics of this breeding blanket are discussed from the viewpoint of the breeding and continuous recovery of tritium, the ease of manufacture and the maintenance of soundness. The breeding blanket is composed of breeding region, tritium purge region, cooling region, plasma stabilizing conductors and blanket container. Li 2 O is excellent in its tritium breeding performance and heat conductivity. The functions required for the breeding blanket, the fundamental structure, the examples of breeding blanket concept, the selection of breeding blanket concept, the characteristics of Li 2 O pebble type blanket and its future prospect are described. (Kako, I.)

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

International Nuclear Information System (INIS)

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

1985-04-01

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

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)

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

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

International Nuclear Information System (INIS)

Garber, H.J.

1975-10-01

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

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

Long-term tritium monitoring to study river basin dynamics: case of the Danube River basin

Science.gov (United States)

Aggarwal, Pradeep; Araguas, Luis; Groening, Manfred; Newman, Brent; Kurttas, Turker; Papesch, Wolfgang; Rank, Dieter; Suckow, Axel; Vitvar, Tomas

2010-05-01

During the last five decades, isotope concentrations (O-18, D, tritium) have been extensively measured in precipitation, surface- and ground-waters to derive information on residence times of water in aquifers and rivers, recharge processes, and groundwater dynamics. The unique properties of the isotopes of the water molecule as tracers are especially useful for understanding the retention of water in river basins, which is a key parameter for assessing water resources availability, addressing quality issues, investigating interconnections between surface- and ground-waters, and for predicting possible hydrological shifts related to human activities and climate change. Detailed information of the spatial and temporal changes of isotope contents in precipitation at a global scale was one of the initial aims of the Global Network of Isotopes in Precipitation (GNIP), which has provided a detailed chronicle of tritium and stable isotope contents in precipitation since the 1960s. Accurate information of tritium contents resulting of the thermonuclear atmospheric tests in the 1950s and 1960s is available in GNIP for stations distributed world-wide. Use of this dataset for hydrological dating or as an indicator of recent recharge has been extensive in shallow groundwaters. However, its use has been more limited in surface waters, due to the absence of specific monitoring programmes of tritium and stable isotopes in rivers, lakes and other surface water bodies. The IAEA has recently been compiling new and archival isotope data measured in groundwaters, rivers, lakes and other water bodies as part of its web based Water Isotope System for Data Analysis, Visualization and Electronic Retrieval (WISER). Recent additions to the Global Network of Isotopes in Rivers (GNIR) contained within WISER now make detailed studies in rivers possible. For this study, we are re-examining residence time estimates for the Danube in central Europe. Tritium data are available in GNIR from 15

Tritium migration in the Twin Lake 260-metre natural-gradient dispersion test

International Nuclear Information System (INIS)

Killey, R.W.D.; Wills, C.A.; Moltyander, G.L.

1990-01-01

The experiment reported here is an expansion of studies of dispersive processes in an unconfined sand aquifer on the property of Chalk River Laboratories near Twin Lake, covering a 270 m flow path between the injection well and the groundwater discharge area. Previous experience had shown that the use of a non-reactive tracer that emits moderate-energy gamma rays provides much more information than can be gleaned from tracers that require actual water sample collection. At the time of this experiment non-reactive gamma-emitting tracers with half-life long enough to undertake the 270 m experiment had not been developed. Tritium was used so some information on large-scale dispersion phenomena could be collected and instrumentation would be properly placed for a subsequent experiment that would use a gamma-emitting tracer. Because of the scoping character of the experiment, only limited data were collected. The experiment involved the controlled injection of a relatively large (60 m 3 ) volume of water labelled with tritiated water, and the subsequent tracking of the slug during natural gradient convective transport to the discharge area. This paper describes the hydrogeologic setting and the experimental and analytical methods, and presents and discusses the findings. (L.L.) (8 refs., 11 figs., tab.)

Tritium levels in milk in the vicinity of chronic tritium releases

International Nuclear Information System (INIS)

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

2016-01-01

Tritium is the radioactive isotope of hydrogen. It can be integrated into most biological molecules. Even though its radiotoxicity is weak, the effects of tritium can be increased following concentration in critical compartments of living organisms. For a better understanding of tritium circulation in the environment and to highlight transfer constants between compartments, we studied the tritiation of different agricultural matrices chronically exposed to tritium. Milk is one of the most frequently monitored foodstuffs in the vicinity of points known for chronic release of radionuclides firstly because dairy products find their way into most homes but also because it integrates deposition over large areas at a local scale. It is a food which contains all the main nutrients, especially proteins, carbohydrates and lipids. We thus studied the tritium levels of milk in chronic exposure conditions by comparing the tritiation of the main hydrogenated components of milk, first, component by component, then, sample by sample. Significant correlations were found between the specific activities of drinking water and free water of milk as well as between the tritium levels of cattle feed dry matter and of the main organic components of milk. Our findings stress the importance of the metabolism on the distribution of tritium in the different compartments. Overall, dilution of hydrogen in the environmental compartments was found to play an important role dimming possible isotopic effects even in a food chain chronically exposed to tritium. - Highlights: • Tritium can be incorporated in all the hydrogenated components of milk. • Components' isotopic ratios T/H of chronically exposed milk remain in the same range. • In environmental conditions, distribution of tritium in milk components varies. • Metabolism plays a role in the distribution of tritium in the components of milk. • In environmental conditions, dilution of hydrogen dims possible isotopic effects.

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

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

Plasma-material interactions

International Nuclear Information System (INIS)

Wilson, K.L.

1984-01-01

Plasma-interactive components must be resistant to erosion processes, efficient in heat removal, and effective in minimizing tritium inventory and permeation. As long as plasma edge temperatures are 50 eV, no one material can satisfy the diverse requirements imposed by these plasma materials interactions. The only solution is the design of duplex, or even more complicated, structures. The material that faces the plasma should be low atomic number, with acceptable erosion and evaporation characteristics. The substrate material must have high thermal conductivity for heat removal. Finally, materials must be selected judiciously for tritium compatibility. In conclusion, materials play a critical role in the achievement of safe and economical magnetic fusion energy. Improvements in materials have already led to many advances in present day device operation, but additional innovative materials solutions are required for the critical plasma materials interaction issues in future power reactors

Evaluation of Parameter Uncertainty Reduction in Groundwater Flow Modeling Using Multiple Environmental Tracers

Science.gov (United States)

Arnold, B. W.; Gardner, P.

2013-12-01

Calibration of groundwater flow models for the purpose of evaluating flow and aquifer heterogeneity typically uses observations of hydraulic head in wells and appropriate boundary conditions. Environmental tracers have a wide variety of decay rates and input signals in recharge, resulting in a potentially broad source of additional information to constrain flow rates and heterogeneity. A numerical study was conducted to evaluate the reduction in uncertainty during model calibration using observations of various environmental tracers and combinations of tracers. A synthetic data set was constructed by simulating steady groundwater flow and transient tracer transport in a high-resolution, 2-D aquifer with heterogeneous permeability and porosity using the PFLOTRAN software code. Data on pressure and tracer concentration were extracted at well locations and then used as observations for automated calibration of a flow and transport model using the pilot point method and the PEST code. Optimization runs were performed to estimate parameter values of permeability at 30 pilot points in the model domain for cases using 42 observations of: 1) pressure, 2) pressure and CFC11 concentrations, 3) pressure and Ar-39 concentrations, and 4) pressure, CFC11, Ar-39, tritium, and He-3 concentrations. Results show significantly lower uncertainty, as indicated by the 95% linear confidence intervals, in permeability values at the pilot points for cases including observations of environmental tracer concentrations. The average linear uncertainty range for permeability at the pilot points using pressure observations alone is 4.6 orders of magnitude, using pressure and CFC11 concentrations is 1.6 orders of magnitude, using pressure and Ar-39 concentrations is 0.9 order of magnitude, and using pressure, CFC11, Ar-39, tritium, and He-3 concentrations is 1.0 order of magnitude. Data on Ar-39 concentrations result in the greatest parameter uncertainty reduction because its half-life of 269

Plasma norepinephrine in humans: limitations in assessment of whole body norepinephrine kinetics and plasma clearance

DEFF Research Database (Denmark)

Christensen, N J; Henriksen, Jens Henrik Sahl

1989-01-01

]IP and 131I-hippurate, whole body clearance from plasma of [3H]NE, as obtained from infusion rate divided by plasma concentration of tracer [1.74 +/- 0.64 (SD) 1/min] was significantly higher than the value obtained by total tracer infusion divided by total plasma area of tracer (1.27 +/- 0.51, P less than 0...... irreversible removal of NE, is smaller than previously estimated due to recycling through the plasma space. Attention has been drawn to limitations of [3H]NE kinetics....

