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Sample records for ceramic tritium breeders

  1. Tritium recovery from ceramic breeder blanket

    International Nuclear Information System (INIS)

    It is known that chemical forms of tritium released from ceramic breeders are T2O and T2. Among issues relevant to the tritium chemical form, tritium inventory is one of the major criteria in the selection of breeder material. The primary purpose of this report is to study the dependence of tritium inventory in a blanket with ceramic solid breeder on the tritium chemical form. In this light, tritium inventory in a Li2O blanket has been evaluated as a function of tritium chemical form under the conditions of the Japanese Fusion Experimental Reactor (FER). It was shown that in a blanket with Li2O as a breeder, which has a strong affinity to water vapor, the inventory due to T2O adsorption becomes quite large. In order to reduce the T2O adsorption inventory, conversion of the tritium chemical form through an isotope exchange reaction with hydrogen added to the sweep gas (T2O + 2 H2 → H2O + 2 HT) has been proposed, and its advantages and problems have been examined. Lithium hydroxide formation and mass transfer, which are considered to be inherent in the Li2O-T2O system and to be critical issues for the feasibility of a Li2O blanket, have been also discussed. (author)

  2. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Modeling of tritium behavior in ceramic breeder materials

    International Nuclear Information System (INIS)

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

  6. Extraction of tritium from ceramic breeder material

    International Nuclear Information System (INIS)

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

  7. Modeling of tritium behavior in ceramic breeder materials

    International Nuclear Information System (INIS)

    The model described in this paper considers diffusion and desorption as the rate-controlling mechanisms for tritium release from a ceramic breeder material. This model was used to predict the tritium release from samples of Li2SiO3 and LiAlO2, given the temperature history of the samples. The diffusion-desorption model did a better job of predicting the tritium release for these samples under pure helium purge gas than did a pure diffusion model using the best values for the diffusivity of these materials available. The activation energies of desorption found from the best fit of the predicted tritium release to the observed release were 105-108 kJ/mol for Li2SiO3 and 85.7 kJ/mol for LiAlO2. These values are in fair agreement with activation energies reported in the literature. 13 refs., 6 figs

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

    International Nuclear Information System (INIS)

    Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic 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

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

    International Nuclear Information System (INIS)

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

  10. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

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

    International Nuclear Information System (INIS)

    The potential sources of tritium contamination of the helium coolant of ceramic breeder blankets have previously been evaluated for the specific case of the European BIT DEMO blanket. This confirmed that the control of tritium losses to the steam circuit is a critical issue which demands development concerning (a) permeation barriers, (b) tritium recovery processes maintaining a very low tritium activity in the coolant, and (c) control of the coolant chemistry. The specifications of these developments required the evaluation of the tritium losses through the steam generators, and includes the definition of their operating conditions by thermodynamic cycle calculations, and their thermal-hydraulic design. For both tasks, specific computer tools were developed. The geometry obtained, the surface area and the 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. Our results indicate that in nominal steady-state operation the tritium escape into the steam cycle could be restricted to less than 10 Ci per day. The conditions for this are specified, but their feasibility demands, in particular, the resolution of certain gas chemistry problems, and their validation in the more stringent environment of an operating blanket. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) was identified as bearing a large tritium release potential. The problems associated with such transients are discussed and possible solutions are proposed. (orig.)

  12. Production of nuclear fusion reactor fuel by ceramic tritium breeder material

    International Nuclear Information System (INIS)

    Fuel tritium is generated from the nuclear reaction between the fusion neutron and the lithium of the breeder material arranged in the blanket that encloses the fusion plasma in the fusion reactor. However, the release process of the generated tritium has not been completely clarified. Recently, Japan Atomic Energy Agency started the tritium generation and recovery experiment in using nuclear fusion neutron source (FNS). In this report, the recent results of study on breeder material and its manufacturing technology is presented. (author)

  13. 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)

  14. Tritium percolation through porous ceramic breeders - a random-lattice approach

    International Nuclear Information System (INIS)

    Among the major processes leading to tritium transport through Li ceramic breeders the percolation of gaseous tritium species through the connected porosity remains the least amenable to a satisfactory treatment. The combination of diffusion and reaction through the convoluted transport pathways prescribed by the system of pores poses a formidable challenge. The key issue is to make the fundamental connection between the tortuousity of the medium with the transport processes in terms of only basic parameters (e.g., molecular diffusion coefficient and porosity distribution) that are amenable to fundamental understanding and experimental determinations. This fundamental challenge is met within the following approaches. On the microscale the short range transport is modeled via a connection-diffusion-reaction approach. On a macro scale the long range transport is described within a matrix formalism. The convoluted microstructure of the pore system as prescribed from experimental measurements is synthesized into the present approach via Monte Carlo simulation techniques. In this way the approach requires as inputs only physical-chemical parameters that are amenable to clear basic understanding and experimental determination. In this sense it provides predictive capability. Using this approach the concept of residence time has been analyzed in a critical manner. Implication for tritium release experiments was discussed. (orig.)

  15. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  18. Design, synthesis and characterization of the advanced tritium breeder: Li4+xSi1-xAlxO4 ceramics

    Science.gov (United States)

    Zhao, Linjie; Long, Xinggui; Chen, Xiaojun; Xiao, Chengjian; Gong, Yu; Guan, Qiushi; Li, Jiamao; Xie, Lei; Chen, Xiping; Peng, Shuming

    2015-12-01

    Li4+xSi1-xAlxO4 solid solutions which were designed as the advanced tritium breeder were obtained by solid state reactions. Samples were systematically characterized by various techniques. XRD, neutron diffraction and Raman results showed that the Aluminum substituted silicon into the Li4SiO4 lattice and Li+ interstitials formed as a result of charge compensation. Rietveld refinements of neutron diffraction showed that the crystalline structure had been expanded as Al-doped. Moreover, the lithium atom density, thermal conductivity and the mechanical property of the Li4+xSi1-xAlxO4 ceramics were improved relative to the Li4SiO4.

  19. Evaluation of tritium release properties of advanced tritium breeders

    International Nuclear Information System (INIS)

    Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li2TiO3) advanced tritium breeders with excess Li (Li2+xTiO3+y) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li2+xTiO3+y pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li2+xTiO3+y pebbles exhibited good tritium release properties similar to the Li2TiO3 pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)

  20. Research and development status of ceramic breeder materials

    International Nuclear Information System (INIS)

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was also recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option breeder material. Blanket design studies have indicated areas in the properties data base that need further investigation. Current studies are focusing on issues such as tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests are underway, some as part of an international collaboration for development of ceramic breeder materials. 36 refs

  1. Tritium transport in lithium ceramics porous media

    Energy Technology Data Exchange (ETDEWEB)

    Tam, S.W.; Ambrose, V.

    1991-12-31

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

  2. Tritium transport in lithium ceramics porous media

    Energy Technology Data Exchange (ETDEWEB)

    Tam, S.W.; Ambrose, V.

    1991-01-01

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

  3. Tritium transport in lithium ceramics porous media

    International Nuclear Information System (INIS)

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

  4. Solid breeder blanket design and tritium breeding

    International Nuclear Information System (INIS)

    Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements - to breed one tritium per fusion neutron - are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li2O, LiAlO2, Li2ZrO3, Li4SiO4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the amount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket. (orig.)

  5. Preliminary test for reprocessing technology development of tritium breeders

    International Nuclear Information System (INIS)

    In order to develop the reprocessing technology of lithium ceramics (Li2TiO3, CaO-doped Li2TiO3, Li4SiO4 and Li2O) as tritium breeder materials for fusion reactors, the dissolution methods of lithium ceramics to recover 6Li resource and the purification method of their lithium solutions to remove irradiated impurities (60Co) were investigated. In the present work, the dissolving rates of lithium from each lithium ceramic powder using chemical aqueous reagents such as HNO3, H2O2 and citric acid (C6H8O7 . H2O) were higher than 90%. Further the decontamination rate of 60Co added into the solutions dissolving lithium ceramics was higher than 97% using the activated carbon impregnated with 8-hydroxyquinolinol as chelate agent.

  6. Lithium reprocessing technology for ceramic breeders

    Science.gov (United States)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Saito, Minoru; Tatenuma, Katuyashi; Kainose, Mitsuru

    1995-03-01

    Lithium ceramics have been receiving considerable attention as tritium breeding materials for fusion reactors. Reprocessing technology development for these materials is proposed to recover lithium, as an effective use of resources and to remove radioactive isotopes. Four potential ceramic breeders (Li 2O, LiAlO 2, Li 2ZrO 3 and Li 4SiO 4) were prepared in order to estimate their dissolution properties in water and various acids (HCl, HNO 3, H 2SO 4, HF and aqua regia). The dissolution rates were determined by comparing the weight of the residue with that of the starting powder (the weight method). Recovery properties of lithium were examined by the precipitation method.

  7. Neutron irradiation of candidate ceramic breeder materials of fusion reactors

    International Nuclear Information System (INIS)

    In the context of the European programs for the future fusion reactors, the Process Chemistry Department of ENEA, Casaccia Center (Rome), has been involved in preparing ceramic blanket materials as tritium breeders; a special consideration has been addressed to the nuclear characterization of LiAlO2 and Li2ZrO3. In this paper are reported neutron irradiation of ceramic specimens in TRIGA reactor and γ-spectrometric measurements for INAA purposes; and isothermal annealing of the irradiated samples and tritium extraction, by using an 'out of pile' system. (author) 3 refs.; 4 figs.; 4 tabs

  8. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of ''the Sixth International Workshop on Ceramic Breeder Blanket Interactions'' which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: 1) fabrication and characterization of ceramic breeders, 2) properties data for ceramic breeders, 3) tritium release characteristics, 4) modeling of tritium behavior, 5) irradiation effects on performance behavior, 6) blanket design and R and D requirements, 7) hydrogen behavior in materials, and 8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li2TiO3, tritium release behavior of Li2TiO3 and Li2ZrO3 including tritium diffusion, modeling of tritium release from Li2ZrO3 in ITER condition, helium release behavior from Li2O, results of tritium release irradiation tests of Li4SiO4 pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  9. Proceedings of the fifteenth international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Fifteenth International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors. This workshop was held in Sapporo, Japan on 3-4, Sept. 2009. Twenty six participants from EU, Japan, India, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket development. By this workshop, advance of key technologies for solid breeder blanket development was shared among the participants. Also, desired direction of further investigation and development was recognized. The 20 of the presented papers are indexed individually. (J.P.N.)

  10. Tritium dynamics in fusion reactor solid breeder

    International Nuclear Information System (INIS)

    In the field of the NET research progrm, the chemical and diffusive processes involved in solid ceramic breeder materials have been analysed. A mathematical model describing the phenomena has been developed to obtain a quantitative evaluation for a first design approach. The data obtained by means of the above mentioned model are in good agreement with the data obtained by other research groups working in Europe and in United States. The computer codes BLANKET2, MC2, FWBC, have been developed to simulate the phenomena

  11. Simulation of physical parameter for in-pipe tritium breeder

    International Nuclear Information System (INIS)

    It is necessary to build in-pipe tritium breeder in our country in order to assess breeder material of tritium breeder module (TBM) and to find the release law of tritium. The irradiation vessel is one of the key components of TBM. The physical parameters about in-pipe tritium breeder were simulated with MCNP code. The values of the self-shielding factor, equivalent cross-section, daily production of tritium and total heating power are separately 0.435, 1.09 x 10-22 cm2, 2.8 x 1010 Bq and 8.2 kW. And they would provide necessary data for designing the irradiation vessel. (authors)

  12. Status of advanced tritium breeder development for DEMO in the broader approach activities in Japan

    International Nuclear Information System (INIS)

    DEMO reactors require '6Li-enriched ceramic tritium breeders' which have high tritium breeding ratios (TBRs) in the blanket designs of both EU and JA. Both parties have been promoting the development of fabrication technologies of Li2TiO3 pebbles and of Li4SiO4 pebbles including the reprocessing. However, the fabrication techniques of tritium breeders pebbles have not been established for large quantities. Therefore, these parties launch a collaborative project on scaleable and reliable production routes of advanced tritium breeders. In addition, this project aims to develop fabrication techniques allowing effective reprocessing of 6Li. The development of the production and 6Li reprocessing techniques includes preliminary fabrication tests of breeder pebbles, reprocessing of lithium, and suitable out-of-pile characterizations. The R and D on the fabrication technologies of the advanced tritium breeders and the characterization of developed materials has been started between the EU and Japan in the DEMO R and D of the International Fusion Energy Research Centre (IFERC) project as a part of the Broader Approach activities from 2007 to 2016. The equipment for production of advanced breeder pebbles is planned will be installed in the DEMO R and D building at Rokkasho, Japan. The design work in this facility was carried out. The specifications of the pebble production apparatuses and related equipment in this facility were fixed, and the basic data of these apparatuses was obtained. In this design work, the preliminary investigations of the dissolution and purification process of tritium breeders were carried out. From the results of the preliminary investigations, lithium resources of 90% above were recovered by the aqueous dissolving methods using HNO3 and H2O2. The removal efficiency of 60Co by the addition in the dissolved solutions of lithium ceramics were 97-99.9% above using activated carbon impregnated with 8-hydroxyquinolinol. In this report, preparation status

  13. Preliminary test for reprocessing technology development of tritium breeders

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi; Tsuchiya, Kunihiko; Hayashi, Kimio [Blanket Irradiation and Analysis Group, Directorates of Fusion Energy Research, Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Higashi Ibaraki-gun, Ibaraki 311-1393 (Japan); Nakamura, Mutsumi; Terunuma, Hitoshi [KAKEN Co., Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan); Tatenuma, Katsuyoshi [KAKEN Co., Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan)], E-mail: tatenuma@kakenlabo.co.jp

    2009-04-30

    In order to develop the reprocessing technology of lithium ceramics (Li{sub 2}TiO{sub 3}, CaO-doped Li{sub 2}TiO{sub 3}, Li{sub 4}SiO{sub 4} and Li{sub 2}O) as tritium breeder materials for fusion reactors, the dissolution methods of lithium ceramics to recover {sup 6}Li resource and the purification method of their lithium solutions to remove irradiated impurities ({sup 60}Co) were investigated. In the present work, the dissolving rates of lithium from each lithium ceramic powder using chemical aqueous reagents such as HNO{sub 3}, H{sub 2}O{sub 2} and citric acid (C{sub 6}H{sub 8}O{sub 7} . H{sub 2}O) were higher than 90%. Further the decontamination rate of {sup 60}Co added into the solutions dissolving lithium ceramics was higher than 97% using the activated carbon impregnated with 8-hydroxyquinolinol as chelate agent.

  14. 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)

    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)

  15. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

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

  16. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    International Nuclear Information System (INIS)

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C β-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ∼3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation

  17. Proceedings of the sixth international workshop on ceramic breeder blanket interactions

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji [ed.

    1998-03-01

    This report is the Proceedings of `the Sixth International Workshop on Ceramic Breeder Blanket Interactions` which was held as a workshop on ceramic breeders under Annex II of IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials, and Japan-US Workshop 97FT4-01. This workshop was held in Mito city, Japan on October 22-24, 1997. About forty experts from EU, Japan, USA, and Chile attended the workshop. The scope of the workshop included the following: (1) fabrication and characterization of ceramic breeders, (2) properties data for ceramic breeders, (3) tritium release characteristics, (4) modeling of tritium behavior, (5) irradiation effects on performance behavior, (6) blanket design and R and D requirements, (7) hydrogen behavior in materials, and (8) blanket system technology and structural materials. In the workshop, information exchange was performed for fabrication technology of ceramic breeder pebbles in EU and Japan, data of various properties of Li{sub 2}TiO{sub 3}, tritium release behavior of Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} including tritium diffusion, modeling of tritium release from Li{sub 2}ZrO{sub 3} in ITER condition, helium release behavior from Li{sub 2}O, results of tritium release irradiation tests of Li{sub 4}SiO{sub 4} pebbles in EXOTIC-7, R and D issues for ceramic breeders for ITER and DEMO blankets, etc. The 23 of the papers are indexed individually. (J.P.N.)

  18. Low temperature tritium release experiment from lithium titanate breeder material

    International Nuclear Information System (INIS)

    Engineering data of neutron irradiation performance are needed to design a fusion blanket. Of the engineering data, tritium release characteristic is one of the most important data. Therefore, tritium release experiments of the tritium breeding materials were carried out to evaluate the effects of various parameters, i.e. sweep-gas flow rate, irradiation temperature, hydrogen content in sweep gas and so on, on tritium release. Lithium titanate (Li2TiO3) is a candidate tritium breeding material for the blanket design of International Thermonuclear Experimental Reactor (ITER). As for the shape of the breeder material, a small spherical form is preferred to enhance tritium release from the breeder and to reduce the induced thermal stress in the breeder. Li2TiO3 pebbles with a diameter of 1mm and a total weight of ∼134g have been fabricated, and a pebble-pac assembly of the Li2TiO3 pebbles was irradiated in the Japan Materials Testing Reactor (JMTR), for 3 cycles (about 75 days). The tritium generated in breeder, and released from the breeder was swept downstream by the sweep gas for on-line analysis of tritium content. The total concentration and gaseous concentration of tritium released from the Li2TiO3 pebbles were measured, and HT/(HT+HTO) ratio was evaluated. The sweep-gas flow rate was changed from 10 to 1,000cm3/min, and hydrogen concentration in the sweep gas was changed from 100 to 10,000 ppm. The irradiation temperature of the outer region of the pebble-pac assembly was held below 450degC. The results showed that tritium release from the Li2TiO3 pebbles was started between 100 and 140degC and that the amount of released with increasing the irradiation temperature. The sweep-gas flow rate did not have an effect on tritium release from the Li2TiO3 pebble bed in the steady state. On the other hand, the hydrogen content in the sweep gas had an effect on the tritium release from the Li2TiO3 pebble bed. (author)

  19. Filbe molten salt research for tritium breeder applications

    International Nuclear Information System (INIS)

    This paper presents an overview of Flibe (2Lif·BeF2) 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)

  20. Modeling of tritium transport in a pin-type solid breeder blanket

    International Nuclear Information System (INIS)

    This section of the pin-type solid breeder blanket study is the first detailed attempt at modeling tritium inventory and release within a fusion reactor blanket based on actual tritium generation and thermal hydraulic profiles, rather, than a simple average unit cell extrapolation or some assumed exponential profiles. The DIFFUSE 83 code was found to give inventory results consistent with previous modeling efforts and with the general spherical grain model. The inventories for this blanket design calculated using DIFFUSE were found to be very satisfactory, less than 14 g at steady-state for a 129,000 kg LiAlO2 blanket. This result is reasonable compared with a BCSS LiALO2 blanket inventory calculated by GA Technologies. DIFFUSE was found to be very useful in approximating tritium inventories during transient startup/shutdown modes. The evaluations of transient inventories in this study appear to be the most detailed to date. The results suggest the need for controlling coolant flow during start-up to maintain high breeder temperatures and low tritium inventory, and the use of pre-heated coolant to bake out tritium inventory after shut-down. DIFFUSE modeling of breeder pins of 100% theoretical density indicates very limited tritium release from the LiAlO2 ceramic, suggesting batch processing of the pins for tritium extraction at the end of blanket lifetime. Preliminary analysis of other surface and radiative trapping effects shows DIFFUSE to be potentially a very useful tool in approximating and evaluating experimental results. Additional DIFFUSE analysis of these effects given the available experimental data is warranted

  1. In-situ tritium release (CORELLI-2 experiment) and ex-reactor ionic conductivity of substoichiometric LiAlO2 breeder ceramics

    International Nuclear Information System (INIS)

    LiAlO2 pellets with about 5% Li deficiency, prepared by a ''wet'' and a ''dry'' route were tested in situ for tritium release properties in nearly the same environmental conditions (CORELLI-2 experiment). Both the ''wet'' and ''dry'' route specimens were characterized by 80% of theoretical density (TD), almost fully open porosity and grain size ≤0.5 μm. The tritium removal rate evolution, following temperature or sweep gas changes during the irradiation, were observed to be nearly the same for both materials, in spite of their different preparation routes and impurities concentration. The ionic conductivities, as determined by impedance spectroscopy, were also similar. The presence of LiAl5O8 spinel phase in both samples apparently influenced the defect structure related transport properties of both lithium and tritium in these materials. (orig.)

  2. EXOTIC-7: irradiation of ceramic breeder materials to high lithium burnup

    International Nuclear Information System (INIS)

    The EXOTIC-7 irradiation experiment in the high flux reactor (HFR) has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characteristics of candidate ceramic breeder materials, originating from the fusion programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li2ZrO3, LiAlO2 and Li8ZrO6 and pebbles of Li4SiO4 and Li2ZrO3, with a variety of characteristics, like grain size and porosity. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules contained a mixture of Li4SiO4 and beryllium pebbles. The experimental design, sample loading and main irradiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented. (orig.)

  3. Proceedings of the eleventh international workshop on ceramic breeder blanket interactions

    International Nuclear Information System (INIS)

    This report is the Proceedings of 'the Eleventh International Workshop on Ceramic Breeder Blanket Interactions' which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. This workshop was held in Tokyo, Japan on December 15-17, 2003. About thirty experts from China, EU, Japan, Korea, Latvia, Russia and USA attended the workshop. The scope of the workshop included 1) evolutions in ceramic breeder blanket design, 2) progress in ceramic breeder material development, 3) irradiation testing, 4) breeder material properties, 5) out-of-pile pebble bed experiment, 6) modeling of the thermal, mechanical and tritium transfer behavior of pebble beds and 7) interfacing issues of solid breeder blanket. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of Li2TiO3 and so on, fabrication technology developments and characterization of the Li2TiO3 and Li4SiO4 pebbles, research on measurements and modeling of thermo-mechanical behaviors of Li2TiO3 and Li4SiO4 pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system. The 26 of the presented papers are indexed individually. (J.P.N.)

  4. A comprehensive model for the prediction of tritium behavior in solid breeder materials during steady-state and transient conditions

    International Nuclear Information System (INIS)

    In recent years, the area of tritium transport and release from Li-base ceramics in fusion blankets has become increasingly important particularly in conjunction with the growing amount of data available from in-pile tritium recovery experiments. Key variables that can strongly affect the tritium inventory and the kinetics of release, such as purge gas composition, temperature, solid breeder microstructure and activation energies for bulk diffusion and for desorption have been identified. Therefore, in the current phase of research and development, there is a strong incentive to develop comprehensive predictive capabilities in order to understand the new experimental data, to extrapolate these data to different ranges of conditions of interest, and to provide a necessary tool for fusion blanket design analysis. The objectives of this research are: (1) to develop new models for tritium transport in solid breeders to better describe the complex multistep phenomena that characterize tritium release, (2) to develop a computer code to predict tritium behavior, as a function of different variables and for a wide range of operating conditions, (3) to calibrate such models with existing experimental results. A comprehensive model is proposed. The sequence of transport processes leading to tritium release includes diffusion through the grain and along the grain boundaries, adsorption and desorption at the breeder surface and diffusion through the pore. A computer code called MISTRAL has been developed based on this model. The results obtained are in reasonable agreement with the experimental results, for the available set of property data, and indicate a fairly good predictive capability of the model for the analysis of several transients of interest for solid breeder fusion blankets

  5. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    International Nuclear Information System (INIS)

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits. (fusion engineering)

  6. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    Science.gov (United States)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.

  7. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    Science.gov (United States)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  8. Lithium ceramic of blankets intend for Russian fusion reactors and an influence of the ceramic properties on parameters of reactor tritium systems

    International Nuclear Information System (INIS)

    Russian Controlled Fusion Program involves development of a DEMO design and participation in ITER Project. A solid breeder blanket in DEMO contains a ceramic orthosilicate lithium breeder and a beryllium multiplier. Test Modules of the blanket are developed in a frame of ITER activities. Experimental models of tritium breeding zones (TBZ) for the Modules, materials and technology fabrication of the TBZ, tritium reactor systems to control and treat of gases released from lithium ceramic being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ were designed, manufactured and have been tested already in IVV-2M nuclear reactor. The first model consists of lithium orthosilicate ceramic sphere pebbles (1-1.5 mm diameter) and beryllium sphere (0.1 and 1.0 mm diameter). Ceramic cylindrical pellets (11 mm diameter and 10 mm height) and porous beryllium (20% porosity) are in the second model. Some properties and microstructure of the ceramic elements are performed. Initial results of some changes of ceramic structure and in-pile experimental ratio of hydrogen and oxygen form of tritium release in helium/neon purge gas are presented. These results and outcome of irradiated LiAlO2, Li4SiO4 and Li2SiO3 ceramics in a water-graphite nuclear reactor are considered to be a DATE BASE for development of the Test Modules and the DEMO blanket and influence of the kinetic tritium release parameters on DEMO tritium systems are discussed. (author)

  9. EXOTIC-7: Irradiation of ceramic breeder materials to high lithium burnup

    International Nuclear Information System (INIS)

    The EXOTIC-7 irradiation experiment in the High Flux Reactor (HFR) at Petten has been completed. Its aim has been to investigate the effects of high lithium-burnup on the mechanical stability and tritium release characterisitcs of candidate ceramic breeder materials, originating from the Fusion Programmes of CEA, FZK, ENEA, AECL and ECN. The tested ceramic breeder materials were pellets of Li2ZrO3, LiAlO2 and Li8ZrO6 and pebbles of Li4SiO4 and Li2ZrO3, with a variety of characteristics, like grain size and porostiy. The test matrix provided the simultaneous irradiation of eight independent capsules with on-line tritium monitoring. Two capsules containd a mixture of Li4SiO4 and beryllium pebbles. The experimental design, sample loading and main riiadiation parameters are described. Some PIE results and analysis of in-situ tritium release behaviour are presented. (orig.)

  10. Fluorine 18 in tritium generator ceramic materials

    International Nuclear Information System (INIS)

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

  11. Research of tritium gas releasing from lithium ceramics in the course of reactor irradiation

    International Nuclear Information System (INIS)

    The results of experimental researches of tritium gas releasing from ceramic breeder ITER material- lithium ceramics Li2TiO3 - under the conditions of irradiation on the reactor IVG.1M are presented in the work. The lithium ceramics samples were of sphere shape with diameter 2 mm with enrichment on isotope 6Li 7%. Gas releasing was measured using radio-frequency mass-spectrometer MH-6407P at the temperature range from 20 deg.C to 1000 deg.C. The outcomes of the researches showed that tritium gas releasing process depends on temperature and fluence of neutrons and at that helium generated in the samples simultaneously with tritium almost does not release. author)

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

    International Nuclear Information System (INIS)

    A first estimate for the tritium release behaviour of a ceramic breeder blanket in pulsed operation is obtained by assuming a linear steady state temperature distribution and taking into account the time constant of the thermal behaviour. The release behaviour of the breeder exposed to consecutive periods of tritium generation is described with an analytical solution of the diffusion equation. The results are compared with a simple exponential approach valid for surfacte desorption controlled release. The exponential model is used to simulate a blanket with aluminate as breeder material, which takes longest to reach steady state. The simulation demonstrates that a significant fraction (>67%) of steady state can be achieved after a testing time of about one day. (author). 7 refs.; 8 figs.; 3 tabs

  13. Tritium solubility and permeation in high retention fusion reactor breeder elements

    International Nuclear Information System (INIS)

    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 104 Pa and 1 Pa appear possible with an element containing a high solubility material provided it is kept below about 4000C. This should lead to a leakage into the coolant of between 10 Ci day-1 and 104 Ci day-1 which is considerably less than the 107 Ci day-1 in present designs. (author)

  14. Lithium-based oxide ceramics for tritium-breeding applications

    International Nuclear Information System (INIS)

    Material preparation techniques, crystallographic data, phase diagrams, metal compatibility, and thermal properties have been assembled for the lithium-based oxide ceramics designated as potential solid tritium breeders for fusion devices. The materials discussed in this report include: Li2O, β-Li5AlO4, γ-LiAlO2, Li4SiO4, Li2SiO3, Li4TiO4, Li2TiO3, Li8ZrO6, Li4ZrO4, and Li2ZrO3. The thermal properties covered were vaporization, thermal conductivity, specific heat, and linear thermal expansion. There has been no attempt to rank the above mentioned candidates, but rather to merely indicate points that must be considered when using the various materials as solid breeders. These encompass low lithium atom densities, destructive phase transformations, a higher thermal expansion, low thermal conductivity, excessive vaporization at low temperatures, corrosive nature toward metals and difficulty in sample preparation

  15. Exotic development of ceramic tritium breeding materials

    International Nuclear Information System (INIS)

    In the near future fusion reactors will be based on the tritium-deuterium plasma reaction. As such the production of tritium, a non-natural element, becomes of crucial importance in fusion technology. An experimental programme, EXOTIC, is in progress since 1983 in which the laboratories of SNL-Springfields, ECN-Petten, JRC-Petten and SCK-CEN Mol work in close collaboration within the framework of the manufacture, characterization and irradiation of ceramic lithium compounds. This programme must result in the understanding of the tritium release processes and the effect of material characteristics on this release. Up to now three irradiation experiments in the High Flux Reactor-Petten were scheduled (EXOTIC I, II and III). The Annual Progress Report 1985 summarizes information on these experiments during the period of 1985. The reader is also referred to the previous Annual Progress Report 1984. During the EXOTIC I experiment lithium silicate pellets and lithium aluminate pellets were irradiated. The resulting tritium release data are still to be interpreted in full detail. Some preliminary observations are presented in this Report. In the EXOTIC II experiment lithium oxide, lithium aluminate and lithium silicate pellets were used. In the EXOTIC III experiment lithium oxide, lithium zirconate and lithium silicate pellets were used. (Author)

  16. R and D status on Water Cooled Ceramic Breeder Blanket Technology

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio, E-mail: enoeda.mikio@jaea.go.jp; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu; Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji

    2014-10-15

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li{sub 2}TiO{sub 3} pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

  17. R and D status on Water Cooled Ceramic Breeder Blanket Technology

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-02-01

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

  19. Recovery Method of Bred Tritium from Solid Breeders

    International Nuclear Information System (INIS)

    It is required to develop an efficient tritium fueling cycle keeping the overall tritium breeding ratio larger than 1.0 and a reliable tritium confinement system assuring the radiation safety of tritium in construction of the D-T fusion reactor. The blanket is the place where the tritium recovery system has contact with the cooling system for electricity generation at the elevated temperature. Therefore, design of efficient means to recover bred tritium with minimum permeation loss is to be made.It is proposed in this study to construct a recovery system using the Pd alloy with adsorption bed after a precious metal catalyst bed. Effects of existence of water on dissociation reaction of hydrogen on palladium alloy membrane and on recombination reaction are discussed in this study for the case when 800 Pa of water vapor is introduced to the permeation primary side and/or permeation secondary side for the case when water vapor co-existed, and it was observed that water vapor prevents hydrogen permeation through palladium alloy at the lower temperature than 473K

  20. Trial examination of direct pebble fabrication for advanced tritium breeders by the emulsion method

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi, E-mail: hoshino.tsuyoshi@jaea.go.jp

    2014-10-15

    Highlights: • The integration of raw material preparation and granulation is proposed as a new direct pebble fabrication process. • The emulsion method granulates gel spheres of Li{sub 2}CO{sub 3} and TiO{sub 2} or SiO{sub 2}. • The gel spheres are calcined and sintered in air. • The crush load of the sintered Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} pebbles obtained is 37.2 or 59.3 N, respectively. - Abstract: Demonstration power plant reactors require advanced tritium breeders with high thermal stability. For the mass production of advanced tritium breeder pebbles, pebble fabrication by the emulsion method is a promising technique. To develop the most efficient pebble fabrication method, a new direct pebble fabrication process utilizing the emulsion method was implemented. A prior pebble fabrication process consisted of the preparation of raw materials followed by granulation. The new process integrates the preparation and granulation of raw materials. The slurry for the emulsion granulation of Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} as a tritium breeder consists of mixtures of Li{sub 2}CO{sub 3} and TiO{sub 2} or SiO{sub 2} at specific ratios. Subsequently, gel spheres of tritium breeders are fabricated by controlling the relative flow speeds of slurry and oil. The average diameter and crush load of the obtained sintered Li{sub 2}TiO{sub 3} or Li{sub 4}SiO{sub 4} pebbles were 1.0 or 1.5 mm and 37.2 or 59.3 N, respectively. The trial fabrication results suggest that the new process has the potential to increase the fabrication efficiency of advanced tritium breeder pebbles.

  1. Trial examination of direct pebble fabrication for advanced tritium breeders by the emulsion method

    International Nuclear Information System (INIS)

    Highlights: • The integration of raw material preparation and granulation is proposed as a new direct pebble fabrication process. • The emulsion method granulates gel spheres of Li2CO3 and TiO2 or SiO2. • The gel spheres are calcined and sintered in air. • The crush load of the sintered Li2TiO3 or Li4SiO4 pebbles obtained is 37.2 or 59.3 N, respectively. - Abstract: Demonstration power plant reactors require advanced tritium breeders with high thermal stability. For the mass production of advanced tritium breeder pebbles, pebble fabrication by the emulsion method is a promising technique. To develop the most efficient pebble fabrication method, a new direct pebble fabrication process utilizing the emulsion method was implemented. A prior pebble fabrication process consisted of the preparation of raw materials followed by granulation. The new process integrates the preparation and granulation of raw materials. The slurry for the emulsion granulation of Li2TiO3 or Li4SiO4 as a tritium breeder consists of mixtures of Li2CO3 and TiO2 or SiO2 at specific ratios. Subsequently, gel spheres of tritium breeders are fabricated by controlling the relative flow speeds of slurry and oil. The average diameter and crush load of the obtained sintered Li2TiO3 or Li4SiO4 pebbles were 1.0 or 1.5 mm and 37.2 or 59.3 N, respectively. The trial fabrication results suggest that the new process has the potential to increase the fabrication efficiency of advanced tritium breeder pebbles

  2. Development of fabrication technologies for advanced tritium breeder pebbles by the sol–gel method

    International Nuclear Information System (INIS)

    Highlights: • Li2TiO3 with excess Li (Li2+xTiO3+y) was developed as an advanced tritium breeder. • Pebble fabrication by the sol–gel method is a promising technique for the mass production of advanced tritium breeder pebbles. • To increase the density of the sintered Li2+xTiO3+y pebbles, the sintering temperature was changed. • At 1353 K, the density of the pebbles increased to approximately 85% T.D. without any increase in the grain size. -- Abstract: Demonstration power plant (DEMO) reactors require advanced tritium breeders with high thermal stability. Li2TiO3 with excess Li (Li2+xTiO3+y) was developed as an advanced tritium breeder. Pebble fabrication by the sol–gel method is a promising technique for the mass production of advanced tritium breeder pebbles. I have been developing a sol–gel technique for fabricating Li2+xTiO3+y pebbles, and the next step is to optimize the granulation conditions to reach the target value. In a previous study, the average grain size on the surfaces and cross sections of sintered Li2+xTiO3+y pebbles was large whereas the theoretical density (T.D.) of these pebbles was small. To increase the density of the sintered Li2+xTiO3+y pebbles, the sintering temperature was changed, and at 1353 K, the density of the pebbles increased to approximately 85% T.D. without any increase in the grain size. This suggests that the pore size in the sintered Li2+xTiO3+y pebbles decreased because of sintering in vacuum and argon

  3. Particle flow of ceramic breeder pebble beds in bi-axial compression experiments

    International Nuclear Information System (INIS)

    Pebble beds of Tritium breeding ceramic material are investigated within the framework of developing solid breeder blankets for future nuclear fusion power plants. For the thermo-mechanical characterisation of such pebble beds, bed compression experiments are the standard tools. New bi-axial compression experiments on 20 and 30 mm high pebble beds show pebble flow effects much more pronounced than in previous 10 mm beds. Owing to the greater bed height, conditions are reached where the bed fails in cross direction and unhindered flow of the pebbles occurs. The paper presents measurements for the orthosilicate and metatitanate breeder materials that are envisaged to be used in a solid breeder blanket. The data are compared with calculations made with a Drucker-Prager soil model within the finite-element code ABAQUS, calibrated with data from other experiments. It is investigated empirically whether internal bed friction angles can be determined from pebble beds of the considered heights, which would simplify, and broaden the data base for, the calibration of the Drucker-Prager pebble bed models

  4. Numerical simulation of ceramic breeder pebble bed thermal creep behavior

    International Nuclear Information System (INIS)

    The evolution of ceramic breeder pebble bed thermal creep deformation subjected to an external load and a differential thermal stress was studied using a modified discrete numerical code previously developed for the pebble bed thermomechanical evaluation. The rate change of creep deformation was modeled at the particle contact based on a diffusion creep mechanism. Numerical results of strain histories have compared reasonably well with those of experimentally observed data at 740 C using activation energy of 180 KJ/mole. Calculations also show that, at this activation energy level, a particle bed at an elevated temperature of 800 C may cause undesired local sintering at a later time when it is subjected to an external load of 6.3 MPa. Thus, by tracking the stress histories inside a breeder pebble bed the numerical simulation provides an indication of whether the bed may encounter an undesired condition under a typical operating condition. (orig.)

