Temporal sealing material of tritium-contaminated stainless steel

International Nuclear Information System (INIS)

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

2010-01-01

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

Fluorine 18 in tritium generator ceramic materials

International Nuclear Information System (INIS)

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

1992-01-01

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

Modeling of tritium behavior in ceramic breeder materials

International Nuclear Information System (INIS)

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

1988-11-01

Computer models are being developed to predict tritium release from candidate ceramic breeder materials for fusion reactors. Early models regarded the complex process of tritium release as being rate limited by a single slow step, usually taken to be tritium diffusion. These models were unable to explain much of the experimental data. We have developed a more comprehensive model which considers diffusion and desorption from the grain surface. In developing this model we found that it was necessary to include the details of the surface phenomena in order to explain the results from recent tritium release experiments. A diffusion-desorption model with a desorption activation energy which is dependent on the surface coverage was developed. This model provided excellent agreement with the results from the CRITIC tritium release experiment. Since evidence suggests that other ceramic breeder materials have desorption activation energies which are dependent on surface coverage, it is important that these variations in activation energy be included in a model for tritium release. 17 refs., 12 figs

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

International Nuclear Information System (INIS)

Hassanein, A.

1999-01-01

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

Tritium saturation in plasma-facing materials surfaces

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.; Causey, R.A.; Federici, G.; Haasz, A.A.

1998-01-01

Plasma-facing components in the international thermonuclear experimental reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10 20 -10 23 particles/m 2 s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments. (orig.)

Tritium saturation in plasma-facing materials surfaces

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J. [Idaho Nat. Eng. and Environ. Lab., Idaho Falls, ID (United States); Causey, R.A. [Sandia National Labs., Livermore, CA (United States); Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Haasz, A.A. [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

1998-10-01

Plasma-facing components in the international thermonuclear experimental reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10{sup 20}-10{sup 23} particles/m{sup 2}s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments. (orig.) 39 refs.

Tritium interactions with steel and construction materials in fusion devices

International Nuclear Information System (INIS)

Dickson, R.S.

1990-11-01

The literature on the interactions of tritium and tritiated water with metals, glasses, ceramics, concrete, paints, polymers and other organic materials is reviewed in this report Some of the processes affecting the amount of tritium found on various materials, such as permeation, sorption and the conversion of tritium found on various materials, such as permeation, sorption and conversion of elemental tritium (T 2 ) to tritiated water (HTO), are also briefly outlined. Tritium permeation in steels is fairly well understood, but effects of surface preparation and coatings on sorption are not yet clear. Permeation of T 2 into other metals with cleaned surfaces has been studied thoroughly at high temperature, and the effect of surface oxidation has also been explored. The room-temperature permeation rates of low-permeability metals with cleaned surfaces are much faster than indicated by high-temperature results, because of grain-boundary diffusion. Elastomers have been studied to a certain extent, but some mechanisms of interaction with tritium gas and sorbed tritium are unclear. Ceramics have some of the lowest sorption and permeation rates, but ceramic coatings on stainless steels do not lower permeation or tritium as effectively as coatings obtained by oxidation of the steel, probably because of cracking caused by differences in thermal expansion coefficient. Studies on concrete are in their early stages; they show that sorption of tritiated water on concrete is a major concern in cleanup of releases of elemental tritium into air in tritium handling facilities. Some of the codes for modelling releases and sorption of T 2 and HTO contain unproven assumptions about sorption and T 2 → HTO conversion. Several experimental programs will be required in order to clear up ambiguities in previous work and to determine parameters for materials which have not yet been investigated. (146 refs., tab.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-06-15

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

Corrosion behaviour of Al based tritium permeation barriers in flowing Pb-17Li

International Nuclear Information System (INIS)

Glasbrenner, H.; Konys, J.; Voss, Z.; Wedemeyer, O.

2002-01-01

Tritium permeation barriers on low-activation steels are required in fusion technology in order to reduce the tritium permeation rate through the structural material into the cooling water system. Al-Fe layers with alumina on top can fulfil the required reduction rate. Three techniques were selected to produce such a multi-layered coating system: chemical vapour deposition (CVD) by CEA, hot-dip aluminising (HDA) by FZK and vacuum plasma spraying (VPS) by JRC Ispra. A sufficient corrosion resistance against Pb-17Li attack is also required for the coating. Therefore, the corrosion behaviour of these three coatings on ferritic-martensitic steels was studied in the PICOLO loop of FZK in flowing Pb-17Li at 480 deg. C up to 10 000 h. Corrosion effects could not be found on HDA and VPS coated specimens even up to the longest time of exposure. The total thickness of the two-layered system remained unchanged at around 130 μm for all examined HDA and VPS specimens. In contrast to this, corrosion effects could be inspected on CVD coated specimens

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Tritium permeation barriers in contact with liquid lithium-lead eutectic (Pb-17Li)

International Nuclear Information System (INIS)

Forcey, K.S.; Perujo, A.

1995-01-01

The permeation of deuterium through coated stainless steel tubes containing liquid lithium-lead eutectic (Pb-17Li) has been studied and compared to measurements through tubes without the lithium compound. The measurements form part of an investigation into the effect of a potential tritium breeder material on permeation barriers for fusion reactors. The coatings studied were CVD TiC and Al 2 O 3 and a pack aluminised layer. Without the lithium-lead, the CVD coatings reduced the permeation rate up to 1 order of magnitude, and the aluminised layer up to 2 orders of magnitude. A CVD layer was unaffected by Pb-17Li whilst in the case of the aluminised tube, the lithium-lead completely removed the permeation barrier, presumably by attacking the surface oxide. Furthermore, the aluminised sample presented a large number of cracks and poor adheren ce to the substrate. ((orig.))

Helium effects on tritium storage materials

International Nuclear Information System (INIS)

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

2008-01-01

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

Helium effects on tritium storage materials

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-15

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

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.)

Analysis of tritium migration and deposition in fusion-reactor systems

International Nuclear Information System (INIS)

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

1981-01-01

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

Experimental studies of tritium barrier concepts for fusion reactors

International Nuclear Information System (INIS)

Maroni, V.A.; Van Deventer, E.H.; Renner, T.A.; Pelto, R.H.; Wierdak, C.J.

1976-01-01

Ongoing experimental studies at ANL aimed at the development of methods to reduce tritium migration in fusion reactor systems currently include (1) work on the development of multilayered metal composites and impurity-coated refractory metals as barriers to tritium permeation in elevated temperature (greater than 300 0 C) structures and (2) investigations of the kinetics of tritium trapping reactions in inert gas purge streams under conditions that emulate fusion reactor environments. Significant results obtained thus far are (1) demonstration of greater than 50-fold reductions in the hydrogen permeability of stainless steel structures by using stainless steel-clad composites containing an intermediate layer of a selected copper alloy and (2) verification that surface-oxide coatings lead to greater than 100-fold reductions in the hydrogen permeability of vanadium, but that severe oxygen penetration and embrittlement of the vanadium occur at temperatures in the range from 300 to 800 0 C and under conditions of extremely low oxygen potential. Other considerations pertaining to the large-scale use of metal composites in fusion reactors are discussed, and progress in efforts to demonstrate the fabricability of metal composites is reviewed. Also presented are results of studies of the efficiencies of (1) CuO and CuO--MnO 2 beds in converting HT to HTO and (2) magnesium metal beds in converting HTO to HT

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

International Nuclear Information System (INIS)

Gacs, I.; Mlinko, S.

1978-01-01

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

Diamond and Diamond-Like Materials as Hydrogen Isotope Barriers

International Nuclear Information System (INIS)

Foreman, L.R.; Barbero, R.S.; Carroll, D.W.; Archuleta, T.; Baker, J.; Devlin, D.; Duke, J.; Loemier, D.; Trukla, M.

1999-01-01

This is the final report of a two-year, Laboratory Directed Research and Development (LDRD) project at Los Alamos National Laboratory (LANL). The purpose of this project was to develop diamond and diamond-like thin-films as hydrogen isotope permeation barriers. Hydrogen embrittlement limits the life of boost systems which otherwise might be increased to 25 years with a successful non-reactive barrier. Applications in tritium processing such as bottle filling processes, tritium recovery processes, and target filling processes could benefit from an effective barrier. Diamond-like films used for low permeability shells for ICF and HEDP targets were also investigated. Unacceptable high permeabilities for hydrogen were obtained for plasma-CVD diamond-like-carbon films

Influence of surface roughness and melt superheat on HDA process to form a tritium permeation barrier on RAFM steel

Energy Technology Data Exchange (ETDEWEB)

Purushothaman, J. [B.S. Abdur Rahman University, Chennai 600048 (India); MTD, MMG, IGCAR, Kalpakkam 603102 (India); Ramaseshan, R., E-mail: seshan@igcar.gov.in [TFCS, SND, MSG, IGCAR, Kalpakkam 603102 (India); Albert, S.K. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Rajendran, R. [B.S. Abdur Rahman University, Chennai 600048 (India); Gowrishankar, N. [IP Rings Ltd., Maraimalainagar, Chennai 603209 (India); Ramasubbu, V. [MTD, MMG, IGCAR, Kalpakkam 603102 (India); Murugesan, S.; Dasgupta, Arup [PMG, MMG, IGCAR, Kalpakkam 603102 (India); Jayakumar, T. [MTD, MMG, IGCAR, Kalpakkam 603102 (India)

2015-12-15

Highlights: • Surface modified RAFMS samples were subjected to HDA and thermal oxidation. • Sample modified by SB process showed better coating and interface morphology. • Aluminized samples at 740 °C for 2 min showed Fe{sub 2}Al{sub 9}Si{sub 2} intermetallic phase. • Oxidized samples showed Fe{sub 2}Al{sub 8}Si, Fe{sub 2}Al{sub 3}Si{sub 3} and Fe{sub 3}Al{sub 2}Si{sub 3} intermetallic phases. • A uniform permeation barrier Al{sub 2}O{sub 3} was formed on the coating of oxidized HDA samples. - Abstract: The most optimal candidate material for fabrication of Test Blanket Module (TBM) in the installation of ITER and future fusion reactors is Reduced Activation Ferritic Martensitic (RAFM) steel, yet one of the major challenges that need to be addressed with RAFM is minimizing the loss of tritium in a reactor environment through the formation of tritium permeation barrier. One of the most promising methods for the tritium permeation barrier is through duplex coating with Al{sub 2}O{sub 3}/Fe–Al which is well known to reduce tritium permeation rate by several orders of magnitude. The present work aims to form an alumina layer on RAFM steel by a two-step method, which consists of (i) Hot Dip Aluminizing (HDA) and (ii) conversion of Al into alumina by a subsequent oxidation process. In addition, the influence of surface roughness of the substrate, superheat condition of the Al alloy melt and its composition on microstructural properties of coating before and after oxidation were investigated using OM, SEM–EDS, XRD, indentation micro hardness and scratch test. The experimental results confirmed the formation of alumina layer on RAFM steel after the HDA and oxidation process. Moreover, the surface roughness of the substrate, melt superheat of Al alloy and its composition are found to have a significant influence on the microstructure, thickness, micro-hardness, nature of intermetallic compounds formed and adhesion strength of the coating.

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

International Nuclear Information System (INIS)

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

1994-01-01

The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab.

Measurement of tritium permeation through resistant materials near room temperature

International Nuclear Information System (INIS)

Maienschein, J.; DuVal, V.; McMurphy, F.; Uribe, F.; Musket, R.; Brown, D.

1985-01-01

To measure tritium permeation through low-permeability materials at 50 to 170 0 C, we use highly-sensitive liquid scintillation counting to detect the permeating tritium. To validate our method, we conducted extensive experiments with copper, for which much data exists for comparison. We report permeability of tritium through copper at 50, 100, and 170 0 C, and discuss details of the experimental technique. Further plans are outlined. 15 refs., 5 figs., 1 tab

Lithium aluminate/zirconium material useful in the production of tritium

Science.gov (United States)

Cawley, W.E.; Trapp, T.J.

A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

EXOTIC: Development of ceramic tritium breeding materials

International Nuclear Information System (INIS)

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

1989-03-01

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

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

Tritium transport in subsurface soil and its retentivity in cement waste forms

International Nuclear Information System (INIS)

Varghese, C.; Khan, Z.A.

1990-01-01

Tritiated aqueous wastes generated from power reactors and reprocessing plants need fixation and containment in solid media. Incorporating tritium in inorganic hydrates like cement retards its movements substantially. Naturally occurring clay mineral like vermiculite which has high water holding capacity was used as a filler material. To improve the retention of tritium in the cement blocks, certain coating materials were used after curing the blocks for 28 days. Long time leach tests for these barrier incorporated wastes forms have shown tritium releases of 17% loaded tritium over a period of 97 days. But the selected coating reduced it to 0.9%. Investigations on the movements of tritium at Tarapur and Narora revealed that in the unsaturated zone, tritium percolation was low, whereas in the saturated zone, it was found to travel with higher velocities, limiting to ground water movement. (author). 5 refs., 3 tabs

Electric-Discharge Machining Techniques for Evaluating Tritium Effects on Materials

International Nuclear Information System (INIS)

Morgan, M.J.

2003-01-01

In this investigation, new ways to evaluate the long-term effects of tritium on the structural properties of components were developed. Electric-discharge machining (EDM) techniques for cutting tensile and fracture toughness samples from tritium exposed regions of returned reservoirs were demonstrated. An existing electric discharge machine was used to cut sub-size tensile and fracture toughness samples from the inside surfaces of reservoir mockups. Tensile properties from the EDM tensile samples were similar to those measured using full-size samples cut from similar stock. Although the existing equipment could not be used for machining tritium-exposed hardware, off-the shelf EDM units are available that could. With the right equipment and the required radiological controls in place, similar machining and testing techniques could be used to directly measure the effects of tritium on the properties of material cut from reservoir returns. Stress-strain properties from tritium-exposed reservoirs would improve finite element modeling of reservoir performance because the data would be representative of the true state of the reservoir material in the field. Tensile data from samples cut directly from reservoirs would also complement existing shelf storage and burst test data of the Life Storage Program and help answer questions about a specific reservoir's processing history and properties

Development of advanced tritium breeding material with added lithium for ITER-TBM

Energy Technology Data Exchange (ETDEWEB)

Hoshino, Tsuyoshi, E-mail: hoshino.tsuyoshi@jaea.go.jp [Blanket Irradiation and Analysis Group, Fusion Research and Development Directorate, Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Higashi Ibaraki-gun, Ibaraki 311-1393 (Japan); Kato, Kenichi; Natori, Yuri; Oikawa, Fumiaki; Nakano, Natsuko; Nakamura, Mutsumi [Kaken, Co. Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan); Sasaki, Kazuya [Institute of Engineering Innovation and Department of Nuclear Engineering and Management School of Engineering, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-8656 (Japan); Suzuki, Akihiro [Nuclear Professional School, School of Engineering, University of Tokyo, 2-22 Shirakata-Shirane, Ibaraki 319-1188 (Japan); Terai, Takayuki [Institute of Engineering Innovation and Department of Nuclear Engineering and Management School of Engineering, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-8656 (Japan); Tatenuma, Katsuyoshi [Kaken, Co. Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan)

2011-10-01

Lithium titanate (Li{sub 2}TiO{sub 3}) is one of the most promising candidates among tritium breeding materials because of its good tritium release characteristics. However, the mass of Li{sub 2}TiO{sub 3} decreased with time in a hydrogen atmosphere by the reduction of Ti and Li evaporation. In order to prevent the mass decrease at high temperatures, advanced tritium breeding material with added Li (Li{sub 2+x}TiO{sub 3+y}) should be developed. For this purpose, an advanced Li{sub 2}TiO{sub 3} with added Li was synthesized from proportionally mixed LiOH.H{sub 2}O and H{sub 2}TiO{sub 3} with a Li/Ti ratio of 2.2. The results of X-ray diffraction measurement showed that this advanced tritium breeding material existed as the non-stoichiometric compound Li{sub 2+x}TiO{sub 3+y}. The desired molar ratio of Li/Ti was achieved by appropriate mixing of LiOH.H{sub 2}O and H{sub 2}TiO{sub 3}. Therefore, synthesis by mixing LiOH.H{sub 2}O and H{sub 2}TiO{sub 3} is a promising mass production method for the advanced tritium breeding material with added Li for the test blanket module of ITER.

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.)

Control of tritium permeation through fusion reactor strucural materials

International Nuclear Information System (INIS)

Maroni, V.A.

1978-01-01

The intention of this paper is to provide a brief synopsis of the status of understanding and technology pertaining to the dissolution and permeation of tritium in fusion reactor materials. The following sections of this paper attempt to develop a simple perspective for understanding the consequences of these phenomena and the nature of the technical methodology being contemplated to control their impact on fusion reactor operation. Considered in order are: (1) the occurrence of tritium in the fusion fuel cycle, (2) a set of tentative criteria to guide the analysis of tritium containment and control strategies, (3) the basic mechanisms by which tritium may be released from a fusion plant, and (4) the methods currently under development to control the permeation-related release mechanisms. To provide background and support for these considerations, existing solubility and permeation data for the hydrogen isotopes are compared and correlated under conditions to be expected in fusion reactor systems

In-pile tritium permeation experiment

International Nuclear Information System (INIS)

Longhurst, G.R.; Miller, L.G.; Watts, K.D.; Kershner, C.J.; Rogers, M.L.

1982-01-01

The experiments in progress are examining various aspects of the permeation of hydrogen isotopes through fusion materials. Of particular importance will be the measurement of permeation due to ion implantation in the presence of a neutron radiation field. Theoretical and early experimental results for these experiments have suggested that sufficient tritium will permeate fusion reactor interior structures that development of a permeation barrier will be needed. (orig.)