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

International Nuclear Information System (INIS)

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

1993-11-01

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

Overview of tritium processing development at the tritium systems test assembly

International Nuclear Information System (INIS)

Anderson, J.L.

1986-01-01

The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been operating with tritium since June 1984. Presently there are some 50 g of tritium in the main processing loop. This 50 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation system and to integrate the fuel cleanup system into the main fuel processing loop. In January 1986 two major experiments were conducted. During these experiments the fuel cleanup system was integrated, through the transfer pumping system, with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems leaves only the main vacuum system to be integrated into the TSTA fuel processing loop. In September 1986 another major tritium experiment was performed in which the integrated loop was operated, the tritium inventory increased to 50 g and additional measurements on the performance of the distillation system were taken. In the period June 1984 through September 1986 the TSTA system has processed well over 10 8 Ci of tritium. Total tritium emissions to the environment over this period have been less than 15 Ci. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flows have been uncovered

Tritium permeation model for plasma facing components

Science.gov (United States)

Longhurst, G. R.

1992-12-01

This report documents the development of a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. The model is developed for solution using commercial spread-sheet software such as Lotus 123. Comparison calculations are provided with the verified and validated TMAP4 transient code with good agreement. Results of calculations for the ITER CDA diverter are also included.

Tritium permeation model for plasma facing components

International Nuclear Information System (INIS)

Longhurst, G.R.

1992-12-01

This report documents the development of a simplified one-dimensional tritium permeation and retention model. The model makes use of the same physical mechanisms as more sophisticated, time-transient codes such as implantation, recombination, diffusion, trapping and thermal gradient effects. It takes advantage of a number of simplifications and approximations to solve the steady-state problem and then provides interpolating functions to make estimates of intermediate states based on the steady-state solution. The model is developed for solution using commercial spread-sheet software such as Lotus 123. Comparison calculations are provided with the verified and validated TMAP4 transient code with good agreement. Results of calculations for the ITER CDA diverter are also included

Reprocessing of spent plasma

International Nuclear Information System (INIS)

Pierini, G.

1981-01-01

This invention relates to a process for removing helium and other impurities from a mixture containing deuterium and tritium, a deuterium/tritium mixture when purified in accordance with such a process and, more particularly, to a process for the reprocessing of spent plasma removed from a thermofusion reactor. (U.K.)

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

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

International Nuclear Information System (INIS)

Brown, D.H.

1998-01-01

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

Tritium labeling of amino acids and peptides with liquid and solid tritium

International Nuclear Information System (INIS)

Peng, C.T.; Hua, R.L.; Souers, P.C.; Coronado, P.R.

1988-01-01

Amino acids and peptides were labeled with liquid and solid tritium at 21 K and 9 K. At these low temperatures radiation degradation is minimal, and tritium incorporation increases with tritium concentration and exposure time. Ring saturation in L-phenyl-alanine does not occur. Peptide linkage in oligopeptides is stable toward tritium. Deiodination in 3-iodotyrosine and 3,5-diiodotyrosine occurs readily and proceeds in steps by losing one iodine atom at a time. Nickel and noble metal supported catalysts when used as supports for dispersion of the substrate promote tritium labeling at 21 K. Our study shows that both liquid and solid tritium are potentially useful agents for labeling peptides and proteins. 11 refs., 1 fig., 3 tabs

Tritium labeling of amino acids and peptides with liquid and solid tritium

International Nuclear Information System (INIS)

Souers, P.C.; Coronado, P.R.; Peng, C.T.; Hua, R.L.

1988-01-01

Amino acids and peptides were labeled with liquid and solid tritium at 21/degree/K and 9/degree/K. At these low temperatures radiation degradation is minimal, and tritium incorporation increases with tritium concentration and exposure time. Ring saturation in L-phenylalanine does not occur. Peptide linkage in oligopeptides is stable toward tritium. Deiodination in 3-iodotyrosine and 3,5-diiodotyrosine occurs readily and proceeds in steps by losing one iodine atom at a time. Nickel and noble metal supported catalysts when used as supports for dispersion of the substrate promote tritium labeling at 21 K. Our study shows that both liquid and solid tritiums are potentially useful agents for labeling peptides and proteins

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

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

A sensitive radioimmunoassay of atrial natriuretic peptide in human plasma, using a tracer with an immobilized glycouril agent

International Nuclear Information System (INIS)

Rosmalen, F.M.A.; Tan, A.C.I.T.L.; Benraad, T.J.

1987-01-01

A highly specific and sensitive radioimmunoassay (RIA) for alpha-human atrial natriuretic peptide (hANP[1-28]) in plasma was developed. The assay used a [ 125 I]monoiodotyrosyl-hANP[1-28] tracer, prepared with an immobilized glycouril agent (Protag) and purified by high pressure liquid chromatography (HPLC), and a highly specific antiserum raised against hANP[1-28], coupled to keyhole limpet haemocyanin, in sheep. Plasma was extracted using C-18 Seppak cartridges. A good parallelism was found after dilution prior to extraction of plasma of patients with congestive heart failure (CHF) or of plasma of healthy subjects. Recovery of hANP[1-28] added to plasma was 96%. The limit of detection was 0.8 pg/tube, intra- and inter-assay variation were 9 and 12%, respectively. Mean plasma ANP values in 25 normal persons with a normal salt intake was 26.0 ± 15.5 (± SD) pg/ml. Plasma levels of 18 subjects (7 normals, 11 CHF) were measured using four different antisera after the extraction step. High correlations were found between the values obtained with these four antisera. (Auth.)

Data Needs for Erosion and Tritium Retention in Beryllium Surfaces

International Nuclear Information System (INIS)

Braams, B.J.