  5. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    International Nuclear Information System (INIS)

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D2 gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  6. Studies on tritium breeding ratio for solid breeder blanket cooled by pressurized water through nuclear and thermal analyses

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) has been performing the research, development and design of blankets with water-cooled solid breeder for fusion power plant as a leading institute in Japan, according to the long-term R and D program established by the Fusion Council in 1999. For our design, pebbles of a ceramic tritium breeder (Li2TiO3) and a beryllium neutron multiplier (Be) are packed in the constitutive layer structures of a test blanket module (TBM) for ITER. These reports are results of one-dimensional nuclear and thermal analyses on the TBM emphasizing on optimized configuration of the breeder and multiplier layers. Taking into account increment of TBR, the radial widths of the breeder and multiplier layers are optimized. The main results of our study are as follows: (1) In multilayered structures of pebble beds, existence of the peak of the TBR was revealed within the range of the volume ratio R=V(Be)/V(Li2TiO3)=4-5. (2) In the case of optimized layer structure for the single packing, a layer of Be was set to be the two layers behind a layer of Li2TiO3. The R became available for staying in the range of R=4-5. Consequently, the TBR respectively increased by 2.0%, 3.2% and 4.0% with 7.5%(nature), 40% and 90% of enrichment of 6Li compared to TBR of TBM in which the layers of Be and Li2TiO3 were interlaminated. This database of TBR for optimized layer structure contributes to the estimation of TBR at the design stage of the TBM and demonstration blanket aimed to strengthen the commercial competitiveness and technical feasibility. (author)

  7. Progress of R and D on the technology of In-pile irradiation and tritium In-situ extraction experiment of solid breeder pebble bed for CN HCCB in CARR

    International Nuclear Information System (INIS)

    The progress of the key technology of the In-Pile Irradiation and Tritium In-Situ Extraction (IPITISE) experiment was introduced. According to the design and requirements of the Helium cooled ceramic breeder (HCCB) tritium system, the scheme of the IPITISE experiment was established. The primary components of this apparatus included pebble bed assembly (PBA), tritium extraction system (TES), and tritium measurement system (TMS). The primary design and calculation of the structure, nuclear physics, and thermo-hydraulics of the PBA were carried out. It can be concluded that the max weight load of the PBA was 150 g above. The effective thermal neutron flux rate of the PBA would be high as 1.73 E +14 n/cm2 · s. The power of irradiation heat could be reached 18.47 W/cm3. After optimization of the design and parameter of PBA structure, the temperature of reduced activation ferrite/martensite (RAFM) steel with tritium permeation barrier (TPB) coating and Li4SiO4 could be respective controlled under 150∼550℃ and 200∼ 920℃. The in-situ tritium release behaviors would be studied by this experiment as well as the tritium permeation through the structure materials under irradiation condition or the reactor was shut down. Consequently, the irradiation performance of the key materials, the retention characteristics and release behaviors of Li4SiO4 ceramic breeder pebbles, the tritium permeation data of RAFM with TPB coating and credible evaluation of in-situ tritium extraction technology would be provided for China tritium breeder test blanket module in the CIPITISE experiment. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-04-30

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

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

    International Nuclear Information System (INIS)

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

  10. Tritium permeation and recovery for the helium-cooled molten salt fusion breeder

    International Nuclear Information System (INIS)

    Design concepts are presented to control tritium permeation from a molten salt/helium fusion breeder reactor. 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. 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. A research and development effort is needed to prove feasibility or to develop alternative barriers. A 1 mm aluminum sleeve is proposed to suppress permeation through the steam generator tubes. This gives a calculated reduction factor of more than 500 relative to bare steel, including a factor of 30 due to an assumed oxide layer. The permeation equations are developed in detail for a multi-layer tube wall including a frozen salt layer and with two fluid boundary-layer resistances. Conditions are discussed for which Sievert's or Henry's Law materials become flux limiters. An analytical model is developed to establish the tritium split between wall permeation and reactor-tube flow

  11. Numerical simulation of ceramic breeder pebble bed thermal creep behavior

    International Nuclear Information System (INIS)

    The evolution of ceramic breeder pebble bed thermal creep deformation subjected to an external load and a differential thermal stress was studied using a modified discrete numerical code previously developed for the pebble bed thermomechanical evaluation. The rate change of creep deformation was modeled at the particle contact based on a diffusion creep mechanism. Numerical results of strain histories have shown lower values as compared to those of experimentally observed data at 740 deg. C using an activation energy of 180 kJ/mol. Calculations also show that, at this activation energy level, a particle bed at an elevated temperature of 800 deg. C may cause too much particle overlapping with a contact radius growth beyond 0.65 radius at a later time, when it is subjected to an external load of 6.3 MPa. Thus, by tracking the stress histories inside a breeder pebble bed the numerical simulation provides an indication of whether the bed may encounter an undesired condition under a typical operating condition

  12. 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.

  13. A uranium bed with ceramic body for tritium storage

    International Nuclear Information System (INIS)

    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

  14. Tritium concentration monitoring of the purge gas stream of HCPB breeder blankets in future fusion reactors

    International Nuclear Information System (INIS)

    In fusion technology it is necessary to monitor tritiated gases for process monitoring. Such a system should be able to monitor the gas without taking samples. It should also be compact, cheap, the system stability should be excellent and it should recognize changes in the activity fast. Standard tools for activity measurements are ionization chambers and calorimeters. Ionization chambers work without sample taking but they are gas species dependent. Also pressures in the 100 mbar range are needed. Calorimeters are not suitable to be used as process monitors and it takes several hours to get a result. For activity measurements with a calorimeter it is necessary to extract gas samples. The Tritium Activity Chamber Experiment (TRACE) is a specially designed prototype to monitor traces of tritium in a gas sample utilizing Beta Induced X-Ray Spectroscopy (BIXS). Future fusion plants like ITER or DEMO could use such a system to monitor the purge gas streams in HCPB breeder blankets. TRACE will explore the possibility to monitor the expected 10 ppm tritium in the helium purge gas stream. We will evaluate if a BIXS system can be used as a standard monitoring system for tritiated gases in the range of (10-5-100) mbar tritium partial pressure.

  15. Zeolite membranes and palladium membrane reactor for tritium extraction from the breeder blankets of ITER and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D., E-mail: david.demange@kit.edu; Borisevich, O.; Gramlich, N.; Wagner, R.; Welte, S.

    2013-10-15

    Highlights: • We present a new concept to recover tritium from the helium in breeder blankets. • Zeolite membranes are fully tritium compatible and can pre-concentrate tritiated molecules. • PERMCAT catalytic membrane reactor recovers tritium to be reused in the fuel cycle. -- Abstract: While the tritium technology for the inner DT fuel cycle of fusion reactors shall be demonstrated in ITER, the tritium management in the breeder blanket remains very challenging. Most of the process options rely on ad(b)sorption/desorption cycles, using dedicated packed beds to handle separately the molecular and oxide forms of tritium. This approach seems satisfactory for ITER, but seems difficult to scale up to DEMO. The alternative use of a catalytic membrane reactor in combination with inorganic membranes would simplify and improve the overall tritium management. Zeolite membranes should enable in a single step the pre-concentration of all tritiated species. This tritium enriched stream could be afterwards processed using PERMCAT (catalytic Pd-based membrane reactor) to finally recover the tritium in its pure molecular form. This paper discusses at the conceptual level such approach. The latest experimental results on zeolite membrane and multi-tube PERMCAT reactor are presented. Next R and D activities for technical scale demonstrations and refined simulation tools are proposed to finally estimate the sizes of the components to be operated in ITER and DEMO.

  16. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    Science.gov (United States)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  17. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    International Nuclear Information System (INIS)

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of 115In(n, n′)115mIn reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured 115In(n, n′)115mIn reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from 6Li and 7Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% 6Li and 7.54% 6Li) in Li2CO3. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from 6Li at one location in the breeder layer was also measured by direct online measurement of tritons from 6Li(n, t)4He reaction using silicon surface barrier detector and 6Li to triton converter. Additional verification of neutron spectra (En > 0.35 MeV) in the mock-up zones were obtained by measuring 115In(n, n′)115mIn reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li2CO3 pellets was 1.11 in first breeder zone and 1.09 in second breeder zone with uncertainty 8.3% at 1σ level. The experimental details

  18. Tritium breeding mock-up experiments containing lithium titanate ceramic pebbles and lead irradiated with DT neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jakhar, Shrichand; Abhangi, M.; Tiwari, S. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Makwana, R. [Department of Physics, MS University, Vadodara (India); Chaudhari, V.; Swami, H.L.; Danani, C.; Rao, C.V.S.; Basu, T.K. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India); Mandal, D.; Bhade, Sonali; Kolekar, R.V.; Reddy, P.J. [Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Bhattacharyay, R.; Chaudhuri, P. [Institute for Plasma Research, Bhat, Gandhinagar 382 428 (India)

    2015-06-15

    Highlights: • Breeding benchmark experiment on LLCB TBM in ITER was performed. • Nuclear responses measured are TPR and reaction rate of {sup 115}In(n, n′){sup 115m}In reaction. • Measured responses are compared with calculations by MCNP and FENDL 2.1 library. • TPR measurements agree with calculations in the estimated error bar. • Measured {sup 115}In(n, n′){sup 115m}In reaction rates are underestimated by the calculations. - Abstract: Experiments were conducted with breeding blanket mock-up consisting of two layers of breeder material lithium titanate pebbles and three layers of pure lead as neutron multiplier. The radial dimensions of breeder, neutron multiplier and structural material layers are similar to the current design of the Indian Lead–Lithium cooled Ceramic Breeder (LLCB) blanket. The mock-up assembly was irradiated with 14 MeV neutrons from DT neutron generator. The local tritium production rates (TPR) from {sup 6}Li and {sup 7}Li in breeder layers were measured with the help of two different compositions of Li isotopes (60.69% {sup 6}Li and 7.54% {sup 6}Li) in Li{sub 2}CO{sub 3}. Tritium production in the multiplication layers were also measured with above mentioned two types of pellets to compare the experimental tritium production with calculations. TPR from {sup 6}Li at one location in the breeder layer was also measured by direct online measurement of tritons from {sup 6}Li(n, t){sup 4}He reaction using silicon surface barrier detector and {sup 6}Li to triton converter. Additional verification of neutron spectra (E{sub n} > 0.35 MeV) in the mock-up zones were obtained by measuring {sup 115}In(n, n′){sup 115m}In reaction rate and comparing it with calculated values in all five layers of mock-up. All the measured nuclear responses were compared with transport calculations using code MCNP with FENDL2.1 and FENDL3.0 cross-section libraries. The average C/E ratio for tritium production in enriched Li{sub 2}CO{sub 3} pellets was 1

  19. Influence of chemisorption products of carbon dioxide and water vapour on radiolysis of tritium breeder

    International Nuclear Information System (INIS)

    Highlights: • Chemisorption products affect formation proceses of radiation-induced defects. • Radiolysis of chemisorption products increase amount of radiation-induced defects. • Irradiation atmosphere influence radiolysis of lithium orthosilicate pebbles. - Abstract: Lithium orthosilicate pebbles with 2.5 wt% excess of silica are the reference tritium breeding material for the European solid breeder test blanket modules. On the surface of the pebbles chemisorption products of carbon dioxide and water vapour (lithium carbonate and hydroxide) may accumulate during the fabrication process. In this study the influence of the chemisorption products on radiolysis of the pebbles was investigated. Using nanosized lithium orthosilicate powders, factors, which can influence the formation and radiolysis of the chemisorption products, were determined and described as well. The formation of radiation-induced defects and radiolysis products was studied with electron spin resonance and the method of chemical scavengers. It was found that the radiolysis of the chemisorption products on the surface of the pebbles can increase the concentration of radiation-induced defects and so could affect the tritium diffusion, retention and the released species

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

    International Nuclear Information System (INIS)

    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 H2/H2O with relative concentrations varying from pure H2 to pure H2O) on (a) Tritium removal from LiAlO2 and Li2ZrO3; (b) Tritium permeation through AISI-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. (orig.)

  1. Current status of safety design and safety analysis for China ITER helium coolant ceramic breeder test blanket system long

    International Nuclear Information System (INIS)

    Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) designed by China are planned to be tested in ITER to validate key technologies, including demonstration of nuclear safety, for future fusion reactor breeding blankets. Furthermore, in order to be operated in ITER, a nuclear facility (INB) recognized by French nuclear safety authority, safety design and safety analysis of the TBS are mandatory for the licensing procedures. This paper summarizes the status at current design phase with following main elements: The main radiological source terms in the system are tritium and activation products. Nuclear and tritium analysis are performed to identify their inventories and distributions in system. Multiple confinement barriers are considered to be the most essential safety feature. French regulation for pressure equipment and nuclear equipment (ESP/ESPN regulations) will be followed to ensure the system integrities. ALARA principle is kept in mind during the whole safety design phases. Protective actions including choice of advanced materials, improvement of shielding, optimization of operation and maintenance activities, usage of remote handling operations, zoning and access control have been considered. Passive safety is emphasized in the system design, only minimal active safety functions including call for fusion plasma shutdown and isolation of TBM from ex-vessel ancillary systems. High reliability and redundancies are required for components related to these functions. Several accidents have been identified and analyzed. Consider the limited inventories in the system and the intrinsic safety of fusion device, positive conclusions have been obtained. (author)

  2. Activation characteristics and waste management options for some candidate tritium breeders

    International Nuclear Information System (INIS)

    Activation and transmutation characteristics are calculated for the candidate breeder compositions Li2O, LiAlO2, Li2SiO3, Li2ZrO3, LiVO3 and 17Li-83Pb. Irradiation conditions comprise a 2.5 y continuous exposure to the neutron flux appropriate to the outboard blanket zone of the EEF reference reactor with an assumed first wall neutron loading of 5 MW m-2. Results are presented for specific activity, surface γ-dose rate, ingestion and inhalation doses and compositional changes. Neglecting any retained tritium, activity is least for Li2 and LiVO3 and greatest for Li2ZrO3 and 17Li-83Pb. The silicate and aluminate are intermediate in level. Following reactor service, all the materials should be suitable, after appropriate conditioning, for geological disposal as Intermediate Level Waste. Alternatively, they could be considered for recycling to reclaim the unused lithium. In all cases, recycling is probably feasible within 10 y of removal from service and should be easier for the oxide silicate and vanadate. (orig.)

  3. Status and perspective of the R and D on ceramic breeder materials for testing in ITER

    International Nuclear Information System (INIS)

    The main line of ceramic breeder materials research and development is based on the use of the breeder material in the form of pebble beds. At present, there are three candidate pebble materials (Li4SiO4, and two forms of Li2TiO3) for DEMO reactors that will be used for testing in ITER. This paper reviews the R and D of as-fabricated pebble materials against the blanket performance requirements and makes recommendations on necessary steps toward the qualification of these materials for testing in ITER

  4. Corrosion susceptibility of EUROFER97 in lithium ceramics breeders

    International Nuclear Information System (INIS)

    EUROFER97 specimens were exposed in vacuum to lithium silicate pebbles at 550 °C for up to 2880 h, to evaluate its corrosion susceptibility in a simulated breeder blanket environment. The specimens and pebble bed were then analyzed and characterized by SEM-EDX, XRD, and HR-TEM. The results revealed the formation of a double chromium/iron oxide corrosion layer. HR-TEM also showed that the inner layer was amorphous, while the outer was crystalline. The amorphous layer was brittle, broke easily, and became detached from the steel

  5. Tritium release from beryllium discs and lithium ceramics irradiated in the SIBELIUS experiment

    International Nuclear Information System (INIS)

    The SIBELIUS experiment was designed to obtain information on the compatibility between beryllium and ceramics, as well as beryllium and steel, in a neutron environment. This experiment comprised irradiation of eight capsules, seven of which were independently purged with a He/0.1% H2 gas mixture. Four capsules were used to examine beryllium/ceramic (Li2O, LiAlO2, Li4SiO4, and Li2ZrO3) and beryllium/steel (Types 316L and 1.4914) compacts. Isothermal anneal experiments have been run on representative beryllium and ceramic disks from each of the four capsules at 550 degrees C to 850 degrees C in steps of 100 degrees C. The results indicate that tritium release from the beryllium did not exhibit burst release behavior, as previously reported, but rather a progressive release with increasing temperature. Generally, ∼99% of the tritium was released by 850 degrees C. Tritium release from the ceramic discs was quite similar to the behavior shown in other dynamic tritium release experiments on lithium ceramics. The tritium content in beryllium discs adjacent to a steel sample was found to be significantly lower than that found in a beryllium disc adjacent to a ceramic sample. Recoil of tritium from the ceramic into the beryllium appears to be the source of tritium entering the beryllium, probably residing in the beryllium oxide layer

  6. Thermodynamics of ceramic breeder materials for fusion reactors

    International Nuclear Information System (INIS)

    Based on known or deduced phase relationships in ternary lithium oxygen systems such as Li-Al-O, Li-Si-O and Li-Zr-O, the unknown free enthalpy of formation values of ternary compounds are calculated starting from the known data of the compounds of the binary border systems. Criterion for the data assessment is interconsistency of the data of all the compounds within a given multi-component system. With the help of these data the development of partial pressures during the breeding process can be calculated for all the compounds of interest. In order to facilitate a compatibility assessment the quaternary systems Cr-Li-Si-O, Fe-Li-Si-O and Be-Li-Si-O were also investigated and thermodynamic data of pertinent ternary and quaternary compounds determined. Both the partial pressure development and the compatibility behaviour of a lithium containing compound are criteria for its qualification as a breeder material for a fusion reactor. (orig.)

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

    International Nuclear Information System (INIS)

    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 H2/H2O with relative concentrations varying from pure H2 to pure H2O) on tritium removal from LiAlO2 and Li2ZrO3; 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

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

    International Nuclear Information System (INIS)

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

  9. Sensitivity and Uncertainty Analyses of the Tritium Production in the HCPB Breeder Blanket Mock-up Experiment

    International Nuclear Information System (INIS)

    Dedicated computational methods, tools and data have been recently developed in the framework of the European Fusion Technology Programme to enable sensitivity and uncertainty analyses of fusion neutronics experiments. severely limited due to these two requirements. (author)er productgeneration and the associated uncertainties against the experimental data provided in the neutronics experiment at the Frascati Neutron Generator on a mock-up of the HCPB (Helium-Cooled Pebble Bed) breeder test blanket. This work is devoted to the computational analyses of this experiment comprising the following steps: (i) Calculation of the Tritium production rates (TPR) in the Li2CO3 pellets using a detailed 3D model of the experimental set-up; the Monte Carlo code MCNP and the discrete ordinates code TORT were applied for these calculations with EFF-3 and FENDL-2.0/2.1 nuclear data. (ii) Sensitivity calculations for the Li2CO3 pellets stacks to assess the sensitivity of the Tritium production to the reactions cross-sections of the involved nuclides Be, 6,7Li, C and O; the calculations were performed with the MCSEN Monte Carlo code using the track length estimator and, in parallel, with the deterministic SUSD3D code using neutron fluxes calculated by TORT in forward and adjoint mode. (iii) Calculations of the data related uncertainties of the TPR using co-variance data from EFF (9Be, 6Li, 12C), FENDL-2 (7Li) and JENDL-3.3 (16O); both probabilistic (MCNP/MCSEN) and deterministic (TORT/SUSD3D) approaches were applied. (iv) Assessment of the total uncertainties for the TPR including uncertainties of the measurements, the nuclear data and the calculations. The data related uncertainties of the calculated Tritium generation are in the order of 4 - 5 % (2 sigma). The main uncertainties are due to the Be cross-section data. The total uncertainties of the predicted TPR including data uncertainties, statistical uncertainties of the Monte Carlo calculation and the experimental uncertainties

  10. Fabrication of porous LiAlO2 ceramic breeder material

    International Nuclear Information System (INIS)

    The gamma-LiAlO2 ceramic material is the reference candidate for the solid breeder option of the Next European Torus Program. The experiments and methodologies developed in Italy to produce high surface area gamma-LiAlO2 powders to be compacted by cold pressing and sintering at 70 to 90% of the theoretical density, keeping a near fully open porosity is presented. The lithiating step was assessed for the Li2CO3 and Li2O2 precursors reacting with Al2O3 having submicron grain size. Sol-gel methodologies were also developed for the gamma-LiAlO2 preparation by which very high surface area ceramic grade powders were obtained

  11. Post-irradiation exam of tritium release from long-term irradiated Li2TiO3 ceramics

    International Nuclear Information System (INIS)

    Full text of publication follows: Lithium ceramics is planned to be used in tritium breeding systems of future fusion reactors. To provide effective tritium generation while obeying the ecological and safety restrictions on tritium processing it is necessary to investigate tritium interaction with elements of proposed breeding systems. Therefore tritium-ceramics interaction is of most interest in such systems. Presented work describes experimental studies of tritium yield from lithium ceramics (Li2TiO3+5 mol.% TiO2 ) after long-term neutron irradiation. Initially ceramics was 96% enriched with Li6 and irradiated with neutrons (about 220 days) in research water-water reactor of Kazakh National Nuclear Center (WWRK) till the 20% burn-up of Li6. Examinations of residual tritium yield from irradiated lithium ceramics were conducted using thermodesorption method with linear heating rates from 2 to 10 K/min up to ceramics melting point temperature. The experiments were carried-out under continuous pump-out and mass-analysis of desorbed gases in experimental chamber. As the result the data on tritium (and other gases) release rates from irradiated ceramics are obtained. Preliminary results on estimations of residual tritium content in irradiated lithium ceramics and its thermodesorption data are presented in given report. (authors)

  12. Tauro: a ceramic composite structural material self-cooled Pb-17Li breeder blanket concept

    International Nuclear Information System (INIS)

    The use of a low-activation (LA) ceramic composite (CC) as structural material appears essential to demonstrate the potential of fusion power reactors for being inherently or, at least, passively safe. Tauro is a self-cooled Pb-17Li breeder blanket with a SiC/SiC composite as structure. This study determines the required improvements for existing industrial LA composites (mainly SiC/SiC) in order to render them acceptable for blanket operating conditions. 3D SiC/SiC CC, recently launched on the market, is a promising candidate. A preliminary evaluation of a possible joining technique for SiC/SiC is also described. (orig.)

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

    International Nuclear Information System (INIS)

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

  14. Neutron activation, gamma spectrometry and tritium measurements on Italian lithium aluminate and zirconate, as selection means of candidate breeders for fusion reactors

    International Nuclear Information System (INIS)

    This paper discusses measurements of impurities and tritium releasing characteristics of Lithium Aluminate and zirconate, prepared by ENEA in the frame of the European Program on Fusion Technology, performed, respectively, by neutron activation analysis (NAA) and out of pile annealing. The resulting tritium removing rate from the ceramics was interpreted in terms of surface desorption kinetics. With reference purge gas (He + 0.1% H2), the predominant form of tritium, released by lithium aluminate is HT/T2, HTO/T2O by lithium zirconate. The latter was found to have a better performance in tritium release than aluminate. The presence of moisture was found to catalyze the tritium release at lower temperatures

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

    International Nuclear Information System (INIS)

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

  16. 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

  17. Process for obtaining tritium as solid oxide material containing lithium by neutron irradiation

    International Nuclear Information System (INIS)

    The process consists of: a) using a glass containing Li2O or a glass ceramic containing Li2O as breeder material, b) irradiating the breeder material in filled perforated pipes, c) having the breeder material in direct contact with the cooling gas accepting and transporting the reaction products resulting from irradiation, d) separating the forms of tritium with the cooling gas continuously from the breeding material. (orig./PW)

  18. Irradiation of Li-ceramics: tritium production for fusion energy

    International Nuclear Information System (INIS)

    After an introduction to fusion energy, this paper describes the processes to organize, plan and execute CRITIC-1, an irradiation of annular lithium oxide pellets in NRU, with on-line tritium release and measurement. Responsibilities were split in two main areas: in-core capsule design and tritium measurement and recovery. Additional areas or tasks addressed were: location of the tritium analysis and recovery system in the reactor, gas systems operation and control, data logging and safety. The paper will include a description of the issues in each of these areas, and the solutions taken in CRITIC-1. Where appropriate, solutions used in other similar experiments, such as BEATRIX-II, an irradiation in a fast flux test reactor at Westinghouse/Battelle PNL, will be described. Because NRU had not previously been used for instrumented irradiations (i.e., with thermocouples, etc), planning began at a very basic level. Problems and solutions will be described, and photographs at various stages will be presented. (author)

  19. Tritium production from ceramic targets: a summary of the Hanford Coproduct Program

    International Nuclear Information System (INIS)

    Early-generation fusion reactors will require tritium breeding from lithium or lithium compounds. Some current fusion reactor conceptual designs specify ceramic lithium compounds (lithium aluminate, lithium silicate, lithium oxide) for blanket materials. One potential source of information is the technology developed in support of large-scale tritium production programs in fission reactors. Much of that work has been classified. However, recent declassification of documents containing information no longer regarded as sensitive has provided much information of potential value to fusion reactor designers. This report summarizes the tritium production technology developed at Hanford in the mid-1960s under the so-called Coproduct Program. Information of potential value to the fusion community has been extracted from declassified and unclassified reports, summarized, and referenced

  20. Lithium orthosilicate ceramics: sol-gel preparation, lithium dynamics and tritium release

    International Nuclear Information System (INIS)

    Ceramics based on the lithium orthosilicate (Li4SiO4) are candidates as blanket materials for forthcoming fusion reactors. Lithium orthosilicate powders, with controlled stoichiometry, were prepared from sol-gel route. This method of processing powders makes possible the preparation of monophase ceramics with fine-grained uniform microstructure by sintering at 650-8000C, without prior calcination. Lithium transport properties were investigated from complex impedance spectroscopy and 7Li NMR spin-lattice relaxation measurements. The enhancement of the lithium conductivity in the orthosilicate type structure was realized by introducing mobile ion vacancies in the lithium sites, as noted in the Li4SiO4-Li3PO4 system. Concerning tritium release properties, deduced from out-of-pile experiments, no relation was found between the tritium behavior and the lithium bulk-diffusion within the grains. However, a large effect of the microstructure was displayed. The release rate appeared much faster for microporous fine-grained ceramics than for dense coarse-grained ones. In fact, the tritium release is controlled, at least at low temperature, by water chemistry and can be very well described by OH-/OT- recombination and desorption

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

    International Nuclear Information System (INIS)

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

  2. 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)

    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

  3. Study of lithium materials for tritium release, evolution in the CCHEN

    International Nuclear Information System (INIS)

    The Chilean Nuclear Energy Commission decided in 1993 to study lithium ceramic materials with tritium properties, ceramics breeder, using neutron irradiation in order to allow the transition from the conceptual to the engineering design of the ceramic breeder blanket fusion reactor. The project 'Development of Lithium Ceramic Materials for Fusion Reactors' was defined leading to the construction of the Tritium Handling Laboratory with the technical assistance of the Argonne National Laboratory and with Canadian equipment. Later, it was presented to the International Atomic Energy agency, IAEA, as a technical cooperation project for obtaining a tritium device in real time, irradiation loop. This was approved with a significant budget for equipment, training and experts. The project was presented to the international Ceramic Breeder Blanket Interactions community at the International Workshop on Ceramic Breeder Blanket Interactions. To date several investigations in this subject have been carried out, producing the release of tritium in the laboratory, in batch, and recently in continuous with the construction, installation and operation of an irradiation loop in the RECH-1, loading with ceramic material prepared and fabricated in the our laboratories.This work presents the genesis, evolution and current state, of this study, from 1993 to date

  4. Tritium

    International Nuclear Information System (INIS)

    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.)

  5. Integral experiments for verification of tritium production on the beryllium/lithium titanate blanket mock-up with a one-breeder layer

    International Nuclear Information System (INIS)

    The first series of integral experiments on the blanket mock-up with a one breeder layer was performed in support of the concept of the solid breeding blanket cooled with water, proposed by JAERI for application in the DEMO reactor. The mock-up for the first series of experiments was designed to be as simple as possible within the proposed blanket concept. Key objectives of the experiments were: to check how correctly the tritium production rate can be predicted in the breeder layer closest to the first wall, since this particular location is greatly affected by changes of incoming neutron spectra; to validate the modified experimental techniques for measurements of tritium production rate in conditions of quick gradient thermal neutron field inside the lithium titanate layer. The mock-up contains F82H, lithium titanate and beryllium layers, with respective thicknesses of 16 mm, 12 mm and 203 mm. An additional tungsten layer was installed in front of the first layer in order to simulate armor material. The mock-up, being placed inside the SS316 cylindrical enclosure, is shaped as a pseudo-cylindrical slab with an area-equivalent diameter of 628 mm. Integral experiments on the blanket mock-up irradiated by neutrons from the D-T source with and without the source reflector were executed. A detailed description of experimental results and an example of calculation analysis are presented. (author)

  6. European DEMO BOT solid breeder blanket

    International Nuclear Information System (INIS)

    The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.)

  7. Status of the database for solid breeder materials

    International Nuclear Information System (INIS)

    The databases for solid breeder ceramics (Li2O, Li4SiO4, Li2ZrO3 and LiAl02) and beryllium multiplier material were critically reviewed and evaluated as part of the ITER/CDA design effort (1988-1990). The results have been documented in a detailed technical report which includes progress made in expanding the solid breeder and beryllium databases up through September 1993. Emphasis was placed on the physical, thermal, mechanical, chemical-stability/compatibility, tritium retention/release and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Materials properties correlations were selected for use in design analysis, and ranges of input parameters (e.g., temperature, porosity, etc.) were established. The need for updating the ceramic breeder database was discussed at the Third Ceramic Breeder Blanket Interactions (CBBI-3) workshop at UCLA in June 1994. Progress made in expanding the ceramic breeder database and plans for updating the database are discussed

  8. Exchange reaction of hydrogen isotopes on proton conductor ceramic of hydrogen pump for blanket tritium recovery system

    International Nuclear Information System (INIS)

    Electrochemical hydrogen pump using ceramic proton conductor has been investigated to discuss its application for the blanket tritium recovery system of the nuclear fusion reactor. As the series of those work, the transportation experiments of H2-D2 mixture via ceramic proton conductor membrane have been carried out. Then, the phenomenon that was caused by the exchange reaction between the deuterium in the ceramic and the hydrogen in the gas phase has been observed. So, the ceramic proton conductor which doped deuterium was exposed to hydrogen under the control of zero current, and the effluent gas was analyzed. It is considered that the hydrogen in the gas phase is taken as proton to the ceramic by isotope exchange reaction, and penetrates to the ceramic by diffusion with replacement of deuteron. (author)

  9. ICF tritium production reactor

    International Nuclear Information System (INIS)

    The conceptual design of an ICF tritium production reactor is described. The chamber design uses a beryllium multiplier and a liquid lithium breeder to achieve a tritium breeding ratio of 2.08. The annual net tritium production of this 532 MW/sub t/ plant is 16.9 kg, and the estimated cost of tritium is $8100/g

  10. Development of ceramic-coated lithium particles for tritium production tests in high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Ceramic-coated lithium particles (CCLP) are proposed for the tritium production tests in High Temperature Engineering Test Reactor (HTTR). The CCLP-production method and its tritium-retention as well as -release capability are mainly studied. The CCLP consists of a lithium compound kernel coated with ceramic layers. The kernel is made from lithium compound like LiAlO2. Ceramic materials like Al2O3 are used for the coating layers. The present study has made clear that tritium leakage from CCLP is less than 1% after 400-days-heating at 1,000K, and the tritium can be released completely in a short time by heating over 1,400K. From these characteristics, it is expected that the tritium can be well extracted from the CCLP by heating over 1,400K in the post irradiation test after irradiation under 1,000K in the HTTR. Furthermore, a good chemical stability between the kernel and coatings was confirmed through X-ray diffraction tests after heating of their powders. The mechanical integrity of coating layers for inner gas pressure in the CCLP was evaluated to be good on the basis of material data. (author)

  11. ITER solid breeder blanket materials database

    Energy Technology Data Exchange (ETDEWEB)

    Billone, M.C. [Argonne National Lab., IL (United States); Dienst, W. [Kernforschungszentrum Karlsruhe GmbH (Germany). Inst. fuer Material- und Festkoerperforschung; Flament, T. [CEA Centre d`Etudes de Fontenay-aux-Roses (France). Commissariat A L`Energie Atomique; Lorenzetto, P. [NET Team, Garching (Germany); Noda, K. [Japan Atomic Energy Research Inst., Takai, Ibaraki, (Japan); Roux, N. [CEA Centre d`Etudes et de Recherches Les Materiaux (France). Commissariat a L`Energie Atomique

    1993-11-01

    The databases for solid breeder ceramics (Li{sub 2},O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAlO{sub 2}) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized.