In-pile tritium permeation experiment

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R.; Miller, L.G.; Watts, K.D. (Idaho National Engineering Lab., Idaho Falls (USA)); Kershner, C.J.; Rogers, M.L. (Monsanto Research Corp., Miamisburg, OH (USA). Mound Facility)

The experiments in progress are examining various aspects of the permeation of hydrogen isotopes through fusion materials. Of particular importance will be the measurement of permeation due to ion implantation in the presence of a neutron radiation field. Theoretical and early experimental results for these experiments have suggested that sufficient tritium will permeate fusion reactor interior structures that development of a permeation barrier will be needed.

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

Effect of hydrophobic paints coating for tritium reduction in concrete materials

International Nuclear Information System (INIS)

Edao, Y.; Fukada, S.; Nishimura, Y.; Katayama, K.; Takeishi, T.; Hatano, Y.; Taguchi, A.

2012-01-01

Highlights: ► Effects of hydrophobic paint coating in tritium transport are investigated. ► Two kinds of paints, acrylic-silicon resin and epoxy resin are used. ► The hydrophobic paints are effective to reduce tritium permeation. ► The effect of tritium reduction of epoxy paint is higher than that of silicon. - Abstract: The effects of hydrophobic paint coating on a concrete material of cement paste on the tritium transport are investigated. The cement paste is coated with two kinds of paints, acrylic-silicon resin paint and epoxy resin paint. We investigated the amount of tritium trapped in the samples exposed to tritiated water vapor by means of sorption and release. It was found that both the hydrophobic paints could reduce effectively tritium permeation during 50 days exposure of tritiated water vapor. The effect of tritium reduction of the epoxy paint was higher than that of silicon while the amount of tritium trapped in the epoxy paint was larger than that of silicon due to difference of the structure. Based on an analysis of a diffusion model, the rate-determining step of tritium migration through cement paste coated with the paints is diffusion through the paints respectively. It was found that tritium was easy to penetrate through silicon because there were many pores or voids in the silicon comparatively. In the case of tritium released from the epoxy paint, it is considered that tritium diffusion in epoxy is slow due to retardation by isotope exchange reaction to water included in epoxy paint.

Design and construction of thermal desorption measurement system for tritium contained materials

International Nuclear Information System (INIS)

Hara, M.; Hatano, Y.; Calderoni, P.; Shimada, M.

2014-01-01

The dual-mode thermal desorption analysis system was designed and built in Idaho National Laboratory (INL) to examine the evolution of the hydrogen isotope gas from materials. The system is equipped with a mass spectrometer for stable hydrogen isotopes and an ionization chamber for tritium components. The performance of the system built was tested with using tritium contained materials. The evolution of tritiated gas species from contaminated materials was measured successfully by using the system. (author)

Simulation of thermal stresses in SiC-Al2O3 composite tritium penetration barrier by finite-element analysis

International Nuclear Information System (INIS)

Liu, Hongbing; Tao, Jie; Gautreau, Yoann; Zhang, Pingze; Xu, Jiang

2009-01-01

Tritium penetration barrier (TPB) composed of Al 2 O 3 and SiC on 316L stainless steel was proposed to improve the tritium penetration resistance of the substrate in this work. At the same time, the concept of functionally graded materials (FGM) was applied to manage to decrease residual stresses between Al 2 O 3 and 316L stainless steel substrate due to the mismatch of their thermal expansion coefficients. The effects of system architecture on the residual stresses developed in the composite coatings were investigated numerically by means of finite-element analysis (FEA). Modeling results showed that the presence of the graded properties and the compositions within the coating did reduce the stress discontinuity at the interfaces between the coating and the substrate. Also, the magnitudes of the residual stresses on the coating surface and at the coating/substrate interface were dependent on the Al 2 O 3 and SiC coating thickness.

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)

Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

International Nuclear Information System (INIS)

Reiter, F.; Tominetti, S.; Perujo, A.

1992-01-01

The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-12-01

This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 6}Zr{sub 2}O{sub 7}, Li{sub 8}ZrO{sub 6}, Li{sub 2}O and Li{sub 2}SiO{sub 3} have been irradiated at different temperature levels and up to a maximum lithium burnup of about 3%. Tritium residence times of the various breeding materials have been determined from temperature transients performed during irradiation. After irradiation the tritium inventory has been determined from small samples of the various materials. From the out-of-pile tritium release experiments activation energies were determined. These activities have been performed at ECN within the framework of the European Fusion Technology Programme on Breeding Blankets. (orig.).

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.)

Problems bound to the tritium in materials for the nuclear - some illustrations; Problematiques liees au tritium dans les materiaux dans le domaine nucleaire - quelques illustrations

Energy Technology Data Exchange (ETDEWEB)

Gastaldi, O. [CEA Cadarache (DTN/STPA/LPC), 13 - Saint-Paul-lez-Durance (France)

2007-07-01

The tritium control takes more and more importance in the nuclear industry because of the release more and more limited, in the environment. After a presentation on the tritium sources in the environment, the author presents the different ways of its production. Then for each reactor channel, the main problems are presented (fission and fusion). The last part deals with the behavior of the tritium in materials: the tritium inventory control in a fusion system, the tritium management after the reactor exploitation. (A.L.B.)

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

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

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

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)

PRODUCTION OF TRITIUM

Science.gov (United States)

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

1963-02-26

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

Tritium accountancy in fusion systems

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

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

Handling of tritium contaminated effluents and wastes: Final Report

International Nuclear Information System (INIS)

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

1983-01-01

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

A preliminary study on PVDC modified composite materials of protective for tritium

International Nuclear Information System (INIS)

Wan Xiaoli; Dan Guiping; Li Ye; Wen Wei; Zhang Dong

2012-01-01

Through the experimental device, the HTO permeation performances of two kinds of PVDC modified composite materials were studied. The characteristic curves of the two composite materials were ascertained, and various other packing materials with anti-tritium permeation performance were compared. (authors)

Improved permeation barriers for tritiated waste packaging

International Nuclear Information System (INIS)

Vassallo, G.; Van Den Bergh, R.; Forcey, K.S.; Perujo, A.

1994-01-01

High-density polyethylene (HDPE) is extensively used as flexible bagging or packaging for soft tritiated waste in the tritium community because of its low permeability to the more radiotoxic form of tritium, i.e., tritiated water (HTO). However, HDPE does not represent a perfect barrier to HTO nor does it effectively hinder the permeation of elemental tritium, i.e, HT. This latter drawback is particularly important considering that the elemental form may readily convert to HTO outside of the waste package. The possible use of a multilayer film as packing material for the conditioning of tritiated waste is assessed, and its capability to hinder the permeation of elemental tritium is measured and compared with that of bare HDPE. The material investigated is readily available from the food industry. 5 refs., 1 tab

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

Tritium protective clothing

International Nuclear Information System (INIS)

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

1979-06-01

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

Tritium protective clothing

Energy Technology Data Exchange (ETDEWEB)

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

1979-06-01

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

Tritium transport and release from lithium ceramic breeder materials

International Nuclear Information System (INIS)

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

1994-01-01

In an operating fusion reactor,, the tritium breeding blanket will reach a condition in which the tritium release rate equals the production rate. The tritium release rate must be fast enough that the tritium inventory in the blanket does not become excessive. Slow tritium release will result in a large tritium inventory, which is unacceptable from both economic and safety viewpoints As a consequence, considerable effort has been devoted to understanding the tritium release mechanism from ceramic breeders and beryllium neutron multipliers through theoretical, laboratory, and in-reactor studies. This information is being applied to the development of models for predicting tritium release for various blanket operating conditions

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Design and tritium permeation analysis of China HCCB TBM port cell

International Nuclear Information System (INIS)

Jiangfeng, S.; Guoqiang, H.; Zhiyong, H.; Chang'an, C.; Deli, L.

2015-01-01

China is planning to develop a helium-cooled ceramic breeder (HCCB) test blanket module (TBM) on ITER to test key blanket technologies. In this paper, the design and tritium permeation analysis of China HCCB TBM port cell are introduced. A theoretical model has been developed to estimate tritium permeation rates and leak rates from the components and pipes which China has scheduled to house in the port cell. It is shown that on normal working conditions, the permeation and leak rate of the systems in the port cell will be no higher than 1.58 Ci/d without the use of tritium permeation barriers, and 0.10 Ci/d with the use of tritium permeation barriers. It also appears that tritium permeation barriers are necessary for high temperature components such as the reduction bed and the heater

Design and tritium permeation analysis of China HCCB TBM port cell

Energy Technology Data Exchange (ETDEWEB)

Jiangfeng, S.; Guoqiang, H.; Zhiyong, H.; Chang' an, C.; Deli, L. [China Academy of Engineering Physics, Mianyang, Sichuan (China)

2015-03-15

China is planning to develop a helium-cooled ceramic breeder (HCCB) test blanket module (TBM) on ITER to test key blanket technologies. In this paper, the design and tritium permeation analysis of China HCCB TBM port cell are introduced. A theoretical model has been developed to estimate tritium permeation rates and leak rates from the components and pipes which China has scheduled to house in the port cell. It is shown that on normal working conditions, the permeation and leak rate of the systems in the port cell will be no higher than 1.58 Ci/d without the use of tritium permeation barriers, and 0.10 Ci/d with the use of tritium permeation barriers. It also appears that tritium permeation barriers are necessary for high temperature components such as the reduction bed and the heater.

Tritium Management In HCLL-PPCS Model AB Blanket

International Nuclear Information System (INIS)

Ricapito, I.; Aiello, A.; Benamati, G.; Utili, M.; Ciampichetti, A.; Zucchetti, M.

2006-01-01

One the main issues in the HCLL blanket development for a prototype fusion reactor is the technical feasibility of the bred tritium processing system. The basis of such concern lies in the very low tritium-Pb17Li Sieverts' constant, as measured by different scientists in the past years. In the PPCS reactor 650 g/d of tritium must be generated in the breeding blanket while less than 1 g/y of tritium has to be released to the environment through the secondary cooling circuit. As a consequence, CPS (Coolant Purification System) plays a fundamental role because it has to keep at an acceptable level the tritium partial pressure in the primary HCS (Helium Cooling Circuit) limiting, therefore, the tritium environmental release through leakage and permeation into the secondary cooling circuit. On the other hand, the He mass flow-rate to be processed by CPS is linear with the tritium permeation rate from the breeder into HCS. Therefore, with the above mentioned low Sieverts' constant values and the consequent high tritium partial pressure in the liquid metal, the possibility to keep acceptable the CPS capacity depends on a highly efficient and stable performance of tritium permeation barriers, to be applied not only on the blanket cooling plates but also on the steam generator walls. However, the experimental results on the tritium permeation barriers under relevant operative conditions were so far quite disappointing. The new data on the Sieverts' constant achieved at ENEA CR Brasimone, one order of magnitude higher than those founding the past, have a big impact in relaxing the above mentioned requirements for the tritium management in PPCS model AB reactor. Besides presenting and discussing these recent experimental results, an updated assessment of the tritium permeation rate from the liquid breeder into HCS through the cooling plates and from HCS into the environment through the steam generators is given in this paper. The consequent new constraints in terms of tritium

Tritium immobilisation

International Nuclear Information System (INIS)

Bridger, N.J.

1982-01-01

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

Tritium diffusion in nonmetallic solids of interest for fusion reactors. Final report

International Nuclear Information System (INIS)

Elleman, T.

1979-01-01

Tritium diffusion measurements have been conducted in Al 2 O 3 , BeO, Y 2 O 3 , SiC, B 4 C, Si 3 N 4 and pyrolytic carbon as a basis for evaluating these materials as potential tritium barriers in fusion reactors. Deuterium solubility measurements were conducted with Al 2 O 3 , SiC and pyrolytic carbon to establish the pressure and temperature dependence of solubility and to identify solubility ranges. Hydrogen permeability measurements on commercially available Al 2 O 3 and SiC materials were used as a check on calculated permeabilities and to provide data on hydrogen permeation rates in polycrystalline materials. The diffusion, solubility and permeation results are presented and discussed in terms of fusion reactor applications

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

Detection of tritium sorption on four soil materials

International Nuclear Information System (INIS)

Teng Yanguo; Zuo Rui; Wang Jinsheng; Hu Qinhong; Sun Zongjian; Zeng Ni

2011-01-01

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

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

International Nuclear Information System (INIS)

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

1990-09-01

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

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 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)

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.

The Use of Subsurface Barriers to Support Treatment of Metals and Reduce the Flux of Tritium to Fourmile Branch at the Savannah River Site in South Carolina - 13358

Energy Technology Data Exchange (ETDEWEB)

Blount, Gerald; Thibault, Jeffrey; Wells, Leslie [Savannah River Nuclear Solutions LLC, 730-4B, Aiken, SC 29808 (United States); Prater, Phillip [Department of Energy, Savannah River Site, Aiken, SC 29808 (United States)

2013-07-01

Fourmile Branch. The operating cost, including waste disposal, for the two systems was ∼$1.3 M/month. Both systems employed reinjection of tritiated water up gradient of the extraction, and produced large quantities of waste from non-tritium isotopes and metals removal prior to reinjection. Both systems were determined to be ineffective and potentially detrimental with respect to limiting the flux of contaminants to Fourmile Branch. After it became apparent that there was very little benefit to continued operation of the systems, and the staggering cost of operations was recognized by the SRS and regulators, a new remedy was developed. The new system uses vertical subsurface barriers to redirect groundwater flow to limit the transport of contaminants to the stream. The barriers were constructed of acid resistant grout using deep soil mixing techniques. The grout mixture used low swelling clay, fly ash, and sodium hydroxide to form a pozzolana material with low permeability and low strength. The SRS and regulators agreed to a series of remedial goals, with the first goal to reduce tritium flux to the stream by 70% and bring constituents other than tritium to groundwater protection standards. (authors)

Radiation durability of polymeric materials in solid polymer electrolyzer for fusion tritium plant

International Nuclear Information System (INIS)

Iwai, Yasunori; Yamanishi, Toshihiko; Hiroki, Akihiro; Tamada, Masao

2009-02-01

This document presents the radiation durability of various polymeric materials applicable to a solid-polymer-electrolyte (SPE) water electrolyzer to be used in the tritium facility of fusion reactor. The SPE water electrolyzers are applied to the water detritiation system (WDS) of the ITER. In the ITER, an electrolyzer should keep its performance during two years operation in the tritiated water of 9TBq/kg, the design tritium concentration of the ITER. The tritium exposure of 9TBq/kg for two years is corresponding to the irradiation of no less than 530 kGy. In this study, the polymeric materials were irradiated with γ-rays or with electron beams at various conditions up to 1600 kGy at room temperature or at 343 K. The change in mechanical and functional properties were investigated by stress-strain measurement, thermogravimetric analysis (TGA), differential scanning calorimetry (DSC), X-ray photoelectron spectra (XPS), and so on. Our selection of polymeric materials for a SPE water electrolyzer used in a radiation environment was Pt + Ir applied Nafion N117 ion exchange membrane, VITON O-ring seal and polyimide insulator. (author)

Tritium/hydrogen barrier development

International Nuclear Information System (INIS)

Hollenberg, G.W.; Simonen, E.P.; Kalinen, G.; Terlain, A.

1994-06-01

A review of hydrogen permeation barriers that can be applied to structural metals used in fusion power plants is presented. Both implanted and chemically available hydrogen isotopes must be controlled in fusion plants. The need for permeation barriers appears strongest in Li17-Pb blanket designs, although barriers also appear necessary for other blanket and coolant systems. Barriers that provide greater than a 1000 fold reduction in the permeation of structural metals are desired. In laboratory experiments, aluminide and titanium ceramic coatings provide permeation reduction factors, PRFS, from 1000 to over 100,000 with a wide range of scatter. The rate-controlling mechanism for hydrogen permeation through these barriers may be related to the number and type of defects in the barriers. Although these barriers appear robust and resistant to liquid metal corrosion, irradiation tests which simulate blanket environments result in very low PRFs in comparison to laboratory experiments, i.e., <150. It is anticipated from fundamental research activities that the REID enhancement of hydrogen diffusion in oxides may contribute to the lower permeation reduction factors during in-reactor experiments

Sorption of tritium and tritiated water on construction materials

International Nuclear Information System (INIS)

Dickson, R.S.; Miller, J.M.

1991-11-01

Sorption and desorption of tritium (HT) and tritiated water (HTO) on materials to be used in the construction of fusion facilities were studied. In ∼ 24-hour exposures in argon or room air, metal samples sorbed 8-200 μCi/m 2 of tritium from atmospheres of 5-9 Ci/m 3 HT, and non-metallic samples sorbed 60-800 μCi/m 2 from atmospheres of 14 Ci/m 3 HT. Sorption of HTO varied much more widely than HT sorption for different samples, ranging from 4 μCi/m 2 for glass to 1,300,000 μCi/m 2 for concrete samples, in 24-hour exposures to 1 Ci/m 3 HTO in room air. Time dependence of desorption in dry air showed a rapid initial process and a slower secondary process. (Author) (10 refs., 4 figs., 2 tabs.)

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

International Nuclear Information System (INIS)

Conrad, R.; Debarberis, L.

1991-01-01

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

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.

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

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.)

EXOTIC: Development of ceramic tritium breeding materials

International Nuclear Information System (INIS)

Kwast, H.; Conrad, R.

1989-09-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (author) 8 refs.; 2 figs

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.

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

Progress report on the accelerator production of tritium materials irradiation program

International Nuclear Information System (INIS)

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

1997-01-01

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

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)

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 breeding in fusion reactors

International Nuclear Information System (INIS)

Abdou, M.A.