2011-07-01

A Consultants' Meeting was held at IAEA Headquarters 30-31 May 2011 with the aim to provide advice about the scope and aims of a planned IAEA coordinated research project on erosion and tritium retention in beryllium plasma-facing materials and about other activities of the A+M Data Unit in the area of plasma interaction with beryllium surfaces. The present report contains the proceedings, recommendations and conclusions of that Consultants' Meeting. (author)

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

Several experiments on the movement of tritium in a tropical ecosystem have been conducted in the montane rainforest of Eastern Puerto Rico by the Bio-Medical Division of the Lawrence Radiation Laboratory, Livermore, in cooperation with the Puerto Rico Nuclear Center. Tritiated whaler was used as a tracer for water movement in: a) mature evergreen trees of the climax rainforest; b) soil and substory vegetation and c) rapidly growling successional species. A feasibility study on the Atlantic Pacific Interoceanic Canal is currently being conducted. If thermonuclear explosives were used in constructing the canal, tritium would be deposited as tritiated water and distributed among the several biological compartments of the tropical ecosystem in that area. The main hydrogen compartments are water in the soil and in leaves, limbs and wood of forest trees. Organic tissue hydrogen comprises another compartment. In the tree experiment, tritiated water was injected directly into several species of mature, broad leaved evergreen tropical trees. Transpiration and residence time for tritium was determined from analyses of leaves sampled during a several month period. Transpiration ranged from 4 ml/day/gm dry leaf for an understory Dacryodes excelsa to 10.0 and 13.8 ml/day/gm dry leaf for a mature Sloanea berteriana and D. excelsa, respectively. Mean residence time for the S. berteriana was 3.9 {+-} 0.2 days and the understory and mature D. excelsa values were 9.5 {+-} 0.4 and 11.0 {+-} 0. 6 days, respectively. In another experiment, tritiated water was sprinkled over a 3.68 m{sup 2} plot and its movement down into the soil and up into the vegetation growing on the plot was traced. The pattern of water movement in the soil was clearly demonstrated. The mean residence time for tritium in the soil and in trees was found to be 42 {+-} 2 days and 67 {+-} 9 days, respectively. The residence time for tritium in the trees in this experiment was considerably longer than for the single

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

Several experiments on the movement of tritium in a tropical ecosystem have been conducted in the montane rainforest of Eastern Puerto Rico by the Bio-Medical Division of the Lawrence Radiation Laboratory, Livermore, in cooperation with the Puerto Rico Nuclear Center. Tritiated whaler was used as a tracer for water movement in: a) mature evergreen trees of the climax rainforest; b) soil and substory vegetation and c) rapidly growling successional species. A feasibility study on the Atlantic Pacific Interoceanic Canal is currently being conducted. If thermonuclear explosives were used in constructing the canal, tritium would be deposited as tritiated water and distributed among the several biological compartments of the tropical ecosystem in that area. The main hydrogen compartments are water in the soil and in leaves, limbs and wood of forest trees. Organic tissue hydrogen comprises another compartment. In the tree experiment, tritiated water was injected directly into several species of mature, broad leaved evergreen tropical trees. Transpiration and residence time for tritium was determined from analyses of leaves sampled during a several month period. Transpiration ranged from 4 ml/day/gm dry leaf for an understory Dacryodes excelsa to 10.0 and 13.8 ml/day/gm dry leaf for a mature Sloanea berteriana and D. excelsa, respectively. Mean residence time for the S. berteriana was 3.9 ± 0.2 days and the understory and mature D. excelsa values were 9.5 ± 0.4 and 11.0 ± 0. 6 days, respectively. In another experiment, tritiated water was sprinkled over a 3.68 m 2 plot and its movement down into the soil and up into the vegetation growing on the plot was traced. The pattern of water movement in the soil was clearly demonstrated. The mean residence time for tritium in the soil and in trees was found to be 42 ± 2 days and 67 ± 9 days, respectively. The residence time for tritium in the trees in this experiment was considerably longer than for the single injected input

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

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

Estimation of the contribution of gaps to tritium retention in the divertor of ITER

Czech Academy of Sciences Publication Activity Database

Matveev, D.; Kirschner, A.; Schmid, K.; Litnovsky, A.; Borodin, D.; Komm, Michael; Van Oost, G.; Samm, U.

-, T159 (2014), 014063-014063 ISSN 0031-8949 Institutional support: RVO:61389021 Keywords : plasma * tokamak * tritium retention * ITER * castellated surfaces * gaps * divertor * impurity deposition Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.126, year: 2014 http://iopscience.iop.org/1402-4896/2014/T159/014063/

Suppression of Tritium Retention in Remote Areas of ITER by Nonperturbative Reactive Gas Injection

NARCIS (Netherlands)

Tabares, F. L.; Ferreira, J. A.; Ramos, A.; van Rooij, G. J.; Westerhout, J.; Al, R.; Rapp, J.; Drenik, A.; Mozetic, M.

2010-01-01

A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4 nm/min

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)

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)

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)

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

Mechanism of Wandoan coal liquefaction by the use of tritium and 14C tracer method

International Nuclear Information System (INIS)

Kabe, Toshiaki; Nitoh, Osamu; Kawakami, Akira; Marumoto, Motoi; Nakagawa, Kouhei

1986-01-01

In order to make the behavior of hydrogen donor solvent clear, Wandoan coal was liquefied in tritium labeled tetralin solvent contained a small amount of 14 C labeled naphthalene, under initial H 2 pressure : 5.9 MPa, reaction temperature range : 400-440 deg C and with or without Ni-Mo-Al 2 O 3 catalyst. The concentration of 14 C in tetralin indicated that the hydrogenation of naphthalene to tetralin occurred. From tritium and hydrogen distributions in coal products, solvents and molecular hydrogen, the amounts of hydrogen which transferred by hydrogen addition and exchange reactions were estimated, and the effects of the catalyst and reaction temperature were examined. Without catalyst, the coal liquefaction proceeded mainly by the hydrogen addition from hydrogen donor solvent to coal and the hydrogen addition from molecular hydrogen to coal products hardly occurred. The catalyst was effective in the hydrocracking of preasphaltenes, but did not promote the hydrocracking of oil. Furthermore, the catalyst promoted the hydrogen addition from molecular hydrogen to coal products and solvents, and activated the hydrogen exchange between molecular hydrogen and solvents, but the hydrogen exchanges did not reach to equilibrium under the condition of 440 deg C. (author)

Investigating sources and pathways of perfluoroalkyl acids (PFAAs) in aquifers in Tokyo using multiple tracers

International Nuclear Information System (INIS)

Kuroda, Keisuke; Murakami, Michio; Oguma, Kumiko; Takada, Hideshige; Takizawa, Satoshi

2014-01-01

We employed a multi-tracer approach to investigate sources and pathways of perfluoroalkyl acids (PFAAs) in urban groundwater, based on 53 groundwater samples taken from confined aquifers and unconfined aquifers in Tokyo. While the median concentrations of groundwater PFAAs were several ng/L, the maximum concentrations of perfluorooctane sulfonate (PFOS, 990 ng/L), perfluorooctanoate (PFOA, 1800 ng/L) and perfluorononanoate (PFNA, 620 ng/L) in groundwater were several times higher than those of wastewater and street runoff reported in the literature. PFAAs were more frequently detected than sewage tracers (carbamazepine and crotamiton), presumably owing to the higher persistence of PFAAs, the multiple sources of PFAAs beyond sewage (e.g., surface runoff, point sources) and the formation of PFAAs from their precursors. Use of multiple methods of source apportionment including principal component analysis–multiple linear regression (PCA–MLR) and perfluoroalkyl carboxylic acid ratio analysis highlighted sewage and point sources as the primary sources of PFAAs in the most severely polluted groundwater samples, with street runoff being a minor source (44.6% sewage, 45.7% point sources and 9.7% street runoff, by PCA–MLR). Tritium analysis indicated that, while young groundwater (recharged during or after the 1970s, when PFAAs were already in commercial use) in shallow aquifers (< 50 m depth) was naturally highly vulnerable to PFAA pollution, PFAAs were also found in old groundwater (recharged before the 1950s, when PFAAs were not in use) in deep aquifers (50–500 m depth). This study demonstrated the utility of multiple uses of tracers (pharmaceuticals and personal care products; PPCPs, tritium) and source apportionment methods in investigating sources and pathways of PFAAs in multiple aquifer systems. - Highlights: • Aquifers in Tokyo had high levels of perfluoroalkyl acids (up to 1800 ng/L). • PFAAs were more frequently detected than sewage-tracer

Investigating sources and pathways of perfluoroalkyl acids (PFAAs) in aquifers in Tokyo using multiple tracers

Energy Technology Data Exchange (ETDEWEB)