  12. In situ tritium recovery from LiAlO2 pellets

    International Nuclear Information System (INIS)

    TEQUILA-1 is the first phase of an in situ tritium release experiment performed in MESULINE reactor, at CEN Grenoble, with almost the same facilities and irradiation conditions as used for the LISA series experiments. Three couples of LiAlO2 specimens with the same density (80% of TD) but quite different microstructures (grain size) were tested in six vented capsules by thermal step cycling in the temperature range 400 C - 700 C. The reference He + 0.1% H2 pruging gas was used, but an oxidizing mixture containing moisture, of composition He + O2 (37 vpm) + H2O (115 vpm), was also successfully tested to extract the tritium from the ceramic breeders. The best tritium releasing performance was achieved by the P-type ceramic breeder, with the smalles grain size (0.3 Lm), for which a Tritium retention time less than one day could be measured at 450 C

  13. Fusion breeder sphere - PAC blanket design

    International Nuclear Information System (INIS)

    There is a considerable world-wide effort directed toward the production of materials for fusion reactors. Many ceramic fabrication groups are working on making lithium ceramics in a variety of forms, to be incorporated into the tritium breeding blanket which will surround the fusion reactor. Current blanket designs include ceramic in either monolithic or packed sphere bed (sphere-pac) forms. The major thrust at AECL is the production of lithium aluminate spheres to be incorporated in a sphere-pac bed. Contemporary studies on breeder blanket design offer little insight into the requirements on the sizes of the spheres. This study examined the parameters which determine the properties of pressure drop and coolant requirements. It was determined that an optimised sphere-pac bed would be composed of two diameters of spheres: 75 weight % at 3 mm and 25 weight % at 0.3 mm

  14. Crystal chemistry of immobilization of Fast Breeder Reactor (FBR) simulated waste in Sodium Zirconium Phosphate (NZP) based ceramic matrix

    International Nuclear Information System (INIS)

    Full text: Sodium zirconium phosphate (hereafter NZP) is a potential material for immobilization of long lived heat generating radio nuclides. Possibility for the incorporation of simulated waste of fast breeder reactor origin in NZP was examined. It was found that most of the elements could be immobilized in this ceramic matrix without significant changes of the three-dimensional framework of the host material. All simulated waste forms synthesized by ceramic route at 1200 deg C crystallize in the rhombohedral system (space group R-3c). The crystal chemistry of 0-35% waste loaded NZP waste forms have been investigated using General Structure Analysis System (GSAS) programming of the step analysis powder diffraction data of the waste forms. Rietveld refinement of crystal data on the WOx loaded waste forms (NZPI-NZPVII) gives a satisfactory convergence of R-factors. The particle size along prominent reflecting planes calculated by Scherrer's formula varies between 68-141nm. The polyhedral distortions and effective valence calculations from bond strength data are also reported. Morphological examination by SEM reveals that the size of almost rectangular parallelepiped shaped grains varies between 0.2 and 5 μm. The EDX analysis provides analytical evidence of immobilization of effluent cations in the matrix

  15. Tritium safety of fusion power plants

    International Nuclear Information System (INIS)

    Tritium systems of a nuclear fusion plant, using the deuterium-tritium fuel cycle, has to ensure tritium safety during plant operation, and activated/tritiated materials management, during plant decommissioning. Accidents resulting in tritium releases to the environment may occur. It is important, therefore, to minimize the mobile tritium inventory by means of an adequate design and optimization of the plant tritium-bearing systems. Furthermore, the behaviour of tritium in the environment must be accurately studied, in order to take into account its oxidation and absorption by different materials. An experimental tritium-breeding module of DEMO fusion reactor is now under development in Russia. We plan to test it in the International Thermonuclear Experimental Reactor (ITER). The ceramic lithium orthosilicate will be used in it as tritium breeder. The Tritium Cycle System (TCS) will ensure tritium extracting and processing of gaseous mixtures containing tritium. The report contains the flow chart of the TCS using alloys producing hydrides. TCS assures highest possible autonomy and independence on ITER tritium plant at technological operations. The classification of the TCS modes of operation, adopted at the present stage of the module development, is described. The main initial events that may result in accidents are analysed. The maximum design accident and its consequences are considered. In particular, the maximum effective dose equivalent to the most exposed individual is calculated by means of the GEN II/FRAMES code. The flow sheet of technological operations at the maintenance and repair works and the System of Radiological Safety ensuring safety during these works is analyzed. Finally, some aspects of tritium decontamination from standpoint of waste handling are developed. In particular, material detritiation should be sufficient to allow clearance and recycling of less activated fusion materials. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-30

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

  17. Tritium transport analysis in HCPB DEMO blanket with the FUS-TPC Code (KIT Scientific Reports ; 7642)

    OpenAIRE

    Franza, Fabrizio

    2013-01-01

    In thermonuclear fusion reactors, the fuel is an high temperature deuterium-tritium plasma, in which tritium is bred by lithium isotopes present inside solid ceramic breeder (e.g. Li-Orthosilicate) or inside liquid eutectic alloys (e.g. Pb-16Li alloy). In the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the Helium Cooled Pebble Bed (HCPB) blanket concept or transported in solution by the owi...

  18. Development of the lithium fuel particles with ceramic coating for testing the tritium production in the HTTR reactor

    International Nuclear Information System (INIS)

    The method for obtained tritium by means of the lithium fuel particles with ceramic coating (LPCC), irradiated in the HTTR high-temperature gaseous reactor, and the results of tests on the hermeticity of such particles at the temperature of 1400 K are described. The basic characteristics of the LPCC with two coating layers and coefficients of the lithium diffusion in certain metals and ceramic materials are presented. The design for applying coatings on the LPCC by means of a pseudoliquified layer is described. The chemical stability and mechanical integrity of the LPCC were subjected to tests on their hermeticity at the temperature, expected in the irradiation zone. It is also shown, that 0.1 g of tritium may be obtained annually in one LPCC

  19. BEATRIX-II: A multi-national solid breeder experiment

    International Nuclear Information System (INIS)

    BEATRIX-II is an IEA program focused on tritium recovery experiments on lithium ceramic materials in a fast neutron reactor which partially simulates the environment of a fusion blanket. In addition to data on the performance of Li2O and Li2ZrO3, the BEATRIX-II program offers information on innovative technologies associated with tritium recovery. The successful execution of the BEATRIX-II program also offers a precedent for the structure, schedule and interfaces that other international programs should consider. Japan, Canada, and USA re participants in the BEATRIX-II program with primary responsibilities being assigned to Japan Atomic Research Institute, Atomic Energy of Canada Ltd., Battelle Pacific Northwest Laboratory, and Westinghouse Hanford Company. The purpose of the BEATRIX-II experiment is to conduct in situ tritium recovery experiments on ceramic solid breeder materials under irradiation conditions which expanded the burnup, irradiation damage, tritium production, and temperature regimes previously investigated. A liquid metal, fast neutron reactor was selected because spatial variations in tritium and heat production are minimized and temporal variations in the lithium burnup rate (burnout) are also minimal. The Fast Flux Test Facility was selected because it possessed a high neutron flux, excellent control and monitoring capabilities and ready access for a tritium recovery experiment

  20. Sources of tritium

    International Nuclear Information System (INIS)

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

  1. Interactions of tritium and materials

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

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

  2. Development of hi-tech ceramics fabrication technology

    International Nuclear Information System (INIS)

    There are some ceramic materials being used in the nuclear energy such as nuclear fuel, coolant pump seals, tritium breeder materials, a high temperature absorber, and the solid electrolyte for recovering tritium. In addition, lots of researches recently have been conducted on the development of highly functional ceramics such as highly efficient shielding materials, functional graded materials and radioactive isotopes-separating materials. Therefore, one of the objectives of this project is to develop ultra-fine and pure powder manufacturing technology. Tritium breeder materials, LiAlO2, Li2ZrO3 and Li2TiO3 were made with a combustion process of mixed fuels that is developed indigenously in this project. Additionally, this study also focused on the development of promising low temperature electrolytes of ceria. By using the ceria powder made by the combustion process of GNP was investigated their sinterability and the electrolytic characteristics. (author). 167 refs., 74 tabs., 91 figs

  3. Application of non-porous alumina based ceramics as structural material for devices handling tritium at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Yukhimchuk, A.A.; Maksimkin, I.P.; Baluev, V.V.; Boitsov, I.E.; Vertey, A.V.; Malkov, I.L.; Musyaev, R.K.; Popov, V.V.; Sitdikov, D.T. [Russian Federal Nuclear Centre, All-Russian Research Institute of Experimental Physics - RFNC-VNIIEF, Sarov (Russian Federation); Khapov, A.S.; Grishechkin, S.K.; Kiselev, V.G. [All-Russian Research Institute of Automatics - FSUE-VNIIA, Moscow (Russian Federation)

    2015-03-15

    The article presents results of comparative tests for the determination of deuterium fluxes permeating through walls of austenitic stainless steel AISI304 (DIN 1.4301) chamber and Al{sub 2}O{sub 3} based ceramic F99.7 chamber. Both chambers represent a piece of φ(ext)=26*φ(int)=22*117 mm{sup 3} tube with spherical bottom ending. It is shown that at 773 K and deuterium pressure of 1200 mbar the permeated deuterium flux through the stainless steel chamber constituted 8*10{sup -5} cm{sup 3}/s, while the flux through ceramic one it did not exceed the sensitivity of the measurement method threshold, namely about 1.5*10{sup -7} cm{sup 3}/s. The ceramic chamber turned out to survive more than 10{sup 3} cycles of heating up to 773 K with no damages. It did not lose its tightness up to 10 bar of internal deuterium pressure. The authors also present test results of a prototype bed for reversible tritium storage. The bed's case was made of alumina based ceramic F99.7, titanium being used as tritide making metal and high frequency induction used for heating the tritide metal. (authors)

  4. Application of non-porous alumina based ceramics as structural material for devices handling tritium at elevated temperatures

    International Nuclear Information System (INIS)

    The article presents results of comparative tests for the determination of deuterium fluxes permeating through walls of austenitic stainless steel AISI304 (DIN 1.4301) chamber and Al2O3 based ceramic F99.7 chamber. Both chambers represent a piece of φ(ext)=26*φ(int)=22*117 mm3 tube with spherical bottom ending. It is shown that at 773 K and deuterium pressure of 1200 mbar the permeated deuterium flux through the stainless steel chamber constituted 8*10-5 cm3/s, while the flux through ceramic one it did not exceed the sensitivity of the measurement method threshold, namely about 1.5*10-7 cm3/s. The ceramic chamber turned out to survive more than 103 cycles of heating up to 773 K with no damages. It did not lose its tightness up to 10 bar of internal deuterium pressure. The authors also present test results of a prototype bed for reversible tritium storage. The bed's case was made of alumina based ceramic F99.7, titanium being used as tritide making metal and high frequency induction used for heating the tritide metal. (authors)

  5. Effect of purge gas oxidizing potential on tritium release from Li-ceramics and on its permeation through 316L SS clads under irradiation (TRINE experiment)

    International Nuclear Information System (INIS)

    The effect of red-ox potential of helium purge gas (variously doped with H2, H2O and O2) was examined on tritium release from Li-ceramics (LiAlO2 and Li2ZrO3 pellets) and on its permeation rate through the 316L stainless steel clads (bare and coated) held at 500 C. Decreasing the H2 content from 1000 vpm (reference 'R' gas mixture) to 100 vpm, and substituting H2O for H2, the tritium permeation rate (ca. 1.41010 atoms cm-2 s-1 in R-gas) increases. Tritium inventories in the Li ceramics were increased too. When a strong oxidizing purge (1000 vpm O2 added to He containing 100 vpm H2O) was used, a retention time (τ) of two days at 400 C was measured for Li2ZrO3. In this oxidizing environment the tritium permeation loss dropped by a factor five for the uncoated capsules while an aluminide coating became a very effective tritium barrier: tritium permeation flux at 550 C fell below the measurable limit. (orig.)

  6. Modeling tritium behavior in Li2ZrO3

    International Nuclear Information System (INIS)

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

  7. Modeling tritium behavior in Li2ZrO3

    International Nuclear Information System (INIS)

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

  8. Design and tritium permeation analysis of China HCCB TBM port cell

    International Nuclear Information System (INIS)

    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

  9. Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jie; Wu, Yingwei, E-mail: wyw810@mail.xjtu.edu.cn; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-03-15

    Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li{sub 4}SiO{sub 4} lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition.

  10. Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

    International Nuclear Information System (INIS)

    Highlights: • A supercritical-water cooled solid breeder test blanket module (SWCB TBM) was designed. • The neutronics calculations show that the tritium breeding ratio (TBR) of SWCB TBM is 1.17. • The outlet temperature of SWCB TBM can reach as high as 500 °C. • Both thermal stress and deformation of the SWCB TBM design are within safety limits. - Abstract: In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li4SiO4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition

  11. Conceptual design of a water cooled breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Pu, Yong; Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Li, Jia; Peng, ChangHong [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China); Ma, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Chen, Lei [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui 230027 (China)

    2014-10-15

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by {sup 6}Li(n,α)T reaction. Li{sub 2}TiO{sub 3} pebbles and Be{sub 12}Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li{sub 2}TiO{sub 3} and Be{sub 12}Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be{sub 12}Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option

  12. Conceptual design of a water cooled breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • We proposed a water cooled ceramic breeder blanket with superheated steam. • Superheated steam is generated at the first wall and the front part of breeder zone. • Superheated steam has negligible impact on neutron absorption by coolant in FW and improves TBR. • The superheated steam at higher temperature can improve thermal efficiency. - Abstract: China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by 6Li(n,α)T reaction. Li2TiO3 pebbles and Be12Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li2TiO3 and Be12Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be12Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Canadian fusion breeder blanket program: Irradiation facilities at chalk river*1

    Science.gov (United States)

    Hastings, I. J.; Burton, D. G.; Celli, A.; Delaney, R. D.; Fehrenbach, P. J.; Howe, L. M.; Larson, L. L.; MacEwen, S. R.; Miller, J. M.; Naeem, T. A.; Sawicki, J. A.; Swanson, M. L.; Verrall, R. A.; Zee, R. H.

    1986-11-01

    The major irradiation facility at Chalk River Nuclear Laboratories (CRNL) is the NRU research reactor. Both unvented and vented capsule experiments on candidate blanket ceramics can be performed. In the unvented tests, tritium release data (HT-to-HTO ratio, tritium retention) are obtained by post-irradiation heating of the breeder ceramic in the presence of a sweep gas. Four tests have been completed on Li 2O and LiAlO 2. Effects of sweep gas composition, extraction vessel material and ceramic properties have been determined. Two unvented irradiations under the BEATRIX international breeder exchange program have been completed; analysis is underway. The vented tests involve long-term irradiation of candidate blanket materials. CRITIC-I, scheduled for mid-1986 under BEATRIX, will examine ANL-fabricated Li 2O in a six-month irradiation at 700-1200 K, varying sweep gas composition, with on-line HT/HTO measurement. Additionally, accelerator simulation techniques are available, using 70 kV and 2.0 MV mass separators, a 2.5 MV Van de Graaff accelerator and a tandem accelerator super-conducting cyclotron, the latter allowing irradiation with protons, deuterons or helium at 18-20 MeV.

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

    International Nuclear Information System (INIS)

    This report describes the results of six EXOTIC experiments comprising a total of 48 capsules. Samples of the candidate tritium breeding materials LiAlO2, Li2ZrO3, Li4SiO4, Li6Zr2O7, Li8ZrO6, Li2O and Li2SiO3 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

  20. New progress on design and R and D for solid breeder test blanket module in China

    Energy Technology Data Exchange (ETDEWEB)

    Feng, K.M., E-mail: fengkm@swip.ac.cn; Zhang, G.S.; Hu, G.; Chen, Y.J.; Feng, Y.J.; Li, Z.X.; Wang, P.H.; Zhao, Z.; Ye, X.F.; Xiang, B.; Zhang, L.; Wang, Q.J.; Cao, Q.X.; Zhao, F.C.; Wang, F.; Liu, Y.; Zhang, M.C.

    2014-10-15

    Highlights: • The new progress on design and R and D of Chinese solid breeder TBM are introduced. • The mock-up fabrication and component tests for Chinese HCCB TBM have being developed. • The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CFL-1 are being prepared. • The fabrication of 1/3 sized mock-up is being carried-out. • The key technology development is proceeding to the large-scale mock-up fabrication. - Abstract: ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R and D activities for each TBM module with the auxiliary system are introduced. The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li{sub 4}SiO{sub 4} pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R and D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.

  1. Development of online measurement system of tritium in out-of-pile tritium release experiment

    International Nuclear Information System (INIS)

    It is very important to accurately measure tritium concentration and morphology for mastering tritium release behavior of tritium production breeder and improving the performance of tritium production breeder in out-of-pile tritium release experiment. According to the characteristics of small flow of carrier gas, small quantity of gas and argon as carrier gas, based on the ionization chamber, the digital online tritium measurement system was developed. The sensitive volume of ionization chamber is 50 mL, the digital instrument can automatically obtain the tritium concentration, at the same time, process and display it. The results show that the saturated zone of ionization chamber is about 35 V in argon, and the detection limit is 3.7 × 107 Bq/m3, which is satisfactory for online tritium measurement in out-of-pile tritium release experiment. (authors)

  2. Fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs

  3. Neutronic optimization of solid breeder blankets for STARFIRE design

    International Nuclear Information System (INIS)

    Extensive neutronic tradeoff studies were carried out to define and optimize the neutronic performance of the different solid breeder options for the STARFIRE blanket design. A set of criteria were employed to select the potential blanket materials. The basic criteria include the neutronic performance, tritium-release characteristics, material compatibility, and chemical stability. Three blanket options were analyzed. The first option is based on separate zones for each basic blanket function where the neutron multiplier is kept in a separate zone. The second option is a heterogeneous blanket type with two tritium breeder zones. In the first zone the tritium breeder is assembled in a neutron multiplier matrix behind the first wall while the second zone has a neutron moderator matrix instead of the neutron multiplier. The third blanket option is similar to the second concept except the tritium breeder and the neutron multiplier form a homogeneous mixture

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

    Institute of Scientific and Technical Information of China (English)

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

    2006-01-01

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

  5. Neutronics Experiment on A HCPB Breeder Blanket Mock-Up

    International Nuclear Information System (INIS)

    A neutronics experiment has been performed in the frame of European Fusion Technology Program on a mock-up of the EU Test Blanket Module (TBM), Helium Cooled Pebble Bed (HCPB) concept, with the objective to validate the capability of nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. The experiment has been carried out at the FNG 14-MeV neutron source in collaboration between ENEA, Technische Universitaet Dresden, Forschungszentrum Karlsruhe, J. Stefan Institute Ljubljana and with the participation of JAEA. The mock-up, designed in such a way to replicate all relevant nuclear features of the TBM-HCPB, consisted of a steel box containing beryllium block and two intermediate steel cassettes, filled with of Li2CO3 powder, replicating the breeder insert main characteristics: radial thickness, distance between ceramic layers, thickness of ceramic layers and of steel walls. In the experiment, the TPR has been measured using Li2CO3 pellets at various depths at two symmetrical positions at each depth, one in the upper and one in the lower cassette. Twelve pellets were used at each position to determine the TPR profile through the cassette. Three independent measurements were performed by ENEA, TUD/VKTA and JAEA. The neutron flux in the beryllium layer was measured as well using activation foils. The measured tritium production in the TBM (E) was compared with the same quantity (C) calculated by the MCNP.4c using a very detailed model of the experimental set up, and using neutron cross sections from the European Fusion File (EFF ver.3.1) and from the Fusion Evaluated Nuclear Data Library (FENDL ver. 2.1, ITER reference neutron library). C/E ratios were obtained with a total uncertainty on the C/E comparison less than 9% (2 s). A sensitivity and uncertainty analysis has also been performed to evaluate the calculation uncertainty due to the uncertainty on neutron cross sections. The results of such analysis

  6. New simulations to qualify eutectic lithium-lead as breeder material

    OpenAIRE

    Fraile García, Alberto; Cuesta Lopez, Santiago; Caro, Alfredo; Iglesias, R.; Perlado Martin, Jose Manuel

    2011-01-01

    Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. One of the main issues is the problem of liquid metals breeder blanket behavior. The knowledge of eutectic properties like optimal composition, physical and thermodynamic behavior or diffusion coefficients of Tritium are extremely necessary for current designs. In particular, the knowledge of the function linking the tritium concentration dissolved in liquid materials with the tritium partial pressure at...

  7. Monitoring of tritium

    Science.gov (United States)

    Corbett, James A.; Meacham, Sterling A.

    1981-01-01

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

  8. First adaptation of the European ceramic B.I.T. blanket design to the updated DEMO specifications

    International Nuclear Information System (INIS)

    The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Programme have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects. (orig.)

  9. The development of breeder reactors in the US

    International Nuclear Information System (INIS)

    This article discusses the early history of breeder development in the US, the early history of the fast reactor in the US, changes during the Carter administration, and the development of LMFBR technology. Topics considered include the intermediate-energy plutonium breeder, the molten plutonium breeder, the aqueous homogeneous reactor, the molten-salt reactor, the liquid metal-fueled reactor, electronuclear breeding, the Experimental Breeder Reactor-I, the Experimental Breeder Reactor-II, the Enrico Fermi Reactor, a programmatic change to ceramic fuel, the South East Fast Oxide Reactor, the sodium void coefficient, the 1000-MWe studies of 1964, the 1000-MWe studies of 1967-1969, the FARET design, the Fast Flux Test Facility, the Clinch River Breeder Reactor (CRBR), the gas-cooled fast breeder, the light-water breeder, materials for cladding and duct walls, and reactor safety. It is pointed out that the Congress opposes the construction of the CRBR, while the Reagan administration strongly supports it

  10. Development of hi-tech ceramics fabrication technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Park, Ji Yeon; Kim, Sun Jai; Jung Choong Hwan; Oh, Seok Jin

    1997-07-01

    There are some ceramic materials being used in the nuclear energy such as nuclear fuel, coolant pump seals, tritium breeder materials, a high temperature absorber, and the solid electrolyte for recovering tritium. In addition, lots of researches recently have been conducted on the development of highly functional ceramics such as highly efficient shielding materials, functional graded materials and radioactive isotopes-separating materials. Therefore, one of the objectives of this project is to develop ultra-fine and pure powder manufacturing technology. Tritium breeder materials, LiAlO{sub 2}, Li{sub 2}ZrO{sub 3} and Li{sub 2}TiO{sub 3} were made with a combustion process of mixed fuels that is developed indigenously in this project. Additionally, this study also focused on the development of promising low temperature electrolytes of ceria. By using the ceria powder made by the combustion process of GNP was investigated their sinterability and the electrolytic characteristics. (author). 167 refs., 74 tabs., 91 figs

  11. Fusion breeder neutronics. Final report

    International Nuclear Information System (INIS)

    Research efforts in fusion breeder neutronics have been focused on two tasks that are strongly related. Efforts in Task 1 concentrate on examining the required conditions to sustain fuel self-sufficiency in fusion reactors operated on a D-T fuel cycle. In this respect, in-depth and detailed engineering analyses have been performed on various blanket and reactor concepts to verify the potential of each blanket concept to exhibit a tritium breeding ratio (TBR) in excess of unity by a margin that compensates for losses, radioactive decay and other inventory requirements. Efforts in Task 2 concentrate on evaluating the overall uncertainties (both experimental and analytical) associated with the TBR

  12. Breeding blanket development; Tritium release from breeder

    OpenAIRE

    土谷 邦彦; 河村 弘; 長尾 美春

    2006-01-01

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

  13. Comparison of lithium and the eutectic lead-lithium alloy, two candidate liquid metal breeder materials for self-cooled blankets

    International Nuclear Information System (INIS)

    Liquid metals are attractive candidates for both near-term and long-term fusion applications. The subjects of this comparison are the differences between the two candidate liquid metal breeder materials Li and LiPb for use in breeding blankets in the areas of neutronics, magnetohydrodynamics, tritium control, compatibility with structural materials, heat extraction system, safety and required research and development program. Both candidates appear to be promising for use in self-cooled breeding blankets which have inherent simplicity with the liquid metal serving as both breeder and coolant. Each liquid metal breeder has advantages and concerns associated with it, and further development is needed to resolve these concerns. The remaining feasibility question for both breeder materials is the electrical insulation between the liquid metal and the duct walls. Different ceramic coatings are required for the two breeders, and their crucial issues, namely self-healing of insulator cracks and tolerance to radiation-induced electrical degradation, have not yet been demonstrated. (orig.)

  14. Comparison analysis of fusion breeder blanket concepts

    International Nuclear Information System (INIS)

    Based on the wide survey, the development status and key issues of fusion breeder blanket concepts are summarized. Two types of blanket concepts, i.e. solid and liquid breeder blanket, were compared and assessed in terms of engineering feasibility, tritium recovery and control, economic and safety aspects, etc. The advantages and disadvantages of the two types of blanket concepts were clarified from the viewpoint of technology realization and development potential. This study may act as a valuable reference for fusion blanket concept selection and design. (authors)

  15. The cost of tritium production in a fusion reactor

    International Nuclear Information System (INIS)

    In this paper, a computational model is presented in order to assess the cost of tritium breeding in a fusion power reactor. This model compares the differential cost of the Li-bearing breeder blanket with that of a steel shield and adds the loss of revenue due to the lower energy multiplication of the breeder blanket compared to the steel shield. The cost of tritium production ranges from $215-$300/g for a simple breeder up to $1420/g for a high temperature breeder

  16. Tritium breeding materials

    International Nuclear Information System (INIS)

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

  17. Global depletion analysis of Korean helium cooled solid breeder TBM model for demo fusion reactor

    International Nuclear Information System (INIS)

    The Korean HCSB (helium cooled solid breeder) TBM (test blanket module) is proposed with its specific compositions of lithium ceramic, beryllium and graphite in pebble form. In the Korean HCSB TBM, the amount of beryllium is reduced and the reduction is replaced by graphite for a neutron reflector, while tritium breeding ratio (TBR) remains almost unchanged with relatively low Li6 enrichment of ∼40%. However, the previous Korean HCSB was designed based on the LOCAL assumption, in which the surroundings are assumed by the reflective boundary condition. In this research, we establish a simple GLOBAL neutronics model based on demo fusion reactor and perform neutronics analyses including depletion (transmutation) calculation during 100 EFPDs (effective full power days) using the modified MONTEBURNS code.

  18. Development of Liquid Type Breeder Technology for ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Sok; Hong, Bong Geun; Lee, Dong Won

    2008-07-15

    In relation to liquid type TBM technology development, various works are performed. We established a test loop concept to test the MHD effects and materials compatibility for the Pb-17Li breeder material. For the loop construction, electromagnetic pump and storage tank for the Pb-17Li loop was manufactured and some technical requirements are summarised. As a reference, technical literatures relevant to the liquid type TBM materials and the tritium extraction from breeder materials are also surveyed.

  19. BEATRIX: The international breeder materials exchange

    International Nuclear Information System (INIS)

    The BEATRIX experiment is an IEA-sponsored effort that involves the exchange of solid breeder materials and shared irradiation testing among research groups in several countries. The materials will be tested in both closed capsules (to evaluate material lifetime) and opened capsules (to evaluate purge-flow tritium recovery). Pre- and post-irradiation measurement of thermophysical and mechanical properties will also be carried out

  20. Solid breeder blanket concepts

    International Nuclear Information System (INIS)

    An investigation is made of a mechanical concept for the blanket with solid breeders in view of the possible adaptation to power reactor. A special arrangement of the multiplier and breeder materials is developed to permit a further neutronic optimisation

  1. Breeder Reprocessing Engineering Test

    Energy Technology Data Exchange (ETDEWEB)

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  2. Preparation and characterization of Li4SiO4 ceramic pebbles by graphite bed method

    International Nuclear Information System (INIS)

    Highlights: • Lithium orthosilicate pebbles were fabricated by a new graphite bed process. • Two routes using different raw materials have been conducted in this work. • The fabricated pebbles exhibit a high relative density with uniform microstructure. • This method is short and simple as the pebbles could be fabricated in a continuous process. - Abstract: Lithium-based ceramics have long been recognized as tritium breeding materials in fusion reactor blankets. Lithium orthosilicate (Li4SiO4) is one of these materials and has been recommended by many ITER research teams as the first selection for the solid tritium breeder. In this paper, the fabrication of Li4SiO4 pebbles used as tritium breeder by a graphite bed method was studied for the first time. Ceramic powders and deionized water were mixed and ball milled to obtain homogeneous suspensions. And then the ceramic suspensions were dispersed on spread graphite powder through nozzles. Spherical droplets with highly uniform size were formed by the surface tension of the liquid droplets. The droplets converted into green pebbles after drying. After calcination and sintering, Li4SiO4 pebbles with desired size and shape were prepared. The obtained Li4SiO4 pebbles had narrow size distribution and favorable sphericity. Thermal analysis, phase analysis and microstructure observation of the pebbles were carried out systematically. Properties of the prepared pebbles were also characterized for crushing load strength, density and porosity, etc. The values were found to be conforming to the desired properties for used as solid breeder

  3. Tritium self-sufficiency of HCPB blanket modules for DEMO considering time-varying neutron flux spectra and material compositions

    Energy Technology Data Exchange (ETDEWEB)

    Aures, A., E-mail: Alexander.Aures@ccfe.ac.uk; Packer, L.W.; Zheng, S.

    2013-10-15

    Highlights: • Simulations on the tritium breeding performance of HCPB blanket modules were done. • MCNP5 and FISPACT were used for coupled transport and activation calculations. • Material transmutation affects the neutron flux spectra within the blanket modules. • The consequences of time-dependent spectra on TBR and tritium self-sufficiency were investigated. -- Abstract: Significant transmutation of solid-type breeding blanket materials affects the time and spatial variation of neutron energy within such materials. This has an impact on simulation assumptions required to accurately assess tritium surplus quantities for conceptual power plant devices. This paper details an investigation, via simulation, of the consequences for the tritium breeding ratio and the tritium self-sufficiency of a DEMO concept with homogeneous Helium-Cooled Pebble Bed blanket modules containing Li{sub 4}SiO{sub 4} ceramic breeder material. For this purpose, a code was developed to couple MCNP5 and FISPACT to supply material compositions from activation calculations to the neutron transport calculation in an iterative loop covering several time steps. Simulation results are presented for a simple 1D spherical device model and a DEMO tokamak model.

  4. Thermal behaviour and tritium management for in-pile testing of the pebble bed assemblies in the HFR in Petten

    International Nuclear Information System (INIS)

    Four pebble-bed assemblies are to be irradiated in the HFR in Petten with the objective to study the thermo-mechanical behaviour of the breeder ceramic pebble beds during irradiation. The thermo-mechanical behaviour of the pebble bed assemblies was calculated in a 2D axi-symmetric model in MARC. In this approach there could not be accounted for the influence of thermocouple tubes on the temperature distribution in the assembly, because these are distributed in the assembly in a non axi-symmetric manner. The solution for this problem was to expand the model to a 3D model used for thermal computations only. For safety reasons the tritium production in the breeder and permeation through the first and second containment must be estimated before the in-pile experimentation begins. In order to do so, the calculated thermal distribution is used as input for the enhanced two-dimensional finite element model in MARC. Adaptations are made in the 2D model by adding the capability of performing mass flux calculations. This paper describes the finite element models used for computation of the temperature distribution and the tritium flux through the pebble bed assembly. The results of these calculations are critical for a safety assessment of the in-pile operation of the experiment and will give a better understanding of the in-pile behaviour on temperature and tritium management in advance. (orig.)

  5. In-pile Tritium Permeation through F82H Steel with and without a Ceramic Coating of Cr2O3-SiO2 Including CrPO4

    International Nuclear Information System (INIS)

    Development of coating on blanket structural materials with significant reduction capability of tritium permeation is highly required in order to realize a reasonable design of a tritium recovery and processing system of demonstration (DEMO) fusion reactors. An effective coating has been developed in Japan Atomic Energy Agency (JAEA) using a ceramic material of Cr2O3-SiO2 including CrPO4. In previous out-of-pile deuterium permeation experiments at 600 oC [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701], a significant permeation reduction factor (PFR) of about 300 was obtained for the coating on the inner-side surface of tubular diffusion cells made by ferritic steel (F82H). In the present study, in-pile experiments on tritium permeation were conducted for F82H steel with and without the same coating, using a testing reactor IGV-1M in Kazakhstan. The tritium source used was liquid lithium-lead eutectics, Pb17Li, which was poured into a space around a tubular diffusion cell (specimen) of F82H steel with or without the coating on the inner side the cell. The irradiation time was about 4 hours, which corresponds to a fast-neuron fluence of about 2x1021m-2 (E > 1.1 MeV). The permeation reduction factor (PRF) was obtained by comparison of kinetics curves of tritium permeation through the diffusion cell of F82H steel with and without the coating. The PRFs at 600 and 500 oC were 292 and 30, respectively. These values are close to corresponding PRF values of 307 and 45, which had been obtained at 600 and 500 oC, respectively, in the previous out-of-pile experiments [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701]. (author)

  6. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain); Sedano, L. A. [Asociacion Euratom-Ciematpara Fusion, Av. Complutense 22, 28040 Madrid (Spain); Perlado, J. M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain)

    2008-07-15

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T{sub 2} permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H{sub 2} tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations

  7. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T2 permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H2 tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations. Dose

  8. Preparation and characterization of Li{sub 4}SiO{sub 4} ceramic pebbles by graphite bed method

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Ming; Zhang, Yingchun, E-mail: zycustb@163.com; Xiang, Maoqiao; Liu, Zhiang

    2015-06-15

    Highlights: • Lithium orthosilicate pebbles were fabricated by a new graphite bed process. • Two routes using different raw materials have been conducted in this work. • The fabricated pebbles exhibit a high relative density with uniform microstructure. • This method is short and simple as the pebbles could be fabricated in a continuous process. - Abstract: Lithium-based ceramics have long been recognized as tritium breeding materials in fusion reactor blankets. Lithium orthosilicate (Li{sub 4}SiO{sub 4}) is one of these materials and has been recommended by many ITER research teams as the first selection for the solid tritium breeder. In this paper, the fabrication of Li{sub 4}SiO{sub 4} pebbles used as tritium breeder by a graphite bed method was studied for the first time. Ceramic powders and deionized water were mixed and ball milled to obtain homogeneous suspensions. And then the ceramic suspensions were dispersed on spread graphite powder through nozzles. Spherical droplets with highly uniform size were formed by the surface tension of the liquid droplets. The droplets converted into green pebbles after drying. After calcination and sintering, Li{sub 4}SiO{sub 4} pebbles with desired size and shape were prepared. The obtained Li{sub 4}SiO{sub 4} pebbles had narrow size distribution and favorable sphericity. Thermal analysis, phase analysis and microstructure observation of the pebbles were carried out systematically. Properties of the prepared pebbles were also characterized for crushing load strength, density and porosity, etc. The values were found to be conforming to the desired properties for used as solid breeder.

  9. Fast breeder reactor

    International Nuclear Information System (INIS)

    The fluid-cooled fast breeder reactor described includes an outer cylindrical boundary wall, a plurality of canless fuel elements and breeder material elements received within the boundary wall and being in an array therein forming a fissionable fuel zone and a breeder material zone coaxially surrounding the fissionable fuel zone, a coolant supply system for applying fluid coolant at uniform pressure to the entire cross section within the cylindrical boundary wall, and flow guide devices extending substantially horizontally and disposed at different levels one above the other within the breeder material zone which coaxially surrounds the fissionable fuel zone, means for elastically securing the flow guide devices at alternate levels within the breeder material to the boundary wall, the flow guide devices at the levels intermediate the alternate levels being spaced by an annular gap from the boundary wall. 7 claims, 7 drawing figures

  10. Embattled breeder reactor

    International Nuclear Information System (INIS)

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  11. Low technology high tritium breeding blanket concept

    International Nuclear Information System (INIS)

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

  12. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  13. Conceptual design of solid breeder blanket system cooled by supercritical water

    International Nuclear Information System (INIS)

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li2TiO3 or Li2O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  14. Neutronics design for a fusion breeder

    International Nuclear Information System (INIS)

    As a fusion breeder, one of the most important figure is support ratio which reflects the economic and fuel production performance of the system to a great extent. In this paper, the support ratio is calculated by using one dimension transport program ANISN and optimized by adjusting 6Li enrichment and blanket arrangement. The radial distribution of producted U-233 is also taken into account. Measures are taken for better blanket design, and satisfactory results are obtained. Tritium breeding ratio T reaches 1.11 and the support ratio is enhanced from 11 to 14. The engineering, safety and environment performance are improved

  15. Current operations and experiments at the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    The Tritium Systems Test Assembly (TSTA) has continued to move toward operation of a fully-integrated, full-sized, computer-controlled fusion fuel processing loop. Concurrent, nonloop experiments have answered important questions on new components and issues such as palladium diffusion membranes, ceramic electrolysis cells, regenerable tritium getters, laser Raman spectroscopy, unregenerable tritium inventory on molecular sieves, tritium contamination problems and decontamination methods, and operating data on reliability, emissions, doses, and wastes generated. 4 refs., 2 figs

  16. Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 2

    International Nuclear Information System (INIS)

    The BOT (Breeder Outside Tube) Helium Cooled Solid Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test blankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of the test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.)

