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Sample records for radionuklide tritium sr-90

  1. Determination of Sr-90 in the air

    International Nuclear Information System (INIS)

    Marovic, G.; Bajlo, M.; Bauman, A.

    1983-01-01

    Sr-90 concentrations were determined in high-volume air samples (up to 10000 m 3 ) during 12 months. In 1981 radioactivity varied around a value of 2.7 x 10 - 5 Bq m 3 . The results are in agreement with the activities recorded in the same year in New York, USA and in France. (author)

  2. Preparation of 90Sr-90Y generator

    International Nuclear Information System (INIS)

    Jin Xiaohai; Yu Haibin; Zhang Jinming; Zhang Peixin; Lin Qiongfang

    1990-01-01

    In recent years, 90 Y has been considered as one of the best radionuclides for tumor radioimmunotherapy when chelated to tumor-associated antibodies. This evaluation is based on the superior properties of this radionuclide (suitable half-life, pure β-ray emitter of intermediate energy, stable daughters, and suitable chemical properties) and because it is available as a radionuclide generator product by decay of its 28a parent 90 Sr. The experimental conditions of 90 Sr- 90 Y generator are described. The elution efficiency of 90 Sr- 90 Y generator reaches 98%. One of the most important problems is the 90 Sr contamination breakthrough from the generator. The level of 90 Sr contamination must be controlled to the clinical standard. The cation exchange resin 732 (100-150 mesh) was successfully used for the separation of 90 Y from 90 Sr. The method used by the authors provides a 90 Y-HAc solution which is very simple and safe for administration to the patients. 90 Y was separated from 90 Sr almost completely, the level of the 90 Sr contamination per 740 MBq 90 Y product was only 0.74 kBq. However the toxicity of 90 Sr is extremely high, the human life-time permissible dose is 74 kBq, then 740 MBq of 90 Y is allowed to be administrated to a patient for 50-100 times

  3. Dosimetry of Sr-90 ophthalmic applicators

    International Nuclear Information System (INIS)

    Reft, C.S.; Kuchnir, F.T.; Rosenberg, I.; Myrianthopoulos, L.C.

    1990-01-01

    Sr-90 ophthalmic applicators are commonly used for the treatment of superficial eye disorders. Although a variety of dosimetric devices such as film, thermoluminescent dosimeters (TLD's), ion chambers, and radiochromic foils have been used to measure the peak dose at the applicator surface, there is no internationally agreed upon calibration procedure. Recently, large discrepancies among calibrations of the same applicator at three institutions have been reported. Here we describe a technique to obtain the peak dose rate at the applicator surface using LiF TLD's. The technique can be used for the calibration of flat as well as curved surface applicators. Results for two flat and three concave applicators are presented. Our measurement of the surface dose rate for one of the flat applicators is compared with those obtained by four other institutions, each using different dosimetric devices

  4. Development of 90Sr-90Y generators

    International Nuclear Information System (INIS)

    Barrio, Graciela

    2007-01-01

    Yttrium-90 is a radioisotope of great interest in the field of Nuclear Medicine. It is considered one of the most important and most used radionuclides for radioimmunotherapeutical applications, especially promising for the treatment of certain types of cancer. Another important application of 90 Y is radio synovectomy. This radionuclide has a half-life of 64 hours, emits long range beta particles (maximum energy of 2.3 MeV) and decays, without intermediate nuclides, to a stable daughter. 90 Y may be obtained carrier-free, generated by the decay of its parent 90 Sr (half-life=28 years). 90 Sr is a product from uranium fission, and due to its long half-life, can be indefinitely used, which is certainly advantageous. It is present in great amounts, and needs to be processed and purified in order to be used as raw material for the generators. Generators of 90 Sr- 90 Y may thus be used during various months, due to 90 Sr long half-life. Several methods for the separation of 90 Y from 90 Sr by solvent extraction and ion exchange have been reported in literature. Thanks to its simplicity, ion exchange techniques have been more commonly used for this generator system. The main objective of this work was to develop a methodology for the preparation of 90 Sr- 90 Y generators, using cationic exchange resins. In such method, 90 Sr is strongly adsorbed in the resin and 90 Y is eluted by a 0.003 M EDTA solution. According to the quality control carried out, results showed that elution yields are greater than 65%, thus confirming the efficiency of the separation method used.

  5. Peculiarities of Sr-90 migration in the environment

    International Nuclear Information System (INIS)

    Romanov, G.N.; Stukin, D.A.; Aleksakhin, R.M.