Kuroda, Keisuke, E-mail: keisukekr@gmail.com [Department of Urban Engineering, Graduate School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-8656 (Japan); Murakami, Michio [Institute of Industrial Science, The University of Tokyo, 4-6-1 Komaba, Meguro, Tokyo 153-8505 (Japan); Oguma, Kumiko [Department of Urban Engineering, Graduate School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-8656 (Japan); Takada, Hideshige [Laboratory of Organic Geochemistry (LOG), Institute of Symbiotic Science and Technology, Tokyo University of Agriculture and Technology, Fuchu, Tokyo 183-8509 (Japan); Takizawa, Satoshi [Department of Urban Engineering, Graduate School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-8656 (Japan)

2014-08-01

We employed a multi-tracer approach to investigate sources and pathways of perfluoroalkyl acids (PFAAs) in urban groundwater, based on 53 groundwater samples taken from confined aquifers and unconfined aquifers in Tokyo. While the median concentrations of groundwater PFAAs were several ng/L, the maximum concentrations of perfluorooctane sulfonate (PFOS, 990 ng/L), perfluorooctanoate (PFOA, 1800 ng/L) and perfluorononanoate (PFNA, 620 ng/L) in groundwater were several times higher than those of wastewater and street runoff reported in the literature. PFAAs were more frequently detected than sewage tracers (carbamazepine and crotamiton), presumably owing to the higher persistence of PFAAs, the multiple sources of PFAAs beyond sewage (e.g., surface runoff, point sources) and the formation of PFAAs from their precursors. Use of multiple methods of source apportionment including principal component analysis–multiple linear regression (PCA–MLR) and perfluoroalkyl carboxylic acid ratio analysis highlighted sewage and point sources as the primary sources of PFAAs in the most severely polluted groundwater samples, with street runoff being a minor source (44.6% sewage, 45.7% point sources and 9.7% street runoff, by PCA–MLR). Tritium analysis indicated that, while young groundwater (recharged during or after the 1970s, when PFAAs were already in commercial use) in shallow aquifers (< 50 m depth) was naturally highly vulnerable to PFAA pollution, PFAAs were also found in old groundwater (recharged before the 1950s, when PFAAs were not in use) in deep aquifers (50–500 m depth). This study demonstrated the utility of multiple uses of tracers (pharmaceuticals and personal care products; PPCPs, tritium) and source apportionment methods in investigating sources and pathways of PFAAs in multiple aquifer systems. - Highlights: • Aquifers in Tokyo had high levels of perfluoroalkyl acids (up to 1800 ng/L). • PFAAs were more frequently detected than sewage-tracer

Tritium effects on germ cells and fertility

International Nuclear Information System (INIS)

Dobson, R.L.; Kwan, T.C.; Straume, T.

1982-01-01

Primordial oocytes in juvenile mice show acute gamma-ray LD 50 as low as 6 rad. This provides opportunities for determining dose-response relations at low doses and chronic exposure in the intact animal - conditions of particular interest for hazard evaluation. Examined in this way, 3 HOH in body water is found to kill murine oocytes exponentially with dose, the LD 50 level for chronic exposure being only 2μCi/ml (delivering 0.4 rad/day). At very low doses and dose rates, where comparisons between tritium and other radiations are of special significance for radiological protection, the RBE of tritium compared with 60 Co gamma radiation reaches approximately 3. Effects on murine fertility from tritium-induced oocyte loss have been quantified by reproductive capacity measurements. Chronic low-level exposure has been examined also in three primate species - squirrel, rhesus, and bonnet monkeys. In squirrel monkeys the ovarian germ-cell supply is 99% destroyed by the time of birth from prenatal exposure to body-water levels of 3 HOH (administered in maternal drinking water) of only 3 μCi/ml, the LD 50 level being 0.5 μCi/ml (giving 0.1 rad/day), one fourth that in mice. Though not completely ruled out, similar high sensitivity of female germ cells has not been found in macaques; and it probably does not occur in man. The exquisite radiosensitivity of primordial oocytes in mice is apparently due to vulnerability of the plasma membrane (or something of similar geometry and location), not DNA. Evidence for this comes from tritium data as well as neutron studies. Tritium administered as 3 HOH, and therefore generally distributed, is much more effective in killing murine oocytes than is tritium administered as 3 H-TdR, localized in the nucleus. This situation in the mouse may have implications for estimating radiation genetic risk in the human female

Applying tracer techniques to NPP liquid effluents for estimating the maximum concentration of soluble pollutants in a man-made canal

International Nuclear Information System (INIS)

Varlam, Carmen; Stefanescu, Ioan; Varlam, Mihai; Raceanu, Mircea; Enache, Adrian; Faurescu, Ionut; Patrascu, Vasile; Bucur, Cristina

2006-01-01

Full text: The possibility of a contamination agent being accidentally or intentionally spilled upstream from a water supply is a constant concern to those diverting and using water from a channel. A method of rapidly estimating the travel-time or dispersion is needed for pollution control or warning system on channels where data are scarce. Travel-time and mixing of water within a stream are basic streamflow characteristics needed in order to predict the rate of movement and dilution of pollutants that could be introduced in the stream. In this study we propose using tritiated liquid effluents from CANDU type nuclear power plant as a tracer, to study hydrodynamics on Danube-Black Sea Canal. This canal is ideal for this kind of study, because wastewater evacuations occur occasionally due to technical operations of nuclear power plant. Tritiated water can be used to simulate the transport and dispersion of solutes in Danube-Black Sea Canal because they have the same physical characteristics as the water. Measured tracer-response curves produced from injection of a known amount of soluble tracer provide an efficient method of obtaining the necessary data. This method can estimate: (1) the rate of movement of a solute through the canal reach: (2) the rate of peak attenuation concentration of a conservative solute in time; and (3) the length of time required for the solute plume to pass a point in the canal. This paper presents the mixing length calculation for particular conditions (lateral branch of the canal, and lateral injection of wastewater from the nuclear power plant). A study of published experimentally-obtained formulas was used to determine proper mixing length. Simultaneous measurements in different locations of the canal confirm the beginning of the experiment. Another result used in a further experiment concerns the tritium level along the Danube-Black Sea Canal. We measured tritium activity concentration in water sampled along the Canal between July

A novel method for trace tritium transport studies

Czech Academy of Sciences Publication Activity Database

Bonheure, G.; Mlynář, Jan; Murari, A.; Giroud, C.; Belo, P.; Bertalot, L.; Popovichev, S.; JET EFDA, Contributors.

2009-01-01

Roč. 49, č. 8 (2009), 085025 ISSN 0029-5515 R&D Projects: GA MŠk LA08048 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * transport * tritium * tomography Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 4.270, year: 2009 http://www.iop.org/EJ/article/0029-5515/49/8/085025/nf9_8_085025.pdf

A low inventory adsorptive process for tritium extraction and purification

International Nuclear Information System (INIS)

Keefer, B.; Bora, B.; Chew, M.; Rump, M.; Kveton, O.K.

1990-08-01

The fuel cycles of future fusion power systems present a diverse spectrum of challenges to gas separation technology, for extraction, concentration, purification and confinement of tritium in fusion fuel cycles. Economic and safety factors motivate process design for minimum tritium inventory, functional simplicity, and overall reliability. A new gas separation process with some features of interest to fusion has been demonstrated under the auspices of the Canadian Fusion Fuels Technology Project. This process (Thermally Coupled Pressure Swing Adsorption or 'TCPSA') is potentially applicable to several fusion applications for separation purification of hydrogen, notably for tritium extraction from breeder blanket purge helium. Recent experimental tests have been directed toward fusion applications, primarily extraction and concentration of tritium-rich hydrogen from the blanket purge helium stream, and also considering purification of this and other hydrogen isotope streams such as the plasma exhaust. For example, hydrogen at 0.1% concentration in helium has been extracted in a TCPSA module operating at 195 K, with the process performed in a single working space to achieve simultaneous high extraction and concentration of the hydrogen. With methane or carbon oxides as the impurities, substantially complete separation is achieved by the same apparatus at ambient temperature. Engineering projections for scale-up to ITER blanket purge extraction and purification applications indicate a low working inventory of tritium

Development of tritium permeation barriers on Al base in Europe

Science.gov (United States)

Benamati, G.; Chabrol, C.; Perujo, A.; Rigal, E.; Glasbrenner, H.