  17. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  18. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  19. Estimation of the tritium production and inventory in beryllium

    International Nuclear Information System (INIS)

    Beryllium has been proposed as a candidate material for the neutron multiplier in fusion blanket designs. Tritium will be produced and will accumulate in beryllium under neutron irradiation. The tritium production and inventories under 1.5 and 3.0GW fusion power operation were calculated for a layered pebble bed blanket with lithium oxide (Li2O) breeder and beryllium (Be) multiplier. Neutronics calculations were carried out using the one-dimensional transport code ANISN, and the tritium production due to direct reaction of 9Be(n,T)7Li and the two-step reactions 9Be(n,α)6Li(n,α)Twas taken into account. The tritium production due to the two-step reaction was calculated to be 50% of the total tritium production after 1 year full power operation (FPY). The tritium inventory was estimated by considering three kinetic parameters, the permeability from the breeder region, diffusivity in a beryllium matrix, and solubility. Tritium permeation from the breeder region to the beryllium region through a 316SS wall was as much as 3gh-1, which is 30% of the tritium production (9.6gh-1) in the breeder region. Using the diffusion coefficient of beryllium with no oxide layer on its surface, the total tritium inventory was calculated to be 7gFPY-1, mainly owing to solubility. The content of beryllium oxide significantly affects the effective diffusion coefficient. Using a diffusion coefficient for beryllium with beryllium oxide layer on its surface, the tritium inventory was found to be equal to the amount produced. (orig.)

  20. Some issues in two-dimensional modeling of tritium transport

    International Nuclear Information System (INIS)

    Among the major processes leading to tritium transport through Li ceramic breeders the percolation of gaseous tritium species through the connected porosity remains the lest amenable to a satisfactory treatment. The combination of diffusion and reaction through the convoluted transport pathways prescribed by the system of pores poses a formidable challenge. The key issue is to make the fundamental connection between the tortuousity of the medium with the transport processes in terms of only basic parameters that are amenable to fundamental understanding and experimental determinations. This fundamental challenges is met within the following approaches. The technique that we have employed is a random network percolation model. Local transport in each individual pore channel is described by a set of convection-diffusion-reaction equations. Long range transport is described by a matrix technique. The heterogeneous structure of the medium is accounted for via Monte Carlo methods. In this way the approach requires as inputs only physical-chemical parameters that are amenable to clear basic understanding and experimental determination. In the sense it provides predictive capability. The approach has been applied to an analysis of the concept of tritium residence time which is associated with the first passage time, a direct output of our analysis. In the next stage of our work the tool that we have developed would be employed to investigate the issues of vary large networks, realistic microstructural information and the effect of varying pressure gradient along the purge channels. We have demonstrated that the approach that has been adopted can be utilized to analyze in a very illuminating way the underlying issues of the concept of residence time. We believe that the present approach is ideally suited to tackle these very important yet difficult issues

  1. A description of the tritium facility at the Chalk River Laboratories

    International Nuclear Information System (INIS)

    AECL's Tritium Facility is located at its Chalk River Laboratories (CRL). The Tritium Facility was originally built to support the tritium technology needs for CANDU reactors and Canadian fusion program. The Tritium Facility commenced its operation in 1979. Since its inception, it has been involved in the development of heavy water detritiation and upgrading processes, development and testing of tritium-breeder materials and design and testing of fusion-fuel cleanup systems for fusion reactor applications, investigation of tritium-materials interactions, tritium storage getters etc. The Tritium Facility also contributed to the design, construction and commissioning activities of the Combined Electrolysis and Catalytic Exchange Upgrading and Detritiation (CECE-UD) Facility at CRL and the Wolsong Tritium Removal Facility (WTRF) in Korea. This paper describes the general set-up of the laboratory, its capabilities and the current tritium-related activities. (author)

  2. Development of the irradiation assembly used in-pile tritium production

    International Nuclear Information System (INIS)

    The irradiation assembly of in-pile tritium production is main part of the first in-pile tritium demonstration apparatus for hybrid reactor in China. The paper describes its principle configuration and specifications. Design calculation are given in which include tritium production calculation, heat transfer calculation and stress calculation. Key technology in development of the assembly is explained. Operation and in-situ tritium release experiments of the assembly show that thermal and nuclear characteristics of the assembly is good. Adjustable temperature range of the tritium breeder is 250∼700 degree C. Radius temperature difference in the tritium breeder is less than 70 degree C. Tritium production efficiency of the assembly under thermal neutron irradiation is 0.183 x 10-7 Bq·cm2. The assembly and its utilization is discussed and reviewed

  3. Welsh tritium

    International Nuclear Information System (INIS)

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

  4. The fusion breeder

    International Nuclear Information System (INIS)

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U3O8). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-pathitem for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices

  5. Conceptual design study for a mirror fusion breeder

    International Nuclear Information System (INIS)

    A mirror fusion breeder, CHD, has been designed for providing plenty of nuclear fuel for light water reactors to meet the needs for rapid development of nuclear power in the first half of next century. The breeder is able to support the nuclear fuel needs for more than 10 LWRs of equal scale in power with fuel enriched directly in CHD without reprocessing. Measures are taken to flatten the power density distribution in the blanket so that fission is suppressed in the region close to the plasma, and by this way fuel production is enhanced for this direct enriched fusion breeder. In order to reduce the MHD pressure drop, LiPb flows in the blanket axially. Though the tritium inventory in the reactor is very low, special material and design have to be developed to reduce the permeation of tritium through the coolant pipes. The cost of electricity from the system, consisting of 11 LWR plants and one fusion breeder is predicted to be 1.05 times of that from a traditional LWR plant. This figure is insensitive both to the cost of CHD and its support ratio

  6. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  7. Tritium calorimetry

    International Nuclear Information System (INIS)

    Complete text of publication follows. Future deuterium-tritium fusion experiments (like ITER) will use large amount of tritium. Therefore, it is very important to develop better tritium accountancy methods. Tritium calorimetry is used to measure the heat produced by the beta-decay of tritium. If we consider that all the decay energy is converted into thermal heat, we can calculate the tritium activity and mass from calorimetric measurements. The advantages of calorimetry are that it measures absolute activity, and the physical or chemical composition of the sample is not relevant. For example, tritiated structural components can only be measured in a non-destructive way with calorimeters. Disadvantages are: long measurement time for large sample volumes, and offline sampling. The accepted conversion factor is 0.324W/g ± 0.3%. I have started participation from ATOMKI in an EFDA-GOT program, called TRI-TOFFY (TRITium fOr Fusion Fuel cYcle), in 2010. I have spent 8 months at Tritium Laboratory Karlsruhe (TLK), Germany in 2011, and 9 months in 2012. TLK is a semi-industrial scale facility for processing tritium, the radioactive hydrogen isotope. The main tasks of TLK are fusion research (ITER) and neutrino physics (KATRIN), but also EU projects. The present site inventory is ∼ 25 g T2 (8914 TBq). There are four calorimeters are used for tritium analytics at TLK. My main work was to carry out upgrade on these devices, to deploy new modern control and data acquisition (DAQ) software, and to partly change their hardware. I worked on three calorimeters at the laboratory. The ANTECH-351 is a commercial 20 years old calorimeter. It is a power compensation type isothermal calorimeter. Useful sample volume is 1.2 dm3. This is not a sensitive device (power range is 1 mW - 5 W), mainly used for tritium shipment (from Canadian CANDU reactors) validation, but can measure tritium samples very fast: less than 8 hours. The IGC-V0.5 is a custom made heat flow calorimeter, using a

  8. Tritium contamination and decontamination

    International Nuclear Information System (INIS)

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

  9. A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

    International Nuclear Information System (INIS)

    Highlights: • The issue of a potentially producing a large tritium surplus inventory, within a solid breeder, is addressed. • A possible solution to this problem is presented in the form of a neutron poison based tritium production controller. • The tritium surplus inventory has been modelled by the FATI code for a simplified WCCB model and as a function of time. • It has been demonstrated that the tritium surplus inventory can be managed, which may impact on safety considerations. - Abstract: The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist. This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into

  10. MHI(Mitsubishi Heavy Industries)'s activities on tritium technology

    International Nuclear Information System (INIS)

    MHI has been developing tritium technology for more than 20 years, mainly in the following fields concerning thermonuclear fusion reactors. 1 Isotope separation system by cryogenic distillation. 2 Fuel clean up system by palladium permeation and electrolysis cell method. 3 Tritium recovery system from the blanket by palladium permeation method. 4 Blanket materials, mainly the development and characterization of Li ceramic. 5 Tritium removal system by tritium oxidation catalysis. Based on the tritium technology recently attained through the above developments, MHI has built a tritium treatment installation where MHI can treat 2.2x1012 Bq(6Ci) of tritium per year. (author)

  11. Description of tritium release from lithium titanate at constant temperature

    Energy Technology Data Exchange (ETDEWEB)

    Pena, L.; Lagos, S.; Jimenez, J.; Saravia, E. [Comision Chilena de Energia Nuclear, Santiago (Chile)

    1998-03-01

    Lithium Titanate Ceramics have been prepared by the solid-state route, pebbles and pellets were fabricated by extrusion and their microstructure was characterized in our laboratories. The ceramic material was irradiated in the La Reina Reactor, RECH-1. A study of post-irradiation annealing test, was performed measuring Tritium release from the Lithium Titanate at constant temperature. The Bertone`s method modified by R. Verrall is used to determine the parameters of Tritium release from Lithium Titanate. (author)

  12. Development of Solid Breeder Blanket at JAERI

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI

  13. Swiss breeder research programme

    International Nuclear Information System (INIS)

    A new initiative for a Swiss Fast Breeder Research Program has been started during 1991. This was partly the consequence of a vote in Fall 1990, when the Swiss public voted for maintaining nuclear reactors in operation, but also for a moratorium of 10 years, within which period no new reactor project should be proposed. On the other hand the Swiss government decided to keep the option 'atomic reactors' open and therefore it was essential to have programmes which guaranteed that the knowledge of reactor technology could be maintained in the industry and the relevant research organisations. There is also motivation to support a Swiss Breeder Research Program on the part of the utilities, the licensing authorities and the Paul Scherrer Institute (PSI). The utilities recognise the breeder reactor as an advanced reactor system which has to be developed further and might be a candidate, somewhere in the future, for electricity production. In so far they have great interest that a know-how base is maintained in our country, with easy access for technical questions and close attention to the development of this reactor type. The licensing authorities have a legitimate interest that an adequate knowledge of the breeder reactor type and its functions is kept at their disposal. PSI and the former EIR have had for many years a very successful basic research programme concerning breeder reactors, and were in close cooperation with EFR. The activities within this programme had to be terminated owing to limitations in personnel and financial resources. The new PSI research programme is based upon two main areas, reactor physics and reactor thermal hydraulics. In both areas relatively small but valuable basic research tasks, the results of which are of interest to the breeder community, will be carried out. The lack of support of the former Breeder Programme led to capacity problems and finally to a total termination. Therefore one of the problems which had to be solved first was

  14. Tritium test of the tritium processing components under the Annex III US-Japan Collaboration

    International Nuclear Information System (INIS)

    The process ready components for Fuel Cleanup System were tested at the TSTA under the US-Japan Collaboration program. Palladium diffuser for tritium purification and Ceramic Electrolysis Cell for decomposition of tritiated water respectively were tested with pure tritium for years. The characteristics of the components with hydrogen isotopes, effects of impurities, and long-term reliability of the components were studied. It was concluded that these components are suitable and attractive for fusion fuel processing systems. (author)

  15. R and D activities of the liquid breeder blanket in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Won, E-mail: dwlee@kaeri.re.kr [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Eo Hwak; Kim, Suk Kwon; Yoon, Jae Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer MARS and GAMMA were developed for He coolant and liquid breeder analysis. Black-Right-Pointing-Pointer FMS/FMS and Be/FMS joining methods were developed and verified with high heat flux test. Black-Right-Pointing-Pointer High temperature and pressure nitrogen and He loops were constructed for heat transfer experiment for developed codes validation. Black-Right-Pointing-Pointer A PbLi breeder loop was constructed for components, MHD, and corrosion tests. Black-Right-Pointing-Pointer A chamber for tritium extraction with a gas-liquid contact method was constructed. - Abstract: A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5-1.0 MW/m{sup 2}. Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and

  16. R and D activities of the liquid breeder blanket in Korea

    International Nuclear Information System (INIS)

    Highlights: ► MARS and GAMMA were developed for He coolant and liquid breeder analysis. ► FMS/FMS and Be/FMS joining methods were developed and verified with high heat flux test. ► High temperature and pressure nitrogen and He loops were constructed for heat transfer experiment for developed codes validation. ► A PbLi breeder loop was constructed for components, MHD, and corrosion tests. ► A chamber for tritium extraction with a gas–liquid contact method was constructed. - Abstract: A liquid breeder blanket has been developed in parallel with the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) program in Korea. The Korea Atomic Energy Research Institute (KAERI) has developed the common fields of a solid TBM such as design tools, structural material, fabrication methods, and He cooling technology to support this concept for the ITER. Also, other fields such as a liquid breeder technology and tritium extraction have been developed from the designed liquid TBM. For design tools, system codes for safety analysis such as Multi-dimensional Analysis of Reactor Safety (MARS) and GAs Multi-component Mixture Analysis (GAMMA) were developed for He coolant and liquid breeder. For the fabrication methods, Ferritic Martensitic Steel (FMS) to FMS and Be to FMS joinings with a Hot Isostatic Pressing (HIP) were developed and verified with a high heat flux test of up to 0.5–1.0 MW/m2. Moreover, three mockups were successfully fabricated and a 10-channel prototype is being fabricated to make a rectangular channel FW. For the integrity of the joining, two high heat flux test facilities were constructed, and one using an electron beam has been constructed. With the 6 MPa nitrogen loop, a basic heat transfer experiment for code validation was performed. From the verification of the components such as preheater and circulator, a 9 MPa He loop was constructed, and it supplies high temperature (500 °C) and pressure (8 MPa) He to the high

  17. Remote fabrication of pellet fuels for United States breeder reactors

    International Nuclear Information System (INIS)

    Goal of the program is to demonstrate the feasibility of fabricating breeder fuel in a remotely operated and maintained mode by 1985. Development for pellet fuel fabrication is in the engineering stage with much of the equipment for ceramic unit operations in final design or currently under testing. Results to date confirm that remote fabrication of pellet fuels is feasible. Several of the processes and equipment items are described in this report

  18. Tritium handling in vacuum systems

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-10-01

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

  19. Magmatic tritium

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  1. Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

    Science.gov (United States)

    Li, Jia; Zhang, Xiaokang; Gao, Fangfang; Pu, Yong

    2016-02-01

    China Fusion Engineering Test Reactor (CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2 to ensure tritium self-sufficiency. A concept design for a water cooled ceramics breeding blanket (WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR. Based on this concept, a one-dimensional (1D) radial built breeding blanket was first designed, and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build. A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models, addressing neutron wall loading (NWL), tritium breeding ratio (TBR), fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components. The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  2. Breeding zone models of DEMO ceramic helium cooled blanket test module for testing in IVV-2M reactor

    International Nuclear Information System (INIS)

    The goal of DEMO ceramic helium cooled blanket test module (CHC BTM) is to demonstrate a breeding capability that would lead to tritium self-sufficiency in ITER reactor and to extract a high-grade heat suitable for electricity generation. Experimental validation of all the adopted design solutions is main important problem at design and calculation works carrying out in order to develop the CHC BTM. One important task for breeding zones feasibility validation is in-pile tests. Two models were developed and fabricated for testing in the fission IVV-2M reactor. Breeding zone is based on poloidal BIT-conception. The models structural material is ferrito-martensitic steel. Breeder material is lithium orthosilicate in pebble beds and pellet forms. Multiplier material is beryllium in pebble beds and porosity forms. The cooling is provided by helium at 10 MPa. The tritium produced in the breeder material is purged by the helium flow at 0.1-0.2 MPa. Designs of model description and experimental channel, results of neutronic and thermo-hydraulic calculations are presented in the paper. (orig.)

  3. Further neutronic analyses of the European ceramic B.I.T. blanket for Demo

    International Nuclear Information System (INIS)

    The present study concerns the most recent neutronic analyses of two design versions of the european ceramic B.I.T. blanket, jointly developed by ENEA and CEA since few years. The last year developments required a new 3-D geometry evaluations of the global TBR (Tritium Breeding Ratio). The results indicated that the ENEA version reaches a global TBR value of 1.13. The CEA version, in a 3-D model using a simplified description of the breeder module layout, reaches a TBR value of 1.12. Nuclear heat deposition density has been determined for all blanket components as a function of the poloidal co-ordinate. Shielding properties of this type of blanket have been analyzed

  4. Can the breeder go commercial

    International Nuclear Information System (INIS)

    Contrary to some beliefs in the electric utility industry that ERDA is committed to developing a commercial breeder economy, it is pointed out that ERDA isn't even willing to pay the total cost of the R and D program--and unless there is a major commitment from the private sector (the electric utility industry, in particular) the breeder program will die. The schedule as of Fall 1976 called for: (1) Fast Flux Test Facility (scheduled to go critical in 1979, operate in 1980); (2) Clinch River Breeder Reactor Project (CRBRP) (1/3 commercial size plant hopefully operating by 1983); (3) Prototype Large Breeder Reactor (planned construction starting in 1981, operating in 1988); and (4) Commercial Breeder Reactor (CBR-1 design work to start in 1983, construction in 1986, and operation in 1993). The $257 million the utility industry has pledged to the CRBRP was just for openers. The $2 billion follow-on breeder project being designed calls for massive capital input from a utility (or utility consortium)--and if that is not forthcoming, then in the words of an ERDA official, ''we'll have to reassess the whole breeder program.''

  5. Magmatic tritium

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  6. Development of a hydrogen permeation sensor for liquid breeder type TBMs

    International Nuclear Information System (INIS)

    Korea has developed Test Blanket Modules (TBMs) for ITER and DEMO fusion reactor. The tritium extraction from a breeder is one of the key technologies and its methods have been investigated. For developing the tritium extraction methods and evaluating the amount of tritium in the system, a reliable and correct sensor is required to measure the hydrogen concentration in liquid metal breeder. There are several researches for developing the sensors in the ITER participants and especially, EU has developed the permeation sensors trying to selecting materials with low Sievert's constant and high hydrogen diffusivity coefficient. When it comes to geometry, cylindrical and annulus shape of permeable sensors were invented to measure the hydrogen concentration in the liquid metal breeder. The annulus type was finally chosen to reduce the time necessary to measure the concentration. However, this response time is still too long time about tens of minutes to measure the tritium concentration in the online system. To solve this problem, we designed and fabricated the several sensors, with various materials and shapes, the results are introduced in the present study

  7. The Karlsruhe solid breeder blanket and the test module to be irradiated in ITER/NET

    International Nuclear Information System (INIS)

    The blanket for the DEMO reactor should operate at an average neutron flux of 2.2 MW/m2 for 20000 h. This requires the use of a structural material which can withstand high neutron fluences without swelling. The ferritic steel Manet was chosen for this purpose. The breeder material is in the form of Li4SiO4 pebbles of 0.35 to 0.6 mm diameter. The 6 mm thick beds of pebbles are placed between beryllium plates which are cooled by high pressure helium flowing inside steel tubes. Breeder material and beryllium are contained in radial canisters, placed inside boxes. The coolant helium enters the blanket at 250deg C, cools first the box walls and then the breeder and multiplier, and leaves the blanket at 450deg C. The maximum temperature in the first wall steel is 550deg C, while the minimum and maximum temperatures in the breeder are 380 and 820deg C, respectively. The resulting total tritium inventory in the breeder is only 10 g, and the real tridimensional tritium breeding ratio is 1.11. The conceptual design of the test module, of its extraction system and of the required out-of-reactor ancillary systems has allowed an estimate of the time constants of the various components and thus allowed an assessment of the requirements given by the testing of the modules on the NET/ITER machine. (orig.)

  8. International strategies for breeder development

    International Nuclear Information System (INIS)

    This paper studies the perspectives of breeder reactors development. The near term context has led some experts to the conclusion that breeder reactor technology is too far ahead of its time. Some have compared breeders to the supersonic airplane, Concorde: good technical performance but failure in its economic dimensions. In this paper, the author points out the major shortcomings of such an assessment which may be valid in the short time. However, with a short-term market-dominated perspective that uses an 8% discount rate, one can neglect every thing that is going to happen in 50 years. 6 refs., 11 figs

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

    International Nuclear Information System (INIS)

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

  10. Modeling unusual tritium release behavior from Li2O

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  12. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  13. Modelling for near-surface interaction of lithium ceramics and sweep-gas by use of cellular automation

    International Nuclear Information System (INIS)

    Tritium release from the lithium ceramics as a fusion reactor breeder material is strongly affected by the composition of the sweep-gas as result of its influences with the material's surface. The typical surface processes which play important roles are adsorption, desorption and interaction between vacancy site and the constituents of the sweep-gas. Among a large number of studies and models, yet it seems to be difficult to model the overall behaviour of those processes due to its complex time-transient nature. In the present work the coarse grained atomic simulation based on the Cellular Automaton (CA) is used to model the dynamics of near-surface interaction between Li2O surface and sweep-gas that is consisting of a noble gas, hydrogen gas and water vapour. (author)

  14. Inner Breeding Tritium Cycle Conceptual Design and Tritium Control Strategies for HCLL Blankets

    International Nuclear Information System (INIS)

    Design of the Inner Breeding Tritium Cycle (IBTC) for DEMO-like He-Cooled Lithium-Lead (HCLL) breeding blankets presents many open questions on solutions and choice of operational modes and parameters. Tritium transfer limits to the environment is the top design constraint for IBTC conceptual design. Among the options, Rankine cycle is the most conservative choice for Power Conversion Cycle in terms of technology maturity and tritium control requirements. Optimization of GC-HTR designs adaptation to DEMO primary coolant (PC) [300/500 oC, 80 bar] permit one to assess the two general diverse coolant chemistry options (HT oxidation or H2 isotopic swamping). Both options are discussed in terms of tritium control, and internal and external IBTC processing demands. Permeation from breeder into the He primary coolant and extraction of bred tritium out from the Pb15.7Li act as input givens of the IBTC conception. Dynamic tritium transfer under imposed MHD advection regimes coupling with convection fields in channel thermal steady-state distributions and radial breeding sources are inputs for actual assessments based on 2D moving-slab numerical techniques. IBTC relevant polarimetric runs showing the evolution of tritium poloidal-toroidal BB-in/BB-out concentration planes in LM channels are given. Ultimate tritium processing technologies performance (CPS: Coolant Purification System, TES: Tritium Extraction System from Pb15.7Li and TRS: Tritium Recovery System from TES purging columns) acts as boundary IBTC design constraints. Actual limits for transient modes are discussed. The IBTC design variables concern: i) system disposition in the IBTC lay-out, ii) use of tritium control solution at BB design level (ex. anti-permeation barrier), (iii) selection of system processing variables (ex. LM flowing velocities) and (iv) external effluents inputs for PC chemistry control. High processing efficiencies of CPS for relatively low flow rates means by-passing IHEx does not have a

  15. Tritium handling and vacuum considerations for the STARFIRE commercial tokamak reactor

    International Nuclear Information System (INIS)

    Tritium processing and vacuum pumping requirements were analyzed for the STARFIRE commercial fusion reactor design. It was found that vacuum pumps having a helium capture probability of 0.5 (total helium pump speed 1.2 x 104 m3/s) in combination with the proposed STARFIRE limiter-vacuum concept is sufficient to achieve plasma impurity control and, simultaneously, high fractional burnup (11%). The high fractional burnup and minimum fuel recycle time result in a very low fuel cycle tritium inventory, approx. 1300 g. A Lean-T burn method that can further reduce the fuel cycle inventory by 30 to 50% is discussed. D2O is proposed as a first wall coolant from considerations of plasma contamination (due to hydrogen isotope permeation through coolant tubes) and enrichment of recycled tritium from the coolant circuit. Tritium recovery from solid breeders, under realistic structural and breeder materials constraints, appears to represent a formidable task. The tritium inventory in the solid breeder is estimated to be as high as 10 kg, which would make the blanket the largest single hold-up point for tritium in the plant

  16. The Tritium White Paper

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    JAEA (Japan Atomic Energy Agency) manages 2 tritium handling laboratories: Tritium Processing Laboratory (TPL) in Tokai and DEMO-RD building in Rokkasho. TPL has been accumulating a gram level tritium safety handling experiences without any accidental tritium release to the environment for more than 25 years. Recently, our activities have focused on 3 categories, as follows. First, the development of a detritiation system for ITER. This task is the demonstration test of a wet Scrubber Column (SC) as a pilot scale (a few hundreds m3/h of processing capacity). Secondly, DEMO-RD tasks are focused on investigating the general issues required for DEMO-RD design, such as structural materials like RAFM (Reduced Activity Ferritic/Martensitic steels) and SiC/SiC, functional materials like tritium breeder and neutron multiplier, and tritium. For the last 4 years, we have spent a lot of time and means to the construction of the DEMO-RD facility and to its licensing, so we have just started the actual research program with tritium and other radioisotopes. This tritium task includes tritium accountancy, tritium basic safety research such as tritium interactions with various materials, which will be used for DEMO-RD and durability. The third category is the recovery work from the Great East Japan earthquake (2011 earthquake). It is worth noting that despite the high magnitude of the earthquake, TPL was able to confine tritium properly without any accidental tritium release

  18. International cooperation on breeder reactors

    International Nuclear Information System (INIS)

    In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

  19. Safe handling of tritium

    International Nuclear Information System (INIS)

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

  20. TFTR tritium handling concepts

    International Nuclear Information System (INIS)

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

  1. Tritium conference days

    International Nuclear Information System (INIS)

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

  2. Diverse ceramics of lithium synthesized by the combustion method; Diversos ceramicos de litio sintetizados por el metodo de combustion

    Energy Technology Data Exchange (ETDEWEB)

    Cruz G, D. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)

    2006-07-01

    Lithium ceramics would be applied as tritium breeder materials in the future fusion nuclear reactors. The present study shows a modified combustion synthesis to produce lithium ceramics using urea (CO(NH{sub 2}){sub 2}) - oxides (TiO{sub 2}, ZrO{sub 2}, Al{sub 2}O{sub 3})- hydroxide (LiOH) mixtures, that differ from the traditional combustion synthesis which utilizes, metal nitrates and fuels (urea/hydrazide, oxalyl dihydrazide, malonic acid dihydrazide, glycine, tetra formal tris azine, etc) mixtures in stoichiometric molar ratios to produce lithium ceramics In the present work, the modified combustion synthesis was performed to produce Li{sub 4}SiO{sub 4}, Li{sub 2}SiO{sub 3}, {beta}-Li{sub 2}TiO{sub 3}, m- Li{sub 2}ZrO{sub 3}, and {gamma}-LiAIO{sub 2}. It was necessary to add LiOH excess to balance Li{sub 2}O sublimation. The advantages and disadvantages of the modified combustion synthesis to prepare {beta}-Li{sub 2}TiO{sub 3} and m-Li{sub 2}ZrO{sub 3} ceramics were also studied. During synthesis were used insoluble oxide compounds. Although thermodynamic properties have been studied extensively from first principles, only limited insight exists about the kinetic properties of decomposition of lithium ceramics. In several works, Li{sub 4}SiO{sub 4} and Li{sub 2}SiO{sub 3} had shown high tritium solubility at lower temperatures than other tritium breeding materials. Therefore, we examined the thermal stability of these lithium silicates. Finally, the effect of 12000 kGy of {gamma} rays irradiation was analyzed in the lithium ceramics produced. The XRD analyses of irradiated samples showed decomposition of Li{sub 2}SiO{sub 3} to Li{sub 2}Si{sub 2}O{sub 5} due to radiolysis processes. Li{sub 4}SiO{sub 4} was decomposed to Li{sub 2}SiO{sub 3}. {beta}Li{sub 2}TiO{sub 3} did not decompose under {gamma} irradiation but m-Li{sub 2}ZrO{sub 3} decomposed to ZrO{sub 2}. Finally, {gamma}-LiAIO{sub 2} was stable to {gamma} irradiation. In general, consolidation effects

  3. Fusion Breeder Program interim report

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.; Lee, J.D.; Neef, W.

    1982-06-11

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83.

  4. Fusion Breeder Program interim report

    International Nuclear Information System (INIS)

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83

  5. Tritium contamination control

    International Nuclear Information System (INIS)

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

  6. Production behavior of irradiation defects in solid breeder materials

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Hirotake; Moritani, Kimikazu [Kyoto Univ. (Japan)

    1998-03-01

    The irradiation effects in solid breeder materials are important for the performance assessment of fusion reactor blanket systems. For a clearer understanding of such effects, we have studied the production behavior of irradiation defects in some lithium ceramics by an in-situ luminescence measurement technique under ion beam irradiation. The luminescence spectra were measured at different temperatures, and the temperature-transient behaviors of luminescence intensity were also measured. The production mechanisms of irradiation defects were discussed on the basis of the observations. (author)

  7. Tritium pellet injector results

    International Nuclear Information System (INIS)

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

  8. Study on refueling flow field of CITP-Ⅱ tritium production irradiation device

    International Nuclear Information System (INIS)

    The paper introduces the structure of CITP-Ⅱ tritium production irradiation device, presents the basic process of refueling breeders, studies the device's Solid-Gas phase flow, and computes the flow field parameters, such as pressure, velocity, and buoyancy. The result shows that the reliable structural design and reasonable gas differential pressure could realize the online refueling of irradiation device. (authors)

  9. Deuterium desorption behavior of solid tritium breeding material, lithium titanate

    International Nuclear Information System (INIS)

    Several types of blanket module, solid breeder/water or helium cooling, LiPb breeder/helium cooling, liquid LiN and Flibe, have been developed toward both ITER and a demonstration reactor in Japan. In the solid breeder blanket cooled by water, pellets of Li2TiO3 will be employed as tritium breeding material. Structure material in this blanket is low activation ferritic steel, F82H. The operation temperature is limited below approximately 820 K owing to swelling caused by neutron irradiation. Tritium produced by fusion neutrons in this breeding material has to be desorbed under a blanket operation for tritium recovery to be easy. The blanket module, however, has a spatial distribution of temperature. Thus, the tritium desorption behavior has to be clarified in order to make a scheme for tritium recovery. In the present study, a solid breeding material, Li2TiO3, was irradiated by 1.7 keV deuterium ions, and an amount of retained deuterium and deuterium desorption behavior were investigated using a thermal desorption. Dependence of deuterium fluence on amount of retained deuterium was also obtained. In order to examine trapping mechanisms of deuterium in Li2TiO3, similar experiments were conducted for Li and Ti. Deuterium implanted to Li2TiO3 desorbed in forms of HD, D2, HDO and D2O. The amount of deuterium desorbed in form of HD was a few times larger than those of other gas species. The desorption peak appeared at 600 K, but significant desorption up to 900 K was observed. The range of temperature in the lithium titanate of the blanket module is assumed from 550 K to 1200 K. These results suggest that the tritium produced in the blanket is partly not desorbed. Thus, the spatial distribution of temperature in the blanket has to be controlled for the tritium to be desorbed during the operation. The desorption spectra of deuterium in Li2TiO3 were similar to those of Li. This suggests that most of implanted deuterium is trapped in forms of Li-D and Li-OD. Based upon the

  10. Tritium confinement requirements for the water-cooled Pb-17Li blanket for demo

    International Nuclear Information System (INIS)

    In the water-cooled liquid Pb-17Li blanket concept for DEMO the limitation of tritium permeation from the breeder material into the cooling water will be required. In order to find out on what conditions this tritium permeation remains within reasonable limits, a 1-d FEM code was developed which evaluates the tritium partial pressure in Pb-17Li, the tritium inventory in the blanket material, and the tritium permeation from the Pb-17Li into the cooling water as a function of the permeation reduction factor of a barrier and the efficiency of the tritium extraction from Pb-17Li. With a parametric study the conditions were identified which allow a permeation rate of as little as 1 g.d-1 without pushing the requirements for permeation barriers and extraction efficiencies excessively far. an example is a barrier with a permeation reduction factor of 75 together with an extractor efficiency of approximately 83%. In these conditions the expected tritium inventory is attributed to approximately one third to the martensitic blanket structure (some ten grams) while two thirds will be found in the Pb-17Li. These inventory values are two orders of magnitude lower than in solid breeder blankets and are thus not considered a critical issue. 6 refs., 5 figs., 2 tabs

  11. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Zhongliang; Chen, Hongli, E-mail: hlchen1@ustc.edu.cn; Chen, Chong; Li, Min; Zhou, Guangming

    2015-06-15

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements.

  12. Preliminary neutronics design and analysis of helium cooled solid breeder blanket for CFETR

    International Nuclear Information System (INIS)

    Highlights: • Neutronics design of a helium cooled solid breeder blanket for CFETR was presented. • The breeding zones parallel to FW and perpendicular to FW were optimized. • A series of neutronics analyses for the proposed blanket were shown. - Abstract: Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor being designed in China to bridge the gap between ITER and future fusion power plant. Tritium self-sufficiency is one of the most important issues for CFETR and the tritium breeding ratio (TBR) is recommended not less than 1.2. As one of the candidates, a helium cooled solid breeder blanket for CFETR superconducting tokamak option was proposed. In the concept, radial arranged U-shaped breeding zones are adopted for higher TBR and simpler structure. In this work, three-dimensional neutronics design and analysis of the blanket were performed using the Monte Carlo N-Particle transport code MCNP with IAEA data library FENDL-2.1. Tritium breeding capability of the proposed blanket was assessed and the breeding zones parallel to first wall (FW) and perpendicular to FW were optimized. Meanwhile, the nuclear heating analysis and shielding performance were also presented for later thermal and structural analysis. The results showed that the blanket could well meet the tritium self-sufficiency target and the neutron shield could satisfy the design requirements

  13. Estimation on tritium production and inventory in beryllium

    International Nuclear Information System (INIS)

    Beryllium has been proposed as a candidate material for neutron multiplier on fusion blanket design, tritium will be produced and accumulated in beryllium during neutron irradiation. It is very important to estimate the tritium inventory on blanket design. Tritium production and inventory under 3.0 GW fusion power were calculated for the layered pebble bed blanket, with Li2O breeder and beryllium multiplier. Neutronics calculations were carried out by one-dimensional transport code, ANISN and tritium production was calculated by direct reaction of 9Be(n,T)7Li and two-step reactions of 9Be(n,α)6Li(n,α)T. At the position near first wall, the direct reaction occupied the majority of tritium production. However, at the position of the mid-depth, contribution of the two-step reaction was included the production from neutron slow down by beryllium itself. The ratio accounting for 50% of total tritium production of two-step reaction production over whole blanket for one year full power operation (FPY) was resulted

  14. Tritium extraction from neutron-irradiated lithium aluminate

    International Nuclear Information System (INIS)

    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 1012 -1013 n/cm2 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 % H2 that with Ar plus 0.1 % H2. 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)

  15. Tritium removing device

    International Nuclear Information System (INIS)

    Tritium-containing gases in a reactor container are discharged to a gas pressurizer and the gases pressurized there are sent to the primary side of a tritium separation device under a high or low pressure. Polyimide polymer separation membranes having selective permeability to elemental tritium and tritium vapor are coated in the tritium separation device. The separation device is divided into primary and secondary sides by the separation membranes and the pressure in the secondary side is lowered by a vacuum pump, etc. Tritium contained in the tritium-containing gases passes through the separation membranes selectively to be moved into the secondary side. Accordingly, tritium is treated in the elemental form and equipments for regeneration such as an adsorption column, etc. are no more necessary and the space can be saved due to minimization of the removing device. Further, since tritium can be removed continuously without storing a great amount of tritium, it is preferable in view of safety. (T.M.)