1982-10-01

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

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

Tritium in Meteorites and in Recovered Satellite Material; Tritium dans les meteorites et dans les matieres provenant d'un satellite recupere; Tritij v meteoritakh i v vozvrashchennom sputnike; Tritio en meteoritos y en el material de un satelite recuperado

Energy Technology Data Exchange (ETDEWEB)

Fireman, E L [Smithsonian Astrophysical Observatory, Cambridge, MA (United States)

1962-01-15

Tritium was measured in separated phases and in whole rock samples of the Bruderheim chondritic meteorite and in samples of lead and iron material of recovered satellites. Radioactive isotopes of argon were also measured. The tritium and argon radioactivity in the Bruderheim meteorite can be reasonably well explained by the interaction of cosmic ray particles of some thousand million volts energy with the meteoritic material. The tritium content of the recovered satellite material was more than a factor of a hundred too large to be explained by the interactions of cosmic rays or by the interactions of solar flare particles with the satellite. The high tritium content of the satellite material must result from a flux of incident tritium particles that stop in the satellite. (author) [French] Tritium dans les meteorites et dans les matieres provenant d'un satellite recupere. On a mesure la teneur en tritium d'echantillons de diverses parties et de l'ensemble de la roche formant la chondrite de Bruderheim, ainsi que celle d'echantillons de plomb et de fer preleves sur des satellites recuperes. On a egalement mesure la teneur en radioargon dans la meteorite de Bruderheim. La presence de tritium et de radioargon peut fort bien s'expliquer par l'interaction d'elements de la meteorite et de rayons cosmiques corpusculaires ayant une energie de l'ordre du milliard de volts. La teneur en tritium des satellites recuperes etait plus de cent fois trop elevee pour pouvoir etre expliquee par une interaction entre les rayons cosmiques ou les particules provenant d'eruptions solaires et le satellite etudie. La forte teneur en tritium des matieres provenant des satellites est certainement due a un flux incident de particules de tritium qui sont retenues dans le satellite. (author) [Spanish] Se midieron las concentraciones de tritio en fases separadas y en muestras enteras de roca del meteorito condritico de Bruderheim, y en muestras de plomo y de hierro de satelites recuperados

Simulations about self-absorption of tritium in titanium tritide and the energy deposition in a silicon Schottky barrier diode

International Nuclear Information System (INIS)

Li, Hao; Liu, Yebing; Hu, Rui; Yang, Yuqing; Wang, Guanquan; Zhong, Zhengkun; Luo, Shunzhong

2012-01-01

Simulations on the self-absorption of tritium electrons in titanium tritide films and the energy deposition in a silicon Schottky barrier diode are carried out using the Geant4 radiation transport toolkit. Energy consumed in each part of the Schottky radiovoltaic battery is simulated to give a clue about how to make the battery work better. The power and energy-conversion efficiency of the tritium silicon Schottky radiovoltaic battery in an optimized design are simulated. Good consistency with experiments is obtained. - Highlights: ► Simulation of the energy conversion inside the radiovoltaic battery is carried out. ► Energy-conversion efficiency in the simulation shows good consistency with experimental result. ► Inadequacy of the present configuration is studied in this work and improvements are proposed.

An assembly of tritium production experiment

International Nuclear Information System (INIS)

Abe, Toshihiko

1981-01-01

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

Development of ITER Tritium Storage Material

Energy Technology Data Exchange (ETDEWEB)

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

2007-01-15

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

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

Science.gov (United States)

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

1994-09-01

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

A small and compact AMS facility for tritium depth profiling

Indian Academy of Sciences (India)

employing diamond-like carbon (DLC) stripper foils at this accelerator, another ... the switching magnet the tritium ions are counted with a surface barrier detector. .... AMS has been successfully applied to depth profiling of tritium in graphite ...

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

Gas-handling system for studies of tritium-containing materials

International Nuclear Information System (INIS)

Carstens, D.H.W.

1975-01-01

A gas handling system for preparation and study of tritium containing compounds and materials is described. The system at any one time can handle amounts of DT gas up to about 3 moles and has provisions for purification, storage, and measurement of the gas. Experimental conditions covering the ranges 20 to 800 0 C and 0.1 Pa to 137 MPa (10 -2 torr to 20,000 psi) can be maintained. (auth)

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 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

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.)

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

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

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

Tritium and helium behavior in irradiated beryllium

International Nuclear Information System (INIS)

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

1990-11-01

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

Present status of tritium research activities at universities in Japan

International Nuclear Information System (INIS)

Kawamura, K.

1983-01-01

The behaviours of tritium towards various materials are very similar to those of hydrogen, since tritium is one of the hydrogen isotope. In addition to those properties, tritium shows the radiochemical and radiological reactivities due to an emitted #betta#-ray. The permeability of tritium through various materials is the example of the former. The formation of tritiated methane in tritium stored in stainless steel vessels and the increase of helium content in tritium-bearing metallic materials are the examples of the latter. For these reasons, advanced and somewhat more complicated techniques are required for handling tritium. After the Ministry of Education, Science and Culture (MOE) made an appropriation on Grant-in-Aid for Fusion Research in 1975 year's budget, development of tritium handling technology for fusion reactors have been actively pursued. The specific experiments to be embodied in present research activities are: 1. Measurements of tritium permeation rate through various materials. 2. Fundamental studies on tritium containment materials. 3. Fundamental studies of tritium waste treatment and storage. In this paper, the works achieved under the above research activities are described and some results obtained from experiments are reported. (author)

Tritium containment of controlled thermonuclear fusion reactor

International Nuclear Information System (INIS)

Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

1979-01-01

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

Prospects of ceramic tritium breeder materials

International Nuclear Information System (INIS)

Roth, E.; Roux, N.; Conservatoire National des Arts et Metiers; CEA Centre d'Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette

1989-01-01

In this paper the authors examine the prospects of the main ceramics proposed as breeder materials for fusion reactors, i.e. Li-2O, Li-2ZrO-3, LiAlO-2, Li-4SiO-4. To do so they review terms of reference of contemplated blankets for NET, ITER and DEMO, and the proposed blanket concepts and materials. Issues respective to the use of each breeder material are examined, and from this review it is concluded that ceramics are the most favorable breeder materials whose use can be contemplated as well for a driver blanket for NET or ITER and for a DEMO blanket. Ceramics are then compared between themselves and it is seen that, subject to the confirmation of recent experimental results, lithium zirconate could be used with advantage in any of the present blanket concepts, except in those employing lithium at its natural isotopic abundance, in which case only Li-2O can be used. However in specific cases, or in parts of a blanket, other ceramics may be profitably employed. As a general conclusion suggestions are made to further improve ceramic breeder performances, and it is recommended to intensify also work on problems that have to be solved in order to operate ceramic breeder blankets e.g. tritium extraction and recovery systems and conditions of beryllium use. (author). 37 refs.; 12 tabs

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Experiments on tritium behavior in beryllium, (1)

International Nuclear Information System (INIS)

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

1989-06-01

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

Review of general tritium accountancy techniques

International Nuclear Information System (INIS)

Vassallo, G.; Engelmann, U.

1995-01-01

The accountancy of material in any facility forms an integral part of good housekeeping practices. However, for materials such as tritium, a combination of safety, security and economic reasons often demands that a comprehensive material control program be implemented. Within a tritium facility, the isotope is usually stored at a central magazine from where it can be distributed to and collected from process plant and experiments and received from external suppliers. This paper outlines the routine magazine measurement techniques employed for quantitatively assaying tritium for such control purposes and reviews the advantages and drawbacks of various methods. 10 refs., 2 figs., 2 tabs

Tritium sources; Izvori tricijuma

Energy Technology Data Exchange (ETDEWEB)

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

1993-07-01

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

A new tritium process monitor based on scintillating fibres

International Nuclear Information System (INIS)

Pacenti, P.; Edwards, R.A.H.; Monte, A. de; Campi, F.

1998-01-01

The main requirements for tritium monitoring in processes related with fusion fuel cycle are low tritium memory, fast response and accuracy, in decreasing order of importance. At present, in-line tritium monitoring in such tritium processing is done mostly using ionization chambers, which suffer a number of drawbacks: output and sensitivity depends on total gas pressure, composition and flow, etc., and have problems such as tritium memory and generally of saturation effect at high tritium concentrations. Solid scintillators can only work well with tritium if they offer a large surface area, because tritium is absorbed within the first microns of material. The present design uses entirely inorganic scintillator and construction materials, chosen to minimize tritium memory. The described on line and real time tritium detector presents some advantages in comparison with well established flow-through tritium process monitors, such as ionization chambers and thermal conductivity detectors. (authors)

Conceptual design of tritium accountancy system for LLCB TBM

International Nuclear Information System (INIS)

Patel, Rudreksh; Sircar, Amit

2017-01-01

Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) will be tested in ITER for performance evaluation of high grade of heat extraction and tritium breeding. The bred tritium in the breeder materials is extracted and recovered by Tritium Extraction System (TES), whereas tritium permeated from breeder materials to helium coolants, viz., primary coolant and secondary coolant, is recovered by Coolant Purification System (CPS). This recovered tritium has to be accounted before transferring it to tritium plant (i.e., ITER inner fuel). This tritium accountancy is performed by Tritium Accountancy System (TAS). In addition to tritium accountancy, TAS also provides necessary data for the validation of design and modelling tools.In this work, we have presented conceptual design of TAS. It also describes operational philosophy, process parameters, process flow diagram, and interface details with ITER tritium plant. (author)

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

Novel Methods of Tritium Sequestration: High Temperature Gettering and Separation Membrane Materials Discovery for Nuclear Energy Systems

Energy Technology Data Exchange (ETDEWEB)

Chen, Franglin [Univ. of South Carolina, Columbia, SC (United States); Sholl, David [Georgia Inst. of Technology, Atlanta, GA (United States); Brinkman, Kyle [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Lyer, Ratnasabapathy [Claflin Univ., Orangeburg, SC (United States); Iyer, Ratnasabapathy [Claflin Univ., Orangeburg, SC (United States); Reifsnider, Kenneth [Univ. of South Carolina, Columbia, SC (United States)

2015-01-22

This project is aimed at addressing critical issues related to tritium sequestration in next generation nuclear energy systems. A technical hurdle to the use of high temperature heat from the exhaust produced in the next generation nuclear processes in commercial applications such as nuclear hydrogen production is the trace level of tritium present in the exhaust gas streams. This presents a significant challenge since the removal of tritium from the high temperature gas stream must be accomplished at elevated temperatures in order to subsequently make use of this heat in downstream processing. One aspect of the current project is to extend the techniques and knowledge base for metal hydride materials being developed for the ''hydrogen economy'' based on low temperature absorption/desorption of hydrogen to develop materials with adequate thermal stability and an affinity for hydrogen at elevated temperatures. The second focus area of this project is to evaluate high temperature proton conducting materials as hydrogen isotope separation membranes. Both computational and experimental approaches will be applied to enhance the knowledge base of hydrogen interactions with metal and metal oxide materials. The common theme between both branches of research is the emphasis on both composition and microstructure influence on the performance of sequestration materials.

Novel Methods of Tritium Sequestration: High Temperature Gettering and Separation Membrane Materials Discovery for Nuclear Energy Systems

International Nuclear Information System (INIS)

2015-01-01

This project is aimed at addressing critical issues related to tritium sequestration in next generation nuclear energy systems. A technical hurdle to the use of high temperature heat from the exhaust produced in the next generation nuclear processes in commercial applications such as nuclear hydrogen production is the trace level of tritium present in the exhaust gas streams. This presents a significant challenge since the removal of tritium from the high temperature gas stream must be accomplished at elevated temperatures in order to subsequently make use of this heat in downstream processing. One aspect of the current project is to extend the techniques and knowledge base for metal hydride materials being developed for the ''hydrogen economy'' based on low temperature absorption/desorption of hydrogen to develop materials with adequate thermal stability and an affinity for hydrogen at elevated temperatures. The second focus area of this project is to evaluate high temperature proton conducting materials as hydrogen isotope separation membranes. Both computational and experimental approaches will be applied to enhance the knowledge base of hydrogen interactions with metal and metal oxide materials. The common theme between both branches of research is the emphasis on both composition and microstructure influence on the performance of sequestration materials.

Use of tritium in the production of selfluminous compounds

International Nuclear Information System (INIS)

Jung, Heung Suk; Ahn, Doh Heui; Baek, Seung Woo; Koo, Je Hyoo; Kook, Il Hyun; Lee, Han Soo; Kim, Kwang Lak

1994-12-01

In a Pressurized Heavy Water Reactor, about one MCi of tritium is produced annually. Tritium is a very useful resource as an essential material for selfluminous compounds. In this report, in order to manufacture selfluminous compounds by using the tritium in Wolsung nuclear power plant, the pretreatment technology of materials and the coating technology of selfluminous compounds was investigated and its raw cost was estimated. It was confirmed that tritium can be used as a very useful industrial material. 5 figs., 15 tabs., 35 refs. (Author)

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 control and accountability instructions

International Nuclear Information System (INIS)

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

1985-08-01

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

Tritium control and accountability instructions

International Nuclear Information System (INIS)

Wall, W.R.

1981-03-01

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

A low tritium hydride bed inventory estimation technique

Energy Technology Data Exchange (ETDEWEB)

Klein, J.E.; Shanahan, K.L.; Baker, R.A. [Savannah River National Laboratory, Aiken, SC (United States); Foster, P.J. [Savannah River Nuclear Solutions, Aiken, SC (United States)

2015-03-15

Low tritium hydride beds were developed and deployed into tritium service in Savannah River Site. Process beds to be used for low concentration tritium gas were not fitted with instrumentation to perform the steady-state, flowing gas calorimetric inventory measurement method. Low tritium beds contain less than the detection limit of the IBA (In-Bed Accountability) technique used for tritium inventory. This paper describes two techniques for estimating tritium content and uncertainty for low tritium content beds to be used in the facility's physical inventory (PI). PI are performed periodically to assess the quantity of nuclear material used in a facility. The first approach (Mid-point approximation method - MPA) assumes the bed is half-full and uses a gas composition measurement to estimate the tritium inventory and uncertainty. The second approach utilizes the bed's hydride material pressure-composition-temperature (PCT) properties and a gas composition measurement to reduce the uncertainty in the calculated bed inventory.

Tritium behaviour in ceramic breeder blankets

International Nuclear Information System (INIS)

Miller, J.M.

1989-01-01

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

Decontamination tests on tritium-contaminated materials

International Nuclear Information System (INIS)

Boutot, P.; Schipfer, P.

1967-01-01

These tests are designed to try out various processes liable to be applied to the decontamination of a material contaminated with tritium. The samples are thin stainless- steel slabs contaminated in the laboratory with elements extracted from industrial installations. The measurement of the initial and residual activities is carried out using an open-window BERTHOLD counter. The best results are obtained by passing a current of pre-heated (300 deg. C) air containing water vapour. This process makes it possible to reach a decontamination factor of 99.5 per cent in 4 hours. In a vacuum, the operation has to be prolonged to 100 hours in order to obtain a decontamination factor of 99.2 per cent. Wet-chemical or electrolytic treatments are efficient but their use is limited by the inherent corrosion risks. A study of the reappearance of the contamination has made it possible to observe that this phenomenon occurs whatever the process used. (authors) [fr

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

International Nuclear Information System (INIS)

Diabate, S.; Strack, S.

1990-12-01

This study gives an overview on the current knowledge about the behaviour of tritium in plants and in food chains in order to evaluate the ingestion pathway modelling of existing computer codes for dose estimations. The tritium uptake and retention by plants standing at the beginning of the food chains is described. The different chemical forms of tritium, which may be released into the atmosphere (HT, HTO and tritiated organics), and incorporation of tritium into organic material of plants are considered. Uptake and metabolism of tritiated compounds in animals and man are reviewed with particular respect to organically bound tritium and its significance for dose estimations. Some basic remarks on tritium toxicity are also included. Furthermore, a choice of computer codes for dose estimations due to chronic or accidental tritium releases has been compared with respect to the ingestion pathway. (orig.) [de

Investigations in the area of thermonuclear structural material science in the Republic of Kazakhstan

International Nuclear Information System (INIS)

Tazhibayeva, I.; Shestakov, V.; Cherepnin, Yu.S.

2001-01-01

The investigations in the area of structural materials for fusion program initiated within the framework of ITER project in the Republic of Kazakhstan are devoted basically in the following direction: to studying the behaviour of hydrogen isotopes in structural elements of the first wall and the divertor in conditions simulating real conditions of material operation, accident situations arising during steam interaction with the beryllium armour of the first wall during accidental coolant loss, to establish an experimental facility for study aspects of tritium safety of thermonuclear installations, for example, levels of tritium accumulation and release; efficiency of barrier layers and protective coating; influence of brazing and welding zones on tritium permeation. The work on determination of tritium release from lead/lithium eutectic alloy by mass-spectrometry method and the development of permeation barriers has begun. At present, work has begun to create Kazakhstan's own tokamak type reactor for investigation of the behaviour of various first wall materials and divertor plates during normal and accident conditions. The concept of spherical tokamak will be used in the construction of KTM reactor. (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

Quantitative analysis of tritium distribution in austenitic stainless steels welds

International Nuclear Information System (INIS)

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

1994-01-01

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

Tritium permeation and recovery for the Flibe/He blanket design

International Nuclear Information System (INIS)

Moir, R.

1984-10-01

This study assumes tritium to be a gas dissolved in molten salt, with TF formation suppressed. Tritium permeates readily through the hot steel tubes of the reactor and steam generator and will leak into the steam system at the rate of about one gram per day in the absence of special permeation barriers, assuming that 1% of the helium coolant flow rate is processed for tritium recovery at 90% efficiency per pass. Tritiated water in the steam system is a personnel hazard at concentration levels well below one part per million and this level would soon be reached without costly isotopic processing. Alternatively, including a combination of permeation barriers on reactor and steam generator tubes and molten salt processing is estimated to reduce the leak rate into the steam system by over two orders of magnitude. For the option with the lowest estimated leak rate, 55 Ci/d, it may be possible to purge the steam system continuously to prevent tritiated water buildup. At best, isotopic separation of dilute tritiated water may not be necessary and for higher leak-rate options the isotopic processing rate can be reduced. The proposed permeation barrier for the reactor tubes is a 10 μm layer of tungsten which, in principle, will reduce tritium blanket permeation by a factor of about 300 below the bare-steel rate

Tritium production and processing in a Tokamak reactor

International Nuclear Information System (INIS)

Leger, D.