The development of the water cooled lithium lead (WCLL) DEMO fusion reactor requires the production of a material capable of acting as a tritium permeation barrier (TPB). In the DEMO blanket reactor permeation barriers on the structural material are required to reduce the tritium permeation from the Pb-17Li or the plasma into the cooling water to acceptable levels (HIP) technology and spray (this one developed also for repair) deposition techniques. The final goal is to select a reference technique to be used in the blanket of the DEMO reactor and in the ITER test module fabrication. The activities performed in four European laboratories are summarised here.

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

Science.gov (United States)

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

2016-01-01

Tritium is the radioactive isotope of hydrogen. It can be integrated into most biological molecules. Even though its radiotoxicity is weak, the effects of tritium can be increased following concentration in critical compartments of living organisms. For a better understanding of tritium circulation in the environment and to highlight transfer constants between compartments, we studied the tritiation of different agricultural matrices chronically exposed to tritium. Milk is one of the most frequently monitored foodstuffs in the vicinity of points known for chronic release of radionuclides firstly because dairy products find their way into most homes but also because it integrates deposition over large areas at a local scale. It is a food which contains all the main nutrients, especially proteins, carbohydrates and lipids. We thus studied the tritium levels of milk in chronic exposure conditions by comparing the tritiation of the main hydrogenated components of milk, first, component by component, then, sample by sample. Significant correlations were found between the specific activities of drinking water and free water of milk as well as between the tritium levels of cattle feed dry matter and of the main organic components of milk. Our findings stress the importance of the metabolism on the distribution of tritium in the different compartments. Overall, dilution of hydrogen in the environmental compartments was found to play an important role dimming possible isotopic effects even in a food chain chronically exposed to tritium. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

Role of soil-to-leaf tritium transfer in controlling leaf tritium dynamics: Comparison of experimental garden and tritium-transfer model results.

Science.gov (United States)

Ota, Masakazu; Kwamena, Nana-Owusua A; Mihok, Steve; Korolevych, Volodymyr

2017-11-01

Environmental transfer models assume that organically-bound tritium (OBT) is formed directly from tissue-free water tritium (TFWT) in environmental compartments. Nevertheless, studies in the literature have shown that measured OBT/HTO ratios in environmental samples are variable and generally higher than expected. The importance of soil-to-leaf HTO transfer pathway in controlling the leaf tritium dynamics is not well understood. A model inter-comparison of two tritium transfer models (CTEM-CLASS-TT and SOLVEG-II) was carried out with measured environmental samples from an experimental garden plot set up next to a tritium-processing facility. The garden plot received one of three different irrigation treatments - no external irrigation, irrigation with low tritium water and irrigation with high tritium water. The contrast between the results obtained with the different irrigation treatments provided insights into the impact of soil-to-leaf HTO transfer on the leaf tritium dynamics. Concentrations of TFWT and OBT in the garden plots that were not irrigated or irrigated with low tritium water were variable, responding to the arrival of the HTO-plume from the tritium-processing facility. In contrast, for the plants irrigated with high tritium water, the TFWT concentration remained elevated during the entire experimental period due to a continuous source of high HTO in the soil. Calculated concentrations of OBT in the leaves showed an initial increase followed by quasi-equilibration with the TFWT concentration. In this quasi-equilibrium state, concentrations of OBT remained elevated and unchanged despite the arrivals of the plume. These results from the model inter-comparison demonstrate that soil-to-leaf HTO transfer significantly affects tritium dynamics in leaves and thereby OBT/HTO ratio in the leaf regardless of the atmospheric HTO concentration, only if there is elevated HTO concentrations in the soil. The results of this work indicate that assessment models

Borehole environmental tracers for evaluating net infiltration and recharge through desert bedrock

Science.gov (United States)

Heilweil, V.M.; Solomon, D.K.; Gardner, P.M.

2006-01-01

Permeable bedrock aquifers in arid regions are being increasingly developed as water supplies, yet little is generally known about recharge processes and spatial and temporal variability. Environmental tracers from boreholes were used in this study to investigate net infiltration and recharge to the fractured Navajo Sandstone aquifer. Vadose zone tracer profiles at the Sand Hollow study site in southwestern Utah look similar to those of desert soils at other sites, indicating the predominance of matrix flow. However, recharge rates are generally higher in the Navajo Sandstone than in unconsolidated soils in similar climates because the sandstone matrix allows water movement but not root penetration. Water enters the vadose zone either as direct infiltration of precipitation through exposed sandstone and sandy soils or as focused infiltration of runoff. Net infiltration and recharge exhibit extreme spatial variability. High-recharge borehole sites generally have large amounts of vadose zone tritium, low chloride concentrations, and small vadose zone oxygen-18 evaporative shifts. Annual net-infiltration and recharge rates at different locations range from about 1 to 60 mm as determined using vadose zone tritium, 0 to 15 mm using vadose zone chloride, and 3 to 60 mm using groundwater chloride. Environmental tracers indicate a cyclical net-infiltration and recharge pattern, with higher rates earlier in the Holocene and lower rates during the late Holocene, and a return to higher rates during recent decades associated with anomalously high precipitation during the latter part of the 20th century. The slightly enriched stable isotopic composition of modern groundwater indicates this recent increase in precipitation may be caused by a stronger summer monsoon or winter southern Pacific El Nin??o storm track. ?? Soil Science Society of America.

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

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)

Tracers confirm downward mixing of Tyrrhenian Sea upper waters associated with the Eastern Mediterranean Transient

Directory of Open Access Journals (Sweden)

W. Roether

2011-01-01

Full Text Available Observations of tritium and ³He in the Tyrrhenian Sea, 1987–2009, confirm the enhanced vertical mixing of intermediate waters into the deep waters that has been noted and associated with the Eastern Mediterranean Transient in previous studies. Our evidence for the mixing rests on increasing tracer concentrations in the Tyrrhenian deep waters, accompanied by decreases in the upper waters, which are supplied from the Eastern Mediterranean. The downward transfer is particularly evident between 1987 and 1997. Later on, information partly rests on increasing tritium-³He ages; here we correct the observed ³He for contributions released from the ocean floor. The Tyrrhenian tracer distributions are fully compatible with data upstream of the Sicily Strait and in the Western Mediterranean. The tracer data show that mixing reached to the bottom and confirm a cyclonic nature of the deep water circulation in the Tyrrhenian. They furthermore indicate that horizontal homogenization of the deep waters occurs on a time scale of roughly 5 years. Various features point to a reduced impact of Western Mediterranean Deep Water (WMDW in the Tyrrhenian during the enhanced-mixing period. This is an important finding because it implies less upward mixing of WMDW, which has been named a major process to enable the WMDW to leave the Mediterranean via the Gibraltar Strait. On the other hand, the TDW outflow for several years represented a major influx of enhanced salinity and density waters into the deep-water range of the Western Mediterranean.