  16. Fast breeder reactor research

    International Nuclear Information System (INIS)

    , Italy, in April or May 1977. Recognizing the importance of international co-ope ration within the framework of IWGFR for preparing surveys, proposals and recommendations concerning sodium cooled fast breeder reactors, the Working Group prepared a number of joint documents with the help of experts from the participating countries, discussed them at the Eighth Annual Meeting and made recommendations on the preparation of subsequent joint documents. (author)

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

    International Nuclear Information System (INIS)

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

  18. On electrochemical tritium production

    International Nuclear Information System (INIS)

    This paper reports tritium formed in LiOD-D2O solutions in which Pd cathodes are used to evolve D2. Electrolysis was carried out for up to 4 1/2 months. Excess heat has been observed from 5 electrodes out of 28, tritium in 15 out of 53 but 9 out of 13 if the electrodes are limited to 1 mm diameter. Steady state tritium concentrations were 104-107 disintegrations min-1 ml-1. A weak correlation may exist between heat observed and tritium produced. The rate of production of tritium was ca 1010 atoms cm-2 s-1. The branching ratio of tritium to neutrons was ∼108. A theoretical dendrite enhanced fusion model is suggested. Growing gas layer breakdown occurs at sufficiently high surface potential dendrite tips and correspondingly fusion reactions occur. The model gives quantitative consistence with experiment, especially the sporadic nature and the observed branching ratio. (author)

  19. Tritium monitoring techniques

    International Nuclear Information System (INIS)

    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

  20. Confinement and Tritium Stripping Systems for APT Tritium Processing

    International Nuclear Information System (INIS)

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

  1. Tritium in metals

    International Nuclear Information System (INIS)

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

  2. Nuclear reaction analysis as a tool for the 3He thermal evolution in Li2TiO3 ceramics

    Science.gov (United States)

    Carella, E.; Sauvage, T.; Bès, R.; Courtois, B.; González, M.

    2014-08-01

    Li2TiO3 ceramic is one of the promising solid breeding candidates for fuel generation in deuterium-tritium Fusion reactors. The Tritium (T) release characteristics consist of a complex combination of gas diffusion stages inside the solid. Considering that this ceramic will produce high concentration of gaseous transmutation products (3H and 4He) when exposed to high-energy neutrons, there are considerable interests in studying 3He thermal evolution for the fundamental understanding of the light ion behavior in breeder blanket materials under reactor conditions. 3He atoms used to simulate the 4He incorporation were implanted by a 600 keV ion beam at a fluence of 1017 at/cm2 and the 3He(d,α)1H nuclear reaction analysis (NRA) technique was subsequently used to study depth profiles evolution after different thermal annealing treatments. The release experiments showed that 3He outgassing is not effective at room temperature, remaining quite negligible till 300 °C. After this temperature, the 3He content in the sample reduces steadily with increasing the annealing temperature, and less than 5% of the initial 3He concentration was found at 900 °C after an isochronal annealing, without significant depth-profile broadening. Scanning and transmission electron microscopies characterization highlight the microstructural changes of the implanted and annealed ceramic within the nuclear cascades zone. The correlation of results obtained by electron microscopy and NRA technique leads to the conclusion that the helium release is governed by a transport mechanism that involves rapid migration/diffusion through interconnected gas cavities and resulting microcracks before reaching grain boundaries and opened pores.

  3. The use of passive detectors to monitor tritium on surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Gammage, R.B.; Meyer, K.E.; Brock, J.L. [Oak Ridge National Lab., TN (United States)

    1996-12-31

    Commercially available BeO exoelectron dosemeters and electret ion chambers (EICs) are being adapted and applied to in situ field monitoring of tritium on surfaces. Thin layer BeO on a conductive graphite substrate is of the order of 50 times more sensitive to tritium than the EIC. At the US department of Energy release limit for fixed surface tritium of 5000 dpm per 100 cm{sup 2}, the exposure time for quantification with the exoelectron dosemeter is of the order of one hour. A multipoint Geiger counter was used for reading exoelectron emission. An alternative ceramic BeO dosemeter (Thermalox 995) has low electrical conductivity and will require a different reader to overcome problems of surface charging during exoemission. The electret is very easy to use and read. Its practical use will be for surfaces with relatively high levels of tritium contamination. (author).

  4. Tritium Attenuation by Distillation

    International Nuclear Information System (INIS)

    The objective of this study was to determine how a 100 Area distillation system could be used to reduce to a satisfactory low value the tritium content of the dilute moderator produced in the 100 Area stills, and whether such a tritium attenuator would have sufficient capacity to process all this material before it is sent to the 400 Area for reprocessing

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

    International Nuclear Information System (INIS)

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

  6. Tritium accumulation and release from Li{sub 2}TiO{sub 3} during long-term irradiation in the WWR-K reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibayeva, I., E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy of National Nuclear Center RK, Krasnoarmeyskaya str.-10, 071100 Kurchatov (Kazakhstan); Beckman, I., E-mail: info@rector.msu.ru [Moscow State University, Leninskie Gory, 119991 Moscow (Russian Federation); Shestakov, V. [Kazakh State University, Tole bi str., 96, Almaty (Kazakhstan); Kulsartov, T. [Institute of Atomic Energy of National Nuclear Center RK, Krasnoarmeyskaya str.-10, 071100 Kurchatov (Kazakhstan); Chikhray, E. [Kazakh State University, Tole bi str., 96, Almaty (Kazakhstan); Kenzhin, E. [Institute of Atomic Energy of National Nuclear Center RK, Krasnoarmeyskaya str.-10, 071100 Kurchatov (Kazakhstan); Kiykabaeva, A. [Kazakh State University, Tole bi str., 96, Almaty (Kazakhstan); Kawamura, H.; Tsuchiya, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan)

    2011-10-01

    Proposed mathematical and software analysis of reactor experiments allowed interpretation of the experimental results of a tritium release study. Tritium was continuously generated by the reaction of lithium-6 with thermal neutrons for various thermal conditions of lithium metatitanate (Li{sub 2}TiO{sub 3}). The main gas release parameters were calculated in order to assess the potential use of lithium metatitanate in tritium breeders. These parameters were: gas release rate, tritium retention, retention time, activation energy for thermal desorption as HT, activation energy for volume diffusion as T{sup +}, and the corresponding pre-exponential (frequency) indexes.

  7. Status and prospects of thermal breeders

    International Nuclear Information System (INIS)

    The main objective of this cooperative study and of this report is to evaluate the extent to which thermal breeders might complement or serve as an alternative to fast breeders in solving the long-term nuclear fuel supply problem. A secondary objective is to consider in a general way issues such as proliferation, safety, environmental impacts, economics, power plant availability, and fuel cycle versatility to determine whether thermal breeder reactors offer advantages or disadvantages with respect to such issues

  8. Feasibility study of a fission-suppressed tokamak fusion breeder

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  9. Feasibility study of a fission-suppressed tokamak fusion breeder

    International Nuclear Information System (INIS)

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m2 and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of 233U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U3O8 depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management

  10. R and D activities on helium cooled solid breeder TBM in Korea

    International Nuclear Information System (INIS)

    R and D activities currently being undertaken for HCSB TBM include joining technologies of structural material, breeder and reflector pebble material development, the effect of TBM ferritic-martensitic steel on the ripple of toroidal magnetic field, and ceramic coating on graphite pebble. The HIP joining performance of FM steel is evaluated. Lithium ceramic breeder and graphite reflector pebble fabrication methods are under development using special fabrication process, and the initial characteristics of the pebbles are assessed. Silicon carbide coating on graphite pebble is also investigated and its preliminary results are mentioned. Finally, an accurate evaluation of the effect of TBM and ferromagnetic inserts on magnetic field are implemented. The current results of these R and D issues are addressed in this paper.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  12. Tritium breeding in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.A.

    1982-10-01

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

  13. Design and analysis of breeding blanket with helium cooled solid breeder for ITER-TBM

    International Nuclear Information System (INIS)

    Test blanket module (TBM) is one of important components in ITER. Some of related blanket technologies of future fusion, such as tritium self-sufficiency, the exaction of high-grade heat, design criteria and safety requirements and environmental impacts, will be demonstrated in ITER-TBM. In ITER device, the three equatorial ports have allocated for TBM testing. China had proposed to develop independently the ITER-TBM with helium cooled solid breeder in 12th meeting of test blanket workgroup (TBWG-12). In this work, the preliminary design and analysis for Chinese HCSB TBM will be carried out. The TBM must be contains the function of the first wall, breeding blanket, shield and structure. Finally, in the period of preliminary investigation, HCSB TBM design adopt modularization concept which is helium as coolant and tritium purge gas, ferritic/martensitic steel as structural material, Lithium orthosilicate (Li4SiO4) as tritium breeder, beryllium pebble as neutron multiplier. TBM is allocated in standard vertical frame port. HCSB TBM consist of first wall, backplate, breeding sub-modules, caps, grid and support plate, and breeding sub-modules is arranged by layout of 2 x 6 in blanket box. In this paper, main components of HCSB TBM will be described in detail, also performance analysis of main components have been completed. (authors)

  14. Development of advanced blanket materials for solid breeder blanket of fusion reactor

    International Nuclear Information System (INIS)

    The design of advanced solid breeding blanket in the DEMO reactor requires the tritium breeder and neutron multiplier that can withstand the high temperature and high fluence, and the development of such as advanced blanket materials has been carried out by the cooperation activities among JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by wet process is a reference material as a tritium breeder, but the stability on high temperature has to be improved for application to DEMO blanket. As one of such the improved materials, TiO2-doped Li2TiO3 pebbles were successfully fabricated and TiO2-doped Li2TiO3 has been studied. For the advanced neutron multiplier, the beryllides that have high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that Be12Ti had lower swelling and tritium inventory than that of beryllium metal. The pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. From these activities, the bright prospect was obtained to realize the DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides. (author)

  15. Neutronics and thermal design analyses of US solid breeder blanket for ITER

    International Nuclear Information System (INIS)

    The US Solid Breeder Blanket is designed to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Safety, low tritium inventory, reliability, flexibility cost, and minimum R ampersand D requirements are the other design criteria. To satisfy these criteria, the produced tritium is recovered continuously during operation and the blanket coolant operates at low pressure. Beryllium multiplier material is used to control the solid-breeder temperature. Neutronics and thermal design analyses were performed in an integrated manner to define the blanket configuration. The reference parameters of ITER including the operating scenarios, the neutron wall loading distribution and the copper stabilizer are included in the design analyses. Several analyses were performed to study the impact of the reactor parameters, blanket dimensions, material characteristics, and heat transfer coefficient at the material interfaces on the blanket performance. The design analyses and the results from the different studies are summarized. 6 refs., 3 figs., 3 tabs

  16. TFTR tritium inventory analysis

    Energy Technology Data Exchange (ETDEWEB)

    Pontau, A.E.; Brice, D.K.; Buchenauer, D.A.; Causey, R.A.; Doyle, B.L.; Hsu, W.L.; Lee, S.R.; McGrath, R.T.; Mills, B.; Wampler, W.R.; Wilson, K.L. (Sandia National Labs., Livermore, CA (USA); Sandia National Labs., Albuquerque, NM (USA)); Langley, R. (Oak Ridge National Lab., TN (USA)); Dylla, H.F.; Heifetz, D.B.; Kilpatrick, S.; Lamarche, P.H.; Sissingh, R.A.P.; Ulrickson, M. (Princeton Univ., NJ (USA). Plasma Physics Lab.); Brooks, J.N. (Argonne National Lab., IL (USA))

    1989-06-01

    The Tokamak Fusion Test Reactor (TFTR) is scheduled to begin D-T operation in 1990 with the on-site tritium inventory limited to 5 grams. The physics and chemistry of the in-vessel tritium inventory will impact safety concerns, and also the entire operating schedule of the tokamak. We have investigated plasma-material interaction processes that will affect this first tritium-fueled tokamak. Tritium inventory estimates for TFTR are derived from: (1) Laboratory simulation, (2) in-situ plasma measurements, (3) post-run surface analysis, and (4) modeling. This paper presents the results of these investigations, the derivation of a tritium inventory estimate and its uncertainties, and a discussion of its impact. A particular discharge-by-discharge operating schedule has been developed and evaluated. The major source of in-vessel tritium inventory will be codeposition of tritium and eroded carbon onto surfaces. We find that the on-site limit may be approached unless specific inventory reduction techniques are invoked, e.g., discharge cleaning. (orig.).

  17. Tritium behaviors in plants

    International Nuclear Information System (INIS)

    The tritium intake of plants was briefly reviewed in this report. The major chemical forms of tritium released from nuclear facilities are HTO and HT and in the natural environment, tritium is also found in various OBT such as CH3T. The exposure dose to HTO by inhalation exposure in humans was evaluated by ICRP to be 104 fold higher than HT and 102 fold than CH3T. Whereas for the organic compound binding form, it was evaluated to be 2.3 times higher than that of HTO. To study the tritium transition into plants, especially edible parts such as vegetables and fruits and the transition process were thought important and many studies including theoretical analysis have been done mainly regarding HTO, HT and CH3T. The transition of HT tritium into plants was negligible. However, it was reported that the released HT was converted to HTO by microorganisms in surface soil and incorporated into plants. But, the HTO concentration of the leaves in potted plants always lower than that of water in the soil of the pot, suggesting that tritium was not concentrated by the plant. However, there are few studies on tritium transition via photosynthesis into plant tissues. (M.N.)

  18. Tritium technology. A Canadian overview

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

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

  20. Tritium facility at TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Sissingh, R.A.P. (Canadian Fusion Fuels Technology Project, Mississauga, ON (Canada)); Rossmassler, R.L. (Princeton Univ., NJ (USA). Plasma Physics Lab.)

    1990-06-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton began operation in December 1982. While it has operated successfully with protium and deuterium achieving record energy confinement time and record ion temperatures. TFTR's ultimate goal is to achieve the scientific break even point of Q=1 for which deuterium/tritium injection is needed. This paper will discuss the design parameters resulting from using tritium as a fuel, the design and operating philosophies employed, the additional systems and equipment required, the effect on the heating, ventilating and air conditioning systems, the tritium monitoring system and the personnel training. (orig.).

  1. Tritium powered luminescent signs

    International Nuclear Information System (INIS)

    A tritium powered emergency exit sign is provided comprising an elongated tritium powered light tube mounted at the focus of a parabolic reflector coated with a phosphorescent coating. The tube and reflector are mounted in an elongate trough shaped housing, with a translucent cover. In an emergency, e.g. in the event of a power failure in a building or other structure, the phosphorescent coating will reinforce the light emitted by the tritium tube for a short period whilst the eyes adjust to the darkness. (author)

  2. Experience in handling concentrated tritium

    International Nuclear Information System (INIS)

    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

  3. 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.

  4. Preliminary structural design and thermo-mechanical analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    Highlights: • A helium cooled solid breeder blanket module was designed for CFETR. • Multilayer U-shaped pebble beds were adopted in the blanket module. • Thermal and thermo-mechanical analyses were carried out under normal operating conditions. • The analysis results were found to be acceptable. - Abstract: With the aim to bridge the R&D gap between ITER and fusion power plant, the Chinese Fusion Engineering Test Reactor (CFETR) was proposed to be built in China. The mission of CFETR is to address the essential R&D issues for achieving practical fusion energy. Its blanket is required to be tritium self-sufficient. In this paper, a helium cooled solid breeder blanket adopting multilayer U-shaped pebble beds was designed and analyzed. Thermo-mechanical analysis of the first wall and side wall combined with breeder unit was carried out for normal operating steady state conditions. The results showed that the maximum temperatures of the structural material, neutron multiplier and tritium breeder pebble beds are 523 °C, 558 °C and 787 °C, respectively, which are below the corresponding limits of 550 °C, 650 °C and 920 °C. The maximum equivalent stress of the structure is under the allowable value with a margin about 14.5%

  5. Tests on technology of tritium production under neutron irradiation of liquid metal Pb-17Li

    International Nuclear Information System (INIS)

    Economic aspects and environmental safety of fusion reactor safety of fusion reactor depend on tritium production and retention technologies. Tests on technology of tritium production and permeability through austenite steel were performed on the breeder alloy Pb-17Li (600-800 g, 30% enrichment of Li-6) under irradiation of thermal neutron flux density 2 x 1017 m-2s-1 and temperature 540 K (for two containments) and 670 K (for one containment). Tritium release and permeability kinetics was monitored open-quotes on-lineclose quotes during liquid metal irradiation. Dependencies of tritium release and permeability under Pb-17Li irradiation on time (up to 2000 h) and irradiation temperature were obtained

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-15

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

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

    International Nuclear Information System (INIS)

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

  8. Method of removing tritium

    International Nuclear Information System (INIS)

    Purpose: To remove trituim in airs simply and reliably in a large amount. Constitution: Tritium contained in air is oxidized in an oxidizing column into water and incorporated in the air. The water-air mixture is caused to flow into and cooled in a first freezing type air drier where almost of tritium water in the air are condensated and separated from the air and, after falling through the drier, recovered by way of a drain tube. The air passing through the freezing type air drier in humidified by a humidifier and then caused to flow into the second freezing drier. Then, a slight amount of tritium water remained in the air is mixed with steams by the humidifier for easier separation, dried in a drier and removed with tritium into cleaned air. After properly humidifying the air in the humidifier, it is flown out through the exit. (Kamimura, M.)

  9. TFTR tritium program

    Energy Technology Data Exchange (ETDEWEB)

    Sissingh, R.A.P.; Rossmassler, R.L.

    1988-09-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton began operation in December 1982. Since then it has operated successfully with protium and deuterium achieving energy confinement time at peak electron density of 10/sup 19/ m/sup -3/s, with ion temperatures of 20 keV. This paper describes the systems and preparations required for D-T operation, i.e. introducing and operating the tokamak with tritium in order to achieve the scientific break even point of Q=1. These systems include the tritium storage and delivery system, the tritium injection systems, the tritium clean-up systems, and the plasma exhaust and collection systems. It is expected that TFTR will have these systems fully operational, with trained personnel.

  10. Tritium waste package

    Science.gov (United States)

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

    1995-01-01

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

  11. Tritium catalyzed deuterium tokamaks

    International Nuclear Information System (INIS)

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

  12. Tritium protective clothing

    International Nuclear Information System (INIS)

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

  13. 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)

  14. Nuclear reaction analysis as a tool for the {sup 3}He thermal evolution in Li{sub 2}TiO{sub 3} ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Carella, E., E-mail: elisabetta.carella@ciemat.es [National Fusion Laboratory, CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Sauvage, T., E-mail: thierry.sauvage@cnrs-orleans.fr [CEMTHI-CNRS, Site Cyclotron, 3A Rue de la Ferollerie, 45071 Orléans (France); Bès, R., E-mail: Rene.BES@cea.fr [CEMTHI-CNRS, Site Cyclotron, 3A Rue de la Ferollerie, 45071 Orléans (France); Courtois, B., E-mail: blandine.courtois@cnrs-orleans.fr [CEMTHI-CNRS, Site Cyclotron, 3A Rue de la Ferollerie, 45071 Orléans (France); González, M., E-mail: maria.gonzalez@ciemat.es [National Fusion Laboratory, CIEMAT, Av. Complutense 40, 28040 Madrid (Spain)

    2014-08-01

    Li{sub 2}TiO{sub 3} ceramic is one of the promising solid breeding candidates for fuel generation in deuterium–tritium Fusion reactors. The Tritium (T) release characteristics consist of a complex combination of gas diffusion stages inside the solid. Considering that this ceramic will produce high concentration of gaseous transmutation products ({sup 3}H and {sup 4}He) when exposed to high-energy neutrons, there are considerable interests in studying {sup 3}He thermal evolution for the fundamental understanding of the light ion behavior in breeder blanket materials under reactor conditions. {sup 3}He atoms used to simulate the {sup 4}He incorporation were implanted by a 600 keV ion beam at a fluence of 10{sup 17} at/cm{sup 2} and the {sup 3}He(d,α){sup 1}H nuclear reaction analysis (NRA) technique was subsequently used to study depth profiles evolution after different thermal annealing treatments. The release experiments showed that {sup 3}He outgassing is not effective at room temperature, remaining quite negligible till 300 °C. After this temperature, the {sup 3}He content in the sample reduces steadily with increasing the annealing temperature, and less than 5% of the initial {sup 3}He concentration was found at 900 °C after an isochronal annealing, without significant depth-profile broadening. Scanning and transmission electron microscopies characterization highlight the microstructural changes of the implanted and annealed ceramic within the nuclear cascades zone. The correlation of results obtained by electron microscopy and NRA technique leads to the conclusion that the helium release is governed by a transport mechanism that involves rapid migration/diffusion through interconnected gas cavities and resulting microcracks before reaching grain boundaries and opened pores.

  15. Tritium systems concepts for the next European torus (NET)

    International Nuclear Information System (INIS)

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

  16. The behaviour of diffusion and permeation of tritium through 316L stainless steel with coating of TiC and TiN + TiC

    International Nuclear Information System (INIS)

    The diffusion and permeation behaviour of tritium through the films of TiC and TiN + TiC coated on surface of 316L stainless steel by chemical vapour deposition has been described. The permeability of tritium through these two kinds of films is low and is five to six orders of magnitude lower than that in bulk at 200-500deg C. The films have good compatibility with the bulk, high resistance to thermal shock and irradiation. They can be used as clad materials of the tritium breeder in the breeder irradiation container in a fusion reactor tritium technology study and as candidate materials of the first wall in a fusion reactor study. (orig.)

  17. The gas-cooled Li2O moderator/breeder canister blanket for fusion-synfuels

    International Nuclear Information System (INIS)

    A new integrated power and breeding blanket is described. The blanket incorporates features that make it suitable for synthetic fuel production. It is matched to the thermal and electrical requirements of the General Atomic water-splitting process for producing hydrogen. The fusion reaction is the Tandem Mirror Reactor (TMR) using Mirror Advanced Reactor Study (MARS) physics. The canister blanket is a high temperature, pressure balanced, crossflow heat exchanger contained within a low activity, independently cooled, moderate temperature, first wall structural envelope. The canister uses Li2O as the moderator/breeder and helium as the coolant. ''In situ'' tritium control, combined with slip stream processing and self-healing permeation barriers, assures a hydrogen product essentially free of tritium. The blanket is particularly adapted to synfuels production but is equally useful for electricity production or co-generation

  18. Nuclear, thermo-mechanical and tritium release analysis of ITER breeding blanket

    International Nuclear Information System (INIS)

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li2TiO3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li2TiO3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328degC in

  19. Comparison of tritium production facilities

    International Nuclear Information System (INIS)

    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

  20. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    International Nuclear Information System (INIS)

    This paper presents results of conceptual design activities and associated R and D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R and D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  1. Dielectric and electrical design consideration of ceramics for fusion devices

    International Nuclear Information System (INIS)

    The research and development of high performance ceramics for nculear applications are increasing their importance. Especially in nuclear develoment, innovative and application of ceramics are needed in fusion reactors. Summarized are the develoment of new materials such as silicon nitride with good mechanical and electrical properties and the application of zirconia-based ceramics for high temperature electrolysis of tritiated water in a tritium recycling system. (orig.)

  2. An economic analysis of fusion breeders

    International Nuclear Information System (INIS)

    This paper presents a study of the economic performance of Fission/Fusion Hybrid devices. This work takes fusion breeder cost estimates and applies methodology and cost factors used in the fission reactor programs to compare fusion breeders with Liquid Metal Fast Breeder Reactors (LMFBR). The results of the analysis indicate that the Hybrid will be in the same competitive range as proposed LMFBRs and have the potential to provide economically competitive power in a future of rising uranium prices. The sensitivity of the results to variations in key parameters is included

  3. Tritium analysis at TFTR

    International Nuclear Information System (INIS)

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

  4. Tritium in the water environment

    International Nuclear Information System (INIS)

    Tritium activity concentrations in water environment in China have been summarized. The levels in different water categories can be listed as: precipitation, river, reservoir, tap water, well, sea and spring in the order of decrease of tritium. (5 tabs.)

  5. Helium-cooled molten-salt fusion breeder

    International Nuclear Information System (INIS)

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF2 + ThF4) is circulated through the blanket and to the on-line processing system where 233U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of 233U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the 233U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned

  6. Production of 4He and tritium from Be in the COBRA-1A2 irradiation

    International Nuclear Information System (INIS)

    The production of 4He and tritium has been calculated for beryllium irradiated in the COBRA-1A2 experiment in the Experimental Breeder Reactor II. Reaction rates were based on adjusted neutron spectra determined from reactor dosimetry measurements at three different elevations in the region of the beryllium capsules. Equations are given so that gas production can be calculated for any specific capsule elevation

  7. Production of {sup 4}He and tritium from Be in the COBRA-1A2 irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, L.R. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    The production of {sup 4}He and tritium has been calculated for beryllium irradiated in the COBRA-1A2 experiment in the Experimental Breeder Reactor II. Reaction rates were based on adjusted neutron spectra determined from reactor dosimetry measurements at three different elevations in the region of the beryllium capsules. Equations are given so that gas production can be calculated for any specific capsule elevation.

  8. Tritium - is it underestimated

    International Nuclear Information System (INIS)

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

  9. Tritium in HTR systems

    International Nuclear Information System (INIS)

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

  10. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  11. In-pile tritium release behaviour of lithiummetatitanate produced by extrusion-spheroidisation-sintering process in EXOTIC-9/1 in the high flux reactor, Petten

    Energy Technology Data Exchange (ETDEWEB)

    Peeters, M.M.W. [N.R.G., P.O. Box 25, 1755 ZG Petten (Netherlands)], E-mail: peeters@nrg-nl.com; Magielsen, A.J.; Stijkel, M.P.; Laan, J.G. van der [N.R.G., P.O. Box 25, 1755 ZG Petten (Netherlands)

    2007-10-15

    The irradiation programme EXOTIC (extraction of tritium in ceramics) is carried out within the European framework for the development of the helium cooled pebble bed concept. The EXOTIC-9/1 is the latest experiment in the series of EXOTICs that are irradiated in the high flux reactor in Petten. Tritium release and inventory in lithium containing ceramic pebbles are key properties to be tested in a TBM. New production routes of pebbles are developed, leading to different thermomechanical and tritium release properties. The objective of the EXOTIC-9/1 is to study in-pile tritium release behaviour of the latest developed lithiummetatitanate pebbles (Li{sub 2}TiO{sub 3}). The pebbles are produced by a extrusion-spheroidisation-sintering process at CEA. The new pebbles differ with respect to porosity from the lithiummetatitanate ceramics tested in the previous EXOTIC 8 programme. The pebbles have diameter in the range from 0.6 to 0.8 mm. Irradiation of EXOTC-9/1 started at 24 March 2005, and will continue until the end of 2006, in total about 400 irradiation days. The temperature is varied between 340 and 580 deg. C. Begin of Life (BOL) tritium production rate is 0.56 mCi/min. Based upon the in-pile tritium release measurements and the analysis of the tritium residence time it can be concluded that tritium release in the new batch of the high density Li{sub 2}TiO{sub 3} pebbles irradiated in EXOTIC 9/1 is rather slow compared to the ceramics irradiated in the EXOTIC 8 irradiation campaign. In this paper, the in-pile tritium behaviour will be reported during normal operation and during transients in temperature, purge gas chemistry and gasflow. The collected data is compared to tritium release data from ceramics irradiated in previous EXOTIC experiments with respect to tritium inventory, residence time and porosity.

  12. Tritium accounting during the first tritium experiment at JET

    International Nuclear Information System (INIS)

    This paper summarises the measuring procedures and the results of the tritium accounting during the first tritium experiment at JET, carried out in November 1991. The measurement of the amount of tritium injected into the Torus and of the quantity recovered from the Torus and the Neutral Injector Boxes is described and the accuracy of the data assessed. The new Gas Collection System used during the experiment is briefly described. The tritium recovery data taken in the months following the experiment are reviewed, with special attention to the first three weeks after the experiment. The total amount of tritium collected in the Gas Collection System is compared with the data of tritium release from the Torus and the Neutral Injection Boxes. The analysis of the data allows us to estimate the residual tritium inventory in the Injection System and in the Torus. (orig.)

  13. Development of LiF tile neutron shield and measurement of tritium release from it

    International Nuclear Information System (INIS)

    For neutron capture therapy of cancer, the neutron irradiation field with low gamma-ray is essential for selective treatment. From various lithium compounds, lithium fluoride LiF was selected as the shielding material for the present purpose, because of 1 large lithium density, 2 chemical stability, 3 easy treatment for nonpoison, and 4 small induced activity. In order to utilized LiF in pure chemical form, we have developed LiF tile, although as yet few investigations have been made of sintering a material in fluoride form. From viewpoint of ceramic technology, some new facts have been observed. The behavior of tritium, produced by the reaction of 6Li(n, α)T, in LiF tile was experimentally clarified. Tritium is released from LiF tile in two processes: (1) tritium released by recoil immediately after neutron irradiation, (2) tritium released with temperature condition and elasped time after temporaty containment in LiF tile. The former tritium was trapped in ethanol and measured with a liquid scintillator. While the latter was released in a loop by heating the irradiated tile and measured with a gas-flow-type tritium monitor. The following facts have been clarified: (1) Amount of recoiled tritium from LiF tile is --0.11 μCi/cm2/1014 nvt. (2) Most of tritium produced in LiF tile is contained in the tile itself. (3) Tritium contained in LiF tile is not released at temperature less than 3000C. (4) Contained tritium is released mainly at temperature between 400 -- 6500C. (5) The higher temperature at which LiF tile was sintered, the better containment of tritium. We have finally succeeded in developing LiF tile with low tritium release as neutron shielding material, which is now on sale in commercial base. (author)

  14. ARIES-I tritium system

    International Nuclear Information System (INIS)

    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

  15. The toxicity of tritium

    International Nuclear Information System (INIS)

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

  16. Tritium retention in TFTR

    International Nuclear Information System (INIS)

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

  17. Tritium Exchange in Biological Systems

    International Nuclear Information System (INIS)

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

  18. Recent progress in safety assessments of Japanese water cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Water Cooled Solid Breeder Test Blanket Module (WCSB TBM) is being designed by JAEA for the primary candidate TBM of Japan, and the safety evaluation of WCSB TBM has been performed. This reports presents summary of safety evaluation activities of the Japanese WCSB TBM, including nuclear analysis, source of RI, waste evaluation, occupational radiolysis exposure (ORE), failure mode effect analysis (FMEA) and postulated initiating event (PIE). For the purpose of basic evaluation of source terms on nuclear heating and radioactivity generation, two-dimensional nuclear analysis has been carried out. By the nuclear analysis, distributions of neutron flux, tritium breeding ratio (TBR), nuclear heat, decay heat and induced activity are calculated. Tritium production is calculated by the nuclear analysis by integrating distributions of TBR values, as about 0.2 g-T/FPD. With respect to the radioactive waste, the induced activity of the irradiated TBM is estimated. For the purpose of occupational radiolysis exposure (ORE), RI inventory is estimated. Tritium inventory in pebble bed of TBM is about 3 x 1012 Bq, and tritium in purge gas is about 3 x 1011 Bq. FMEA has been carried out to identify the PIEs that need safety evaluation. PIEs are summarized into three groups, i.e., heating, pressurization and release of RI. PIEs of local heating are converged without any special cares. With respect to heating of whole module, two PIEs are selected as the most severe events, i.e., loss of cooling of TBM during plasma operation and ingress of coolant into TBM during plasma operation. With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated, because rupture of the pipes result pressurization of such compartments, i.e., box structure of TBM, purge gas loop, TRS, VV, port cell and TCWS vault. Box structure of TBM is designed to withstand the maximum pressure of the cooling system. At other compartments

  19. Universal tritium transmitter

    International Nuclear Information System (INIS)

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

  20. Tritium permeation in EUROFER97 steel in EXOTIC-9/1 irradiation experiment

    International Nuclear Information System (INIS)

    This paper presents the results of the tritium permeation study in EUROFER97 carried out within the EXOTIC (EXtraction Of Tritium In Ceramics) irradiation experiment. In the EXOTIC 9/1 experiment, a pebble bed assembly containing Lithium Titanate (Li2TiO3) pebbles is irradiated for 300 days in the High Flux Reactor (HFR), in the temperature range between 340 and 580 °C, reaching a lithium burn up of 3.5% and 1.2 dpa of damage in steel. The primary objective of this experiment was to measure the in-pile tritium release characteristics of Li2TiO3 pebbles. Additionally tritium permeation through the EUROFER97 pebble bed wall was measured on line. The permeation of tritium was studied at steady state conditions, during temperature transients, and at different hydrogen concentrations in the helium purge gas flow. The model used in the analysis of the experimental data which account for co-permeation of tritium and hydrogen is presented. It has been demonstrated that the permeation of tritium under experiment conditions proceeds in the diffusion limited regime. From the analysis of the experimental data the permeability and diffusivity of tritium in EUROFER97 is determined

  1. Tritium in the aquatic environment

    International Nuclear Information System (INIS)

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

  2. Tritium in the aquatic environment

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-02-01

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

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

    International Nuclear Information System (INIS)

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

  4. The lithium ceramics life test in the WWR-K reactor

    International Nuclear Information System (INIS)

    Full text:Lithium ceramics is one of the most prospecting candidates for the material of the fusion reactor tritium-multiplying blanket. This is the compound Li2Ti0s modified with 0-10 mol.% of Li2O. Tritium is produced as a result of the nuclear reaction 6Li+n=3H+a + 4.78 MeV. The required conditions for practical application of a material in the tritium blanket are its mechanical stability, when a high level of the lithium atom burn up ionizes of the atomic percentage) is reached, and the tritium high-efficient yield within the temperature operational range. To study the lithium ceramics stability against radiation/thermal impact and capability to release produced tritium, the lithium ceramics life test has been carried out. In course of trials the lithium ceramic specimens was subject to the high-fluence irradiation under the ceramics temperature within the range 400-900 deg. C with the tritium release process uninterrupted monitoring. Ceramics irradiation has been performed in the specially developed ampoule assemblies, installed in two irradiation channels located in the WWR-K reactor core central region. 6Six ampoules contained 16.2 g of Li2TiOh enriched to 96 % in 6Li isotope. The reactor thermal power is 6 MW. To control irradiation parameters, the universal loop facility system was used. Information collection, representation and storage were implemented by the special computerized measuring system

  5. Measurement of spread of tritium using the tritium labeled compound

    International Nuclear Information System (INIS)

    It is known that a radioisotope disperses in the air from a radioisotope labeled compound in an aqueous solution via the isotopic exchange reaction. In this research, in order to examine the dispersion mechanism of tritium in the air from a tritium labeled compound, the model room which imitated a working room in the controlled area was made, and the radioactivity of the tritium contained in the air of the model room was measured by sampling the air in the model room. It was found that the dispersion rate of tritium in the air increased with the passage time from its purchase. The dispersion rate of tritium from 3H-ATP changed from 0.10% to 0.76% after 17.7 months. Furthermore, the two-dimensional distribution of tritium on the surface of the whole walls in the model room was obtained using an imaging plate technique. (author)

  6. Compatibility of sodium with ceramic oxides employed in nuclear reactors

    International Nuclear Information System (INIS)

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

  7. Tritium transport vessel using depleted uranium

    Energy Technology Data Exchange (ETDEWEB)

    Heung, L.K.

    1995-01-01

    A tritium transport vessel using depleted uranium was tested in the laboratory using deuterium and protium. The vessel contains 0.5 kg of depleted uranium and can hold up to 18 grams of tritium. The conditions for activation, tritium loading and tritium unloading were defined. The safety aspects that included air-ingress, tritium diffusion, temperature and pressure potentials were evaluated.