1986-09-01

Important aspects of the tritium system in Tokamak reactors that have to be controlled are overviewed in this paper. The doubling time is one of them, that is to say the time required to produce, in addition to the tritium burned enough tritium to be able to supply the initial tritium inventory. Another one is the tritium permeation through walls. In addition to the permeation phenomena, large tritium inventories are trapped in the reactor structural material. Finally, the different atmospheres of halls, etc.., that can be contaminated with tritium, have to be reprocessed

Final characterization report for the 104-B-1 Tritium Vault and 104-B-2 Tritium Laboratory

International Nuclear Information System (INIS)

Encke, D.B.; Harris, R.A.

1996-11-01

This report is a compilation of the characterization data collected from the 104-B-1 Tritium Vault and the 104-B-2 Trillium Laboratory. The characterization activities were organized and implemented to evaluate the radiological status and identify any hazardous materials. The data contained in this report reflects the current conditions and status of the 104-B-1 Tritium Vault and 104-B-2 Tritium Laboratory. This information is intended to be utilized in support of future building decontamination and demolition, to allow for proper disposal of the demolition debris as required by the Washington Administrative Code, WAC 173-303, the Hanford Site Solid Waste Acceptance Criteria, WHC-EP-0063, and the Environmental Restoration Disposal Facility Waste Acceptance Criteria, BHI-00139. Based on the historical information and facility inspections, the only hazardous materials sampling and analysis activities necessary were to identify lead paint and asbestos containing materials (ACM) in the 104-B-1 Tritium Vault and the 104-B-2 Tritium Laboratory. Asbestos samples were obtained from the outer boundary of the roof areas to confirm the presence and type of asbestos containing fibers. Lead paint samples were obtained to confirm the presence and quantity of lead paint on the roof trim, doors and vents

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

Energy Technology Data Exchange (ETDEWEB)

Calderoni, P., E-mail: Pattrick.Calderoni@inl.gov [Fusion Safety Program, Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-7113 (United States); Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G. [Fusion Safety Program, Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-7113 (United States); Hatano, Y.; Hara, M. [Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555 (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529 (Japan); Otsuka, T.; Katayama, K. [Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Konishi, S.; Noborio, K.; Yamamoto, Y. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

2011-10-01

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

International Nuclear Information System (INIS)

Calderoni, P.; Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G.; Hatano, Y.; Hara, M.; Oya, Y.; Otsuka, T.; Katayama, K.; Konishi, S.; Noborio, K.; Yamamoto, Y.

2011-01-01

The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-04-15

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

Special purpose materials for the fusion reactor environment: a technical assessment

Energy Technology Data Exchange (ETDEWEB)

None

1978-02-01

This technology assessment considers the following areas: (1) breeding materials, (2) coolants, (3) tritium barriers, (4) graphite and silicon carbide, (5) ceramics, (6) heat-sink materials, and (7) magnet materials. Some questions and analyses forming the assessment are described. (MOW)

The formation of tritium permeation barriers by CVD

International Nuclear Information System (INIS)

Forcey, K.S.; Perujo, A.; Reiter, F.; Lolli-Ceroni, P.L.

1993-01-01

The effectiveness as permeation barriers of the following CVD coatings have been investigated: TiC (1 to 2 μm in thickness); a bi-layer of TiN on TiC (3 μm total thickness) and CVD Al 2 O 3 on a TiN/TiC bi-layer. The substrate materials were TZM (a Mo alloy) and 316L stainless steel in the form of discs of diameter 48 mm and thickness 0.1 or 1 mm. Permeation measurements were performed in the temperature range 515-742 K using deuterium at pressures in the range 1-50 kPa. CVD layers were shown to form reasonably effective permeation barriers. At a temperature of 673 K TiC is around 6000 times less permeable to deuterium than 316L stainless steel. (orig.)

Material Barriers to Diffusive Mixing

Science.gov (United States)

Haller, George; Karrasch, Daniel

2017-11-01

Transport barriers, as zero-flux surfaces, are ill-defined in purely advective mixing in which the flux of any passive scalar is zero through all material surfaces. For this reason, Lagrangian Coherent Structures (LCSs) have been argued to play the role of mixing barriers as most repelling, attracting or shearing material lines. These three kinematic concepts, however, can also be defined in different ways, both within rigorous mathematical treatments and within the realm of heuristic diagnostics. This has lead to a an ever-growing number of different LCS methods, each generally identifying different objects as transport barriers. In this talk, we examine which of these methods have actual relevance for diffusive transport barriers. The latter barriers are arguably the practically relevant inhibitors in the mixing of physically relevant tracers, such as temperature, salinity, vorticity or potential vorticity. We demonstrate the role of the most effective diffusion barriers in analytical examples and observational data. Supported in part by the DFG Priority Program on Turbulent Superstructures.

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

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

International Nuclear Information System (INIS)

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

1993-01-01

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 19 ions/cm 2 · s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment

Tritium stripping in a nitrogen glovebox using SAES St 198

International Nuclear Information System (INIS)

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

1994-01-01

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

The permeation of tritium through 316L stainless steel with multiple coatings

International Nuclear Information System (INIS)

Yao Zhenyu; Hao Jiakun; Zhou Changshan; Shan Changqi; Yu Jinnan

2000-01-01

TiN + TiC + TiN and TiN + TiC + SiO 2 films were deposited on the surface of 316L austenitic stainless steel by means of physical vapor deposition (PVD). The thickness of the films was about 2-3 μm. The film is compact, oxidation-resistant, and has good adherence with the substrate below 500 deg. C. Tritium gas permeation of 316L with multiple films was examined, and it was found that the tritium permeability in 316L with a TiN + TiC + TiN film was 4-5 orders of magnitude lower, and in 316L with a TiN + TiC + SiO 2 film was 4-6 orders of magnitude lower than that in 316L with a Pd film at about 200-500 deg. C. At about 600 deg. C, the permeability of 316L with the multiple coating was 3-4 orders of magnitude lower than that in 316L with a Pd film. The result shows that the tritium permeation barrier is formed by multiple coating above 300 deg. C, and it is stable below 500 deg. C. However, the barrier is partly destroyed at about 600 deg. C because of oxidation; although this results in degradation of the barrier, it still plays a positive role. These films may be useful as coatings for the first wall, tritium blanket, and heat exchanger in fusion reactors for tritium permeation resistance

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)

Transfer of tritium-labeled organic material from grass into cow's milk

International Nuclear Information System (INIS)

van den Hoek, J.; ten Have, M.H.J.; Gerber, G.B.; Kirchmann, R.

1985-01-01

Two lactating cows were given tritiated hay containing organically bound tritium (OBT) only for about 4 weeks. Tritium activity was determined in milk fat, casein, lactose, milk water, and whole milk. In one cow, milk was sampled for approximately 450 days, covering two lactation periods. At steady state, specific tritium activities in casein, lactose, and milk water were 58, 10, and 11%, respectively, of those in milk fat. Some OBT was converted into THO during catabolism and entered the body water pool. This 3 H source accounted for nearly 40% of tritium in lactose, but in casein and milk fat about 97% of tritium was derived from ingested OBT. Comparison of the specific activity of milk constituents with the specific activity of ingested hay showed the following values: 0.84 for milk fat, 0.49 for casein, 0.05 for lactose, 0.10 for milk water. Decrease of tritium activity with time could be represented by three components with different half-lives for the organic milk constituents. Those for milk fat and casein were quite similar, with a slow component of nearly 3 months

Tritium compatibility of alumina and Fosterite

Energy Technology Data Exchange (ETDEWEB)

Coffin, D.O.

1979-09-01

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

Disposal of tritium-exposed metal hydrides

International Nuclear Information System (INIS)

Nobile, A.; Motyka, T.

1991-01-01

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

Thin film tritium dosimetry

Science.gov (United States)

Moran, Paul R.

1976-01-01

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

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

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

International Nuclear Information System (INIS)

Duan Shaojie

1998-03-01

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

Hydrogen permeation preventive structural materials

International Nuclear Information System (INIS)

Fukushima, Kimichika; Nakahigashi, Shigeo; Imura, Masashi; Terasawa, Michitaka; Ebisawa, Katsuyuki.

1986-01-01

Purpose: To provide highly practical wall materials for use in thermonuclear reactors capable of effectively preventing the permeation of hydrogen isotopes such as tritium thereby preventing the contamination of coolants. Constitution: Helium gas is injected into or at the surface of base materials comprising stainless steel plates to form a helium gas region. Alternatively, boron, nitrogen or the compound thereof having a greater helium forming nuclear reaction cross section than that of the base materials is mixed or injected into the base material to form the helium gas region through (n,α) reaction under neutron irradiation. Since the helium gas region constitutes a diffusion barrier for the tritium as the hydrogen isotope, the permeation amount of tritium is significantly suppressed. Helium gas bubbles or lattice defects are formed in the helium gas region under the neutron irradiation, by which the hydrogen isotope capturing effect can also be effected. In this way, permeation of the hydrogen isotope, contamination of the coolants, etc. can be prevented to provide great practical effectives. (Kawakami, Y.)

Solid tritium breeder materials-Li2O and LiAlO2: a data base review

International Nuclear Information System (INIS)

Liu, Y.Y.; Billone, M.C.; Clemmer, R.G.; Fischer, A.K.; Hollenberg, G.W.; Tam, S.W.

1985-01-01

The fabrication, properties, and irradiation behavior of Li 2 O and γ-LiAlO 2 are reviewed and assessed to determine the potential of these materials to satisfy the basic solid breeder blanket performance requirements. Based on the data analysis and theoretical modeling, a set of major technical uncertainties is identified. These uncertainties include: fabricability of sphere-pac solid breeders; high fluence and burnup effects on thermal conductivity and microstructural stability; high fluence and burnup effects on tritium diffusion coefficients at low temperature; relationship among purge flow chemistry, surface adsorption, and species of released tritium; and mechanical properties and the loads imposed on the structural materials by the breeder during blanket operation. Resolution of these issues is important in assuring that solid breeder blankets can be designed with confidence

A uranium bed with ceramic body for tritium storage

Energy Technology Data Exchange (ETDEWEB)

Khapov, A.S.; Grishechkin, S.K.; Kiselev, V.G. [' All Russia Research Institute of Automatics' - FSUE VNIIA, Moscow (Russian Federation)

2015-03-15

It is widely recognized that ceramic coatings provide an attractive solution to lower tritium permeation in structural materials. Alumina based ceramic coatings have the highest permeation reduction factor for hydrogen. For this reason an attempt was made to apply crack-free low porous ceramics as a structural material of a bed body for tritium storage in a setup used for hydrogenating neutron tube targets at VNIIA. The present article introduces the design of the bed. This bed possesses essentially a lower hydrogen permeation factor than traditionally beds with stainless steel body. Bed heating in order to recover hydrogen from the bed is suggested to be implemented by high frequency induction means. Inductive heating allows decreasing the time necessary for tritium release from the bed as well as power consumption. Both of these factors mean less thermal power release into glove box where a setup for tritium handling is installed and thus causes fewer problems with pressure regulations inside the glove box. Inductive heating allows raising tritium sorbent material temperature up to melting point. The latter allows achieving nearly full tritium recovery.

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.)

Tritium extraction from neutron-irradiated lithium aluminate

International Nuclear Information System (INIS)

Garcia H, F.

1995-01-01

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

Modeling tritium transport in the environment

International Nuclear Information System (INIS)

Murphy, C.E. Jr.

1986-01-01

A model of tritium transport in the environment near an atmospheric source of tritium is presented in the general context of modeling material cycling in ecosystems. The model was developed to test hypotheses about the process involved in tritium cycling. The temporal and spatial scales of the model were picked to allow comparison to environmental monitoring data collected in the vicinity of the Savannah River Plant. Initial simulations with the model showed good agreement with monitoring data, including atmospheric and vegetation tritium concentrations. The model can also simulate values of tritium in vegetation organic matter if the key parameter distributing the source of organic hydrogen is varied to fit the data. However, because of the lack of independent conformation of the distribution parameter, there is still uncertainty about the role of organic movement of tritium in the food chain, and its effect on the dose to man

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

Directory of Open Access Journals (Sweden)

H. Ushida

2016-12-01

Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

The LLNL portable tritium processing system

International Nuclear Information System (INIS)

Anon.

1995-01-01

The end of the Cold War significantly reduced the need for facilities to handle radioactive materials for the US nuclear weapons program. The LLNL Tritium Facility was among those slated for decommissioning. The plans for the facility have since been reversed, and it remains open. Nevertheless, in the early 1990s, the cleanup (the Tritium Inventory Removal Project) was undertaken. However, removing the inventory of tritium within the facility and cleaning up any pockets of high-level residual contamination required that we design a system adequate to the task and meeting today's stringent standards of worker and environmental protection. In collaboration with Sandia National Laboratory and EG ampersand G Mound Applied Technologies, we fabricated a three-module Portable Tritium Processing System (PTPS) that meets current glovebox standards, is operated from a portable console, and is movable from laboratory to laboratory for performing the basic tritium processing operations: pumping and gas transfer, gas analysis, and gas-phase tritium scrubbing. The Tritium Inventory Removal Project is now in its final year, and the portable system continues to be the workhorse. To meet a strong demand for tritium services, the LLNL Tritium Facility will be reconfigured to provide state-of-the-art tritium and radioactive decontamination research and development. The PTPS will play a key role in this new facility

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Tritium migration in nuclear desalination plants

International Nuclear Information System (INIS)

Muralev, E.D.

2003-01-01

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

Effects of interfering constituents on tritium smears

International Nuclear Information System (INIS)

Levi, G.D. Jr.; Cheeks, K.E.

1993-01-01

Tritium smears are performed by Health Protection Operations (HPO) to assess transferable contamination on work place surfaces, materials for movement outside Radiologically Controlled Areas (RCA), and product containers being shipped between facilities. Historically, gas proportional counters were used to detect transferable tritium contamination collected by smearing. Because tritium is a low-energy beta emitter, gas proportional counters do not provide the sensitivity or the counting efficiency to accurately measure the tritium activity on the smear. Liquid Scintillation Counters (LSC) provide greater counting efficiency for the low-energy beta particles along with greater reliability and reproducibility compared to gas flow proportional counters. The purpose of this technical evaluation was to determine the effects of interfering constituents such as filters, dirt and oil on the counting efficiency and tritium recoveries of tritium smears by LSC

Transfer of fallout tritium from environment to human body

International Nuclear Information System (INIS)

Hisamatsu, Shun-ichi; Takizawa, Yukio

1989-01-01

A large quntity of tritium will be used as a fuel of nuclear fusion in the future. It is, therefore, considered important to elucidate tritium behavior present in the environment and the process of tritium transfer from the environment to the human body. Fallout tritium is an applicable material in searching for the long term behavior of tritium in the environment. This paper focuses on the American, Italian, Japanese literature concerning fallout tritium in food and in the human body. The specific activity ratio of bound to free tritium poses an important problem. The mechanism of biological concentration must await further studies. (N.K.) 63 refs

Degradation of elastomers by tritium beta radiation

International Nuclear Information System (INIS)

Zapp, P.E.; Tuer, G.L. Jr.

1984-01-01

Based on its tritium radiation resistance, ethylene propylene rubber has been selected as a candidate for replacement of nitrile rubber in the SRP tritium facilities. A specification for flange gasket material has been developed for ethylene propylene such that its mechanical properties are similar to those of nitrile rubber. In-process testing of ethylene propylene and nitrile gaskets will be conducted in the tritium facilities under identical exposure conditions

Tritium permeation losses in HYLIFE-II heat exchanger tubes

International Nuclear Information System (INIS)

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

1990-01-01

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

Tritium transport in HCLL and WCLL DEMO blankets

Energy Technology Data Exchange (ETDEWEB)

Candido, Luigi [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Utili, Marco [ENEA UTIS- C.R. Brasimone, Bacino del Brasimone, Camugnano, BO (Italy); Nicolotti, Iuri [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy); Zucchetti, Massimo, E-mail: massimo.zucchetti@polito.it [DENERG, Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy)

2016-11-01

Highlights: • Tritium inventories and tritium losses are the main output of the presented model for HCLL and WCLL. • A parametric study has been performed, to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and/or losses. • An improved design is needed, in order to reduce the radiological hazard related to tritium activity. According to test number 7, HCLL-BB could be able to have a tritium inventory of 33.05 g and losses of 19.55 Ci/d. • WCLL-BB shows a very low radiological risk, much lower than that suggested (inventory: 17.48 g, losses: 3.2 Ci/d). An ptimization study has been performed aiming to minimize the water flow rate for an upgraded design. • Both for HCLL and WCLL, the most critical parameters able to produce relevant variations in inventories and losses are the helium/water fraction, the CPS/WDS and the permeation reduction factors. - Abstract: The Helium-Cooled Lithium Lead (HCLL) and Water-Cooled Lithium Lead (WCLL) Breeding Blankets are two of the four blanket designs proposed for DEMO reactor. The study of tritium transport inside the blankets is fundamental to assess their preliminary design and safety features. A mathematical model has been derived, in a new form making makes easier to determine the most critical components as far as tritium losses and tritium inventories are concerned, and to model the tritium performance of the whole system. Two cases have been studied, the former with tritium generation rate constant in time and the latter considering a typical pulsed operation for a time span of 100 h. Tritium inventories and tritium losses are the main output of the model. Tritium concentrations, inventories and losses are initially calculated and compared for the two blankets, in a reference case without permeation barriers or cold traps. A parametric study to show the behavior of the two systems when certain parameters are changed, in order to minimize inventories and

Experiments on tritium behavior in beryllium, (2)

International Nuclear Information System (INIS)

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

1990-02-01

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

Tritium immobilization and packaging using metal hydrides

International Nuclear Information System (INIS)

Holtslander, W.J.; Yaraskavitch, J.M.