Impurity transport studies by means of tracer-encapsulated solid pellet injection in neutral beam heated plasmas on LHD

International Nuclear Information System (INIS)

Tamura, N; Sudo, S; Khlopenkov, K V; Kato, S; Sergeev, V Yu; Muto, S; Sato, K; Funaba, H; Tanaka, K; Tokuzawa, T; Yamada, I; Narihara, K; Nakamura, Y; Kawahata, K; Ohyabu, N; Motojima, O

2003-01-01

The quantitative properties of impurity transport in large helical device (LHD) plasmas heated by neutral beam injection have been investigated by means of tracer-encapsulated solid pellet (TESPEL) injection. In the case of a titanium (Ti) tracer, the behaviour of the emission lines from the highly ionized Ti impurity, Ti Kα(E He-like ∼ 4.7 keV) and Ti XIX (λ = 16.959 nm), has been observed clearly by a soft x-ray pulse height analyzer and a vacuum ultraviolet spectrometer, respectively. A fairly longer decay time of the Ti Kα emission lines is obtained above the value of a line-averaged electron density, 3.0x10 19 m -3 . The dependence of the behaviour of the Ti tracer impurity on the line-averaged electron density below the value of that, 3.5x10 19 m -3 is in qualitative agreement with the characteristics obtained from the observation of the behaviour of an intrinsic metallic impurity in neutral beam heated plasmas on LHD. In order to estimate the properties of the Ti impurity transport quantitatively, the one-dimensional impurity transport code, MIST has been used. As a result of the transport analysis with the MIST code, even an small inward convection should be necessary to account for the experimental results with the value of the line-averaged electron density, 3.5x10 19 m -3 . In order to examine the experimentally obtained transport coefficients, neoclassical analysis with respect to the radial impurity flux has been performed. The inferred rise of the inward convection cannot be explained solely by neoclassical impurity transport. Therefore, in order to account for the inward convection, the effect of a radial electric field and/or some other effect must be taken into account additionally

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)

Tritium contamination of concrete walls and floors in tritium-handling laboratory

International Nuclear Information System (INIS)

Kawano, T.; Kuroyanagi, M.; Tabei, T.

2006-01-01

A tritium handling laboratory was constructed at the National Institute for Fusion Science about twenty years ago and it was recently closed down. We completed the necessary work that is legally required in Japan at the laboratory, when the use of radioisotopes is discontinued, involving measurements of radioactive contamination. We mainly used smear and direct-immersion methods for the measurements. In applying the smear method, we used a piece of filter paper to wipe up the tritium staining the surfaces. The filter paper containing the tritium was placed directly into a dedicated vial, a scintillation cocktail was then poured over it, and the tritium was measured with a liquid scintillation counter. With the direct-immersion method, a piece of concrete was placed directly into a vial containing a scintillation cocktail, and the tritium in the concrete was measured with a liquid scintillation counter. As well as these measurements, we investigated water-extraction and heating-cooling methods for measuring tritium contamination in concrete. With the former, a piece of concrete was placed into water in a tube to extract the tritium, the water containing the extracted tritium was then poured into a dedicated vial containing a scintillation cocktail, and the tritium contamination was measured. With the latter, a piece of concrete was placed into a furnace and heated to 800 degrees centigrade to vaporize the tritiated water into flowing dry air. The flowing air was then cooled to collect the vaporized tritiated water in a tube. The collected water was placed in a vial for scintillation counting. To evaluate the direct-immersion method, ratios were determined by dividing the contamination measured with the heating-cooling method by that measured with the direct-immersion method. The average ratio was about 2.5, meaning a conversion factor from contamination obtained with the direct-immersion method to that with the heating-cooling method. We also investigated the

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

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

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

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

Tritium research and technology facilities at the JRC-Ispra

International Nuclear Information System (INIS)

Dworschak, H.; Mannone, F.; Perujo, A.; Pierini, G.; Reiter, F.; Vassallo, G.; Viola, A.; Camposilvan, J.; Douglas, K.; Grassi, G.; Lolli Ceroni, P.; Simonetta, A.; Spelta, B.

1990-01-01

A set of experiments which are of prominent interest for the development of nuclear fusion technology in Europe are planned by the JRC-Ispra for the near future, in the frame of experimental activities to be performed in ETHEL, the European Tritium Handling Experimental Laboratory under construction at the Ispra site. These experiments already included for the most part as JRC-Task Action Sheets in the 1989-1991 European Technology Programme Actions will initiate in ETHEL on a fully active laboratory scale starting mid-1991. They will concern the following research areas: Recycling of tritium from first wall materials; Tritium recovery from water cooled Pb-17Li blankets; Detritiation of ventilation atmospheres; Plasma exhaust processing; Tritiazed waste management. In view of fully active tritium experiments in ETHEL and to obtain information of the basic processes involved, since 1985 preparatory experimental studies are being performed at the JRC-Ispra laboratories using hydrogen and deuterium. Furthermore, always with regard to ETHEL experiments, particular attention is given to possible technical and managerial problems which potentially may arise in this context. To identify at an early stage such problems a questionnaire has been developed and distributed to researchers in conjunction with an ETHEL information packet. The questionnaire demands information regarding the scope, design and operation of the intended experiment as well as planning and required support to be supplied by ETHEL. A brief description of experimental preparatory studies and future tritium handling experiments in ETHEL as well of the ETHEL facility is here presented. (orig.)

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.

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

Tritium control in fusion reactor materials: A model for Tritium Extracting System

International Nuclear Information System (INIS)

Zucchetti, Massimo; Utili, Marco; Nicolotti, Iuri; Ying, Alice; Franza, Fabrizio; Abdou, Mohamed

2015-01-01

Highlights: • A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a Molecular sieve as adsorbent material. • A computational model has been setup and tested in this paper. • The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. • It turns out the capability to model the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT). - Abstract: In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT).

Tritium control in fusion reactor materials: A model for Tritium Extracting System

Energy Technology Data Exchange (ETDEWEB)

Zucchetti, Massimo [DENERG, Politecnico di Torino (Italy); Utili, Marco, E-mail: marco.utili@enea.it [ENEA UTIS – C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino (Italy); Ying, Alice [University of California Los Angeles (UCLA), Los Angeles, CA (United States); Franza, Fabrizio [Karlsruhe Institute of Technology, Karlsruhe (Germany); Abdou, Mohamed [University of California Los Angeles (UCLA), Los Angeles, CA (United States)

2015-10-15

Highlights: • A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a Molecular sieve as adsorbent material. • A computational model has been setup and tested in this paper. • The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. • It turns out the capability to model the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT). - Abstract: In fusion reactors, tritium is bred by lithium isotopes inside the blanket and then extracted. However, tritium can contaminate the reactor structures, and can be eventually released into the environment. Tritium in reactor components should therefore be kept under close control throughout the fusion reactor lifetime, bearing in mind the risk of accidents, the need for maintenance and the detritiation of dismantled reactor components before their re-use or disposal. A modeling work has been performed to address these issues in view of its utilization for the TES (Tritium Extraction System), in the case of the HCPB TBM and for a molecular sieve as adsorbent material. A computational model has been setup and tested. The results of experimental measurement of fundamental parameters such as mass transfer coefficients have been implemented in the model. It turns out the capability of the model to describe the extraction process of gaseous tritium compounds and to estimate the breakthrough curves of the two main tritium gaseous species (H2 and HT).

The JET gas baking plant for DT operation and analysis of tritium permeation and baking gas activation in DTE1

Energy Technology Data Exchange (ETDEWEB)

Pearce, R.J.H.; Andrew, P.; Bryan, S.; Hemmrich, J.L. [JET Joint Undertaking, Abingdon, Oxon (United Kingdom)

1998-07-01

The JET gas baking plant allows the vacuum vessel to be heated for conditioning and plasma operations. The vessel was maintained at 320 deg. C for the JET DT experiments (DTE 1). The design of the plant is outlined with particular reference to the features to provide compatibility with tritium operations. The experience of baking gas activation and tritium permeation into the plant are given, Developmentsto reduce the tritium permeation out of the vessel are considered. (authors)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Study of transport in unsaturated sands using radioactive tracers

International Nuclear Information System (INIS)

Merritt, W.F.; Pickens, J.F.; Allison, G.B.