  8. Tritium release from lithium orthosilicate pebbles deposited with palladium

    International Nuclear Information System (INIS)

    Full text of publication follows: Slightly over-stoichiometric lithium orthosilicate pebbles have been selected as one optional breeder material for the European Helium Cooled Pebble Bed (HCPB) blanket. This material has been developed in collaboration of Research Center Karlsruhe and the Schott Glass, Mainz. The lithium orthosilicate pebbles are fabricated from lithium hydroxide and silica by a melting and spraying method in a semi-industrial scale facility. Lithium hydroxide was selected as the precursor since enriched lithium hydroxide is commercially available. The lithium orthosilicate pebbles produced by the process contains oxide phases besides orthosilicate, but it was also found that the oxide phases can be decomposed by annealing at high temperatures. The lithium orthosilicate pebbles produced in this way possesses satisfactory pebble characteristics. Therefore, the authors performed out-of-pile annealing tests using the lithium orthosilicate pebbles irradiated in a research reactor. Moreover, the effect of the deposition of palladium in the lithium orthosilicate pebbles on the behavior of tritium release was investigated. Palladium was deposited in the lithium orthosilicate pebbles by the incipient wet impregnation method using a solution of a palladium amino complex. The lithium orthosilicate pebbles were submitted to neutron irradiation at the Kyoto university research reactor. In the out-of-pile annealing experiments, the temperature of the breeder material placed in a tubular reactor made of quartz was raised from ambient temperature to 1173 K at a constant rate of 5 K/min under the stream of sweep gases. The tritium concentration in the outlet stream of the reactor was traced with two ionization chambers. The ionization chambers were used with a water bubbler, which enables to measure the concentrations of molecular form of tritium (HT) and tritiated water vapor (HTO) separately. In the experiments, a 0.1 % hydrogen/nitrogen sweep gas was used. The

  9. Measurement of tritium production rate distribution in natural LiAlO2/HDPE assembly irradiated by D-T neutrons

    International Nuclear Information System (INIS)

    A neutronics experiment was performed to measure the tritium production rate (TPR) profile in the breeder assembly with LiAlO2 as breeder and high density polyethylene (HDPE) as neutron reflector. The breeder assembly was irradiated with 14 MeV neutrons from DT neutron generator at IPR Neutronics Laboratory. The objective of the experiment was to validate the tritium production prediction capability of the Monte-Carlo code MCNP and FENDL 2.1 data library. The tritium production rate profile in the breeding assembly was measured by irradiating Li2CO3 pellets kept at various locations and then tritium counting liquid scintillation technique. Experiment was analyzed with 3D Monte-Carlo code MCNP with FENDL 2.1 cross-section data library. The calculation results were found to agree with the measured tritium production rates except one point near to the source. This experiment is a starting experiment in the series of benchmark experiments for the Indian Demo breeding blanket.

  10. Tritium neutrino mass experiments

    International Nuclear Information System (INIS)

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

  11. Environmental monitoring for tritium at tritium separation facility

    International Nuclear Information System (INIS)

    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)

  12. Tritium implantation in the accelerator production of tritium device

    International Nuclear Information System (INIS)

    We briefly describe the methods we have developed to compute the magnitude and spatial distribution of born and implanted tritons and protons in the Accelerator Production of Tritium (AFT) device. The methods are verified against experimental measurements and then used to predict that 16% of the tritium is implanted in the walls of the APT distribution tubes. The methods are also used to estimate the spatial distribution of implanted tritium, which will be required for determining the possible diffusion of tritium out of the walls and back into the gas stream

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

    International Nuclear Information System (INIS)

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

  14. Tritium monitoring : present status

    International Nuclear Information System (INIS)

    The report summarizes the present status of techniques employed for the monitoring of tritium in water, air and other samples. A brief mention of the work done by numerous workers in the field, critical comments about the work and a fairly exhaustive list of references about the work done during the last 4 decades has been presented. On-line monitoring on real time basis in nuclear reactors is also discussed. (author). 83 refs., 10 refs., 2 tabs

  15. Tritium in atmospheric hydrogen

    OpenAIRE

    Martin, J. David; Hackett, Joseph P.

    2011-01-01

    The radioactivity of tritiated hydrogen (HT) in the atmosphere in Westwood, New Jersey was measured at approximately weekly intervals from August 1971 to August 1973. The background level remained constant at approximately 80 tritium atoms per milligram of air. Frequent increases in the activity level of up to an order of magnitude were observed until January 1973. The source(s) of HT which was responsible for the frequent increases apparently ceased as a tropospheric source in January 1973. ...

  16. Muon capture by tritium

    International Nuclear Information System (INIS)

    The muon capture rate is computed with realistic wave function for the initial tritium nuclei (Faddeev equations on configuration space with realistic potentials), and plane wave approximation for the final three neutrons, with the effective Hamiltonian of Fujii and Primakoff for muon capture and via a non energy weighted sum rule. Such a forbidden transition is hoped to be a probe for exchange current contributions

  17. Accelerator breeder with uranium, thorium target

    International Nuclear Information System (INIS)

    An accelerator breeder, that uses a low-enriched fuel as the target material, can produce substantial amounts of fissile material and electric power. A study of H2O- and D2O-cooled, UO2, U, (depleted U), or thorium indicates that U-metal fuel produces a good fissile production rate and electrical power of about 60% higher than UO2 fuel. Thorium fuel has the same order of magnitude as UO2 fuel for fissile-fuel production, but the generating electric power is substantially lower than in a UO2 reactor. Enriched UO2 fuel increases the generating electric power but not the fissile-material production rate. The Na-cooled breeder target has many advantages over the H2O-cooled breeder target

  18. Improved fuel element for fast breeder reactor

    International Nuclear Information System (INIS)

    The invention, in which the United States Department of Energy has participated as co-inventor, relates to breeder reactor fuel elements, and specifically to such elements incorporating 'getters', hereafter designated as fission product traps. The main object of the invention is the construction of a fast breeder reactor fuel pin, free from local stresses induced in the cladding by reactions with cesium. According to the invention, the fast breeder fuel element includes a cladding tube, sealed at both ends by a plug, and containing a fissile stack and a fertile stack, characterized by the interposition of a cesium trap between the fissile and fertile stacks. The trap is effective at reactor operating temperatures in retaining and separating the cesium generated in the fissile material and preventing cesium reaction with the fertile stack. Depending on the construction method adopted, the trap may consists of a low density titanium oxide or niobium oxide pellet

  19. The role and problems of the breeder

    International Nuclear Information System (INIS)

    World uranium resources are discussed and it is concluded that the period of availability of uranium for use in the present type of nuclear power stations is not much greater than that of oil. The neutron economies of fast and thermal reactors are compared, and the advantages of the breeder for the world uranium economy are demonstrated. The main impediments to the use of the fast breeder are considerations of safety, public acceptance and economics. Fast reactor safety is discussed and the health hazards and possible mis-use of plutonium for terrorism and weapons proliferation are considered. It is widely accepted that the U.K. cannot economically justify the development of the breeder alone and is likely to choose to co-operate with Western Europe. A public enquiry in the U.K. seems certain and would be welcomed by the nuclear industry. (author)

  20. A personal tritium monitor

    International Nuclear Information System (INIS)

    A tritium monitor, similar in size to a normal gamma survey meter, is being developed to improve the measurement of tritiated water vapour (HTO) near workers in Candu nuclear power plants. Methods are available for sampling and monitoring on-line from work areas; the instrument described here is intended to complement such monitoring by allowing on-the-spot individual assessment of tritium hazards. Size, mass and cost are more important than sensitivity in an instrument of this kind than in a central monitor. Accordingly, only inexpensive, readily obtainable mechanical and electrical components have been used in a simple assembly needing little machining. The tritium detector is an ionization chamber. A signal proportional to the concentration of HTO in air is obtained as the difference between the currents from two 90 cm3 ionization chambers. Sample air flows directly through one chamber and through the other after being dried by passing through a replaceable desiccant cartridge. This technique reduces the unwanted signals from gamma radiation and radioactive noble gases. The electronics comprise a MOSFET, single chip amplifier and a liquid crystal digital display that indicates concentrations in the range 1-1999 (MPC)sub(a). The mass of the instrument is 2 kg. (H.K.)

  1. A prototype wearable tritium monitor

    International Nuclear Information System (INIS)

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

  2. Ferritic steels for the first generation of breeder blankets

    International Nuclear Information System (INIS)

    Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

  3. A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

    International Nuclear Information System (INIS)

    Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

  4. Ceramic materials for fission and fusion nuclear reactors

    International Nuclear Information System (INIS)

    A general survey on the ceramics for nuclear applications is presented. For the fission nuclear reactor, the ceramics materials are almost totally used as fuel e.g. (U,Pu)O2; other types of ceramics, e.g. Uranium-Plutonium carbide and nitride, have been investigated as potential nuclear fuels. The (U,Pu)N compound is to be the fuel for the space nuclear power reactor in the U.S.A. For the fusion nuclear reactor, the ceramics should be the fundamental materials for many components: first wall, breeder, RF heating systems, insulant and shielding parts, etc. In recent years many countries are involved on the research and development of ceramic compounds with the principal purpose of being used in the fusion powerplant (year 2010-2020 ?). An effort has been even made to verify if it is possible to use more ceramic components in the fission nuclear plant (probably differntly disigned) to improve the safety level

  5. Fast breeder reactors an engineering introduction

    CERN Document Server

    Judd, A M

    1981-01-01

    Fast Breeder Reactors: An Engineering Introduction is an introductory text to fast breeder reactors and covers topics ranging from reactor physics and design to engineering and safety considerations. Reactor fuels, coolant circuits, steam plants, and control systems are also discussed. This book is comprised of five chapters and opens with a brief summary of the history of fast reactors, with emphasis on international and the prospect of making accessible enormous reserves of energy. The next chapter deals with the physics of fast reactors and considers calculation methods, flux distribution,

  6. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  7. An overview of tritium production

    International Nuclear Information System (INIS)

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

  8. Tritium monitor and collection system

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-03-26

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

  9. Tritium monitor and collection system

    Science.gov (United States)

    Bourne, Gary L.; Meikrantz, David H.; Ely, Walter E.; Tuggle, Dale G.; Grafwallner, Ervin G.; Wickham, Keith L.; Maltrud, Herman R.; Baker, John D.

    1992-01-01

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

  10. Tritium accounting for PHWR plants

    International Nuclear Information System (INIS)

    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 D2O 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)

  11. Tritium gas transfer pump development

    International Nuclear Information System (INIS)

    Non-lubricated, hermetically sealed pumps for tritium service have been selected to replace Sprengel pumps in the existing Tritium Facility. These pumps will be the primary gas-transfer pumps in the planned Replacement Tritium Facility. The selected pumps are Metal Bellows Corporation's bellows pumps and Normetex scroll pumps. Pumping range for a Normetex/Metal Bellows system is from 0.01 torr suction to 2300 torr discharge. Performance characteristics of both pumps are presented. 10 figs

  12. Structural Ceramics

    Science.gov (United States)

    1986-01-01

    This publication is a compilation of abstracts and slides of papers presented at the NASA Lewis Structural Ceramics Workshop. Collectively, these papers depict the scope of NASA Lewis' structural ceramics program. The technical areas include monolithic SiC and Si3N4 development, ceramic matrix composites, tribology, design methodology, nondestructive evaluation (NDE), fracture mechanics, and corrosion.

  13. Advanced Ceramics

    International Nuclear Information System (INIS)

    The First Florida-Brazil Seminar on Materials and the Second State Meeting about new materials in Rio de Janeiro State show the specific technical contribution in advanced ceramic sector. The others main topics discussed for the development of the country are the advanced ceramic programs the market, the national technic-scientific capacitation, the advanced ceramic patents, etc. (C.G.C.)

  14. Tritium in the aquatic environment

    International Nuclear Information System (INIS)

    Most of the tritium released from nuclear facilities into the atmosphere eventually reaches the aqueous environment where it is rapidly taken up by aquatic organisms. This paper reviews the current literature on tritium in the aquatic environment. Conclusions from the review, which covered algae, aquatic plants, invertebrates, fish, and food chain studies, were that aquatic organisms incorporate tritium into their tissue free water very rapidly and reach concentrations near that of the external medium. Incorporation of tritium from triated water into the organic matter of cells is at a slower rate than incorporation into the tissue free water. If organisms consume tritiated food, incorporation of tritium into the organic matter is faster and a higher tritium concentration is reached than when the organisms are exposed to only tritiated water. Incorporation of tritium bound to molecules into the organic matter depends on the chemical form of the 'carrier' molecule. No evidence was found that biomagnification of tritium occurs at higher tropic levels. Radiation doses to large populations of humans from tritium releases will most likely be from the consumption of contaminated water rather than contaminated aquatic food products. (author)

  15. Status and prospects of advanced fissile fuel breeders

    International Nuclear Information System (INIS)

    Fusion--fission hybrid systems, fast breeder systems, and accelerator breeder systems were compared on a common basis using a simple economic model. Electricity prices based on system capital costs only were computed, and were plotted as functions of five key breeder system parameters. Nominally, hybrid system electricity costs were about twenty-five percent lower than fast breeder system electricity costs, and fast breeder system electricity costs were about forty percent lower than accelerator breeder system electricity costs. In addition, hybrid system electricity costs were very insensitive to key parameter variations on the average, fast breeder system electricity costs were moderately sensitive to key parameter variations on the average, and accelerator breeder system electricity costs were the most sensitive to key parameter variations on the average

  16. A review of fusion breeder blanket technology, part 1

    International Nuclear Information System (INIS)

    This report presents the results of a study of fusion breeder blanket technology. It reviews the role of the breeder blanket, the current understanding of the scientific and engineering bases of liquid metal and solid breeder blankets and the programs now underway internationally to resolve the uncertainities in current knowledge. In view of existing national expertise and experience, a solid breeder R and D program for Canada is recommended

  17. Technology developments for improved tritium management

    International Nuclear Information System (INIS)

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

  18. Density functional study of lithium vacancy in Li4SiO4: Trapping of tritium and helium

    Science.gov (United States)

    Shi, Yanli; Lu, Tiecheng; Gao, Tao; Xiang, Xiaogang; Zhang, Qinghua; Yu, Xiaohe; Gong, Yichao; Yang, Mao

    2015-12-01

    Li4SiO4 is a solid breeder material that has important applications in future fusion reactors. The interaction of tritium/helium with lithium vacancy is investigated implementing pseudopotential plane wave method within density functional theory. Models of all types of lithium vacancies and vacancy-tritium/helium defect complexes are created. Possible tritium trap sites in lithium vacancies are examined and the formation energies, vacancy-tritium/helium interaction energies and electronic structures of the defects are calculated. The results indicate that the tritium atom trapped in the lithium vacancy bonds with one of the surrounding oxygen atoms. The formation energies of the vacancy-tritium complexes are in the range of 0.41-1.28 eV under oxygen-rich condition. The interaction energy calculation shows the lithium vacancy has strong tritium trapping capabilities. Moreover, the vacancy-helium complex formation energies are in the range of 1.97-3.43 eV under O-rich condition. The vacancy-helium interaction is relatively weaker, suggesting the helium atom may escape the lithium vacancy more easily.

  19. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve reactor doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discused. (Author)

  20. Possible types of breeders with thorium cycle

    International Nuclear Information System (INIS)

    Neutronics calculations of simplified homogeneous reactor models show the possibility that metal-fueled LMFBRs and coated particle fueled gas cooled reactors achieve doubling times of around 10 years with the thorium cycle. Three concepts of gas-cooled thorium cycle breeders are discussed. (Author)

  1. Tritium accountancy in fusion systems

    International Nuclear Information System (INIS)

    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)

  2. Tritium practices past and present

    International Nuclear Information System (INIS)

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

  3. Tritium permeation, contamination and decontamination

    International Nuclear Information System (INIS)

    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 tritium permeation, contamination and decontamination have been conducted by the CO2 team. The results are summarized as follows: (1) Study for the development of the tritium permeation barrier was carried out. A ZrO2 film with a magnesium phosphate layer sintered on a SUS 430 steel plate showed excellent reduction in the hydrogen permeation. (2) The non-destructive method using an imaging plate was proposed to monitor tritium release from contaminated materials. The method was applied to SUS 316 steel and revealed that the tritium release from SUS 316 steel was diffusion-limited. (3) As for contamination-protection and decontamination techniques, improvement in the decontamination rate from SUS 316 steel was obtained by providing CrO2 coating. (J.P.N.)

  4. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

    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

  5. The tritium operations experience on TFTR

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-09-01

    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.

  6. Tritium-surface interactions

    International Nuclear Information System (INIS)

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

  7. Thermal-hydraulic design and analysis of helium cooled solid breeder blanket for Chinese Fusion Engineering Test Reactor

    International Nuclear Information System (INIS)

    To bridge the gap between ITER and DEMO and to realize the fusion energy in China, a fusion device Chinese Fusion Engineering Test Reactor (CFETR) was proposed and being designed aiming at 50-200 MW fusion power, 30-50% duty time factor, and tritium self-sustained. Three kinds of tritium breeding blanket concepts, including helium-cooled solid blanket, water-cooled solid blanket and liquid metal-cooled liquid blanket, have been considered for CFETR. Compared to ITER test blanket module, the blanket design for CFETR is facing much more challenges due to the compulsive requirements of tritium self-sufficiency, nuclear heat removal and the space limitation for blanket installation. In this paper, a kind of helium cooled solid tritium breeder blanket was designed for CFETR full superconducting tokamak. The thermal-hydraulic designs were carried out based on the blanket structure design and neutronics calculation. The performance evaluation was conducted using ANSYS, and three-dimensional fluid-solid coupled models were modeled for the accuracy results. The results showed that the FW and BU can satisfy the design requirements. (author)

  8. Ceramic joining

    Energy Technology Data Exchange (ETDEWEB)

    Loehman, R.E. [Sandia National Lab., Albuquerque, NM (United States)

    1996-04-01

    This paper describes the relation between reactions at ceramic-metal interfaces and the development of strong interfacial bonds in ceramic joining. Studies on a number of systems are described, including silicon nitrides, aluminium nitrides, mullite, and aluminium oxides. Joints can be weakened by stresses such as thermal expansion mismatch. Ceramic joining is used in a variety of applications such as solid oxide fuel cells.

  9. Neutronics R&D efforts in support of the European breeder blanket development programme

    Science.gov (United States)

    Fischer, U.; Batistoni, P.; Klix, A.; Kodeli, I.; Leichtle, D.; Perel, R. L.

    2009-06-01

    The recent progress in the R&D neutronics efforts spent in the EU to support the development of the HCLL and HCPB breeder blankets is presented. These efforts include neutronic design activities performed in the framework of the European DEMO reactor study, validation efforts by means of neutronics mock-up experiments using 14 MeV neutron generators and the development of dedicated computational tools such as the conversion software McCad for the automatic generation of a Monte Carlo geometry model from available CAD data, and the MCSEN code for Monte Carlo based calculations of sensitivities and uncertainties by using the track length estimator. The supporting validation effort is devoted to the capability of the neutronics tools and data to predict the tritium production and other nuclear responses of interest in neutronics mock-up experiments. Such an experiment has been conducted on a HCPB mock-up while another on a HCLL mock-up is in progress.

  10. Neutronics R and D efforts in support of the European breeder blanket development programme

    International Nuclear Information System (INIS)

    The EU fusion technology programme considers two blanket development lines, the Helium-Cooled Pebble Bed (HCPB) blanket with Lithium ceramics pebbles as breeder material and beryllium pebbles as neutron multiplier, and the Helium-Cooled Lithium-Lead (HCLL) blanket with the Pb-Li eutectic alloy acting both as breeder and neutron multiplier. The long-term strategy aims at providing validated engineering designs of breeder blankets for a fusion power demonstration reactor (DEMO). As an important intermediate step, the breeder blankets need to be tested in a real fusion environment as provided by ITER. HCPB and HCLL Test Blanket Modules (TBM) have been accordingly designed for tests in dedicated ITER blanket ports. The nuclear design and performance of the breeder blanket modules rely on the results provided by neutronics design calculations. Validated computational tools and qualified nuclear data are required for high prediction accuracies including reliable uncertainty assessments. Complementary to the application of established standard tools and data for design analysis, a dedicated neutronics R and D effort is therefore conducted in the EU. This includes the development of dedicated computational tools, the generation of high quality nuclear data and their validation through integral experiments. The recent neutronic design efforts have been devoted to the European DEMO reactor study comprising (i) Monte Carlo based pre-analysis for the dimensioning of the shielding system, (ii) the generation of a generic CAD based Monte Carlo geometry model, and (iii) performance analysis for HCLL and HCPB based DEMO variants. The recent focus of the validation effort is on neutronics TBM mock-up experiments. The first experiment of this kind was performed on a TBM mock-up of the HCPB breeder blanket. The follow-up experiment on a neutronics HCLL TBM mock-up is currently under preparation. Computational pre-analysis were performed to optimise the design of the mock

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

    International Nuclear Information System (INIS)

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

  12. Alternate fuel cycles for fast breeder reactors

    International Nuclear Information System (INIS)

    In this contribution to the syllabus for Subgroup 5D, a full range of alternate breeder fuel cycle options is developed and explored as to energy supply capability, resource utilizations, performance characteristics and technical features that pertain to proliferation resistance. Breeding performance information is presented for designs based on Pu/U, Pu/Th, 233 U/U, etc. with oxide, carbide or metal fuel; with lesser emphasis, heterogeneous and homogeneous concepts are presented. A potential proliferation resistance advantage of a symbiotic system of a Pu/U core, Th blanket breeder producing 233 U for utilization in dispersed LWR's is identified. LWR support ratios for various reactor and fuel types and the increase in uranium consumption with higher support ratios are identified

  13. What can fast breeders do for Ontario

    International Nuclear Information System (INIS)

    Fast reactors have the potential of significantly reducing Ontario's demand for natural resources while meeting virtually any requirements for nuclear power this province may have. The breeding efficiency of the fast reactors does not affect the overall uranium consumption of the system to any significant extent. It is, however, an important economic factor in a breeder/burner system. To minimize the resource consumption, the fast reactors should be introduced in Ontario at the onset of the next century. The 'breeder-burner' mix of reactors can effectively reduce the fissile inventory of the whole power system (including the inventory in irradiated fuel storage bays). For the nuclear capacity growth scenarios thought to be applicable in Ontario, the fast reactor systems have about the same or lower requirements for natural uranium as the best (self-sustaining thorium) CANDU cycles. Compared to all other advanced CANDU cycles, the fast reactors yield a substantial resource saving. (auth)

  14. Tritium removal from various lithium aluminates irradiated by fast and thermal neutrons (COMPLIMENT experiment)

    International Nuclear Information System (INIS)

    Within the frame of the COMPLIMENT experiment, γ-LiAlO2 specimens with different microstructures (grain size distributions) were tested in the same environmental conditions to compare the effects caused by 6Li(n, α)T reaction and by fast neutron scattering, the damaging dose being held at about the same level (1.6-1.8 dpa). The tritium retention times were obtained by the tritium removal of isothermal annealing under He + 0.1% H2 sweeping gas. In spite of the different Li burnups (2.5% and 0.25%) and the residual tritium concentrations which were found in the irradiated specimens (4.3 Ci/g and 0.09 Ci/g, respectively, for specimens held at 450 C during the irradiations), the kinetics of tritium removal was not found to be discriminated by the two different irradiations. Moreover, the results were found to agree with those previously obtained by the ''in-situ'' TEQUILA experiment, performed on the same type of Li ceramics. Hence, the apparent first order desorption mechanism has been confirmed to control the kinetics of tritium removal from the porous fine grain γ-LiAlO2 ceramics. (orig.)

  15. Technological questions of the breeder fuel cycle

    International Nuclear Information System (INIS)

    Since the contributions by the Karlsruhe Nuclear Research Center to the construction of SNR 300 have been completed to a large extent and irradiated KNK II fuel subassemblies have now become available, the possibility and necessity arise of concentrating efforts on the breeder fuel cycle. This work was started in 1980. The 17 papers presented at this seminar will provide a survey of intermediate results obtained until today. (orig./HP)

  16. FOWL CHOLERA IN A BREEDER FLOCK

    OpenAIRE

    Z. Parveen, A. A. Nasir, K.Tasneem and A. Shah

    2003-01-01

    During January, 2003 Pasteurella multocida the causative agent of fowl cholera was isolated from a breeder flock in Lahore District. The age of the flock was 245 days. Increased mortality, swollen wattles and lameness were the clinical findings present in almost all the affected birds, while gross lesions were typical of fowl cholera. To prove the virulence of the organism, mice and six-week old cockerals were infected and P. multocida was reisolated.

  17. The fast breeder reactor Rapsodie (1962)

    International Nuclear Information System (INIS)

    In this report, the authors describe the Rapsodie project, the French fast breeder reactor, as it stands at construction actual start-up. The paper provides informations about: the principal neutronic and thermal characteristics, the reactor and its cooling circuits, the main handling devices of radioactive or contaminated assemblies, the principles and means governing reactor operation, the purposes and locations of miscellaneous buildings. Rapsodie is expected to be critical by 1964. (authors)

  18. Fast Breeder Development: EDF's point of view

    International Nuclear Information System (INIS)

    This paper presents EDF's views and contributions to fast breeder development and to the French SFR trilateral program. Utility requirements are first outlined, based on the approach followed for the EPR reactor. R and D contributions are presented in the areas of core physics, safety, technology innovations, materials, deployment and fuel cycle scenarios. The paper also deals with some of the issues of the 2020 French prototype as seen by EDF.

  19. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  20. The breeder's broke - who will save it

    International Nuclear Information System (INIS)

    In February, Bonn must decide on whether to tear down the most expensive building of the country: The Fast Breeder at Kalkar, which once seemed to lead to a glorious future. DM 3.500 million have been spent on it already and DM 1.000 million more will be needed. But the state has no money. The report is given by Wolfgang Hoffmann and Horst Bieber. (orig.)

  1. Future designs of breeder reactors (Europe, USA)

    International Nuclear Information System (INIS)

    Sodium-cooled reactors with a fast neutron core today are the only fission reactors that offer the reactor physics required for the breeding process and the complete conversion of U-238 or Th-232 into fissile fuel. There are currently five prototype breeder reactors in operation in England, France, and the USSR. The trends observable in development work aim at reducing capital cost, enhancing and improving passive shutdown performance, and simplifying the fuel cycle. (orig.)

  2. Utility considerations for tritium production

    International Nuclear Information System (INIS)

    The Southern Nuclear Company has a long-standing commitment to nuclear power and is interested in pursuing the possible use of one of its existing commercial reactors as an alternative for reinitiating the production of tritium to support the nation's defense program requirements. We understand that Congress and the administration agree on the need to replenish the nation's supply of tritium and that a number of production options are under consideration. This paper discusses the financial considerations, legal and regulatory considerations for the production of tritium utilizing a commercial power reactor

  3. Tritium monitor for fusion reactors

    International Nuclear Information System (INIS)

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

  4. Tritium Elimination System Using Tritium Gas Oxidizing Bacteria

    International Nuclear Information System (INIS)

    In order to eliminate atmospheric tritium gas (HT) released from tritium handling apparatus, we proposed to use the HT oxidizing ability (hydrogenase enzyme) of bacterial strains isolated from surface soils instead of a high temperature precious metal catalyst. Among the isolated strains with high HT oxidation activity, several strains were selected to develop a tritium elimination (detritiation) system. Bioreactors were made of bacterial cells grown on agar medium on a cartridge filter and stored in a refrigerator until use. The detritiation ability of these bioreactors at room temperature was investigated during the intentional HT release experiments carried out in the Cassion Assembly for Tritium Safety Study (CATS) in TPL/JAERI. When HT contaminated air from the CATS was introduced into the biological detritiation system, in which three bioreactors were connected in series, 86% of HT in air was removed as tritiated water in these bioreactors at a flow rate of 100 cm3/min for 2 hours

  5. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  6. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    International Nuclear Information System (INIS)

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option

  7. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    The concept of the breeder reactor is almost as old as the idea of the nuclear reactor itself. From the very first years following the discovery of nuclear fission, scientists and technicians tried to turn mankind's eternal dream into reality; that is, enjoy an abundant source of energy without using up our raw material reserves. Nuclear energy offered several solutions to realize this dream. One of them, fusion, seemed out of our grasp in the near future. But fission of 235U was possible, and the Manhattan Project soon furnished ample proof of this theory. However, everyone working in this field was conscious of the fact that thermal neutron reactors make very inefficient use of the energy potential contained in natural uranium. The solution was to use in a core sufficiently rich in fissile matter, the excess neutrons to convert the 238U, so poorly used by other types of reactors, into fissile 239Pu. Regeneration, or 'breeding' of fuel, can multiply the energy drawn from a ton of uranium by a factor of 50 to 100. This would enable us to ward off the specter of an energy shortage and the rapid depletion of uranium mines. As early as 1945 in Los Alamos, Enrico Fermi stated: 'The country which first develops a breeder reactor will have a great competitive edge in atomic energy.' The development of the breeder reactor in the USA and around the world is discussed

  8. Process for conditioning tritium for final storage

    International Nuclear Information System (INIS)

    The process for conditioning tritium for final storage a) in which the tritium is introduced into a zeolite matrix, b) the zeolite matrix is dehydrated at temperatures above 4000C and c) the tritium in vapour form is brought into contact with the zeolite matrix is characterized by the fact that after saturation of the zeolite matrix with tritium, the tritium in the zeolite matrix is enclosed by means of microwave irradiation. (orig.)

  9. Tritium production, recovery and application in Korea.

    Science.gov (United States)

    Son, Soon-Hwan; Lee, Sook-Kyung; Kim, Kwang-Sin

    2009-01-01

    Four CANDU reactors have been operating at the site of Wolsong Nuclear Power Generation in Korea. The Wolsong tritium removal facility was constructed to reduce the tritium levels in heavy water systems. This facility was designed to process 100kg/h of tritiated heavy water feed and to produce 99% pure T(2). This recovered tritium will be made available for commercial applications. The initial phases on the tritium applications are made to establish the infrastructure and the tritium controls. PMID:19307127

  10. Toward the ultimate goal of tritium self-sufficiency: Technical issues and requirements imposed on ARIES advanced power plants

    International Nuclear Information System (INIS)

    Due to the lack of external tritium sources, all fusion power plants must demonstrate a closed tritium fuel cycle. The tritium breeding ratio (TBR) must exceed unity by a certain margin. The key question is: how large is this margin and how high should the calculated TBR be? The TBR requirement is design and breeder-dependent and evolves with time. At present, the ARIES requirement is 1.1 for the calculated overall TBR of LiPb systems. The Net TBR during plant operation could be around 1.01. The difference accounts for deficiencies in the design elements (nuclear data evaluation, neutronics code validation, and 3D modeling tools). Such a low Net TBR of 1.01 is potentially achievable in advanced designs employing advanced physics and technology. A dedicated R and D effort will reduce the difference between the calculated TBR and Net TBR. A generic breeding issue encountered in all fusion designs is whether any fusion design will over-breed or under-breed during plant operation. To achieve the required Net TBR with sufficient precision, an online control of tritium breeding is highly recommended for all fusion designs. This can easily be achieved for liquid breeders through online adjustment of Li enrichment.

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

    Science.gov (United States)

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

    2015-11-01

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

  12. Preparation and analysis of helium purge gas mixture to be used in tritium extraction system of LLCB TBM

    International Nuclear Information System (INIS)

    Hydrogen isotopes are extracted from the ceramic breeder and liquid breeder zones of Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) with Helium purge gas. 1000 ppm of hydrogen gas is mixed with the purge helium gas to facilitate improved extraction of hydrogen isotopes due to hydrogen swamping reactions. An experimental set-up is developed for making up the purge gas mixture with a composition similar to the purge gas composition provided at the inlet of the ceramic breeder zones and detritation column of LLCB TBM. This is achieved by introducing different ppm levels (500-5000 ppm) of hydrogen in helium gas by flow control mechanism. The analysis of the purge gas mixture is performed using a highly sensitive gas chromatograph system. In this work, parametric analysis is performed to optimize the process parameters viz., flow rates, temperatures etc. for achieving the required mixture of hydrogen and helium. This paper describes the detailed design of the experimental set-up along with parametric analysis results leading to optimized experimental conditions. (author)

  13. Current Design of the Flange Type Hydrogen Permeation Sensor in Liquid Breeder

    Energy Technology Data Exchange (ETDEWEB)

    Lee, E. H.; Jin, H. G.; Yoon, J. S.; Kim, S. K.; Lee, D. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, H. G. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In 2004, A. Ciampichetti et al. proposed a hollow capsule shape permeation sensor and they theoretically and experimentally evaluated the performance of the sensor made of Nb membrane at test condition of 500 .deg. C. However, the evaluation result showed the measured hydrogen permeation flux in the sensor much lower than the predicted one and they concluded that, the result is due to the formation of an oxide layer on the sensor membrane surface. Three years later, A. Ciampichetti et al. observed that a hollow capsule shape permeation sensor has too long response time to measure hydrogen concentration in liquid breeder. However, they suggested optimizing the sensor geometry with the reduction of the ratio 'total sensor volume/permeation surface' to overcome the low hydrogen permeating flux. For development of the liquid breeding technologies in nuclear fusion, the permeation sensor to measure tritium concentration in liquid metal breeder has been developed. Lee et al. proposed a flange type permeation sensor to dramatically reduce the ratio sensor 'inside volume/permeation surface' and to remove membrane welding during sensor manufacture process. However, the flange type sensor has problem with sealing. In present study, the modified flange sensor design with a metallic C-ring spring gasket is introduced. The modified sensor will be verified and evaluated under high temperature conditions by end of 2015.

  14. Analysis of the HCPB breeder blanket bock-up experiment for ITER using SUSD3D code

    International Nuclear Information System (INIS)

    In order to validate new nuclear cross-section evaluations, method development and design of the helium-cooled pebble bed (HCPB) test blanket module of ITER a benchmark experiment was performed this year at the Frascati Neutron Generator (FNG) in the scope of the EFF (European Fusion File) project in Europe. The objective of this experiment is to study the tritium breeding ratio and other nuclear quantities in a breeder blanket in order to establish and improve the quality of related JEFF nuclear data. The experiment consists of a metallic beryllium set-up with two double layers of breeder material (Li2CO3 powder). The reaction rate measurements include the Li2CO3 pellets (tritium breeding ratio), activation foils, and neutron and gamma spectrometers inserted at several axial and lateral locations in the block. Our task is to perform the deterministic transport, and cross section sensitivity and uncertainty analysis. The role of the cross-section sensitivity and uncertainty analysis is to optimise the design of the benchmark, and to assist in the interpretation of the measurement results. The paper presents the pre- and post- analysis of the benchmark experiment. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    P. Calderoni; P. Sharpe; M. Shimada

    2009-09-01

    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.

  16. Calculation of the possible tritium production in irradiation positions of the FRJ-2(DIDO) for fusion blanket experiments

    International Nuclear Information System (INIS)

    In the field of tritium and fusion blanket technology possibly an important and early contribution to the development of a fusion reactor blanket can be obtained by irradiation experiments at the research reactor FRJ-2 in Juelich, Federal Republic of Germany. However, the tritium production rate of 0.2 x 1013 to 2 x 1013 cm-3s-1 and the power per volume of 2 to 20 W cm-3 characteristic for a fusion reactor blanket have to be realized. The present report shows the reachable tritium values calculated for different irradiation positions in the FRJ-2 for natural lithium as a breeder material considering the actual existing neutron spectrum. Based on these results we come to the conclusion that the specified blanket data can actually be reached and adjusted. Therefore irradiation experiments at the FRJ-2 would be able to supply basical results for the fusion blanket development. (orig.)