1981-04-01

Tritium recovered from CANDU heavy water reactors will have to be packaged and stored in a safe manner. Tritium will be recovered in the elemental form, T 2 . Metal tritides are effective compounds in which to immobilize the tritium as a stable non-reactive solid with a high tritium capacity. The technology necessary to prepare hydrides of suitable metals, such as titanium and zirconium, have been developed and the properties of the prepared materials evaluated. Conceptual designs of packages for containing metal tritides suitable for transportation and long-term storage have been made and initial testing started. (author)

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)

Economic evaluation of closure cap barrier materials study

Energy Technology Data Exchange (ETDEWEB)

Serrato, M.G.; Bhutani, J.S.; Mead, S.M.

1993-09-01

Volume II of the Economic Evaluation of the Closure Cap Barrier Materials, Revision I contains detailed cost estimates for closure cap barrier materials. The cost estimates incorporate the life cycle costs for a generic hazardous waste seepage basin closure cap under the RCRA Post Closure Period of thirty years. The economic evaluation assessed six barrier material categories. Each of these categories consists of several composite cover system configurations, which were used to develop individual cost estimates. The information contained in this report is not intended to be used as a cost estimating manual. This information provides the decision makers with the ability to screen barrier materials, cover system configurations, and identify cost-effective materials for further consideration.

Economic evaluation of closure cap barrier materials study

International Nuclear Information System (INIS)

Serrato, M.G.; Bhutani, J.S.; Mead, S.M.

1993-09-01

Volume II of the Economic Evaluation of the Closure Cap Barrier Materials, Revision I contains detailed cost estimates for closure cap barrier materials. The cost estimates incorporate the life cycle costs for a generic hazardous waste seepage basin closure cap under the RCRA Post Closure Period of thirty years. The economic evaluation assessed six barrier material categories. Each of these categories consists of several composite cover system configurations, which were used to develop individual cost estimates. The information contained in this report is not intended to be used as a cost estimating manual. This information provides the decision makers with the ability to screen barrier materials, cover system configurations, and identify cost-effective materials for further consideration

Tritium control: October 1982-March 1983

International Nuclear Information System (INIS)

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

1983-01-01

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

Method and plant to remote tritium from the cooling water of a nuclear reactor

International Nuclear Information System (INIS)

O'Brien, C.J.

1976-01-01

A method is proposed for the extraction of tritium from the cooling water of a nuclear reactor, based on the principle of concentrating the tritium by a multi-stage transfer process. The cooling water is brought into contact in each stage with basic, labile, hydrogen-containing material with high pH value, whereby the tritium is transfered into an intermediate solid product and can be separated off. The technical details of the plant are described. Cellulose materials, such as cotton and wood as well as protein-containing material, such as muscle tissue are mentioned as examples of materials with a high affinity to tritium, greater than the affinity of water to tritium. They extract tritium from the cooling water. (HK) [de

Control of the tritium path in process heat HTR's

International Nuclear Information System (INIS)

Kirch, N.; Scheidler, G.

1985-01-01

Nuclear Process Heat plant converting fossil fuels into liquid or gaseous secondary energy carriers generate tritium by several nuclear reactions. Control of the tritium path through the walls of the heat exchanger is highly important to meet regulatory requirements on the acceptable contamination in the product gas or liquid. Therefore, significant effort in the project 'Prototypanlage Nukleare Prozesswaerme' was put not only into generating a data base, but also into means of reducing tritium generation and permeation. Clean graphites with lithium impurities in the ppb level provide a low tritium source term. Realistic modeling of graphite retention and special helium purification systems are essentials. The main barrier to tritium permeation are heat exchanger walls requiring detailed characterization of in-situ surface layers. Studies to optimize the water/steam mass flow in the conversion process offer possibilities for further tritium retention. Progress can be demonstrated as follows: In 1980, between 2 and 8 Bq tritium per gram of product were predicted based on available data and even higher concentrations during startup. However, present day validated code predictions are below required 0.5 Bq/g equilibrium concentration level. During transients - particularly startup - this limit cannot be guaranteed as yet, but further retention potential is being offered by tritium gettering or filtering. An expected increase of the German regulatory requirement to 5 Bq/g will easily be met by present plant design under all operational conditions. (author)

Advanced β-ray-induced X-ray spectrometry for non-destructive measurement of tritium retained in fusion related materials

Energy Technology Data Exchange (ETDEWEB)

Matsuyama, Masao, E-mail: matsu3h@ctg.u-toyama.ac.jp; Abe, Shinsuke

2016-11-01

Highlights: • A new measurement system to measure low-Z elements such as C and O atoms has been constructed for evaluation of tritium trapped by these elements. - Abstract: A new β-ray-induced X-ray measurement system equipped with a silicon drift detector, which was named “Advanced-BIXS”, was constructed to study in detail retention behavior of surface tritium by measurements of low energy X-rays below 1 keV such as C(K{sub α}) and O(K{sub α}) as well as high energy X-rays induced by β-rays from tritium. In this study, basic performance of the present system has been examined using various tritium-containing samples. It was seen that energy linearity, energy resolution and sensitivity were quite enough for measurements of low energy X-rays induced by β-rays. Intensity of characteristic X-rays emitted from the surface and/or bulk of a tritium-containing sample was lowered by argon used as a working gas of the Advanced-BIXS. Pressure dependence of transmittance of C(K{sub α}) and Fe(K{sub α}) was examined as examples of low and high energy X-rays, and it was able to represent by using the mass absorption coefficient in argon. It was concluded, therefore, that the present system has high potentiality for nondestructive measurements of tritium retained in surface layers and/or bulk of fusion related materials.

Tritium release from fast neutron irradiated boron carbide

International Nuclear Information System (INIS)

Hollenberg, G.W.

1977-01-01

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

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...

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.

Comparison of the leading candidate combinations of blanket materials, thermodynamic cycles, and tritium systems for full scale fusion power plants

International Nuclear Information System (INIS)

Fraas, A.P.

1975-01-01

The many possible combinations of blanket materials, tritium generation and recovery systems, and power conversion systems were surveyed and a comprehensive set of designs were generated by using a common set of ground rules that include all of the boundary conditions that could be envisioned for a full-scale commercial fusion power plant. Particular attention was given to the effects of blanket temperature on power plant cycle efficiency and economics, the interdependence of the thermodynamic cycle and the tritium recovery system, and to thermal and pressure stresses in the blanket structure. The results indicate that, of the wide variety of systems that have been considered, the most promising employs lithium recirculated in a closed loop within a niobium blanket structure and cooled with boiling potassium or cesium. This approach gives the simplest and lowest cost tritium recovery system, the lowest pressure and thermal stresses, the simplest structure with the lowest probability of a leak, the greatest resistance to damage from a plasma energy dump, and the lowest rate of plasma contamination by either outgassing or sputtering. The only other blanket materials combination that appears fairly likely to give a satisfactory tritium generation and recovery system is a lithium-beryllium fluoride-Incoloy blanket, and even this system involves major uncertainties in the effectiveness, size, and cost of the tritium recovery system. Further, the Li 2 BeF 4 blanket system has the disadvantage that the world reserves of beryllium are too limited to support a full-blown fusion reactor economy, its poor thermal conductivity leads to cooling difficulties and a requirement for a complex structure with intricate cooling passages, and this inherently leads to an expansive blanket with a relatively high probability of leaks. The other blanket materials combinations yield even less attractive systems

Tritium breeding experiments with lithium titanate in thermal-type mockups

International Nuclear Information System (INIS)

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

2004-01-01

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

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

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.)

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.

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

Energy Technology Data Exchange (ETDEWEB)

Kurinskiy, P., E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Rolli, R. [Karlsruhe Institute of Technology, Institute for Applied Materials – Materials and Biomechanics (IAM-WBM), P.O. Box 3640, Karlsruhe 76021 (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

2014-10-15

Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA.

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

Lay out and materials for in pile tritium transport testing of breeder-inside-tube pin assemblies

International Nuclear Information System (INIS)

Alvani, C.; Casadio, S.; Mancini, M.R.; Nannetti, C.A.; Avon, J.; Ravel, S.; Pruzzo, G.; Terrosi; Roux, N.; Terlain, A.; and others.

1994-01-01

An irradiation experiment (90 FPD in SILOE reactor) has been designed in order to evaluate the in-situ effect of red-ox power of sweeping gas (helium with 100 vpm of H 2 /H 2 O with relative concentrations varying from pure H 2 to pure H 2 O ) on tritium removal from LiAlO 2 and Li 2 ZrO 3 ; and tritium permeation through AlSl-316L SS tubes with bare and coated surfaces. The conditions and materials explored were selected in order to test possible improvements with respect to critical issues for the 'Breeder Inside Tube' (BIT) blanket concept development. (author) 6 refs.; 4 figs.; 2 tabs

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

International Nuclear Information System (INIS)

Noga, J.O.

1981-11-01

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

A study of electrolytic tritium production

International Nuclear Information System (INIS)

Storms, E.K.; Talcott, C.L.

1990-01-01

Tritium production is being investigated using cathodes made from palladium and its alloys with various surface treatments. Three anode materials have been studied as well as different impurities in the electrolyte. Tritium has been produced in about 10% of the cells studied but there is, as yet, no pattern of behavior that would make the effect predictable. 15 refs., 4 figs., 6 tabs

Comparison of inventory of tritium in various ceramic breeder blankets

International Nuclear Information System (INIS)

Nishikawa, M.; Beloglazov, S.; Nakashima, N.; Hashimoto, K.; Enoeda, M.

2002-01-01

It has been pointed out by the present authors that it is essential to understand such mass transfer steps as diffusion of tritium in the grain of breeder material, absorption of water vapor into bulk of the grain, and adsorption of water on surface of the grain, together with the isotope exchange reaction between hydrogen in purge gas and tritium on surface of breeder material and the isotope exchange reaction between water vapor in purge gas and tritium on surface, for estimation of the tritium inventory in a uniform ceramic breeder blanket under the steady-state condition. It has been also pointed out by the present authors that the water formation reaction on the surface of ceramic breeder materials at introduction of hydrogen can give effect on behavior of bred tritium and lithium transfer in blanket. The tritium inventory for various ceramic breeder blankets are compared in this study basing on adsorption capacity, absorption capacity, isotope exchange capacity, and isotope exchange reactions on the Li 2 O, LiAlO 2 , Li 2 ZrO 3 , Li 4 SiO 4 and Li 2 TiO 3 surface experimentally obtained by the present authors. Effect of each mass transfer steps on the shape of release curve of bred tritium at change of the operational conditions is also discussed from the observation at out pile experiment in KUR. (orig.)

F4E R and D programme and results on in-vessel dust and tritium

International Nuclear Information System (INIS)

Le Guern, F.; Gulden, W.; Ciattaglia, S.; Counsell, G.; Bengaouer, A.; Brinster, J.; Dabbene, F.; Denkevitz, A.; Jordan, T.; Kuznetsov, M.; Porfiri, M.T.; Redlinger, R.; Roblin, Ph.; Roth, J.; Segre, J.; Sugiyama, K.; Tkatschenko, I.; Xu, Z.

2011-01-01

In a Tokamak vacuum vessel, plasma-wall interactions can result in the production of radioactive dust and H isotopes (including tritium) can be trapped both in in-vessel material and in dust. The vacuum vessel represents the most important confinement barrier to this radioactive material. In the event of an accident involving ingress of steam to the vacuum vessel, hydrogen could be produced by chemical reactions with hot metal and dust. Hydrogen isotopes could also be desorbed from in-vessel components, e.g. cryopumps. In events where an ingress of air to the vacuum vessel occurs, reaction of the air with hydrogen and/or dust therefore cannot be completely excluded. Due to the radiological risks highlighted by the safety evaluation studies for ITER in normal conditions (e.g. in-vessel maintenance chronic release) and accidental ones (e.g. challenge of vacuum vessel tightness in the event of a hydrogen/dust explosion with air), limitations on the accumulation of dust and tritium in the vacuum vessel are imposed as well as controls over the maximum extent of the quantity of accidental air ingress. ITER IO has defined a strategy for the control of in-vessel dust and tritium inventories below the safety limits based primarily on the measurement and removal of dust and tritium. In this context, this paper will report on the efforts under F4E responsibility to develop a number of the new ITER baseline systems. In particular this paper, after a review of safety constraints and ITER strategy, provides the status of: (1) tasks being launched on diagnostics for in-vessel dust inventory measurement, (2) experiments to enrich the data about the effectiveness of desorption of tritium from Be at 350 o C (divertor baking aiming to release significant amount of tritium trapped in Be co-deposit), (3) on-going R and D programme (experimental and numerical simulation) at FZK, CEA and ENEA on in-vacuum vessel H2 dust explosion.

Special purpose materials for fusion application

International Nuclear Information System (INIS)

Scott, J.L.; Clinard, F.W. Jr.; Wiffen, F.W.

1984-01-01

Originally in 1978 the Special Purpose Materials Task Group was concerned with tritium breeding materials, coolants, tritium barriers, graphite and silicon carbide, ceramics, heat-sink materials, and magnet components. Since then several other task groups have been created, so now the category includes only materials for superconducting magnets and ceramics. For the former application copper-stabilized Nb 3 Sn (Ti) insulated with polyimides will meet the general requirements, so that testing of prototype components is the priority task. Ceramics are required for several critical components of fusion reactors either as dielectrics or as a structural material. Components near the first wall will receive exposures of 5 to 20 MW.year/m"2. Other ceramic applications are well behind the first wall, with lower damage levels. Most insulators operate near room temperature, but ceramic blanket structures may operate up to 1000 0 C. Because of a meager data base, one cannot identify optimum ceramics for structural application; but MgAl 2 O 4 is an attractive dielectric material

Tritium target performance during an LBLOCA in a PWR

International Nuclear Information System (INIS)

Reid, B.D.

1996-01-01

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

Current understanding of organically bound tritium (OBT) in the environment.

Science.gov (United States)

Kim, S B; Baglan, N; Davis, P A

2013-12-01

It has become increasingly recognized that organically bound tritium (OBT) is the more significant tritium fraction with respect to understanding tritium behaviour in the environment. There are many different terms associated with OBT; such as total OBT, exchangeable OBT, non-exchangeable OBT, soluble OBT, insoluble OBT, tritiated organics, and buried tritium, etc. A simple classification is required to clarify understanding within the tritium research community. Unlike for tritiated water (HTO), the environmental quantification and behaviour of OBT are not well known. Tritiated water cannot bio-accumulate in the environment. However, it is not clear whether or not this is the case for OBT. Even though OBT can be detected in terrestrial biological materials, aquatic biological materials and soil samples, its behaviour is still in question. In order to evaluate the radiation dose from OBT accurately, further study will be required to understand OBT measurements and determine OBT fate in the environment. The relationship between OBT speciation and the OBT/HTO ratio in environmental samples will be useful in this regard, providing information on the previous tritium exposure conditions in the environment and the current tritium dynamics. Crown Copyright © 2013. Published by Elsevier Ltd. All rights reserved.

Design options to minimize tritium inventories at Savannah River

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Design options to minimize tritium inventories at Savannah River

International Nuclear Information System (INIS)

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

2016-01-01

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

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

The measurement of tritium in Canadian food items

International Nuclear Information System (INIS)

Brown, R.M.

1995-03-01

Food items locally grown near Perth, Ontario and grocery store produce and locally grown items from the Pickering-Ajax area in the vicinity of the Pickering Nuclear Generating Station (PNGS) have been analyzed for free water tritium (HTO) and organically bound tritium (OBT). The technique of measuring 3 He ingrowth in samples by mass spectrometry has been used because of its sensitivity and freedom from opportunity for contamination during processing and measurement. Concentrations observed at each site were of the order expected on the basis of known levels of tritium in the local atmosphere and precipitation. There was considerable variation between different materials and limited correlation between materials of a single type. (author). 10 refs., 8 tabs., 4 figs

Tritium permeation in fusion reactors: INTOR

International Nuclear Information System (INIS)

Baskes, M.I.; Bauer, W.; Kerst, R.A.; Swansiger, W.A.; Wilson, K.L.

1981-12-01

Tritium permeation through the first wall of advanced fusion reactors is examined. A fraction of the D-T which bombards the first wall as charge exchange neutral particles will permeate through the first wall and enter the coolant. Calculations of the steady state permeation rate for the US INTOR Tokamak design result in values of less than or equal to 0.002 grams of tritium per day under the most favorable conditions. For unfavorable surface conditions the rate is greater than or equal to 0.1 g/day. The magnitude of these permeation rates is critically dependent on the temperatures and surface conditions of the wall. The introduction of permeation barriers at the wall-coolant interface can significantly reduce permeation rates and hence may be desirable for reactor applications

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Tritium interactions of potential importance to fusion reactor systems: technology requirements

International Nuclear Information System (INIS)

Wilkes, W.R.

1976-01-01

The tritium technology requirements created by the controlled thermonuclear research program to develop a demonstration fusion power reactor by the year 2000 are reviewed. It is found that the majority of the technological advances which are needed to ensure adequate tritium containment in a tritium breeding power reactor need to be demonstrated on a pilot scale by approximately 1983, so that they may be incorporated into EPR-II, the second of two planned experimental power reactors. The most important advances include development of containment materials with permeabilities to tritium well below measured values for stainless steel; large scale, low inventory deuterium-tritium separation systems; and improved monitoring and assay systems. There are less critical requirements for information about the effects of tritium and helium on the mechanical properties of materials, the effects of tritium on biological systems, and data on physical and chemical properties of tritium. Substantial progress needs to be made on these problems early enough to permit possible solutions to be tested on EPR-I. In addition, major improvements in tritium handling equipment are required for EPR-I. Those technological problems for which solutions have not yet been demonstrated by EPR-II must be solved by 1989 if they are to be assured successful application in the demonstration reactor

Tritium production in fusion reactors

International Nuclear Information System (INIS)

Roth, E.