1979-01-01

A laboratory experiment was conducted to investigate the mixing that occurs as a series of labelled pulses of water are transported by gravity drainage down through a sand filled column having a water table imposed at the bottom. It also demonstrated the utility of gamma-ray emitting radioactive tracers in studying transport in unsaturated or saturated porous media. The motivation for pursuing this topic was developed from observing that the content of oxygen-18, deuterium and tritium in rainwater shows marked temporal variations whereas their concentrations below the water table in shallow ground water flow systems are generally found to show much less variation. (auth)

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

Tritium in precipitation of Vostok (Antarctica): conclusions on the tritium latitude effect.

Science.gov (United States)

Hebert, Detlef

2011-09-01

During the Antarctic summer of 1985 near the Soviet Antarctic station Vostok, firn samples for tritium measurements were obtained down to a depth of 2.40 m. The results of the tritium measurements are presented and discussed. Based on this and other data, conclusions regarding the tritium latitude effect are derived.

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

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

Tritium inventory in the ITER PFC`s: predictions, uncertainties, R and D status and priority needs

Energy Technology Data Exchange (ETDEWEB)

Federici, G. [ITER, Garching (Germany). JWS; Anderl, R.; Longhurst, G. [Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho 83415 (United States); Brooks, J.N. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, Illinois 60439 (United States); Causey, R.; Cowgill, D.; Wampler, W.; Wilson, K.; Youchison, D. [Sandia National Laboratories, Livermore California and Albuquerque, New Mexico (United States); Coad, J.P.; Peacock, A.; Pick, M. [JET Joint Undertaking, Abingdon, Oxfordshire OX14 3EA (United Kingdom); Doerner, R.; Luckhardt, S. [University of California San Diego, La Jolla, California 92093-0417 (United States); Haasz, A.A. [University of Toronto, Institute for Aerospace Studies, Ontario M3H 5T6 (Canada); Mueller, D.; Skinner, C.H. [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Wong, C. [General Atomics, PO Box 85608, San Diego, California 92186-9784 (United States); Wu, C. [NET Team, Boltzmannstrasse 2, 85748 Garching (Germany)

1998-09-01

New data on hydrogen plasma isotopes retention in beryllium and tungsten are now becoming available from various laboratories for conditions similar to those expected in the International Thermonuclear Experimental Reactor (ITER) where previous data were either missing or largely scattered. Together with a significant advancement in understanding, they have warranted a revisitation of the previous estimates of tritium inventory in ITER, with beryllium as the plasma facing material for the first-wall components, and tungsten in the divertor with some carbon-fibre-composites clad areas, near the strike points. Based on these analyses, it is shown that the area of primary concern with, respect to tritium inventory, remains codeposition with carbon and possibly beryllium on the divertor surfaces. Here, modelling of ITER divertor conditions continues to show potentially large codeposition rates which are confirmed by tokamak findings. Contrary to the tritium residing deep in the bulk of materials, this surface tritium represents a safety hazard as it can be easily mobilised in the event of an accident. It could, however, be possibly removed and recovered. It is concluded that active and efficient methods to remove the codeposited layers are needed in ITER and periodic conditioning/cleaning would be required to control the tritium inventory and avoid exhausting the available fuel supply. Some methods which could possibly be used for in-situ cleaning are briefly discussed in conjunction with the research and development work required to extrapolate their applicability to ITER. (orig.) 53 refs.

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Review of recent japanese activities on tritium accountability in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Fukada, Satoshi, E-mail: sfukada@nucl.kyushu-u.ac.jp [Dept. Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-Koen, Kasuga, 816-8580 (Japan); Oya, Yasuhisa [College of Science, Academic Institute, Shizuoka University, 836 Otani, Suruga-ku, Shizuoka 422-8529 (Japan); Hatano, Yuji [Hydrogen Isotope Research Center, Organization for Promotion Research, University of Toyama, 3190 Gofuku, Toyama 930-8555 (Japan)

2016-12-15

Highlights: • Review of Japanese tritium-safety research is given from several viewpoints. • The keywords are tritium accountability and self sufficiency. • Tritium-relating history, tritium facilities and legal regulation are introduced. - Abstract: After introduction of Japanese history or recent topics on tritium (T)-relating research and T-handling capacity in facilities or universities, present activities on T engineering research in Japan are summarized in short in terms of T accountability on safety. The term of safety includes wide processes from T production, assay, storing, confinement, transfer through safety handling finally to shipment of its waste. In order to achieve reliable operation of fusion reactors, several unit processes included in the T cycle of fusion reactors are investigated. Especially, the following recent advances are focused: T retention in plasma facing materials, emergency detritiation system including fire case, T leak through metal tube walls, oxide coating and water detritiation. Strict control, storing and accurate measurement are especially demanded for T accountability depending on various molecular species. Since kg-order T of vaporable radioisotope (RI) will be handled in a fuel cycle or breeding system of a fusion reactor, the accuracy of <0.1% is demanded far over the conventional technology status. Necessity to control T balance within legal restrictions is always kept in mind for operation of the future reactor.

Review of recent japanese activities on tritium accountability in fusion reactors

International Nuclear Information System (INIS)

Fukada, Satoshi; Oya, Yasuhisa; Hatano, Yuji

2016-01-01

Highlights: • Review of Japanese tritium-safety research is given from several viewpoints. • The keywords are tritium accountability and self sufficiency. • Tritium-relating history, tritium facilities and legal regulation are introduced. - Abstract: After introduction of Japanese history or recent topics on tritium (T)-relating research and T-handling capacity in facilities or universities, present activities on T engineering research in Japan are summarized in short in terms of T accountability on safety. The term of safety includes wide processes from T production, assay, storing, confinement, transfer through safety handling finally to shipment of its waste. In order to achieve reliable operation of fusion reactors, several unit processes included in the T cycle of fusion reactors are investigated. Especially, the following recent advances are focused: T retention in plasma facing materials, emergency detritiation system including fire case, T leak through metal tube walls, oxide coating and water detritiation. Strict control, storing and accurate measurement are especially demanded for T accountability depending on various molecular species. Since kg-order T of vaporable radioisotope (RI) will be handled in a fuel cycle or breeding system of a fusion reactor, the accuracy of <0.1% is demanded far over the conventional technology status. Necessity to control T balance within legal restrictions is always kept in mind for operation of the future reactor.

Technical and Scientific Aspects of the JET Trace-Tritium Experimental Campaign

International Nuclear Information System (INIS)

Jones, T.T.C.; Brennan, D; Pearce, R.J.H.; Stork, D.; Zastrow, K.-D.; Balshaw, N.; Bell, A.C.; Bertalot, L.; Boyer, H.; Butcher, P.R.; Challis, C.D.; Ciric, D.; Clarke, R.; Conroy, S.; Darke, A.C.; Davies, N.; Edlington, T.; Ericsson, G.; Gibbons, C.; Hackett, L.J.; Haupt, T.; Hitchin, M.; Kaye, A.S.; King, R.; Kiptily, V.G.; Knipe, S.; Lawrence, G.; Lobel, R.; Mason, A.; Morgan, P.D.; Patel, B.; Popovichev, S.; Stamp, M.; Surrey, E.; Terrington, A.; Worth, L.; Young, D.