  17. Proceedings of the Office of Fusion Energy/DOE workshop on ceramic matrix composites for structural applications in fusion reactors

    International Nuclear Information System (INIS)

    A workshop to assess the potential application of ceramic matrix composites (CMCs) for structural applications in fusion reactors was held on May 21--22, 1990, at University of California, Santa Barbara. Participants included individuals familiar with materials and design requirements in fusion reactors, ceramic composite processing and properties and radiation effects. The primary focus was to list the feasibility issues that might limit the application of these materials in fusion reactors. Clear advantages for the use of CMCs are high-temperature operation, which would allow a high-efficiency Rankine cycle, and low activation. Limitations to their use are material costs, fabrication complexity and costs, lack of familiarity with these materials in design, and the lack of data on radiation stability at relevant temperatures and fluences. Fusion-relevant feasibility issues identified at this workshop include: hermetic and vacuum properties related to effects of matrix porosity and matrix microcracking; chemical compatibility with coolant, tritium, and breeder and multiplier materials, radiation effects on compatibility; radiation stability and integrity; and ability to join CMCs in the shop and at the reactor site, radiation stability and integrity of joints. A summary of ongoing CMC radiation programs is also given. It was suggested that a true feasibility assessment of CMCs for fusion structural applications could not be completed without evaluation of a material ''tailored'' to fusion conditions or at least to radiation stability. It was suggested that a follow-up workshop be held to design a tailored composite after the results of CMC radiation studies are available and the critical feasibility issues are addressed

  18. Ceramic Methyltrioxorhenium

    CERN Document Server

    Herrmann, R; Eickerling, G; Helbig, C; Hauf, C; Miller, R; Mayr, F; Krug von Nidda, H A; Scheidt, E W; Scherer, W; Herrmann, Rudolf; Troester, Klaus; Eickerling, Georg; Helbig, Christian; Hauf, Christoph; Miller, Robert; Mayr, Franz; Nidda, Hans-Albrecht Krug von; Scheidt, Ernst-Wilhelm; Scherer, Wolfgang

    2006-01-01

    The metal oxide polymeric methyltrioxorhenium [(CH3)xReO3] is an unique epresentative of a layered inherent conducting organometallic polymer which adopts the structural motifs of classical perovskites in two dimensions (2D) in form of methyl-deficient, corner-sharing ReO5(CH3) octahedra. In order to improve the characteristics of polymeric methyltrioxorhenium with respect to its physical properties and potential usage as an inherentconducting polymer we tried to optimise the synthetic routes of polymeric modifications of 1 to obtain a sintered ceramic material, denoted ceramic MTO. Ceramic MTO formed in a solvent-free synthesis via auto-polymerisation and subsequent sintering processing displays clearly different mechanical and physical properties from polymeric MTO synthesised in aqueous solution. Ceramic MTO is shown to display activated Re-C and Re=O bonds relative to MTO. These electronic and structural characteristics of ceramic MTO are also reflected by a different chemical reactivity compared with its...

  19. Tritium transport around nuclear faciliteis

    International Nuclear Information System (INIS)

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

  20. High-concentration tritium sensor

    Energy Technology Data Exchange (ETDEWEB)

    Paglieri, S. N. (Stephen N.); Richmond, S. (Scott); Snow, R. C. (Ronny C.); Morris, J. S. (John S.); Tuggle, D. G. (Dale Glenn)

    2004-01-01

    A bi-layer device was fabricated and tested for the direct collection of electrons emitted by tritium beta decay. The sensor functions at high pressures and concentrations where previously no simple and cost effective direct measurement technique existed for tritium. A polished KOVAR{trademark} (Fe-Ni-Co alloy) rod was coated with a 1-{mu}m thick insulating layer of alumina using electron-beam evaporation, physical vapor deposition (PVD) of aluminum with oxygen dosing. The alumina deposition process was optimized to minimize pinholes and obtain a stable coating with high resistivity. The detector exhibited a nanoampere electrical response over a few decades of tritium concentration, up to pure tritium at 200 kPa. The sensor has been in service for several months now without showing signs of degradation and no discernible physical damage or change in efficiency or linearity has been observed.

  1. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

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

  2. Tritium transport around nuclear facilities

    International Nuclear Information System (INIS)

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

  3. High-concentration tritium sensor

    International Nuclear Information System (INIS)

    A bi-layer device was fabricated and tested for the direct collection of electrons emitted by tritium beta decay. The sensor functions at high pressures and concentrations where previously no simple and cost effective direct measurement technique existed for tritium. A polished KOVAR(trademark) (Fe-Ni-Co alloy) rod was coated with a 1-μm thick insulating layer of alumina using electron-beam evaporation, physical vapor deposition (PVD) of aluminum with oxygen dosing. The alumina deposition process was optimized to minimize pinholes and obtain a stable coating with high resistivity. The detector exhibited a nanoampere electrical response over a few decades of tritium concentration, up to pure tritium at 200 kPa. The sensor has been in service for several months now without showing signs of degradation and no discernible physical damage or change in efficiency or linearity has been observed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  5. Overview of tritium: characteristics, sources, and problems.

    Science.gov (United States)

    Okada, S; Momoshima, N

    1993-12-01

    Tritium has certain characteristics that present unique challenges for dosimetry and health-risk assessment. For example, in the gas form, tritium can diffuse through almost any container, including those made of steel, aluminum, and plastics. In the oxide form, tritium can generally not be detected by commonly used survey instruments. In the environment, tritium can be taken up by all hydrogen-containing molecules, distributing widely on a global scale. Tritium can be incorporated into humans through respiration, ingestion, and diffusion through skin. Its harmful effects are observed only when it is incorporated into the body. Several sources contribute to the inventory of tritium in our environment. These are 1) cosmic ray interaction with atmospheric molecules; 2) nuclear reactions in the earth's crust; 3) nuclear testing in the atmosphere during the 1950s and 1960s; 4) continuous release of tritium from nuclear power plants and tritium production facilities under normal operation; 5) incidental releases from these facilities; and 6) consumer products. An important future source will be nuclear fusion facilities expected to be developed for the purpose of electricity generation. The principal health physics problems associated with tritium are 1) the determination of the parameters for risk estimation with further reduction of their uncertainties (e.g., relative biological effectiveness and dose-rate dependency); 2) risk estimation from complex exposures to tritium in gas form, tritium in oxide form, tritium surface contamination, and other tritium-contaminated forms, with or without other ionizing radiations and/or nonionizing radiations; 3) the dose contributions of elemental tritium in the lung and from its oxidized tritium in the gastrointestinal tract; 4) prevention of tritium (in oxide form) intake and enhancement of tritium (oxide form) excretion from the human body; 5) precise health effects information for low-level tritium exposure; and 6) public

  6. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

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

  7. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

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

  8. Lithium loaded glass detectors for tritium breeding measurements

    International Nuclear Information System (INIS)

    A technique based on the response differences of enriched 6Li and 7Li glass scintillators has been developed for the real time measurement of the tritium production rate from the 6Li isotope in an experimental breeder blanket of a D-T fusion reactor. The effectiveness of the method is dependent on the degree to which the scintillation responses of two glasses can be matched for all interactions arising from the neutron and gamma fields within the assembly. However, the energy resolution of the scintillation pulses contains a component that is dependent on the composition of the glass which cannot be compensated for either electronically or analytically. The glasses used must therefore be carefully selected to ensure that the difference between their intrinsic scintillation resolutions is minimal. This paper describes measurements of the scintillation energy resolution for a range of commerically available glasses as a function of energy for electrons, protons, deuterons, and alpha particles. The results showed that systematic errors in the measurement of tritium breeding from 6Li, using the proposed technique, are minimized if the glasses of different lithium enrichment have the same chemical composition. Analysis of the results indicated that for optimum accuracy the total lithium content of the glasses should be small (about 2%). (orig.)

  9. What determines hatchling weight: breeder age or incubated egg weight?

    OpenAIRE

    AB Traldi; Menten JFM; CS Silva; PV Rizzo; PWZ Pereira; J Santarosa

    2011-01-01

    Two experiments were carried out to determine which factor influences weight at hatch of broiler chicks: breeder age or incubated egg weight. In Experiment 1, 2340 eggs produced by 29- and 55-week-old Ross® broiler breeders were incubated. The eggs selected for incubation weighed one standard deviation below and above average egg weight. In Experiment 2, 2160 eggs weighing 62 g produced by breeders of both ages were incubated. In both experiments, 50 additional eggs within the weight interval...

  10. Tritium concentrations in tree ring cellulose

    International Nuclear Information System (INIS)

    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.)

  11. Cold trapping of traces of tritiated water from the helium loops of a fusion breeder blanket

    International Nuclear Information System (INIS)

    The ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) will comprise three helium loops designed for: tritium extraction from the breeder zone, heat removal, and purification of the coolant. The process step envisaged for tritium extraction as well as for coolant purification includes a cryogenic cold trap as main component for the removal of tritiated water vapour (mainly HTO, H2O). The concentrations of water in the gas streams are expected to be extremely small, i.e. of the order of 10 ppm by volume. In this paper, we describe first runs with a cold trap using helium as the carrier gas at flow rates of 0.1 and 1.0 m3/h. The range of water vapour concentration in the helium carrier gas was 0.5 to >200 ppmv. The experiments have demonstrated the ability of the cold trap to remove water vapour efficiently from the He stream down to concentrations of less than 0.02 ppmv when the inlet water concentration is in the range of 300-650 ppmv or higher

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

    Energy Technology Data Exchange (ETDEWEB)

    Mills, Bernice E.

    2007-11-01

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

  13. Glass-ceramic joining and coating of SiC/SiC for fusion applications

    International Nuclear Information System (INIS)

    The aim of this work is the joining and the coating of SiC/SiC composites by a simple, pressureless, low cost technique. A calcia-alumina glass-ceramic was chosen as joining and coating material, because its thermal and thermomechanical properties can be tailored by changing the composition, it does not contain boron oxide (incompatible with fusion applications) and it has high characteristic temperatures (softening point at about 1400 C). Furthermore, the absence of silica makes this glass-ceramic compatible with ceramic breeder materials (i.e. lithium-silicates, -alluminates or -zirconates). Coatings and joints were successfully obtained with Hi-Nicalon fiber-reinforced CVI silicon carbide matrix composite. Mechanical shear strength tests were performed on joined samples and the compatibility with a ceramic breeder material was examined. (orig.)

  14. Tritium issues for realization of a DT fusion reactor

    International Nuclear Information System (INIS)

    A trend of studies of production and consumption of tritium is described. Realization of DT fusion reactor is discussed by tritium balance obtained from the above studies. It consists of introduction, tritium introduced into plasma vessel, tritium inventory in plasma vessel, tritium loss at fueling cycle system, tritium breeding and loss in blanket system, tritium balance in DT fusion reactor and summary. Investigation of development of external tritium resources has to be started. Tritium flow in DT fusion reactor, comparison of tritium inventory in fusion reactor, schematic diagram of tritium behavior in plasma vessel, change of overall burning efficiency and overall plasma generation rate, tritium inventory in re-deposition layer, effects of recovery efficiency of tritium from re-deposition layer, various breeding efficiencies in solid blanket, tritium flow in inertial confinement reactor with first wall, a tabular comparison of tritium balance calculation values, and comparison between tritium production methods are illustrated. (S.Y.)

  15. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  16. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  17. Large scale breeder reactor pump dynamic analyses

    International Nuclear Information System (INIS)

    The lateral natural frequency and vibration response analyses of the Large Scale Breeder Reactor (LSBR) primary pump were performed as part of the total dynamic analysis effort to obtain the fabrication release. The special features of pump modeling are outlined in this paper. The analysis clearly demonstrates the method of increasing the system natural frequency by reducing the generalized mass without significantly changing the generalized stiffness of the structure. Also, a method of computing the maximum relative and absolute steady state responses and associated phase angles at given locations is provided. This type of information is very helpful in generating response versus frequency and phase angle versus frequency plots

  18. Metabolic models for tritium dosimetry

    International Nuclear Information System (INIS)

    Tritium (3H or T) is the radioactive isotope of hydrogen, which is produced by both natural, and man made sources. Tritium has a small relative natural abundance compared to hydrogen and deuterium (D). As was assessed by the United Nations scientific committee the contribution of cosmogenic tritium to annual effective dose in human is very small, only about 0.01 μSv (UNSCEAR, 2000). In case of heavy water reactors annual tritium doses for critical groups are also very small but theoretically could reach values of several μSv. In this case professionally exposed personnel could be exposed to tritium doses of some mSv. According to these considerations environmental and dosimetric aspects of this radionuclide are of special concern for health physicists. Tritium dose assessment methodologies have special particularities because hydrogen is a chemical element with an important metabolic role in the human body. Operating experience to date of CANDU reactors has indicated that the major contributor to the internal dose of professionally exposed people is the tritiated heavy water (DTO). DTO, like the tritiated water HTO, is assumed to be uniformly mixed with body water pool and reaching equilibrium immediately after the intake. All the statements in this paper related to HTO dosimetry are also considered valid in case of DTO. The results of the computations performed with different retention functions corresponding to different compartment models are presented. The differences between the models due to OBT (Organically Bound Tritium) contribution are 5.6% for two- compartment model and 13% for three-compartment Dunford - Johnson model, respectively. In practice the contribution of OBT is considered to be about 10%. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  20. Engineering ceramics

    CERN Document Server

    Bengisu, Murat

    2001-01-01

    This is a comprehensive book applying especially to junior and senior engineering students pursuing Materials Science/ Engineering, Ceramic Engineering and Mechanical Engineering degrees. It is also a reference book for other disciplines such as Chemical Engineering, Biomedical Engineering, Nuclear Engineering and Environmental Engineering. Important properties of most engineering ceramics are given in detailed tables. Many current and possible applications of engineering ceramics are described, which can be used as a guide for materials selection and for potential future research. While covering all relevant information regarding raw materials, processing properties, characterization and applications of engineering ceramics, the book also summarizes most recent innovations and developments in this field as a result of extensive literature search.

  1. Ceramic glossary

    International Nuclear Information System (INIS)

    This book is a 2nd edition that contains new terms reflecting advances in high technology applications of ceramic materials. Definitions for terms which materials scientists, engineers, and technicians need to know are included

  2. Compatibility problems with beryllium in ceramic blankets

    International Nuclear Information System (INIS)

    Compatibility of beryllium with structural materials (316L austenitic steel and 1.4914 martensitic steel) and with tritium breeding ceramics (lithium aluminate or silicate) has been studied in contact tests between 550 C and 700 C and for durations reaching 3000 hours. Beryllium-ceramic interaction is negligeable in all the temperature range with aluminate and up to 600 C with silicates. On the other hand, noticeable interaction is observed between beryllium and 316L steel at 580 C and above. Beryllium interaction with 1.4914 steel is visible only at 650 C and above and its amplitude is lower than 316L steel one. In these two cases, the superficial layer is brittle, and adherent to the steel. Comparison between beryllium - 0.4 wt% calcium alloy and beryllium at 700 C shows that interaction with steels or ceramics is qualitatively the same but slightly weaker. (author). 6 refs.; 6 figs.; 3 tabs

  3. The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO2, ThO2, ThO2 -- UO2, ZrO2 -- UO2) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO2, ZrO2 -- UO2 fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO2 and ThO2 -- UO2 fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system

  4. Tailored ceramics

    International Nuclear Information System (INIS)

    In polyphase tailored ceramic forms two distinct modes of radionuclide immobilization occur. At high waste loadings the radionuclides are distributed through most of the ceramic phases in dilute solid solution, as indicated schematically in this paper. However, in the case of low waste loadings, or a high loading of a waste with low radionuclide content, the ceramic can be designed with only selected phases containing the radionuclides. The remaining material forms nonradioactive phases which provide a degree of physical microstructural isolation. The research and development work with polyphase ceramic nuclear waste forms over the past ten years is discussed. It has demonstrated the critical attributes which suggest them as a waste form for future HLW disposal. From a safety standpoint, the crystalline phases in the ceramic waste forms offer the potential for demonstrable chemical durability in immobilizing the long-lived radionuclides in a geologic environment. With continued experimental research on pure phases, analysis of mineral analogue behavior in geochemical environments, and the study of radiation effects, realistic predictive models for waste form behavior over geologic time scales are feasible. The ceramic forms extend the degree of freedom for the economic optimization of the waste disposal system

  5. Distillation and measurement of two forms of tritium

    International Nuclear Information System (INIS)

    Two forms of tritium of HTO, free water tritium and bound tritium, exist in the environment. A method was introduced to acquire free water tritium by 8 hour's distilling at 130 degree C and collect bound tritium by oxidative 4 minute's burning at 850 degree C. Methods of measuring two forms of tritium by liquid scintillometer were also discussed

  6. Fast breeder reactor fuel reprocessing in France

    International Nuclear Information System (INIS)

    Simultaneous with the effort on fast breeder reactors launched several years ago in France, equivalent investigations have been conducted on the fuel cycle, and in particular on reprocessing, which is an indispensable operation for this reactor. The Rapsodie experimental reactor was associated with the La Hague reprocessing plant AT1 (1 kg/day), which has reprocessed about one ton of fuel. The fuel from the Phenix demonstration reactor is reprocessed partly at the La Hague UP2 plant and partly at the Marcoule pilot facility, undergoing transformation to reprocess all the fuel (TOR project, 5 t/y). The fuel from the Creys Malville prototype power plant will be reprocessed in a specific plant, which is in the design stage. The preliminary project, named MAR 600 (50 t/y), will mobilize a growing share of the CEA's R and D resources, as the engineering needs of the UP3 ''light water'' plant begins to decline. Nearly 20 tonnes of heavy metals irradiated in fast breeder reactors have been processed in France, 17 of which came from Phenix. The plutonium recovered during this reprocessing allowed the power plant cycle to be closed. This power plant now contains approximately 140 fuel asemblies made up with recycled plutonium, that is, more than 75% of the fuel assemblies in the Phenix core

  7. Prototype fast breeder reactor main options

    International Nuclear Information System (INIS)

    Fast reactor programme gets importance in the Indian energy market because of continuous growing demand of electricity and resources limited to only coal and FBR. India started its fast reactor programme with the construction of 40 MWt Fast Breeder Test Reactor (FBTR). The reactor attained its first criticality in October 1985. The reactor power will be raised to 40 MWt in near future. As a logical follow-up of FBTR, it was decided to build a prototype fast breeder reactor, PFBR. Considering significant effects of capital cost and construction period on economy, systematic efforts are made to reduce the same. The number of primary and secondary sodium loops and components have been reduced. Sodium coolant, pool type concept, oxide fuel, 20% CW D9, SS 316 LN and modified 9Cr-1Mo steel (T91) materials have been selected for PFBR. Based on the operating experience, the integrity of the high temperature components including fuel and cost optimization aspects, the plant temperatures are recommended. Steam temperature of 763 K at 16.6 MPa and a single TG of 500 MWe gross output have been decided. PFBR will be located at Kalpakkam site on the coast of Bay of Bengal. The plant life is designed for 30 y and 75% load factor. In this paper the justifications for the main options chosen are given in brief. (author). 2 figs, 2 tabs

  8. Industrial solar breeder project using concentrator photovoltaics

    Energy Technology Data Exchange (ETDEWEB)

    Hamilton, R; Wohlgemuth, J; Burkholder, J; Levine, A; Storti, G; Wrigley, C; McKegg, A

    1979-08-01

    The purpose of this program is to demonstrate the use of a concentrating photovoltaic system to provide the energy for operating a silicon solar cell production facility, i.e., to demonstrate a solar breeder. Solarex has proposed to conduct the first real test of the solar breeder concept by building and operating a 200 kW(e) (peak) concentrating photovoltaic system based on the prototype and system design developed during Phase I. This system will provide all of the electrical and thermal energy required to operate a solar cell production line. This demonstration would be conducted at the Solarex Rockville facility, with the photovoltaic array located over the company parking lot and on an otherwise unusable flood plain. Phase I of this program included a comprehensive analysis of the application, prototype fabrication and evaluation, system design and specification, and a detailed plan for Phases II and III. A number of prototype tracking concentrator solar collectors were constructed and operated. Extensive system analysis was performed to design the Phase II system as a stand-alone power supply for a solar cell production line. Finally, a detailed system fabrication proposal for Phase II and an operation and evaluation plan for Phase III were completed. These proposals included technical, management, and cost plans for the fabrication and exercise of the proposed system.

  9. Coincidence measurements of FFTF breeder fuel subassemblies

    International Nuclear Information System (INIS)

    A prototype coincidence counter developed to assay fast breeder reactor fuel was used to measure four fast-flux test facility subassemblies at the Hanford Engineering Development Laboratory in Richland, Washington. Plutonium contents in the four subassemblies ranged between 7.4 and 9.7 kg with corresponding 240Pu-effective contents between 0.9 and 1.2 kg. Large count rates were observed from the measurements, and plots of the data showed significant multiplication in the fuel. The measured data were corrected for deadtime and multiplication effects using established formulas. These corrections require accurate knowledge of the plutonium isotopics and 241Am content in the fuel. Multiplication-corrected coincidence count rates agreed with the expected count rates based on spontaneous fission-neutron emission rates. These measurements indicate that breeder fuel subassemblies with 240Pu-effective contents up to 1.2 kg can be nondestructively assayed using the shift-register electronics with the prototype counters. Measurements using the standard Los Alamos National Laboratory shift-register coincidence electronics unit can produce an assay value accurate to +-1% in 1000 s. The uncertainty results from counting statistics and deadtime-correction errors. 3 references, 8 figures, 8 tables

  10. Improved structural materials for fast breeder reactors

    International Nuclear Information System (INIS)

    Electricity plays a crucial role in the economic development of our country. Coal is the primary fuel for generation of electricity in India as in many other countries. In India, generation of power by nuclear reactors is very important because of (i) availability of large thorium resource, (ii) constraints on setting up of fossil fuel based power plants and (iii) the negligibly small green house gas emissions by nuclear energy. The nuclear programme of the country is being implemented in three stages: (i) pressurized heavy water reactors of the CANDU type, (ii) sodium-cooled fast reactors and (iii) thorium-based reactors. Sodium-cooled fast reactor (SFR) technology is envisioned to make use of the large thorium reserves available. India has undertaken and made rapid strides in developing SFR technology and building of fast reactors for energy generation. A Fast Breeder Test Reactor (FBTR) of 40 MWt is operating successfully for over 25 years at Indira Gandhi Centre for Atomic Research. Based on the design, construction and operational experience, a 500 MWe Prototype Fast Breeder Reactor (PFBR) has been designed indigenously and is in an advanced stage of construction. Its design is being further optimised for enhanced economy with respect to cost of electricity production, for use in commercial reactors. Currently, several R and D programmes are under implementation for the development of new materials required for improved economy of commercial fast reactors

  11. The fast breeder reactor. v. 1

    International Nuclear Information System (INIS)

    The Energy Committee's report was prepared after hearing evidence (the minutes of which are published in Volume II) from the Central Electricity Generating Board, the United Kingdom Atomic Energy Authority and the Department of Energy. Memoranda received from other interested bodies or individuals were also considered and members of the Committee visited fast breeder projects in France, West Germany and Japan. As well as the development of the fast reactors, the economics and timescale were reviewed. The particular case of the fast breeder reactor and proposed fuel reprocessing plant at Dounreay was considered. The main conclusion is that major expenditure on fast reactor programmes can only be justified if there is a potential economic case, i.e. if the fuel cycle costs are lower than for PWRs. This would only be the case if uranium costs increased greatly. It is not considered worthwhile to participate in the European Fast Reactor although this should be reviewed in 1993 and 1997. The Committee agree with the Government's decision to cease funding the PFR in 1994 and endorses the need to regenerate the local economy which will be affected by this decision. (UK)

  12. Measurements after irradiation of helium and tritium residual puantities in lithium aluminate (γLiAlO2)

    International Nuclear Information System (INIS)

    Heating between 930OC and 1000OC, under vacuum or in a sweeping gas, of irradiated γLiAlO2 samples allows extraction of the whole residual tritium and helium. The chemical species released by irradiated γAlLiO2 is principally tritiated water. Nevertheless, if ceramics are placed in stainless steel, the residual tritium fraction in tritiated water form decreases when irradiation temperature increases between 250OC and 5200C. The more important factors in tritium retention are irradiation temperature and nature of the can material. Residual tritium amount decreases when irradiation temperature increases between 250OC and 520OC. Beyond 500OC, surface desorption phenomena are important. Retention of helium 4 is weak and not much dependent of irradiation temperature. Residual helium 4 fraction is weaker in big grain samples than in small grain samples. At 975OC, helium 4 release is controlled by diffusion phenomena

  13. Tritium in the environment. Knowledge synthesis

    International Nuclear Information System (INIS)

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

  14. Recommended radiological controls for tritium operations

    International Nuclear Information System (INIS)

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

  15. Activation Calculation for a Fusion Experimental Breeder FEB-E

    Institute of Scientific and Technical Information of China (English)

    FENGKaiming

    2002-01-01

    A fusion breeder might be an essential intermediate application of fusion energy at earlier term, since it has the potential to provide plenty of commercial fissile fuel. Based on fusion physics and technologies available at present and in the near future, the realistic fusion experimental breeder, FEB-E was designed.

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

    International Nuclear Information System (INIS)

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

  17. Linear accelerator for tritium production

    International Nuclear Information System (INIS)

    For many years now, Los Alamos National Laboratory has been working to develop a conceptual design of a facility for accelerator production of tritium (APT). The APT accelerator will produce high energy protons which will bombard a heavy metal target, resulting in the production of large numbers of spallation neutrons. These neutrons will be captured by a low-Z target to produce tritium. This paper describes the latest design of a room-temperature, 1.0 GeV, 100 mA, cw proton accelerator for tritium production. The potential advantages of using superconducting cavities in the high-energy section of the linac are also discussed and a comparison is made with the baseline room-temperature accelerator. copyright 1996 American Institute of Physics

  18. Linear accelerator for tritium production

    International Nuclear Information System (INIS)

    For many years now, Los Alamos National Laboratory has been working to develop a conceptual design of a facility for accelerator production of tritium (API). The APT accelerator will produce high energy protons which will bombard a heavy metal target, resulting in the production of large numbers of spallation neutrons. These neutrons will be captured by a low-Z target to produce tritium. This paper describes the latest design of a room-temperature, 1.0 GeV, 100 mA, cw proton accelerator for tritium production. The potential advantages of using superconducting cavities in the high-energy section of the linac are also discussed and a comparison is made with the baseline room-temperature accelerator

  19. Tritium processing using metal hydrides

    International Nuclear Information System (INIS)

    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

  20. 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. (...

  1. Modeling Tritium Life cycle in Nuclear Plants

    International Nuclear Information System (INIS)

    The mathematical development of a tritium model for nuclear power plants is presented. The model requires that the water and tritium material balance be satisfied throughout normal operations and shutdown. The model results obtained at the time of publishing include the system definitions and comparison of the model predictions of tritium generations compared to the observed plant data of the Braidwood station. A scenario that models using ion exchange resin to remove coolant boron demonstrates the tritium concentration levels are manageable. (authors)

  2. Conceptual Design of Tritium Extraction System

    OpenAIRE

    Miral Thakker, Prof.Amar vaghela

    2012-01-01

    The first generation of fusion reactors will use deuterium and tritium as fuel. Since tritium is not available in nature, it must be produced in the fusion reactor blanket which surrounds the plasma zone. Tritium extraction facility has been designed and fabricated. Calibration procedure has been performed to determine tritium losses, if any during the extraction. Lithium compounds were irradiated using Am-Be neutron source. Out of pile extraction from neutron irradiated lithium compounds was...

  3. Preliminary design of test facilities for tritium breeding blanket development, (1)

    International Nuclear Information System (INIS)

    This report describes the results of the preliminary design of outpile test facilities which are used for development of tritium breeding blanket with ceramic breeding material. The facilities which were designed are as follows; High heat flux test facility, Thermal-hydraulic test facility, Integrity test facility, Fabrication Technology Development Facility. This design study was performed by Kawasaki Heavy Industries, Ltd. under the contract to Fusion Research System Laboratory. (author)

  4. Toxicity and dosimetry of tritium

    International Nuclear Information System (INIS)

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

  5. Tritium oxidation and exchange: preliminary studies

    International Nuclear Information System (INIS)

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

  6. Tritium oxidation and exchange: preliminary studies

    Energy Technology Data Exchange (ETDEWEB)

    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/sup 4/ to 10/sup 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/sup -4/ to 10/sup -1/ mCi/ml) indicate the conversion may proceed more rapidly than would be expected from an extrapolation of previous data taken at higher concentrations (10/sup -1/ to 10/sup 2/ mCi/ml).

  7. 10 CFR 30.55 - Tritium reports.

    Science.gov (United States)

    2010-01-01

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

  8. Tritium turnover in succulent plants

    International Nuclear Information System (INIS)

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

  9. TRICICLO/PB. A computational tool modelling dynamic tritium transfers at HCPB demo blankets systems

    International Nuclear Information System (INIS)

    The design of the cycle and the control of tritium of DEMO breeding blankets (Inner Breeding Tritium Cycle, IBTC) represent a complex and ambitious technological objective of Fusion Nuclear Technology. The IBTC design is today conceptually open to the selection and scale demonstration of tritium processing technologies and to the choice of system design operational modes and parameters. Numerical tools modelling dynamic tritium transfers between IBTC systems based on Flow Process Diagram lay-outs support IBTC developments in many aspects serving to: (1) establish hierarchy for the IBTC design constraints and criteria, (2) to quantify on-diagram system processing technologies, (3) to fix underlying physics needed to express dynamic flux and inventories between systems, and finally (4) to make global parametric tuning and optimization of cycle parameters possible. Among the available options, the Rankine cycle is the most conservative solution for the Power Conversion Cycle in terms of technological maturity and tritium control requirements. Optimization of Gas Cooled-High Temperature Reactor and design adaptation to DEMO primary coolant (PC) [300/500 C, 80bar] permit one to assess the two general diverse coolant chemistry options (HT oxidation or H2 isotopic swamping). Both options are discussed in terms of tritium control, and internal and external IBTC processing requirements for HCPB/DEMO. Permeation from the breeding ceramic into the He primary coolant and extraction of tritium by purge gas act as given inputs for the IBTC concept. Dynamic tritium transfer and radial breeding sources are inputs for actual assessments based on 2D moving-slab numerical techniques. Ultimate tritium processing technologies performance (CPS: Coolant Purification System, TES: Tritium Extraction System from purging lines) acts as boundary IBTC design constraints. Actual limits for transient modes are discussed. The IBTC design variables concern: (i) CPS system disposition in the IBTC

  10. Tritium processing in JT-60U

    International Nuclear Information System (INIS)

    Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7 x 1019 D-D fusion neutrons were produced by D (d, p) T reactions in plasma, which are expected to produce ∼31 GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium Distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ∼ 7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm3 for radiation work permit requirements. (author)

  11. Tritium monitoring in environment at ICIT Tritium Separation Facility

    International Nuclear Information System (INIS)

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

  12. Water chemistry of breeder reactor steam generators

    International Nuclear Information System (INIS)

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed

  13. Safeguards in Prototype Fast Breeder Reactor Monju

    International Nuclear Information System (INIS)

    The assemblies loaded in the core and stored in the ex-vessel storage tank (EVST) are in liquid sodium in the Japanese prototype fast breeder reactor (FBR) Monju. Since it is difficult to apply a direct verification procedure for the fuel assemblies in these areas, a dual containment and surveillance system consisting of two monitoring devices such as surveillance camera and radiation monitor that are functionally independent has been applied. In addition, the Monju Remote Monitoring System was developed to strengthen the continuous surveillance and to reduce the load of the inspection activities. Furthermore, the ex-vessel transfer machine radiation monitor (EVRM) and the exit gate monitor (EXGM) were upgraded to strengthen the monitoring of spent blanket fuel assemblies and to improve the reliability of distinguishing between fuel assemblies and non-fuel items. As the result, the integrated safeguards was introduced in November 2009, and the effective safeguards activities have been implemented in Monju. (author)

  14. The accelerator production of tritium - An overview

    International Nuclear Information System (INIS)

    A reliable supply of tritium is necessary to maintain the U.S. nuclear defense capability. Because tritium decays to 3He at the rate of 5.5% per year, it must be continually replenished. Since the shutdown of the last production reactor in 1988, tritium requirements have been met through reuse of tritium recovered from dismantled nuclear weapons. This is insufficient for future needs, requiring the U.S. Department of Energy to bring a new tritium production capability on-line by 2007

  15. Computer simulation of tritium removal facility design

    International Nuclear Information System (INIS)

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

  16. Management of tritium at nuclear facilities

    International Nuclear Information System (INIS)

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

  17. Tritium hazard via the ingestion pathway

    International Nuclear Information System (INIS)

    The classic methodology for estimating dose to man from environmental tritium ignores the fact that organically bound tritium in foodstuffs may be directly assimilated in the bound compartment of tissues without previous oxidation. We propose a four-compartment model that allows for the ability to input organically bound tritium in foodstuffs directly into the organic compartments of the model. We found that organically bound tritium in foodstuffs can increase the total body dose by a factor of 1.7 to 4.5 times the free body water dose alone, depending on the bound to loose ratio of tritium in the diet. 10 refs., 1 fig., 1 tab

  18. Sorption of tritium by modified natural aluminosilicates

    International Nuclear Information System (INIS)

    Tritium sorption was studied using natural clays, as well as organic compounds intercalated into alumosilicate matrices, i.e. kaolinite and montmorillonite, to consider interaction of tritium-containing underground water with argillaceous geological barriers and potentiality of the water decontamination in terms of tritium. It was ascertained that montmorillonite samples modified by dodecylpyridinium bromide, dimethylsulfoxide and hydrazine-sulfate, as well as kaolinite samples modified by hydrazine-sulfate, feature the highest sorption capacity towards tritium. A mechanism of tritium sorption is suggested, which consists in its isotopic exchange for hydrogen atoms in organic and inorganic compounds containing unshared electron pairs

  19. EFFECTS OF TRITIUM GAS EXPOSURE ON POLYMERS

    Energy Technology Data Exchange (ETDEWEB)

    Clark, E.; Fox, E.; Kane, M.; Staack, G.

    2011-01-07

    Effects of tritium gas exposure on various polymers have been studied over the last several years. Despite the deleterious effects of beta exposure on many material properties, structural polymers continued to be used in tritium systems. Improved understanding of the tritium effects will allow more resistant materials to be selected. Currently polymers find use mainly in tritium gas sealing applications (eg. valve stem tips, O-rings). Future uses being evaluated including polymeric based cracking of tritiated water, and polymer-based sensors of tritium.

  20. Total tritium measurement in atmosphere

    International Nuclear Information System (INIS)

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

  1. 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.