1981-08-01

The present analyses on the possibilities of extracting tritium from the liquid and solid fusion reactor blankets show up many problems. A consistent ensemble of materials and devices for extracting the heat and the tritium has not yet been integrated in a fusion reactor blanket project. The dimensioning of the many pipes required for shifting the tritium can only be done very approximately and the volume taken up by the blanket is difficult to evaluate, etc. The utilization of present data leads to over-dimensioning the installations by prudence and perhaps rejecting the best solutions. In order to measure the parameters of the most promising materials, work must be carried out on well defined samples and not only determine the base physical-chemical coefficients, such as thermal conductivity, scattering coefficients, Sievert parameters, but also the kinetic parameters conventional in chemical engineering, such as the hourly space rates of degassing. It is also necessary to perform long duration experiments under radiation and at operating temperatures, or above, in order to study the ageing of the bodies employed [fr

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

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

International Nuclear Information System (INIS)

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

1983-08-01

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

TRIO-01 experiment: in-situ tritium recovery results

International Nuclear Information System (INIS)

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

1983-10-01

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

Development of organic tritium light technology at Ontario Hydro

International Nuclear Information System (INIS)

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

1992-01-01

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

Breeding blanket development. Tritium release from breeder

International Nuclear Information System (INIS)

Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-28

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

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

Assessment of the significance of organically-bound tritium in environmental materials

International Nuclear Information System (INIS)

Brown, R.M.

1988-09-01

The present state of knowledge of the significance, with respect to dose, or organically-bound tritium (OBT) in diet items has been reviewed. Ratios of the specific activity of the OBT to that of the free water (HTO) in foodstuffs have been commonly reported in the range of 1 to 4. A metabolism model of Etnier, Travis and Hetrick that takes direct assimilation of food OBT into account indicates that such levels result in a dose two to three times greater than that calculated solely on the basis of body water tritium content. Very high OBT/HTO values reported by Italian studies on food items are discounted. It is recommended that OBT/HTO measurements be done on Canadian diet items and that tritium metabolism models be more thoroughly evaluated. 71 refs

Assessment of database for interaction of tritium with ITER plasma facing materials

International Nuclear Information System (INIS)

Dolan, T.J.; Anderl, R.A.

1994-09-01

The present work surveys recent literature on hydrogen isotope interactions with Be, SS and Inconels, Cu, C, and V, and alloys of Cu and V. The goals are (1) to provide input to the International Thermonuclear Experimental Reactor (ITER) team to help with tritium source term estimates for the Early Safety and Environmental Characterization Study and (2) to provide guidance for planning additional research that will be needed to fill gaps in the present materials database. Properties of diffusivity, solubility, permeability, chemical reactions, Soret effect, recombination coefficient, surface effects, trapping, porosity, layered structures, interfaces, and oxides are considered. Various materials data are tabulated, and a matrix display shows an assessment of the quality of the data available for each main property of each material. Recommendations are made for interim values of diffusivity and solubility to be used, pending further discussion by the ITER community

Tritium storage metal-bed pyrophoricity measurements

International Nuclear Information System (INIS)

Longhurst, G.R.; Neilson, R.M. Jr.; Porter, L.J.

1987-01-01

A safety concern for metal-bed tritium storage systems is the possibility of spontaneous combustion and/or explosion if the bed is accidentally exposed to air. This may result in the dispersion of tritium or tritiated compounds. Of several materials being considered for use in tritium storage beds, uranium (U), zirconium-cobalt (ZrCo), and lanthanum-nickel aluminide (LaNi 5-x Al x ) are of particular interest. It is well known that uranium that has been activated by cycles of hydriding and dehydriding is extremely pyrophoric when exposed to air or other oxidizers. Uranium hydride has also been found to be mildly pyrophoric, but less is known about the pyrophoric natures of the hydrides of the other materials. An experiment is in progress to evaluate the pyrophoric response of these materials and their hydrides and deuterides in air. Small (<100 mg) samples of depleted uranium were hydrided and then exposed to atmospheres of air, oxygen, or nitrogen using a thermogravimetric analyzer to monitor the sample weight and temperature. There was not an immediate pyrophoric response at room temperature, but ignition occured at moderately elevated temperatures for air and oxygen atmospheres. The experimental apparatus has been upgraded, and tests are continuing on these materials

Effect of Tritium on Cracking Threshold in 7075 Aluminum

Energy Technology Data Exchange (ETDEWEB)

Duncan, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Morgan, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-02-28

The effect of long-term exposure to tritium gas on the cracking threshold (K_TH) of 7075 Aluminum Alloy was investigated. The alloy is the material of construction for a cell used to contain tritium in an accelerator at Jefferson Laboratory designed for inelastic scattering experiments on nucleons. The primary safety concerns for the Jefferson Laboratory tritium cell is a tritium leak due to mechanical failure of windows from hydrogen isotope embrittlement, radiation damage, or loss of target integrity from accidental excessive beam heating due to failure of the raster or grossly mis-steered beam. Experiments were conducted to investigate the potential for embrittlement of the 7075 Aluminum alloy from tritium gas.

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

Tritium inventory and permeation in liquid breeder blankets

International Nuclear Information System (INIS)

Reiter, F.

1990-01-01

This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

Tritium processing using metal hydrides

International Nuclear Information System (INIS)

Mallett, M.W.

1986-01-01

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

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)

Investigation of tritium release and retention in lithium aluminate

International Nuclear Information System (INIS)

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

1991-01-01

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

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)

IAEA advisory group meeting on: Critical assessment of tritium retention in fusion reactor materials. Summary report

International Nuclear Information System (INIS)

Janev, R.K.; Federici, G.; Roth, J.

1999-07-01

The proceedings, conclusions and recommendations of the IAEA Advisory Group Meeting on 'Critical Assessment of Tritium Retention in Fusion Reactor Materials', held on June 7-8, 1999 at the IAEA Headquarters in Vienna, Austria, are briefly described. The report contains a summary of the presentations of meeting participants, a review of the data status (availability and needs) for the fusion most relevant bulk and mixed materials, and recommendations to the IAEA regarding its future activity in this data area. (author)

Normalized Tritium Quantification Approach (NoTQA) a Method for Quantifying Tritium Contaminated Trash and Debris at LLNL

International Nuclear Information System (INIS)

Dominick, J.L.; Rasmussen, C.L.

2008-01-01

Several facilities and many projects at LLNL work exclusively with tritium. These operations have the potential to generate large quantities of Low-Level Radioactive Waste (LLW) with the same or similar radiological characteristics. A standardized documented approach to characterizing these waste materials for disposal as radioactive waste will enhance the ability of the Laboratory to manage them in an efficient and timely manner while ensuring compliance with all applicable regulatory requirements. This standardized characterization approach couples documented process knowledge with analytical verification and is very conservative, overestimating the radioactivity concentration of the waste. The characterization approach documented here is the Normalized Tritium Quantification Approach (NoTQA). This document will serve as a Technical Basis Document which can be referenced in radioactive waste characterization documentation packages such as the Information Gathering Document. In general, radiological characterization of waste consists of both developing an isotopic breakdown (distribution) of radionuclides contaminating the waste and using an appropriate method to quantify the radionuclides in the waste. Characterization approaches require varying degrees of rigor depending upon the radionuclides contaminating the waste and the concentration of the radionuclide contaminants as related to regulatory thresholds. Generally, as activity levels in the waste approach a regulatory or disposal facility threshold the degree of required precision and accuracy, and therefore the level of rigor, increases. In the case of tritium, thresholds of concern for control, contamination, transportation, and waste acceptance are relatively high. Due to the benign nature of tritium and the resulting higher regulatory thresholds, this less rigorous yet conservative characterization approach is appropriate. The scope of this document is to define an appropriate and acceptable

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

International Nuclear Information System (INIS)

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

1986-01-01

The Fusion Safety Program of EGandG idaho, Inc. at the Idaho National Engineering Laboratory (INEL) is developing a safety analysis code called TMAP (Tritium Migration Analysis Program) to analyze tritium loss from fusion systems during normal and off-normal conditions. TMAP is a one-dimensional code that calculates tritium movement and inventories in a system of interconnected enclosures and wall structures. These wall structures can include composite materials with bulk trapping of the permeating tritium on impurities or radiation induced dislocations within the material. The thermal response of a structure can be modeled to provide temperature information required for tritium movement calculations. Chemical reactions and hydrogen isotope movement can also be included in the calculations. TMAP was used to analyze the movement of tritium implanted into a proposed limiter/first wall structure design

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)

Relation between tritium concentration and chemical composition in rain at Fukuoka

International Nuclear Information System (INIS)

Hayashi, Y.; Momoshima, N.; Maeda, Y.; Kakiuchi, H.

1999-01-01

Tritium concentrations in rain collected at Fukuoka, Japan from 1982 have been measured. From May 1996 tritium concentrations and chemical species have been analyzed for each rain to examine their relationship. Recent rain was concluded not to be affected by tritium from atmospheric nuclear tests. Tritium concentrations showed a seasonal pattern, high during winter and spring and low during summer and fall and had positive correlations with non-sea-salt SO 4 2- , indicating a long distance of acidic materials as well as tritium from continental China. (author)

Studies on steps affecting tritium residence time in solid blanket

International Nuclear Information System (INIS)

Tanaka, Satoru

1987-01-01

For the self sustaining of CTR fuel cycle, the effective tritium recovery from blankets is essential. This means that not only tritium breeding ratio must be larger than 1.0, but also high recovering speed is required for the short residence time of tritium in blankets. Short residence time means that the tritium inventory in blankets is small. In this paper, the tritium residence time and tritium inventory in a solid blanket are modeled by considering the steps constituting tritium release. Some of these tritium migration processes were experimentally evaluated. The tritium migration steps in a solid blanket using sintered breeding materials consist of diffusion in grains, desorption at grain edges, diffusion and permeation through grain boundaries, desorption at particle edges, diffusion and percolation through interconnected pores to purging stream, and convective mass transfer to stream. Corresponding to these steps, diffusive, soluble, adsorbed and trapped tritium inventories and the tritium in gas phase are conceivable. The code named TTT was made for calculating these tritium inventories and the residence time of tritium. An example of the results of calculation is shown. The blanket is REPUTER-1, which is the conceptual design of a commercial reversed field pinch fusion reactor studied at the University of Tokyo. The experimental studies on the migration steps of tritium are reported. (Kako, I.)

An improved air-supplied plastic suit for protection against tritium

International Nuclear Information System (INIS)

Wiernicki, C.

1987-01-01

A newly developed Saran/CPE plastic suit material is described which offers significantly better protection against HTO penetration and permeation than the 12-mil PVC currently used at SRP and most other DOE and commercial sites where tritium and HTO are exposure hazards. Tritium breakthrough time is an important parameter when evaluating the applicability of protective clothing; previously published tritium permeation tests did not measure this parameter. Future studies should quantify steady-state permeation rate and breakthrough time to more fully evaluate potential tritium protective clothing. Saran/CPE has successfully been fabricated into a plastic suit because, in addition to its superior tritium resistance, it has all the characteristics required to construct a rugged, dependable, and comfortable suit. The use of the Saran/CPE suit at SRP reactor and tritium production facilities should be a major contribution to the site As Low As Reasonably Achievable program. Both Saran/CPE have demonstrated excellent resistance to a wide range of chemical contaminants; therefore, this suit material may have applications in the general chemical industry and hazardous waste site cleanup operations. 4 refs., 3 figs., 1 tab

Investigation of tritium inventory and permeation behaviour in the liquid breeder blanket concept of Demo as a function of design and material parameters

International Nuclear Information System (INIS)

Tominetti, S.; Perujo, A.; Reiter, F.

1991-01-01

A numerical code has been used to estimate the time dependence of tritium inventory and of tritium transport into the coolant, into the first wall boxes and through the liquid breeder in the Pb-17Li blanket concept of DEMO. Several issues in both design and material parameters have been considered and the effect on inventory and permeation of coatings with low surface recombination coefficient and/or low diffusivity at various surfaces of the structural material has been studied. TiC has been chosen as reference material for these calculations and a general database on coating efficiency as a function of its properties has also been produced on the basis of TiC data

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)

Analysis of LaNi4.25Al0.75 (LANA.75) tritide after five years of tritium exposure

International Nuclear Information System (INIS)

Wermer, J.R.; Holder, J.S.; Mosley, W.C.

1993-01-01

Tritium aging studies have shown that LaNi 4.25 Al 0.75 (LANA .75) tritide storage material undergoes significant degradation with tritium aging. After 5.4 years of dormant storage at full stoichiometry, which is considered a worst-case condition for this material, the performance is still acceptable for SRS tritium processing applications. The isotherms change, decreasing the desorption pressures, increasing the isotherm plateau slopes, and decreasing the total storage capacity. Eventually, the material will degrade with time to the point where it may no longer be useful for tritium processing applications. At the end of life, the tritium heel can be exchanged with protium or deuterium to produce a final material containing very little tritium

Test device for measuring permeability of a barrier material

Science.gov (United States)

Reese, Matthew; Dameron, Arrelaine; Kempe, Michael

2014-03-04

A test device for measuring permeability of a barrier material. An exemplary device comprises a test card having a thin-film conductor-pattern formed thereon and an edge seal which seals the test card to the barrier material. Another exemplary embodiment is an electrical calcium test device comprising: a test card an impermeable spacer, an edge seal which seals the test card to the spacer and an edge seal which seals the spacer to the barrier material.

Tritium inventory and permeation in the ITER breeding blanket

International Nuclear Information System (INIS)

Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

2000-01-01

A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

Modeling of tritium behavior in Li2O

International Nuclear Information System (INIS)

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

1992-08-01

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

Weapons engineering tritium facility overview

Energy Technology Data Exchange (ETDEWEB)

Najera, Larry [Los Alamos National Laboratory

2011-01-20

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

Study about sorption of protium and mixture protium–tritium on sponge titanium

Energy Technology Data Exchange (ETDEWEB)

Vasut, Felicia, E-mail: feliciavasut@yahoo.com; Stefanescu, Ioan; Bornea, Anisia Mihaela; Zamfirache, Marius; Sofalca, Nicolae; David, Claudia

2013-10-15

The Nuclear Power Plant Cernavoda is equipped with a CANDU reactor and is one of the most powerful tritium sources from Europe. The reactor is moderated and cooled with heavy water that is continuous enriched with tritium. The presence of the tritium decreases the capacity of the heavy water to moderate the nuclear reactions. For this reason, I.C.I.T. Ramnicu Valcea developed a detritiation technology based on catalytic isotopic exchange and cryogenic distillation. At the end of the detritiation process, heavy water is produced with low concentration of tritium (that could be introduced back for the moderation process) and tritium (that have to be stored into a stable form). Tritium is a radioactive material and one of the basic conditions for the operation of the nuclear installations is the security for the operating personnel and for the environment [1]. At I.C.I.T. Ramnicu Valcea were tested materials with high capacity for storage of tritium, like titanium sponge and powder. The first experimental study was made using protium because it was assumed that tritium behaves similar with protium. In addition, it was made experiments of sorption on sponge titanium using a mixture protium–tritium. The result was similar but not identical, titanium sponge absorbing better protium than mixture protium–tritium, resulting different atomic ratios. The paper presents a study about sorption of protium and mixture of protium and tritium on sponge titanium.

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.)

Study of Tritium Behavior in Cement Paste

International Nuclear Information System (INIS)

Takata, H.; Motoshima, T.; Satake, S.; Nishikawa, M.

2005-01-01

The concrete materials are used as the partition wall of the tritium handling facilities. It is important to grasp the tritium behavior in the concrete wall for radiation safety. It is considered in this study that the surface water on the concrete materials consists of physically adsorbed water, chemically adsorbed water and structural water as in the case of porous adsorption materials. The adsorption capacity due to physically and chemically adsorption isotherms observed in this study shows that the amount of water adsorption on the cement paste is a quarter of the amount adsorbed onto the surface of activated alumina or molecular sieves 5A (MS-5A). It shows that concrete is easily contaminated with tritiated water

Analysis of in-situ tritium recovery from solid fusion-reactor blankets

International Nuclear Information System (INIS)

Smith, D.L.; Clemmer, R.G.; Jankus, V.Z.; Rest, J.

1980-01-01

The proposed concept for in-situ tritium recovery from the STARFIRE blanket involves circulation of a low pressure (approx. 0.05 MPa) helium through formed channels in the highly porous solid breeding material. Tritium generated within the grains must diffuse to the grain boundaries, migrate through the grain boundaries to the particle surface and then percolate through the packed bed to the helium purge channel. Highly porous α-LiAlO 2 with a bimodal pore distribution is proposed for the breeding material to facilitate the tritium release

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.)

Estimation of the detection limit of an experimental model of tritium storage bed designed for 'in-situ' accountability

International Nuclear Information System (INIS)

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

2009-01-01

During the water detritiation process most of the tritium inventory is transferred from water into the gaseous phase, then it is further enriched and finally extracted and safely stored. The control of tritium inventory is an acute issue from several points of view: - Financially - tritium is an expensive material; - Safeguard - tritium is considered as nuclear material of strategic importance; - Safety - tritium is a radioactive material: requirements for documented safety analysis report (to ensure strict limits on the total tritium allowed) and for evaluation of accident consequences associated with that inventory. Large amounts of tritium can be stored, in a very safely manner, as metal tritides. A bench-scale experiment of a tritium storage bed with integrated system for in-situ tritium inventory accountancy was designed and developed at ICSI Rm. Valcea. The calibration curve and the detection limit for this experimental model of tritium storage bed were determined. The experimental results are presented in this paper. (authors)

Tritium safety study using Caisson Assembly (CATS) at TPL/JAEA

International Nuclear Information System (INIS)

Hayashi, T.; Kobayashi, K.; Iwai, Y.; Isobe, K.; Nakamura, H.; Kawamura, Y.; Shu, W.; Suzuki, T.; Yamada, M.; Yamanishi, T.