2005-01-01

The JET Trace Tritium (TTE) programme marked the first use of tritium in experiments under the managerial control of UKAEA, which operates the JET Facility on behalf of EFDA. The introduction of tritium into the plasma by gas fuelling and neutral beam injection, even in trace quantities, required the mobilisation of gram-quantities of tritium gas from the Active Gas Handling System (AGHS) product storage units into the supply lines connected to the torus gas valve and the neutral beam injectors. All systems for DT gas handling, recovery and reprocessing were therefore recommissioned and operating procedures re-established, involving extensive operations staff training. The validation of Key Safety Related Equipment (KSRE) is described with reference to specific examples. The differences between requirements for TTE and full DT operations are shown to be relatively small. The scientific motivation for TTE, such as the possibility to obtain high-quality measurements in key areas such as fuel-ion transport and fast ion dynamics, is described, and the re-establishment and development of JET's 14MeV neutron diagnostic capability for TTE and future DT campaigns are outlined. Some scientific highlights from the TTE campaign are presented

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

Tritium metrology within different media: focus on organically bound tritium (OBT); Metrologie du tritium dans differentes matrices: cas du tritium organiquement lie (TOL)

Energy Technology Data Exchange (ETDEWEB)

Baglan, N. [CEA Bruyeres-le-Chatel, DIF, 91 (France); Ansoborlo, E. [CEA Marcoule, DEN/DRCP/CETAMA, 30 (France); Cossonnet, C. [IRSN, DEI/STEME/LMRE, 91 - Orsay (France); Fouhal, L. [CEA Cadarache, DEN/D2S/LANSE, 13 - Saint-Paul-lez-Durance (France); Deniau, I.; Mokili, M. [SUBATECH/IN2P3/CNRS, 44 - Nantes (France); Henry, A. [AREVA-NC/DQSSE/PR - La Hague, 50 - Beaumont-Hague, (France); Fourre, E. [CEA Saclay, DSM/DRECAM/LSCE, 91 - Gif-sur-Yvette (France); Olivier, A. [GEA-Marine nationale, 50 - Cherbourg (France)

2010-07-15

The measurement of tritium in its various forms (mainly gas (HT), water (HTO) or solid (hydrides)), is an important key step for evaluating health and environmental risks and finally, dosimetry assessment. In vegetable or animal samples, tritium is often associated with the free water fraction, but may be included in the organic form as organically bound tritium (OBT). In this case, 2 forms exist: (i) a fraction called exchangeable or labile (E-OBT), bound to oxygen and nitrogen atoms, and (ii) a so-called non-exchangeable fraction (NE-OBT) bound to carbon atoms. The main technique for tritium analysis is liquid scintillation, which enables one to measure concentrations in the range of several Bq.L{sup -1}. The standards (AFNOR, ISO) published to date relate only to tritium analysis in water. Only one CETAMA method addresses OBT analysis in biological environments. This method has been tested since 2001 through intercomparison circuits on grass samples collected from the environment. Regarding tritium analysis in water, the strengths are reliability of this analysis at low concentrations (order of Bq.L{sup -1}), robustness and simplicity, and weaknesses are linked to problems of background, conservation and contamination of samples. Concerning OBT analysis, the analysis is reliable for values around 50 Bq.kg{sup -1} of fresh sample. The weaknesses are problems of contamination, reproducibility, analysis time (2 to 6 days) and lack of reference materials. The difficulty to date is the separation between E-OBT and NE-OBT, that will need experimental validation. (authors)

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.)

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)

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.)

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%

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Tritium labeling of detonation nanodiamonds.

Science.gov (United States)

Girard, Hugues A; El-Kharbachi, Abdelouahab; Garcia-Argote, Sébastien; Petit, Tristan; Bergonzo, Philippe; Rousseau, Bernard; Arnault, Jean-Charles

2014-03-18

For the first time, the radioactive labeling of detonation nanodiamonds was efficiently achieved using a tritium microwave plasma. According to our measurements, the total radioactivity reaches 9120 ± 120 μCi mg(-1), with 93% of (3)H atoms tightly bonded to the surface and up to 7% embedded into the diamond core. Such (3)H doping will ensure highly stable radiolabeled nanodiamonds, on which surface functionalization is still allowed. This breakthrough opens the way to biodistribution and pharmacokinetics studies of nanodiamonds, while this approach can be scalable to easily treat bulk quantities of nanodiamonds at low cost.

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.

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)

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.)

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

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

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.

Tritium in plants; Le tritium dans la matiere organique des vegetaux

Energy Technology Data Exchange (ETDEWEB)

Vichot, L.; Losset, Y. [CEA Valduc, 21 - Is-sur-Tille (France)

2009-07-01

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

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)

Radiation safety and regulatory aspects of tritium used as tracer in oil exploration

International Nuclear Information System (INIS)

Agarwal, S.P.

2005-01-01

Radiotracers have been used to trace the movement of a particular fluid at a well since the turn of the century. It is recognized that a properly designed and implemented radioactive tracing program can highly be cost effective in tracking the movement of oil field waters. Radioactive tracer results provide a practical link between macroscopic field interpretation from seismic and geological studies and localized core production wells

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

Depth profiling of hydrogen in ferritic/martensitic steels by means of a tritium imaging plate technique

International Nuclear Information System (INIS)

Otsuka, Teppei; Tanabe, Tetsuo

2013-01-01

Highlights: ► We applied a tritium imaging plate technique to depth profiling of hydrogen in bulk. ► Changes of hydrogen depth profiles in the steel by thermal annealing were examined. ► We proposed a release model of plasma-loaded hydrogen in the steel. ► Hydrogen is trapped at trapping sites newly developed by plasma loading. ► Hydrogen is also trapped at surface oxides and hardly desorbed by thermal annealing. -- Abstract: In order to understand how hydrogen loaded by plasma in F82H is removed by annealing at elevated temperatures in vacuum, depth profiles of plasma-loaded hydrogen were examined by means of a tritium imaging plate technique. Owing to large hydrogen diffusion coefficients in F82H, the plasma-loaded hydrogen easily penetrates into a deeper region becoming solute hydrogen and desorbs by thermal annealing in vacuum. However the plasma-loading creates new hydrogen trapping sites having larger trapping energy than that for the intrinsic sites beyond the projected range of the loaded hydrogen. Some surface oxides also trap an appreciable amount of hydrogen which is more difficult to remove by the thermal annealing

Transient tracers in the ocean (TTO) program: the North Atlantic study, 1981: the Tropical Atlantic study, 1983

International Nuclear Information System (INIS)

Brewer, P.G.; Sarmiento, J.L.; Smethie, W.M. Jr.

1985-01-01

The two parts of this major geochemical and physical oceanographic expedition took place on the research vessel Knorr of the Woods Hole Oceanographic Institution. The expeditions were designed to observe the passage of man-made geochemical tracers into the interior of the ocean. A systematic survey revealed the penetration into the thermocline and deep ocean of the products of man's military/industrial activities, principally tritium and carbon-14 resulting from atmospheric testing of nuclear weapons. The passage of these tracers documents as nothing else can the manner and time scale of ocean mixing and provides a fundamental calibration for models of ocean circulation. Maps showing the cruise routes are presented. 1 figure, 1 table

Tritium depth profiling by AMS in carbon samples from fusion experiments

International Nuclear Information System (INIS)

Friedrich, M.; Pilz, W.; Sun, G.; Behrisch, R.; Garcia-Rosales, C.

2001-01-01

Tritium depth profiling measurements by accelerator mass spectrometry have been performed at a facility installed at the Rossendorf 3 MV Tandetron. In order to achieve an uniform erosion at the target surface inside of a commercial Cs ion sputtering source and to avoid edge effects, the samples were mechanically scanned inside of a commercial Cs sputter ion source. The sputtered negative ions were mass analysed by the injection magnet of the Tandetron. The tritium ions are counted after the acceleration with semiconductor detectors. Depth profiles have been measured for carbon samples which had been exposed to the plasma at the first wall of the Garching fusion experiment ASDEX-Upgrade and from the European fusion experiment JET, Culham/UK. A dedicated AMS facility with an air-insulated 100 kV tandem accelerator for depth profiling measurements at samples with high tritium concentration is under construction. First results of test operation are presented. (orig.)

Tritium protective clothing

International Nuclear Information System (INIS)