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

    International Nuclear Information System (INIS)

    JAERI designed, fabricated, and installed the JAERI Fuel Cleanup System (J-FCU) as a subsystem of simulated fusion fuel loop at the TSTA. The main function of the J-FCU is to purify and to recover hydrogen isotopes from simulated plasma exhaust while exhausting tritium free impurities. J-FCU has been in tritium test since March, 1991. Ceramic electrolysis cell (CEC) was replaced with its spare on January 1992. The stand-alone tritium test was performed with full impurities (N2, CQ4 and Q2O etc.) on February, 1992. Main purpose of this test was to evaluate the J-FCU total integrity and function with full impurities after replacing CEC. During this test, plugging of Cold Trap (CT) occurred twice and about 500 Ci of tritium exhausted to the Tritium Waste Treatment system (TWT). The safety interlock of the J-FCU acted well, so operation was carried out safely. This report describes the detail results of the above test and discuss its functions and difficulties. (author)

  3. Tritium systems test assembly stabilization

    International Nuclear Information System (INIS)

    The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium technology Research and Development (R and D) primarily for future fusion power reactors. The facility was conceived in mid 1970's, operations commenced in early 1980's, stabilization and deactivation began in 2000 and were completed in 2003. The facility will remain in a Surveillance and Maintenance (S and M) mode until the Department of Energy (DOE) funds demolition of the facility, tentatively in 2009. A safe and stable end state was achieved by the TSTA Facility Stabilization Project (TFSP) in anticipation of long term S and M. At the start of the stabilization project, with an inventory of approximately 140 grams of tritium, the facility was designated a Hazard Category (HC) 2 Non-Reactor Nuclear facility as defined by US Department of Energy standard DOE-STD-1027-92 (1997). The TSTA facility comprises a laboratory area, supporting rooms, offices and associated laboratory space that included more than 20 major tritium handling systems. The project's focus was to reduce the tritium inventory by removing bulk tritium, tritiated water wastes, and tritium-contaminated high-inventory components. Any equipment that remained in the facility was stabilized in place. All of the gloveboxes and piping were rendered inoperative and vented to atmosphere. All equipment, and inventoried tritium contamination, remaining in the facility was left in a safe-and-stable state. The project used the End Points process as defined by the DOE Office of Environmental Management (web page http://www.em.doe.- gov/deact/epman.htmtlo) document and define the end state required for the stabilization of TSTA Facility. The End Points process added structure that was beneficial through virtually all phases of the project. At completion of the facility stabilization project the residual tritium inventory was approximately 3,000 curies, considerably less than the 1.6-gram threshold for a HC 3 facility. TSTA is now

  4. Tritium pellet injection sequences for TFTR

    International Nuclear Information System (INIS)

    Tritium pellet injection into neutral deuterium, beam heated deuterium plasmas in the Tokamak Fusion Test Reactor (TFTR) is shown to be an attractive means of (1) minimizing tritium use per tritium discharge and over a sequence of tritium discharges; (2) greatly reducing the tritium load in the walls, limiters, getters, and cryopanels; (3) maintaining or improving instantaneous neutron production (Q); (4) reducing or eliminating deuterium-tritium (D-T) neutron production in nonoptimized discharges; and (5) generally adding flexibility to the experimental sequences leading to optimal Q operation. Transport analyses of both compression and full-bore TFTR plasmas are used to support the above observations and to provide the basis for a proposed eight-pellet gas gun injector for the 1986 tritium experiments

  5. Fusion tritium program in the United States

    International Nuclear Information System (INIS)

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

  6. NDT and inspection of tritium removal facility

    International Nuclear Information System (INIS)

    CANDU heavy water reactors produce tritium in the moderator and coolant circuits through neutron absorption by the deuterium atoms in heavy water. The concentration of tritium, in the form of DTO molecules builds up slowly with time of reactor operation. A typical yearly production rate of tritium is 2400 curie for each megawatt of electricity produced and as a consequence, a 600 megawatt Candu reactor produces 1.4 million curie of tritium per year. Tritium decays to 3He, a non radioactive species, and has a half life of approximately 12 years. Both Ontario Hydro and AECL are constructing plants to remove tritium from heavy water to maintain the tritium concentration below the equilibrium value. This will result in lower radiation doses to operating personnel and reduce the level of radiation in any releases of heavy water to the environment

  7. Radiobiological characteristic of tritium-labelled lysine

    International Nuclear Information System (INIS)

    Experiments on mice and rats injected with tritium-labeled lysine have revealed that one day after injection about 80% of the label was retained in organs and tissues as tissue-bound tritium. Retention curves for tritium in the body were decomposed into two exponentials. The biological half-lives of tritium-labeled lysine in various tissues exceed half-lives of other tritiated amino acids and of triated water. The average dose in different tissues following injection of tritiated lysine exceeds that from equal of tritium oxide (THO) by 1.5-8 times. Contribution of the tissue-bound tritium in dose is about 90%. radiobiological experiments showed strong genetic and citotoxic effects in male mice after injection of tritium-labeled lysine

  8. A comparison of fusion breeder/fission client and fission breeder/fission client systems for electrical energy production

    International Nuclear Information System (INIS)

    A parametric study that evaluated the economic performance of breeder/client systems is described. The linkage of the breeders to the clients was modelled using the stockpile approach to determine the system doubling time. Since the actual capital costs of the breeders are uncertain, a precise prediction of the cost of a breeder was not attempted. Instead, the breakeven capital cost of a breeder relative to the capital cost of a client reactor was established by equating the cost of electricity from the breeder/client system to the cost of a system consisting of clients alone. Specific results are presented for two breeder/client systems. The first consisted of an LMFBR with LWR clients. The second consisted of a DT fusion reactor (with a 238U fission suppressed blanket) with LWR clients. The economics of each system was studied as a function of the cost of fissile fuel from a conventional source. Generally, the LMFBR/LWR system achieved relatively small breakeven capital cost ratios; the maximum ratio computed was 2.2 (achieved at approximately triple current conventional fissile material cost). The DTFR/LWR system attained a maximum breakeven capital cost ratio of 4.5 (achieved at the highest plasma quality (ignited device) and triple conventional fissile cost)

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

    International Nuclear Information System (INIS)

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

  10. Analysis on tritium permeation in tritium storage bed with gas flowing calorimetry

    International Nuclear Information System (INIS)

    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)

  11. Organically bound tritium level in vegetation at ICIT tritium removal facility

    International Nuclear Information System (INIS)

    In order to evaluate the impact of tritium on wild vegetation around ICIT we have monitored the tritium concentrations in precipitation, air, soil, grass and green wheat from a specific area near the tritium removal facility during vegetation periods in 2012-2013. The tissue free water tritium concentration showed the influence of tritium level in precipitation, with higher values during the summer (around 2 Bq l-1) and lower values during the autumn (around 1.3 Bq l-1). The same behavior was observed for the total organically bound tritium level. (author)

  12. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    International Nuclear Information System (INIS)

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter

  13. In-situ tritium borehole probe for measurement of tritium

    International Nuclear Information System (INIS)

    An apparatus for measuring the in situ levels of tritium in ground water at depth in the earth. A tritium analyzer is made to fit in a sonde or probe which is placed in a borehole. This analyzer can perform a programmed cycle and has a sample intake to allow ambient water to enter; a reaction chamber; a drying chamber; an ion chamber; a cryogenic gas pump, and a spent capsule collection chamber. After the water sample is brought into the unit, it rises into the reaction chamber where it reacts with a preweighed quantity of calcium carbide in a capsule to yield acetylene. Next the acetylene vapor passes through the drying chamber to remove excess water and then flows into the evacuated ion chamber. Following this, the ion chamber is sealed off and a count of tritium beta decay events is started. Following the completion of the count, a valve is opened to remove the acetylene from the ion chamber with the cryogenic gas pump. The spent capsule containing the residue from the reaction is ejected into a collection chamber. Last, the holder for the preweighed calcium carbide capsule is refilled from a stock of such capsules in preparation for a new measurement cycle

  14. The Chalk River Tritium Extraction Plant

    International Nuclear Information System (INIS)

    The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1-35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T2. The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: it produces pure tritium for use in fusion research and development, and it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors (author)

  15. Titanium for long-term tritium storage

    Energy Technology Data Exchange (ETDEWEB)

    Heung, L.K.

    1994-12-01

    Due to the reduction of nuclear weapon stockpile, there will be an excess of tritium returned from the field. The excess tritium needs to be stored for future use, which might be several years away. A safe and cost effective means for long term storage of tritium is needed. Storing tritium in a solid metal tritide is preferred to storing tritium as a gas, because a metal tritide can store tritium in a compact form and the stored tritium will not be released until heat is applied to increase its temperature to several hundred degrees centigrade. Storing tritium as a tritide is safer and more cost effective than as a gas. Several candidate metal hydride materials have been evaluated for long term tritium storage. They include uranium, La-Ni-Al alloys, zirconium and titanium. The criteria used include material cost, radioactivity, stability to air, storage capacity, storage pressure, loading and unloading conditions, and helium retention. Titanium has the best combination of properties and is recommended for long term tritium storage.

  16. Exploring new coolants for nuclear breeder reactors

    International Nuclear Information System (INIS)

    Breeder reactors are considered a unique tool for fully exploiting natural nuclear resources. In current Light Water Reactors (LWR), only 0.5% of the primary energy contained in the nuclei removed from a mine is converted into useful heat. The rest remains in the depleted uranium or spent fuel. The need to improve resource-efficiency has stimulated interest in Fast-Reactor-based fuel cycles, which can exploit a much higher fraction of the energy content of mined uranium by burning U-238, mainly after conversion into Pu-239. Thorium fuel cycles also offer several potential advantages over a uranium fuel cycle. The coolant initially selected for most of the FBR programs launched in the 1960s was sodium, which is still considered the best candidate for these reactors. However, Na-cooled FBRs have a positive void reactivity coefficient. Among other factors, this fundamental drawback has resulted in the canceled deployment of these reactors. Therefore, it seems reasonable to explore new options for breeder coolants. In this paper, a proposal is presented for a new molten salt (F2Be) coolant that could overcome the safety issues related to the positive void reactivity coefficient of molten metal coolants. Although it is a very innovative proposal that would require an extensive R and D program, this paper presents the very appealing properties of this salt when using a specific type of fuel that is similar to that of pebble bed reactors. The F2Be concept was studied over a typical MOX composition and extended to a thorium-based cycle. The general analysis took into account the requirements for criticality (opening the option of hybrid subcritical systems); the requirements for breeding; and the safety requirement of having a negative coolant void reactivity coefficient. A design window was found in the definition of a F2Be cooled reactor where the safety requirement was met, unlike for molten metal-cooled reactors, which always have positive void reactivity coefficients

  17. Fast-breeder-power reactor records in the INIS database

    International Nuclear Information System (INIS)

    This report presents a statistical analysis of more than 19,700 records of publications concerned with research and technology in the field of fast breeder power fission reactors which are included in the INIS Bibliographic Database for the period from 1970. to 1999. The main objectives of this bibliometric study were: to make an inventory of the fast breeder power reactor related records in the INIS Database; to provide statistics and scientific indicators for the INIS users, namely science managers, researchers, engineers, operators, scientific editors and publishers, decision-makers in the field of fast breeder power reactors related subjects; to extract other useful information from the INIS Bibliographic Database about articles published in fast breeder reactors research and technology. The quantitative data in this report are obtained for various properties of relevant INIS records such as year of publication, secondary subject categories, countries of publication, language, publication types, literary types, etc. (author)

  18. Research about the Influence of Environmental Factors on Breeders Quality

    Directory of Open Access Journals (Sweden)

    Adina Popescu

    2011-10-01

    Full Text Available Along the growth period of the breeders, the monitoring of environmental parameters is a fundamental condition toensure the quality of the breeders used for reproduction. The results from the research presented in this paper wereobtained following experimental type investigations developed in vegetation and cold season within Carja 1-Vasluifish farm, on chemical and biological samples which were analyzed within the research laboratory of the Departmentof Aquaculture, Environmental Science and Cadastre. Were analyzed parameters which influence bio-productivity:temperature, oxygen, pH, the concentration of nitrites, nitrates, phosphates, the density and abundance ofphytoplankton and zooplankton, the individual weight and health condition of breeders. Analyzed parametersincluded mean values recorded in the optimal range for fish waters, as reflected in the numerical density andabundance of plankton and the average weight of Asian cyprinids breeders with a plankton nutritional spectrum.

  19. Characterization of the thermal conductivity for ceramic pebble beds

    Science.gov (United States)

    Lo Frano, R.; Aquaro, D.; Scaletti, L.; Olivi, N.

    2015-11-01

    The evaluation of the thermal conductivity of breeder materials is one of the main goals to find the best candidate material for the fusion reactor technology. The aim of this paper is to evaluate experimentally the thermal conductivity of a ceramic material by applying the hot wire method at different temperatures, ranging from 50 to about 800°C. The updated experimental facility, available at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa, used to determine the thermal conductivity of a ceramic material (alumina), will be described along with the measurement acquisition system. Moreover it will be also provided an overview of the current state of art of the ceramic pebble bed breeder thermos-mechanics R&D (e.g. Lithium Orthosilicate (Li4SiO4) and Lithium Metatitanate (Li2TiO3)) focusing on the up-to-date analysis. The methodological approach adopted is articulated in two phase: the first one aimed at the experimental evaluation of thermal conductivity of a ceramic material by means of hot wire method, to be subsequently used in the second phase that is based on the test rig method, through which is measured the thermal conductivity of pebble bed material. In this framework, the experimental procedure and the measured results obtained varying the temperature, are presented and discussed.

  20. Behaviour of tritium in the environment

    International Nuclear Information System (INIS)

    Full text: There is considerable interest in the behaviour of radionuclides of global character that may be released to the environment through the development of nuclear power. Tritium is of particular interest due to its direct incorporation into water and organic tissue. Although there has been a large decrease (more than ten times) in tritium concentration since the stopping of nuclear weapons tests in the atmosphere, the construction in the near future of many water reactors and in the far future of fusion reactors could increase the present levels. Progress has been made during recent years in the assessment of tritium distribution, in detection methods and in biological studies While several meetings have given scientists an opportunity to present papers on tritium, no specific symposium on this topic has been organized by the IAEA since 1961. Thus the purpose of the meeting was to review recent advances and to report on the practical aspects of tritium utilization and monitoring. The symposium was jointly organized with OECD/NEA, in co-operation with the US Department of Energy and the Lawrence Livermore Laboratory. Papers were presented on distribution of tritium, evaluation of future discharges, measurement of tritium, tritium in the aquatic environment, tritium in the terrestrial environment, tritium in man and monitoring of tritium Very interesting papers were given on distribution of tritium and participants got a good idea of the circulation of this radionuclide Some new data were provided on tritium pollution from luminous compounds and we learnt that the tritium release of the Swiss luminous compounds industry is of the same order of magnitude as the tritium release of Windscale. Projections indicate that, in the USA, the total quantity of tritium contained in discarded digital watches will be equal, approximately ten years in the future, to the release of nuclear power reactors Whereas nuclear reactor discharges are controlled there is no control

  1. Reprocessing of fast breeder reactor fuels in France

    International Nuclear Information System (INIS)

    The reprocessing of breeder reactor fuels is a direct technical descendant of the reprocessing of thermal reactor fuels which was developped first. The process used is in both cases the PUREX process, which consists in dissolution by nitric acid followed by selective extraction using TBP. In France, the application of this technique to breeder reactor fuels greatly benefited from the scientific and industrial experience initially acquired with metallic fuels of the MAGNOX type and then with oxide fuels of the LWR type

  2. Exploding the myths about the fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, S.

    1979-01-01

    This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

  3. Group size adjustment to ecological demand in a cooperative breeder

    OpenAIRE

    Zöttl, Markus; Frommen, Joachim G.; Taborsky, Michael

    2013-01-01

    Environmental factors can determine which group size will maximize the fitness of group members. This is particularly important in cooperative breeders, where group members often serve different purposes. Experimental studies are yet lacking to check whether ecologically mediated need for help will change the propensity of dominant group members to accept immigrants. Here, we manipulated the perceived risk of predation for dominant breeders of the cooperatively breeding cichlid fish Neolampro...

  4. Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

    International Nuclear Information System (INIS)

    In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 233U isotope which has very high quality fission cross-section with thermal neutrons. 233U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2D + 3T →? 4He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li2BeF4, LiF-NaF-BeF2, Li20Sn80, natural Lithium and Li17Pb83, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li3N, Li2O, Li2O2, Li2TiO3, Li4SiO3, Li2ZrO3, LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S8-P3 approximation with Gaussian quadratures using

  5. User's manual for the ARMLID (Argonne metallic lithium/isotopic dilution) tritium assay system

    International Nuclear Information System (INIS)

    The Argonne Metallic Lithium - Isotopic Dilution (ARMLID) system described in this report, originally developed at ANL for other purposes, was recently redeployed to measure the tritium production rate (TPR) in a series of US/Japanese collaborative fusion blanket integral experiments, involving large assemblies of fusion breeder blanket materials that were irradiated with a fusion neutron source at FNS/JAERI, Japan. Whereas previous uses of the ARMUD scheme involved just a few samples, its application infusion blanket TPR mapping called for large sample numbers per experiment, implying a commensurate scale of sample fabrication and encapsulation, on one hand, and tritium extraction and counting on the other hand. To shorten the time required for these various tasks, yet still yield reliable and accurate results, both the sample fabrication - encapsulation facility and the tritium extraction system had to be extensively revised from original versions that were designed for accuracy, but not necessarily for speed. The present report describes overall revisions in sufficient detail to serve as a User's Manual for this facility, and/or suggest how a new system might be put together. Either possibility may develop in the near future, in support of ITER design studies. Preliminary and partial descriptions of various aspects and features of the system were presented orally, in the course of annual ANL/JAERI/UCLA ''workshops'', over the last 34 years, as well as elsewhere

  6. Radiological training for tritium facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

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

  7. Radiological training for tritium facilities

    International Nuclear Information System (INIS)

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

  8. The future of the Fast Breeder

    International Nuclear Information System (INIS)

    Fast Breeder Reactors (FBRs) can produce more fissile nuclei than they consume whilst, at the same time, generating energy using fast neutrons. By conversion of uranium isotope 238 into a fissionable fuel, FBRs provide over 60 times more energy than can be extracted from the uranium reserves by thermal reactors. Their development is therefore an essential objective in the next century, particularly for those industrialised countries that have little or no energy resources of their own. The European countries which have been engaged in the development of FBRs for more than 25 years have decided to collaborate in an advanced design, the European Fast Reactor (EFR) which uses the best of previous national projects and draws on extensive operating experience from FBR plants in Europe. The naturally safe characteristics and technological features of sodium-cooled Fast Reactors will be fully utilised in an EFR design which meets the same safety level as the Light Water Reactors (LWRs). Owing to technical progress and series construction effect, the EFR is expected to achieve competitiveness with contemporary LWRs with the higher capital cost of the Fast Reactor offset by its markedly lower fuel cycle cost. (author)

  9. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.)

  10. Development of a tritium monitor combined with an electrochemical tritium pump using a proton conducting oxide

    International Nuclear Information System (INIS)

    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 CaZr0.9In0.1O3-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

  11. A study on tritium separation from LiPb by permeation into Na or NaK and cold trapping

    International Nuclear Information System (INIS)

    The tritium separation and recovery method discussed in this report appears to be a very promising technique for a LiPb self cooled blanket where an intermediate loop is required for safety reasons anyway. This technique can be also of interest for a water cooled LiPb blanket if a tritium purification unit exists in the water loop. Considerable work has been done on cold trapping of hydrogen from Na flows for the fast breeder technology. More work is needed in respect to fusion blanket applications, especially, on hydrogen removal from cold traps by thermal hydride decomposition. This report summarizes the state of the art on the precipitation and decomposition processes and discusses practical experiences with cold traps. Some ideas on a fusion blanket cold trap are outlined and a research program covering the more fundamental aspects for the next future is proposed. (orig.)

  12. Automation of the Tritium Extraction Facility

    International Nuclear Information System (INIS)

    The US Department of Energy has determined its future requirements for tritium will be met using the existing reactors of the Tennessee Valley Authority. Tritium Producing Burnable Absorber Rods (TPBARs) will replace the existing burnable absorber rods in the reactor core to beneficially use excess neutrons to create the tritium. The irradiated TPBARs will be shipped from the reactor to a new facility at the Savannah River Site. This new facility, the Tritium Extraction Facility (TEF), will receive the shipments from the reactor, store the TPBARs, prepare the TPBARs for tritium extraction, extract the tritium, and package the waste for disposal. The high level of gamma radiation emitted from the TPBARs will preclude human contact. Automation and remote handling will be used to accomplish the required operations, while minimizing radiation exposure to workers

  13. DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Blanton, P.

    2013-10-10

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

  14. History of 232-F, tritium extraction processing

    International Nuclear Information System (INIS)

    In 1950 the Atomic Energy Commission authorized the Savannah River Project principally for the production of tritium and plutonium-239 for use in thermonuclear weapons. 232-F was built as an interim facility in 1953--1954, at a cost of $3.9M. Tritium extraction operations began in October, 1955, after the reactor and separations startups. In July, 1957 a larger tritium facility began operation in 232-H. In 1958 the capacity of 232-H was doubled. Also, in 1957 a new task was assigned to Savannah River, the loading of tritium into reservoirs that would be actual components of thermonuclear weapons. This report describes the history of 232-F, the process for tritium extraction, and the lessons learned over the years that were eventually incorporated into the new Replacement Tritium Facility

  15. Tritium proof-of-principle injector experiment

    International Nuclear Information System (INIS)

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

  16. Stockpile tritium production from fusion

    International Nuclear Information System (INIS)

    A fusion breeder holds the promise of a new capability - ''dialable'' reserve capacity at little additional cost - that offers stockpile planners a new way to deal with today's uncertainties in forecasting long range needs. Though still in the research stage, fusion can be developed in time to meet future military requirements. Much of the necessary technology will be developed by the ongoing magnetic fusion energy program. However, a specific program to develop the nuclear technology required for materials production is needed if fusion is to become a viable option for a new production complex around the turn of the century

  17. Industrial ceramics

    International Nuclear Information System (INIS)

    After having given the definition of the term 'ceramics', the author describes the different manufacturing processes of these compounds. These materials are particularly used in the fields of 1)petroleum industry (in primary and secondary reforming units, in carbon black reactors and ethylene furnaces). 2)nuclear industry (for instance UO2 and PuO2 as fuels; SiC for encapsulation; boron carbides for control systems..)

  18. Effluent Treatment Facility tritium emissions monitoring

    International Nuclear Information System (INIS)

    An Environmental Protection Agency (EPA) approved sampling and analysis protocol was developed and executed to verify atmospheric emissions compliance for the new Savannah River Site (SRS) F/H area Effluent Treatment Facility. Sampling equipment was fabricated, installed, and tested at stack monitoring points for filtrable particulate radionuclides, radioactive iodine, and tritium. The only detectable anthropogenic radionuclides released from Effluent Treatment Facility stacks during monitoring were iodine-129 and tritium oxide. This paper only examines the collection and analysis of tritium oxide

  19. Tritium radioluminescent devices, Health and Safety Manual

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-06-01

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

  20. Compatibility problems in tritium breeding blankets

    International Nuclear Information System (INIS)

    Compatibility between tritium breeding materials (liquid or solid), neutron multiplier and structural steels is a concern for the choice of a tritium breeding blanket for NET. For solid tritium breeding blanket, it seems that the more severe compatibility problem is due to the interaction of beryllium with steel. As for the water-cooled Pb17Li blanket, the first results obtained in experimental conditions closed to the concept have evidenced lower corrosion rates than those measured in thermal convection loops

  1. Overview of light sources powered by tritium

    International Nuclear Information System (INIS)

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

  2. The movement of tritium in ecological systems

    International Nuclear Information System (INIS)

    This literature survey summarizes the interaction of tritium gas and tritiated water with various components of the ecological system. The intake of tritium gas and tritiated water in plants and soil is described as well as the location of the highest measurable concentration. This information may serve as a basis for risk assessment from tritium to man through the food chain and enables effective tracing of its concentration in the environment. (author)

  3. Tritium radioluminescent devices, Health and Safety Manual

    International Nuclear Information System (INIS)

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

  4. Compatibility of sodium with ceramic oxides employed in nuclear reactors; Compatibilidad del sodio con oxidos ceramicos utilizados en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Acena Moreno, V.

    1981-07-01

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

  5. Tritium permeation barrier based on self-healing composite materials

    International Nuclear Information System (INIS)

    Pores and cracks in ceramic coatings is one of the most important problems to be solved for the thermally sprayed tritium permeation barriers (TPBs) in fusion reactor. In this work, we developed a self-healing composite coating to address this problem. The coating composed of TiC + mixture(TiC/Al2O3) + Al2O3 was deposited on martensitic steels by means of atmospheric plasma spraying (APS). Before and after heat treatment, the morphology and phase of the coating were comparatively investigated by scanning electron microscopy (SEM) and X-ray diffraction (XRD). In the experiment, NiAl + Al2O3, mixture(TiC/Al2O3) + Al2O3 and NiAl + TiC + mixture(TiC/Al2O3) + Al2O3 films were also fabricated and studied, respectively. The results showed that the TiC + mixture(TiC/Al2O3) + Al2O3 coating exhibited the best self-healing ability and good thermal shock resistance among the four samples after heat treatment under normal atmosphere. The SEM images analyzed by Image Pro software indicated that the porosity of the TiC + mixture(TiC/Al2O3) + Al2O3 coating decreased more than 90% in comparison with the sample before heat treatment. This self-healing coating made by thermal spraying might be a good candidate for tritium permeation barrier in fusion reactors.

  6. Hot isostatic pressing of ceramic waste from spent nuclear fuel

    International Nuclear Information System (INIS)

    Argonne National Laboratory has developed a process to immobilize waste salt containing fission products, uranium, and transuranic elements as chlorides in a glass-bonded ceramic waste form. This salt was generated in the electrorefining operation used in electrometallurgical treatment of spent Experimental Breeder Reactor-II fuel. The ceramic waste process culminated with a hot isostatic pressing operation. This paper reviews the installation and operation of a hot isostatic press in a radioactive environment. Processing conditions for the hot isostatic press are presented for non-irradiated material and irradiated material. Sufficient testing was performed to demonstrate that a hot isostatic press could be used as the final step of the processing of ceramic waste for the electrometallurgical spent fuel treatment process

  7. The introduction of tritium in lactose and saccharose by isotope exchange with gaseous tritium

    International Nuclear Information System (INIS)

    Methods for conducting reactions of catalytic protium-tritium isotopic exchange with gaseous tritium were developed in order to synthesize tritium labelled lactose and saccharose. These methods enabled to prepare these labelled disaccharides with high molar activity. The yield was equal to 50-60%, radiochemical purity ∼ 95%

  8. Portable tritium-in-air monitor (TIAM) for monitoring tritium activity during shutdown in PHWR

    International Nuclear Information System (INIS)

    This paper discusses an overview of Portable Tritium in Air Monitor and usefulness of the system for health safety in Pressured Heavy Water Reactor based Nuclear Power Plants during Shutdown activity. Tritium in Air Monitor is meant for detection of Tritium activity in air in different accessible area and shutdown area inside the Reactor Building and also in atmospheric release. (author)

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

    International Nuclear Information System (INIS)

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

  10. Environmental tritium contamination from a gaseous tritium light device maintenance facility

    International Nuclear Information System (INIS)

    A series of radiological characterization, decontamination works and validation assessments have been conducted for a Gaseous Tritium Light Device (GTLD) maintenance facility in Brisbane, Australia in 2003 and more recently in 2007. The initial survey utilised soil pore water tritium concentration as an indicator of the extent of contamination. Tritium, existing as tritiated water or HTO, is environmentally mobile and tends to behave as non-radioactive water would, i.e. will permeate through soil under normal hydraulic conditions. A more useful measure, and that exploited in this work, for the purposes of dealing with contaminated land is the 'non-aqueous tritium' concentration present in the soil. In addition to tritium existing as HTO, organically bound tritium (OBT), a result of fixation by biological and chemical processes, or tritium diffused into various matrices may be present and represent the non-aqueous component. Quantification of all these variants allows assessment of the reservoir of tritium that may become available in environmental processes. Screening soil contamination levels for non-aqueous tritium concentration of 8.5 Bq.g-1 and 16 Bq.g-1 for residential and industrial/commercial land uses respectively were applied. Comparison of non-aqueous tritium and soil pore water tritium and elemental zinc concentration in soil were made and tritium concentrations in groundwater wells surrounding the facility were determined (tk)

  11. The trends of global tritium precipitations

    International Nuclear Information System (INIS)

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

  12. The Production Rate of Natural Tritium

    OpenAIRE

    Craig, Harmon; Lal, Devendra

    2011-01-01

    A detailed evaluation is made of the production rate of natural tritium in the pre-thermonuclear epoch. Deuterium and tritium analyses on the same precipitation samples are used to establish the uncontaminated tritium levels in precipitation sampled before the Castle tests, and the tritium balance is calculated for the North American troposphere. The global mean production rate Q?, calculated from the geochemical inventory, is found to be 0.5 ± 0.3 atoms T/cm2 sec. This value is three ...

  13. Tritium stripping by a catalytic exchange stripper

    International Nuclear Information System (INIS)

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

  14. Development of tritium handing technology(II)

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. S.; Ahn, D. H.; Kim, K. R. [KAERI, Taejon (Korea, Republic of); Yook, D. S.; Song, K. M.; Son, S. H. [KEPRI, Taejon (Korea, Republic of); Lee, K. J.; Jung, H. Y.; Song, M. C. [KAIST, Taejon (Korea, Republic of)

    2004-02-01

    The buildup rate of tritium in heavy water moderator and coolant of pressurized heavy water reactors in Wolsong Nuclear Power Plant is about 4MCi/a. The control of tritium is of increasing concern to the power reactor industry and general public in Korea. Metal tritides have the advantage of significantly decreasing the volume required to store tritium without increasing the pressure of storage vessel. Titanium hydride was safely used for the long-term storage of tritium. The experimental thermodynamic P-C-T data show that titanium soaks up hydrogen isotope gas at ambient temperature and modest pressures.

  15. Tritium immobilization and packaging using metal hydrides

    International Nuclear Information System (INIS)

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

  16. Mutagenic effect of incorporated tritium amino acids

    International Nuclear Information System (INIS)

    Genetic effect of tritium labelled amino acids was studied. The experiments were carried out on white mongreal rats, genetic effects were evaluated by dominant lethal mutation frequency in male germ cells. It was shown that administration of tritium amino acids results in genetic violations in male germ cells manifested in progeny death. Assessment of integral temporal indices of induced post implantation embryos death showed that 3H-lysine effect exceeds tritium oxide effect by 1.5-2 fold in case of equal absorbed doses. The obtained results are used in alculation of radiation hygienic standards for biogenic tritium compounds. 4 refs.; 1 tab

  17. Tritium proof-of-principle pellet injector

    International Nuclear Information System (INIS)

    The tritium proof-of-principle (TPOP) experiment was designed and built by Oak Ridge National Laboratory (ORNL) to demonstrate the formation and acceleration of the world's first tritium pellets for fueling of future fusion reactors. The experiment was first used to produce hydrogen and deuterium pellets at ORNL. It was then moved to the Tritium Systems Test Assembly at Los Alamos National Laboratory for the production of tritium pellets. The injector used in situ condensation to produce cylindrical pellets in a 1-m-long, 4-mm-ID barrel. A cryogenic 3He separator, which was an integral part of the gun assembly, was capable of lowering 3He levels in the feed gas to <0.005%. The experiment was housed to a glovebox for tritium containment. Nearly 1500 pellets were produced during the course of the experiment, and about a third of these were pure tritium or mixtures of deuterium and tritium. Over 100 kCi of tritium was processed through the experiment without incident. Tritium pellet velocities of 1400 m/s were achieved with high-pressure hydrogen propellant. The design, operation, and results of this experiment are summarized. 34 refs., 44 figs., 3 tabs

  18. Status of fast breeder reactor development in the United States

    International Nuclear Information System (INIS)

    This document was prepared by the Office of the Program Director for Nuclear Energy, U.S. Department of Energy (USDOE). It sets forth the status and current activities for the development of fast breeder technology in the United States. In April 1977 the United States announced a change in its nuclear energy policy. Concern about the potential for the proliferation of nuclear weapons capability emerged as a major issue in considering whether to proceed with the development, demonstration and eventual deployment of breeder reactor energy systems. Plutonium recycle and the commercialization of the fast breeder were deferred indefinitely. This led to a reorientation of the nuclear fuel cycle program which was previously directed toward the commercialization of fuel reprocessing and plutonium recycle to the investigation of a full range of alternative fuel cycle technologies. Two major system evaluation programs, the Nonproliferation Alternative Systems Assessment Program (NASAP), which is domestic, and the International Nuclear Fuel Cycle Evaluation (INFCE), which is international, are assessing the nonproliferation advantages and other characteristics of advanced reactor concepts and fuel cycles. These evaluations will allow a decision in 1981 on the future direction of the breeder program. In the interim, the technologies of two fast breeder reactor concepts are being developed: the Liquid Metal Fast Breeder Reactor (LMFBR) and the Gas Cooled Fast Reactor (CFR). The principal goals of the fast breeder program are: LMFBR - through a strong R and D program, consistent with US nonproliferation objectives and anticipated national electric energy requirements, maintain the capability to commit to a breeder option; investigate alternative fuels and fuel cycles that might offer nonproliferation advantages; GCFR - provide a viable alternative to the LMFBR that will be consistent with the developing U.S. nonproliferation policy; provide GCFR technology and other needed

  19. Operating experience of Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Fast Breeder Test Reactor (FBTR) is a 40 MWt / 13.2 MWe sodium cooled, loop type mixed carbide fuelled reactor. Its main aim is to gain experience in the design, construction and operation of fast reactors and to serve as an irradiation facility for development of fuel and structural material for future fast reactors. The reactor achieved first criticality in October 1985 with small indigenously designed and fabricated Mark I core (70% PuC-30% UC). The reactor power was subsequently raised in steps to 17.4 MWt by addition of Mark II fuel subassemblies (55% PuC-45% UC) and with the Mark I fuel operating at the designed linear heat rating of 400 W/cm. The turbo-generator was synchronized with the grid in July 1997. The achieved peak burn-up is 137 000 MWd/t so far without any fuel-clad failure. Presently the reactor is being operated at a nominal power of 15.7 MWt for irradiation of a test fuel subassembly of the Prototype Fast Breeder Reactor, which is coming up at Kalpakkam. It is also planned to irradiate test subassemblies made of metallic fuel for future fast reactor program. Being a small reactor, all feed back coefficients of reactivity including void coefficient are negative and hence the reactor is inherently safe. This was confirmed by carrying out physics tests. The capability to remove decay heat under various incidental conditions including natural convection was demonstrated by carrying out engineering tests. Thermo couples are provided for on-line monitoring of fuel SA outlet temperature by dedicated real time computer and processed to generate trip signals for the reactor in case of power excursion, increase in clad hot spot temperature and subassembly flow blockage. All pipelines and capacities in primary main circuit are provided with segmented outer envelope to minimize and contain radioactive sodium leak while ensuring forced cooling through reactor to remove decay heat in case of failure of primary boundary. In secondary circuit, provision is

  20. Fast breeder physics and nuclear core design

    International Nuclear Information System (INIS)

    This report gathers the papers that have been presented on January 18/19, 1983 at a seminar ''Fast breeder physics and nuclear core design'' held at KfK. These papers cover the results obtained within about the last five years in the r+d program and give some indication, what still has to be done. To begin with, the ''tools'' of the core designer, i.e. nuclear data and neutronics codes are covered in a comprehensive way, the seminar emphasized the applications, however. First of all the accuracies obtained for the most important parameters are presented for the design of homogeneous and heterogeneous cores of about 1000 MWe, they are based on the results of critical experiments. This is followed by a survey on activities related to the KNK II reactor, i.e. calculations concerning a modification of the core as well as critical experiments done with respect to re-loads. Finally, work concerning reactivity worths of accident configurations is presented: the generation of reactivity worths for the input of safety-related calculations of a SNR 2 design, and critical experiments to investigate the requirements for the codes to be used for these calculations. These papers are accompanied by two contributions from the industrial partners. The first one deals with the requirements to nuclear design methods as seen by the reactor designer and then shows what has been achieved. The latter one presents state, trends, and methods of the SNR 2 design. The concluding remarks compare the state of the art reached within DeBeNe with international achievements. (orig.)