2008-01-01

Tritium confinement is required as the most important safety Junction for a fusion reactor. In order to demonstrate the confinement performance experimentally, an unique equipment, called CATS: Caisson Assembly for Tritium Safety study, was installed in Tritium Process Laboratory of Japan Atomic Energy Agency and operated for about 10 years. Tritium confinement and migration data in CATS have been accumulated and dynamic simulation code was accumulated using these data. Contamination and decontamination behavior on various materials and new safety equipment functions have been investigated under collaborations with a lot of laboratories and universities. (authors)

Tritium sorption in lithium-bismuth and lithium-aluminum alloys

International Nuclear Information System (INIS)

Talbot, J.B.; Smith, F.J.; Land, J.F.; Barton, P.

1976-01-01

The sorption of tritium by molten lithium-bismuth, Li-Bi (15 at.% lithium), and solid equiatomic lithium-aluminum, Li-Al, was investigated to evaluate the potential application of both materials in controlled thermonuclear reactors. The solubility of tritium in molten Li-Bi is less than 0.1 ppb at 500 - 700 0 C and tritium partial pressures of 10 -1 - 10 -3 Torr. Therefore, extraction of tritium from molten Li 2 BeF 4 salt with Li-Bi is not practical. The solubility of hydrogen in solid Li-Al (50 - 50 at.%) at 500 0 C follows Sieverts' Law; the Sieverts' constant was measured to be 1.9 (+-0.1) X 10 4 Torrsup(1/2)/atomic fraction. Tritium sorption in Li-Al ranged from 0.01 to 7 ppm at 400 - 600 0 C at respective tritium partial pressures of 0.14 - 0.52 Torr. (Auth.)

Tritium adsorption/release behaviour of advanced EU breeder pebbles

Science.gov (United States)

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

2017-06-01

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

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)

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

Cosmogenic production of tritium in dark matter detectors

Science.gov (United States)

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

2018-01-01

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

Selection of fluids for tritium pumping systems

International Nuclear Information System (INIS)

Chastagner, P.

1984-02-01

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

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)

Tritium adsorption/release behaviour of advanced EU breeder pebbles

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-15

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

BEATRIX-II: In-situ tritium recovery data correction

International Nuclear Information System (INIS)

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

1993-09-01

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

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

International Nuclear Information System (INIS)

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

1993-01-01

The Tritium Plasma Experiment was assembled at Sandia National Labs., Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Lab. to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 23 ions/m 2 .s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures

Tritium in the burial ground of the Savannah River Site

International Nuclear Information System (INIS)

Hyder, M.L.

1993-06-01

This memorandum reviews the available information on tritium-contaminated material discarded to burial grounds. Tritium was the first isotope studied because it represents the most immediate concern with regard to release to the environment. Substantial amounts of tritium are known to be present in the ground water underneath the area, and outcropping of this ground water in springs and seeps has been observed. The response to this release of tritium from the burial ground is a current concern. The amount of tritium emplaced in the burial ground facilities is very uncertain, however, some general conclusions can be made. In particular, most of the tritium buried is associated with spent equipment and other waste, rather than spent melts. Correspondingly, most of the tritium in the ground water seems to be associated with burials of this type, rather than the spent melts. Maps are presented showing the location of burials of tritiated waste by type, and the location of the largest individual burials according to COBRA records

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

The INEL Tritium Research Facility

International Nuclear Information System (INIS)

Longhurst, G.R.

1990-01-01

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

The INEL Tritium Research Facility

Energy Technology Data Exchange (ETDEWEB)

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

1990-06-01

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

Transient behavior of enrichment of tritium water in adsorption-distillation column

International Nuclear Information System (INIS)

Fukada, Satoshi

2006-01-01

Enrichment of tritium in an adsorption-distillation column was experimentally investigated under the two processes of simple distillation and total-reflux distillation. Adsorption of water on silica-gel pellets enhanced the total isotope separation factor in the water distillation column. The transient behavior of tritium enrichment was analyzed using material balance equations of tritium and water in each cell with a height corresponding to HETP. The experimental transient behavior was well simulated by the material balance equations with additional assumptions on vapor and liquid flow rates regardless of the different processes of simple distillation and total-reflux distillation. (author)

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 sorption on protective coatings for concrete

International Nuclear Information System (INIS)

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

1992-11-01

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

Stainless steel electrochemical behaviour - application to the decontamination of steel parts contaminated by tritium

International Nuclear Information System (INIS)

Bellanger, G.

1991-01-01

This purpose of this work is the study of an electrochemical decontamination process of stainless steel in which tritium is present on the surface of the metal, in the oxide layer and in the metal. We have first investigated the behaviour of the oxide layer. Then we have studied the hydrogen evolution, its diffusion and retrodiffusion in the metal. The results are applied to the decontamination of steel parts contamined by tritium. Part of the tritium can be eliminated by reducing the oxyde layer, which contains large amounts of tritium. However, it is more beneficial to electrolyse at the potential at which the H + ions are reduced. The hydrogen on the steel surface enters in the metal and displaces most of tritium located in the metallic layers near the surface. The tritium surface elimination rate is about 95%. The tritium eliminated through electrolysis is only a small fraction of all the tritium contained in the metal. However, according to conservation experiments of parts after electrolysis, it can be concluded that hydrogen, probably more strongly bound than tritium to steel, forms near the surface a barrier that prevents tritium retrodiffusion. Electrolysis appears as a satisfactory process for the surface decontamination of slightly tritiated steel parts. A decontamination automaton based on the preceding results is described using a pad electrolyser. This type of decontamination is little polluting, and the parts can be recycled after the in situ treatment [fr

Effect of surface water on tritium release behavior from Li4SiO4

International Nuclear Information System (INIS)

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

2010-01-01

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

Current Sandia programs and laboratory facilities for tritium research

International Nuclear Information System (INIS)

Swansiger, W.A.; West, L.A.

1975-01-01

Currently envisioned fusion reactor systems will contain substantial quantities of tritium. Strict control of the overall tritium inventory and environmental safety considerations require an accurate knowledge of the behavior of this isotope in the presence of Controlled Thermonuclear Reactor (CTR) materials. A 14,000 ft 2 laboratory for tritium research is currently under construction at Sandia Laboratories in Livermore. Details about the laboratory in general are provided. Results from studies of hydrogen isotope diffusion in surface-characterized metals will be presented. Details of two permeation systems (one for hydrogen and deuterium, the other for tritium) will be discussed. Data will also be presented concerning the gettering of hydrogen isotopes and application to CTR collector designs. (auth)

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

International Nuclear Information System (INIS)

Pruess, Karsten; Conrad, Mark; Finsterle, Stefan; Kennedy, Mack; Kneafsey, Timothy; Salve, Rohit; Su, Grace; Zhou, Quanlin

2006-01-01

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

Tritium inventory differences: I. Sampling and U-getter pump holdup

International Nuclear Information System (INIS)

Ellefson, R.E.; Gill, J.T.

1986-01-01

Inventory differences (ID) in tritium material balance accounts (MBA) can occur with unmeasured transfers from the process or unmeasured holdup in the system. Small but cumulatively significant quantities of tritium can leave the MBA by normal capillary sampling of process gas operation. A predictor model for estimating the quantity of tritium leaving the MBA by sampling has been developed and implemented. The model calculates the gas transferred per sample; using the tritium concentration in the process and the number of samples, a quantity of tritium transferred is predicted. Verification of the model is made by PVT measurement of process transfer from multiple samplings. Comparison of predicted sample transfers with IDs from several MBA's reveals that sampling typically represents 50% of unmeasured transfers for regularly sampled processes

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.

Comparative study of the more promising combinations of blanket materials, power conversion systems, and tritium recovery and containment systems for fusion reactors

International Nuclear Information System (INIS)

Fraas, A.P.

1975-11-01

The many possible combinations of blanket materials, tritium generation and recovery systems, and power conversion systems were surveyed first by reviewing the principal design studies that have been prepared and then by examining a comprehensive set of designs generated by using a common set of ground rules that included all of the boundary conditions that could be envisioned. The results indicate that, of the wide variety of systems that have been considered, by far the most promising employs lithium recirculated in a closed loop within a niobium blanket structure and cooled with boiling potassium or cesium. This approach gives the simplest and lowest cost tritium recovery system, the lowest pressure and thermal stresses, the simplest structure with the lowest probability of a leak, the greatest resistance to damage from a plasma energy dump, and the lowest rate of plasma contamination by either outgassing or sputtering. The only other blanket materials combination that appears fairly likely to give a satisfactory tritium generation and recovery system is an Li 2 BeF 4 -Incoloy blanket, and even this system involves major uncertainties in the effectiveness, size, and cost of the tritium recovery system. Further, the Li 2 BeF 4 blanket system has the disadvantage that the world reserves of beryllium are too limited to support a full-blown fusion reactor economy, its poor thermal conductivity leads to cooling difficulties and a requirement for a complex structure with intricate cooling passages, and this inherently leads to an expensive blanket with a relatively high probability of leaks. The other blanket materials combinations yield even less attractive systems

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 contamination and monitoring at Frascati Neutron Generator

Energy Technology Data Exchange (ETDEWEB)

Lucci, F.; Sandri, S.; Ianni, A. [ENEA, Frascati (Italy). Dipartimento Ambiente; Vasselli, R. [ANPA, Roma (Italy); Pillon, M.; Bettinali, L. [ENEA, Frascati (Italy). Dipartimento Energia

1994-11-01

The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10{sup 1}1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed.

Tritium contamination and monitoring at Frascati Neutron Generator

International Nuclear Information System (INIS)

Lucci, F.; Sandri, S.; Ianni, A.; Pillon, M.; Bettinali, L.

1994-11-01

The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10 1 1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed

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

Tritium calorimeter setup and operation

CERN Document Server

Rodgers, D E

2002-01-01

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

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

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

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

Operating experience with TFTR's Tritium Storage and Delivery System

International Nuclear Information System (INIS)

Voorhees, D.R.

1995-01-01

The Tritium Storage and Delivery System (TSDS) at TFTR was fabricated at Monsanto Mound Lab in the late 1970's and delivered to PPPL in the early 1980's. Commissioning progressed slowly and was finally completed in 1992 following a series of Preoperational tests and Integrated Systems tests. Those tests included thorough leak testing of glove boxes and process piping, electrical interlocks and controls, instrumentation calibrations, volume determinations and verification of uranium bed capacity. The system accepted tritium in dilute form in May of 1993 and began serious usage of pure tritium in November 1993. As the throughput of high purity tritium increased, shortcomings of the system became evident and extensive repairs were implemented. System leakage and material compatibility were the primary causes of the problems. To date, the system has received, stored and delivered over 500 kCi of tritium and is performing very well. The dedicated quadrupole mass spectrometer and beta scintillator system has been analyzing tritium bearing and pure gas streams for over 3 years with minimal downtime

Tritium Formation and Mitigation in High Temperature Reactors

Energy Technology Data Exchange (ETDEWEB)

Piyush Sabharwall; Carl Stoots

2012-08-01

Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

Synthesis and pharmacokinetics of thiacoccide labeled with tritium

International Nuclear Information System (INIS)

Shestakov, A.D.; Kaminskii, Yu.L.; Nurova, I.M.; Nikitina, V.N.; Zaionts, V.I.; Maksimova, O.V.; Mikhailov, G.A.

1986-01-01

To obtain information on the pharmacokinetics of thiacoccide, the authors effect the synthesis of an analog of thiacoccide, viz., the hydrochloride of 2-methyl-N-(2'-n-propyl-4'-aminopyrimid-5'-ylmethyl)pyridinium chloride (I) labeled with tritium. The simplest way to introduce a tritium label into (I) is by isotopic exchange of hydrogen of (I) with H 3 HO. Typical material obtained from isotopic exchange with water labeled with tritium is shown. The dynamics of 3 H distribution in the organism of the chick and the rate of its elimination on single administration of thiacoccide containing isotope in both the methyl group and in the pyridine ring are shown

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

Accelerator Production of Tritium Programmatic Environmental Impact Statement Input Submittal

International Nuclear Information System (INIS)

Miller, L.A.; Greene, G.A.; Boyack, B.E.

1996-02-01

The Programmatic Environmental Impact Statement for Tritium Supply and Recycling considers several methods for the production of tritium. One of these methods is the Accelerator Production of Tritium. This report summarizes the design characteristics of APT including the accelerator, target/blanket, tritium extraction facility, and the balance of plant. Two spallation targets are considered: (1) a tungsten neutron-source target and (2) a lead neutron-source target. In the tungsten target concept, the neutrons are captured by the circulating He-3, thus producing tritium; in the lead target concept, the tritium is produced by neutron capture by Li-6 in a surrounding lithium-aluminum blanket. This report also provides information to support the PEIS including construction and operational resource needs, waste generation, and potential routine and accidental releases of radioactive material. The focus of the report is on the impacts of a facility that will produce 3/8th of the baseline goal of tritium. However, some information is provided on the impacts of APT facilities that would produce smaller quantities

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

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

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 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.

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

Energy Technology Data Exchange (ETDEWEB)

Renoult, O

1994-04-01

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

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

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

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)

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

2009 EVALUATION OF TRITIUM REMOVAL AND MITIGATION TECHNOLOGIES FOR WASTEWATER TREATMENT

Energy Technology Data Exchange (ETDEWEB)

LUECK KJ; GENESSE DJ; STEGEN GE

2009-02-26

atmosphere, and (4) use of barriers to minimize the transport of tritium in groundwater. Continuing development efforts for tritium separations processes are primarily to support the International Thermonuclear Experimental Reactor (ITER) program, the nuclear power industry, and the production of radiochemicals. While these applications are significantly different than the Hanford application, the technology could potentially be adapted for Hanford wastewater treatment. Separations based processes to reduce tritium levels below the drinking water MCL have not been demonstrated for the scale and conditions required for treating Hanford wastewater. In addition, available cost information indicates treatment costs for such processes will be substantially higher than for discharge to SALDS or other typical pump and treat projects at Hanford. Actual mitigation projects for groundwater with very low tritium contamination similar to that found at Hanford have focused mainly on controlling migration and on evaporation for dispersion in the atmosphere.

Problems in tritium handling in fusion reactors studies at CEA within the european effort

International Nuclear Information System (INIS)

Roth, E.; Hircq, B.; Fidelle, J.P.

1988-01-01

Technological aspects of tritium handling linked with the operation of a fusion reactor are reviewed. Tritium storage is discussed from the point of view of the volumme of a single unit and of the nature of the metal bed. Purification of tritium and recovery from tritiated compounds is studied, including conversion from water to the gaseous form. Interaction of tritium and structural materials is developed from the point of view of corrosion, embrittlement, permeation. A flowsheet displaying a conception of a reference tritium circuit is proposed, and consideration is given to specifications of large components, namely pumps and gatevalves for tritium circuits

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

Design approach for safe tritium handling in ITER

International Nuclear Information System (INIS)

Ohira, Shigeru

2002-01-01

Outlines for tritium handling and a fundamental approach for ensuring safety are presented. The amount of tritium stored and processed in the ITER facility will be much larger than that in the existing facilities for fusion research, though the processing methods and the conditions of processing (e.g., concentration, pressure, etc.) will be similar for those used in those facilities. Therefore, considerations to be taken for tritium handling, such as limitations of tritium permeation and leaks, provision of an appropriate ventilation/detritiation system for maintenance, measures to ensure mechanical integrity, etc., can be provided based on the knowledge obtained in the facilities. The Technical Advisory Committee of the Science and Technology Agency established a fundamental approach in 2000, and set out the basic safety principles and approaches as technical requirements of safety design and assessment, which were derived from the safety characteristics of the ITER plant. Sufficient prevention of accidents can be achieved by ensuring and maintaining the structural integrity of the enclosures containing radioactive materials against the loads anticipated during operation, and a low hazard potential of radioactive materials, sufficiently within prescribed limits, can be maintained by the vitiation and clean-up system even if large release is postulated. (author)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-11-30

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Low-level tritium research facility for the University of Toronto Institute for Aerospace Studies

International Nuclear Information System (INIS)

Kherani, N.P.; Shmayda, W.T.

1984-06-01

The objective of the Low-level Tritium Research Facility for the University of Toronto Institute for Aerospace Studies (UTIAS) is to investigate tritium-material interactions and how they differ with respect to protium and deuterium. The tritium laboratory will also be employed to study tritium retention, tritium imaging, and the effect of tritium on diagnostic devices. This report is a preliminary design document of the UTIAS Low-Level Tritium Research Facility including the fundamentals of tritium, a description of the facility, tritium laboratory requirements and the safety analysis of the laboratory. The facility is designed to handle a total elemental tritium inventory of 10 Ci, though it will initially commence operation with 1 Ci and later increased to the maximum value. In the event of an instantaneous emission of the total tritium inventory within the laboratory, the working personnel would be exposed to an airborne tritium concentration less than the maximum permissible. Moreover, with all the safety features included in this design the likelihood of such an accident is very remote. Thus, the tritium laboratory design is intrinsically safe

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-11-01

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

Filbe molten salt research for tritium breeder applications

International Nuclear Information System (INIS)

Anderl, R.A.; Petti, D.A.; Smolik, G.R.

2004-01-01

This paper presents an overview of Flibe (2Lif·BeF 2 ) molten salt research activities conducted at the INEEL as part of the Japan-US JUPITER-II joint research program. The research focuses on tritium/chemistry issues for self-cooled Flibe tritium breeder applications and includes the following activities: (1) Flibe preparation, purification, characterization and handling, (2) development and testing of REDOX strategies for containment material corrosion control, (3) tritium behavior and management in Flibe breeder systems, and (4) safety testing (e.g., mobilization of Flibe during accident scenarios). This paper describes the laboratory systems developed to support these research activities and summarizes key results of this work to date. (author)

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

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)

Tritium pollution in the Swiss luminous compound industry

International Nuclear Information System (INIS)

Krejci, K.; Zeller, Jr.

1979-01-01

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

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

Tritium calorimeter setup and operation

International Nuclear Information System (INIS)

Rodgers, David E.

2002-01-01

The LBNL tritium calorimeter is a stable instrument capable of measuring tritium with a sensitivity of 25 Ci. Measurement times range from 8-hr to 7-days depending on the thermal conductivity and mass of the material being measured. The instrument allows accurate tritium measurements without requiring that the sample be opened and subsampled, thus reducing personnel exposure and radioactive waste generation. The sensitivity limit is primarily due to response shifts caused by temperature fluctuation in the water bath. The fluctuations are most likely a combination of insufficient insulation from ambient air and precision limitations in the temperature controller. The sensitivity could probably be reduced to below 5 Ci if the following improvements were made: (1) Extend the external insulation to cover the entire bath and increase the top insulation. (2) Improve the seal between the air space above the bath and the outside air to reduce evaporation. This will limit the response drift as the water level drops. (3) Install an improved temperature controller, preferably with a built in chiller, capable of temperature control to ±0.001 C

Progress in tritium retention and release modeling for ceramic breeders

International Nuclear Information System (INIS)

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

1994-01-01

Tritium behavior in ceramic breeder blankets is a key design issue for this class of blanket because of its impact on safety and fuel self-sufficiency. Over the past 10-15 years, substantial theoretical and experimental efforts have been dedicated world-wide to develop a better understanding of tritium transport in ceramic breeders. Models that are available today seem to cover reasonably well all the key physical transport and trapping mechanisms. They have allowed for reasonable interpretation and reproduction of experimental data and have helped in pointing out deficiencies in material property data base, in providing guidance for future experiments, and in analyzing blanket tritium behavior. This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described along with the more recent and sophisticated models developed to help understand them. Recent experimental data are highlighted and model calibration and validation discussed. Finally, example applications to blanket cases are shown as illustration of progress in the prediction of ceramic breeder blanket tritium inventory

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

International Nuclear Information System (INIS)

Quanci, J.F.

1989-01-01

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

Techniques for tritium recovery from carbon flakes and dust at the JET active gas handling system

International Nuclear Information System (INIS)

Gruenhagen, S.; Perevezentsev, A.; Brennan, P. D.; Camp, P.; Knipe, S.; Miller, A.; Yorkshades, J.

2008-01-01

Detritiation of highly tritium contaminated carbon and metal material used as first wall armour is a key issue for fusion machines like JET and ITER. Re-deposited carbon and hydrogen in the form of flakes and dust can lead to a build-up of the tritium inventory and therefore this material must be removed and processed. The high tritium concentration of the flake and dust material collected from the JET vacuum vessel makes it unsuitable for direct waste disposal without detritiation. A dedicated facility to process the tritiated carbon flake material and recover the tritium has been designed and built. In several test runs active material was successfully processed and de-tritiated in the new facility. Samples containing only carbon and hydrogen isotopes have been completely oxidized without any residue. Samples containing metallic impurities, e.g. beryllium, require longer processing times, adjusted processing parameters and yield an oxide residue. The detritiation factor was 2x10 4 . In order to simulate in-vessel and ex-vessel detritiation techniques, the detritiation of a carbon flake sample by isotopic exchange in a hydrogen atmosphere was investigated. 2.8% of tritium was recovered by this means. (authors)

Distribution of tritium in estuarine waters: the role of organic matter

International Nuclear Information System (INIS)

Turner, Andrew; Millward, Geoffrey E.; Stemp, Martin

2009-01-01

Tritium is an important environmental radionuclide whose reactivity with ligands and solids in aquatic systems is assumed to be limited. We studied the fractionation and sorption of tritium (added as tritiated water) in river water and seawater, and found that its distribution appears to be influenced by its affinity for organic matter. Tritium rapidly equilibrates with dissolved organic ligands that are retained by a reverse-phase C18 column, and with suspended sediment particles. Significantly, a measurable fraction of sorbed tritium associates with proteinaceous material that is potentially available to sediment-feeding organisms. These characteristics have not been reported previously and cannot be accounted for solely by isotopic exchange with hydrogen. Nevertheless, they are in qualitative agreement with available measurements of tritium in estuarine and coastal waters where its principal discharge is as tritiated water. Further research into the estuarine biogeochemical behaviour of tritium is required and radiological distribution coefficients and concentration factors that are assumed for this radionuclide may require reconsideration.

Distribution of tritium in estuarine waters: the role of organic matter

Energy Technology Data Exchange (ETDEWEB)

Turner, Andrew [Consolidated Radio-isotope Facility, School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)], E-mail: aturner@plymouth.ac.uk; Millward, Geoffrey E.; Stemp, Martin [Consolidated Radio-isotope Facility, School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)

2009-10-15

Tritium is an important environmental radionuclide whose reactivity with ligands and solids in aquatic systems is assumed to be limited. We studied the fractionation and sorption of tritium (added as tritiated water) in river water and seawater, and found that its distribution appears to be influenced by its affinity for organic matter. Tritium rapidly equilibrates with dissolved organic ligands that are retained by a reverse-phase C18 column, and with suspended sediment particles. Significantly, a measurable fraction of sorbed tritium associates with proteinaceous material that is potentially available to sediment-feeding organisms. These characteristics have not been reported previously and cannot be accounted for solely by isotopic exchange with hydrogen. Nevertheless, they are in qualitative agreement with available measurements of tritium in estuarine and coastal waters where its principal discharge is as tritiated water. Further research into the estuarine biogeochemical behaviour of tritium is required and radiological distribution coefficients and concentration factors that are assumed for this radionuclide may require reconsideration.

Sol-gel synthesis of lithium metatitanate as tritium breeding material under different sintering conditions

Science.gov (United States)

Lu, Wei; Wang, Jing; Pu, Wenjing; Li, Kaiping; Ma, Shubing; Wang, Weihua

2018-04-01

Lithium metatitanate (Li2TiO3) is a promising tritium breeding material candidate for solid blanket of D-T fusion reactors, due to its high mechanical strength, chemical stability, and tritium release rate. In this paper, Li2TiO3 powder with homogeneous crystal structure is synthesized by sol-gel method. The chemical reactions in gel thermal cracking and sintering process are studied by thermo gravimetric/differential scanning calorimetry (TG-DSC). The relationship between the sintering condition and the particle/grain size is characterized by X-ray diffraction (XRD) and scanning electron microscopy (SEM). Results show that below 673 K the gel precursor is completely decomposed and Li2TiO3 phase initially forms. The LiTiO2 by-product formed under the reductive atmosphere in muffle furnace, could be oxidized continually to Li2TiO3 at higher sintering temperature (≥1273 K) for longer sintering time (≥10 h). Both grain and particle sizes rely on a linear growth with the increase of sintering time at 1273 K. Over 1473 K, significant agglomerations exist among particles. The optimal sintering condition is selected as 1273 K for 10 h, for the purer Li2TiO3 phase (>99%), smaller grain and particle size.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-11-14

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

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

A 2D Finite Element Modelling of Tritium Permeation Through Cooling Plates for The HCLL DEMO Blanket Module

International Nuclear Information System (INIS)

Gabriel, F.; Escuriol, Y.; Dabbene, F.; Salavy, J.F.; Giancarli, L.; Gastaldi, O.

2006-01-01

As the Tritium self sufficiency is one of the major challenges for fusion reactor, breeding blankets represent one of the major technological breakthroughs required from passing from ITER to the next step reactor, usually called DEMO. One of the two blanket concepts developed in the EU is the Helium Cooled Lithium Lead (HCLL) blanket which uses the eutectic Pb-15.7Li metal liquid as both breeder and neutron multiplier. The structures, made of EUROFER, a low activation ferritic martensitic steel, are cooled by pressurized helium at 8 MPa and inlet/outlet temperature 300/500 o C. In this concept, the LiPb is fed from the top of the blanket and distributed in parallel vertical channels among pairs of cells (one cell for the radial movement towards the plasma, the other for the return). The liquid metal fills the in-box volume and is slowly re-circulated (few mm per seconds) to remove the produced tritium. In this paper, a local finite element modelling of the tritium permeation rate through the HCLL breeder unit cooling plates is presented. The tritium concentration in the helium circuit and remaining in the lithium lead circuit are evaluated by solving partial differential equations governing the tritium concentration balance, the thermal field and the lithium lead velocity field for a simplified 2D geometrical representation of the breeder unit. This allows estimating the sensitivity effect of coupling these different equations in order to deduce a relevant but simplified modelling for tritium permeation. This is to compare with tritium inventories studies, were the tritium permeation rate is estimated using simplified analytical modelling which generally leads to over estimate the tritium permeation rate to the coolant and so has strong influence on the coolant purification plant design. The finite element modelling performed shows that the Tritium permeation is considerable lower than the one obtained in previous estimations where nominal values of the governing

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

Liquid scintillation cocktails comparison for tritium contamination measurements

International Nuclear Information System (INIS)

Bazzarri, S.; Belloni, P.

1996-01-01

Liquid scintillation counting is one of the most used techniques for the measurements of tritium contamination. Until few years ago a problem related to this kind of measurement was the potential toxicity of the liquid cocktails used to produce the required scintillation. Some new products that guarantee an almost negligible impact on the environment and that are no longer toxic for the operators are now available. Some of this new scintillation cocktail are suitable to be used for tritium measurement. Due to the great benefit from the health point of view of these new materials a test of their scintillation performance has been done at the ENEA centers to select the product having the best characteristics for tritium measurement. (author)

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

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

Tritium waste control: April-September 1982

International Nuclear Information System (INIS)

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

1983-01-01

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

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)

Tritium solubility and permeation in high retention fusion reactor breeder elements

International Nuclear Information System (INIS)

Jakeman, D.

1979-11-01

As an alternative to the current philosophy of reducing the tritium inventory to a minimum by continuously extracting tritium from the breeder of a fusion reactor, an alternative design philosophy is examined in which tritium is contained within high retention breeder elements which can remain in the reactor for a substantial time. To prevent tritium diffusion through the clad of the element it is necessary to maintain a low tritium pressure within the element. Pressures of between 10 4 Pa and 1 Pa appear possible with an element containing a high solubility material provided it is kept below about 400 0 C. This should lead to a leakage into the coolant of between 10 Ci day -1 and 10 4 Ci day -1 which is considerably less than the 10 7 Ci day -1 in present designs. (author)

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

R and D of tritium technology for fusion in CAEP: progress and prospect

International Nuclear Information System (INIS)

Jiangfeng, Song; Daqiao, Meng; Rong, Li; Zhiyong, Huang; Guoqiang, Huang; Chang-an, Chen; Xiaojun, Deng; Cheng, Qin; Xiaojing, Qian; Guikai, Zhang

2015-01-01

China has decided to develop its own fusion engineering test reactor and has also joined ITER. Tritium plant is one of the key systems of fusion system. Programs supposed by China ministry of Science and technology named 'Conceptual design and key technologies research on TBM tritium system' and 'Conceptual design and key technologies research on tritium plant for fusion reactor' were finished in 2013 and 2014. After several years of research, we have finished the design of TBM tritium system, TEP, SDS, WDS, ISS and tritium safety system. The key technologies such as TES, CPS, hydrogen storage materials for SDS, catalysts for WDS, palladium alloy membranes for TEP are under research. In this paper, the progress and prospect of tritium technology for R and D of fusion is introduced. (author)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Analysis of Tritium Breeding in the Test Module

Energy Technology Data Exchange (ETDEWEB)

Hong, SeongHee; Park, YunSeo; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2015-10-15

In this paper, neutronic analyses are conducted on redesign of TMs which have high tritium breeding performance based on results of previous study. Calculation model is simplified, there is no effect to cover very complex geometry of fusion reactor for this study. As spent fuel disposal problem is issued in nuclear industry, FFHR is one of the most fascinating candidates for solving this problem through waste transmutation. Our research team also was designed a full core FFHR for waste transmutation. However, in this study, Test Module (TM) as test bed of FFHR for various purposes are analyzed. Analysis of tritium breeding on the TM was conducted as a first phase among TMs having various purposes. Because there are no fissionable materials in the TM for tritium breeding, geometry and neutronic reactions of its simpler compared to TM for waste transmutation and power production. Additionally, it is important database for tritium self-sufficiency as basic design condition of TM. In the previous study, neutronic analyses are conducted on these various TMs: Helium cooled solid breeder (HCSB), water cooled solid breeder (WCSB), Helium cooled dual breeder (HCDB) and molten-salt cooled liquid breeder (MSLB) in order to understand design characteristics. Neutronics calculations are performed with MCNPX 2.6.0 with ENDF/B-VII.0 neutron cross section library and activity and time-dependent tritium production calculations are performed with CINDER'90. In this paper, analysis of tritium breeding on WCHESL and WCHELL as TM is conducted. WCHESL is designed for effective tritium breeding performance and it satisfies design conditions. On the other hand WCHELL is designed for tritium breeding as much as possible and it also satisfies design conditions. However, neutron multiplication performance with these TM is not outstanding. WCHESL consist ceramic Li breeder, its period is 4.15E+08 sec.

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 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

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Physicochemical processes behind atomic tritium harnessing for investigation into surface of solids

International Nuclear Information System (INIS)

Badun, G.A.; Fedoseev, V.M.

2000-01-01

The thermal dissociation of hydrogen molecules on tungsten wire heated up to 1500 - 2000 K is a comfortable method for the atomic hydrogen production. The role of the different physicochemical processes taking place during dissociation of the molecular tritium interaction, atomic tritium transport to the target and its interaction with the molecules of the target is discussed. High selectivity of the atomic tritium interaction with the components of the different chemical nature target allowed such investigations to be made. The examples of atomic tritium use for the investigation into polymeric materials, absorption layers of surfactants, structure of biological macromolecules and hypomolecular formations are demonstrated [ru

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

Recent progress of China HCCB TBM tritium system

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

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.)

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

International Nuclear Information System (INIS)

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

1983-01-01

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

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

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.)

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 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

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)

Safety and environmental advantages of using tritium-lean targets for inertial fusion

International Nuclear Information System (INIS)

Arzeni, S.; Latkowski, J. F.; Logan, B. G.; Meier, W. R.; Moir, R. W.; Perkins, L. J.; Sanz, J.

1999-01-01

While traditional inertial fusion energy target designs typically use equimolar portions of deuterium and tritium and have areal densities (ρr) of ∼ 3 g/cm 2 , significant safety and environmental (S and E) advantages may be obtained through the use of high-density (ρr ∼ 10 g/cm 2 ) targets with tritium components as low as 0.5%. Such targets would absorb much of the neutron energy within the target and could be self-sufficient from a tritium breeding point of view. Tritium self-sufficiency within the target would free target chamber designers from the need to use lithium-bearing blanket materials, while low inventories within each target would translate into low inventories in target fabrication facilities. Absorption of much of the neutron energy within the target, the extremely low tritium inventories, and the greatly moderated neutron spectrum, make ''tritium-lean'' targets appear quite attractive from an S and E perspective

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

Thermal Conductivity of Ceramic Thermal Barrier and Environmental Barrier Coating Materials

Science.gov (United States)

Zhu, Dong-Ming; Bansal, Narottam P.; Lee, Kang N.; Miller, Robert A.

2001-01-01

Thermal barrier and environmental barrier coatings (TBC's and EBC's) have been developed to protect metallic and Si-based ceramic components in gas turbine engines from high temperature attack. Zirconia-yttria based oxides and (Ba,Sr)Al2Si2O8(BSAS)/mullite based silicates have been used as the coating materials. In this study, thermal conductivity values of zirconia-yttria- and BSAS/mullite-based coating materials were determined at high temperatures using a steady-state laser heat flux technique. During the laser conductivity test, the specimen surface was heated by delivering uniformly distributed heat flux from a high power laser. One-dimensional steady-state heating was achieved by using thin disk specimen configuration (25.4 mm diam and 2 to 4 mm thickness) and the appropriate backside air-cooling. The temperature gradient across the specimen thickness was carefully measured by two surface and backside pyrometers. The thermal conductivity values were thus determined as a function of temperature based on the 1-D heat transfer equation. The radiation heat loss and laser absorption corrections of the materials were considered in the conductivity measurements. The effects of specimen porosity and sintering on measured conductivity values were also evaluated.

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 breeding and release-rate kinetics from neutron-irradiated lithium oxide

International Nuclear Information System (INIS)

Quanci, J.F.

1989-01-01

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

Effect of multiple film on the tritium permeation property in 316L stainless steel

International Nuclear Information System (INIS)

Yao Zhenyu; Hao Jiakun; Zhou Changshan; Shan Changqi

2000-01-01

The films of TiN + TiC + TiN and TiN + TiC + SiO 2 were deposited on the surface of 316L stainless steel by physical vapor deposition technology. The characteristics of films are tested by SEM technology, it shows that the films are compact, thermal shock-resistant, oxidation-resistant and have good compatibility with bulk. the SIMS and IR analysis results show that the tritium permeation barrier is formed when TiC and SiO 2 films are annealed in hydrogen above 300 degree C. The tritium permeability in 316L with film is measured at various temperature, the results show that the tritium permeability in 316L with TiN + TiC + SiO 2 film is 4-6 orders of magnitude lower, and that in 316L with TiN + TiC + TiN film is 4-5 orders of magnitude lower than that in 316L with Pd film at about 200-600 degree C. These films may be used as the surface coating of the first wall, tritium blanket and heat exchanger in fusion reactor

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.)

Tritium system for compact high field devices

International Nuclear Information System (INIS)

Roccella, M.; Bonizzoni, G.; Chiesa, P.; Ghezzi, F.; Nassi, M.; Pavesi, U.; Amedeo, P.; Boschetti, G.; Giffanti, F.; Moriggio, A.

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-01

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

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)

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

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.)

Low technology high tritium breeding blanket concept

International Nuclear Information System (INIS)

Gohar, Y.; Baker, C.C.; Smith, D.L.

1987-10-01

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

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.

Separation of tritium from gaseous and aqueous effluent systems

International Nuclear Information System (INIS)

Kobisk, E.H.

1977-01-01

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

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)

Evaluating All-Metal Valves for Use in a Tritium Environment

Energy Technology Data Exchange (ETDEWEB)

Houk, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Payton, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-09-28