    1991-01-01

    The Eastern Urals radioactive track which formed as a result of the Kyshtym accident constituted a natural experimental base for studying the dynamics of Sr-90 behaviour and migration in the environment. Sr-90 behaviour in soil depends on the processes involved in its physical migration horizontally (water runoff, wind transport) and vertically (effective diffusion), on the intensity with which the physico-chemical forms of strontium change, and primarily on Sr-90 incorporation into the soil-absorptive complex. Over 30 years the effective diffusion processes led to downward displacement of Sr-90 in undisturbed soil to a depth of 30 to 50 cm. In 1988 84-94% of the Sr-90 was located in the 10 cm layer as against 90% in the 2 cm layer in the initial 1 or 2 years. The amount of Sr-90 in mobile forms in leached chernozem and grey forest soil hardly changed over time, amounting to between 76 and 90%. In the first 5-10 years Sr-90 surface water runoff was about 0.2% (and Sr-90 wind resuspension 0.1-1.0%) per year of the Sr-90 inventory per unit area. Due to Sr-90 loss from the top soil layer the water runoff and wind resuspension processes decreased exponentially, halving every 4 to 5 years. In most cases, Sr-90 transfer from soil to plants via the roots does not involve discrimination of Sr-90 with regard to calcium, and therefore the Sr-90 accumulation level in various species of natural and agricultural plants depends on their calcium requirements and the amount of exchangeable calcium in the soil. This made it possible to work out quantitative indices for forecasting Sr-90 accumulation in various species of plants for given types of soil. As Sr-90 moves through the trophic chains (including the agricultural one) we find - depending on whether two adjoining links discriminate against it or not as regards calcium - that some organisms reject it while others concentrate it. When restoring farming on the contaminated territory the fact that cattle discriminated against Sr

  6. The study of accumulation of Sr 90 by plant cells

    International Nuclear Information System (INIS)

    Matusov, G.D.; Kudryashova, N.N.

    2002-01-01

    In this work the absorption and desorption of ions Sr 90 by plant cells and influence of different physical and chemical factors of environment on that processes were investigated. The kinetics of strontium accumulation have been obtained and the factors of accumulation of Sr 90 have been determined for a plant cell itself and its separate compartments

  7. Cs-137 and Sr-90 level in diary products

    International Nuclear Information System (INIS)

    Petukhov, V.L.; Dukhanov, Y.A.; Sevryuk, I.Z.; Patrashkov, S.A.; Korotkevich, O.S.; Gorb, T.S.; Petukhov, I.V.

    2003-01-01

    About 70% of radioactive substances fell on the territory of the Byelorussian Republic after the Chernobyl Atom Power Station Disaster. Cs-137 and Sr-90 accumulation dynamics was studied in milk of the cows from the highest polluted Braginsky area. 408 milk samples of Black and White cows were investigated. In 1995 average Cs-137 and Sr-90 levels were 61.00 and 3.73 Bk/dm 3 respectively. Cs-137 and Sr-90 levels exceeded Byelorussian Republic upper limits RDU - 96 in 10 and 50% of milk samples respectively. After 5 years (by 2000) Cs-137 and Sr-90 levels had become almost 3 and 2 times less (21.70 Bk/dm 3 and 1.72 Bk/dm 3 respectively). Cs-137 and Sr-90 levels exceeded RDU - 96 in 1.5 and 5.5% of milk samples respectively. In the same periods Cs-137 and Sr-90 levels were 7 and 2 times higher than the similar indexes in the relatively clean Novosibirsk area. Thus, radioactive element levels in milk of Black and White cows of the Byelorussian Republic decreased significantly for the past years. (authors)

  8. Sr-90 determination in aqueous and soils samples

    International Nuclear Information System (INIS)

    Gonzalez Sintas, Maria F.; Cerchietti, Maria L.; Arguelles, Maria G.

    2009-01-01

    The main objective of this paper is to evaluate the method for Sr-90 determination in aqueous sample and soils. Area and Personal Dosimetry laboratory (DPA) determines the presence of Sr-90 by Liquid Scintillation (LSC) by applying method of the double window and corresponding adjustments. Calibration is performed by standard solutions of 90 Sr/ 90 Y, where spectral 90 Sr and 90 Y zones are optimized. The initial treatment of the liquid samples includes the concentration for evaporation, while the solid ones dissolve for microwave and acidic digestion. The separation of the analyte involves a selective chromatographic extraction. An average efficiency for 90 Sr of 77 ± 1 % was obtained; the factor a/b was 0,85 ± 0,01 and recovery of 82 ± 8 %. The resultant MAD was 0,10 Bq/L in aqueous samples and 0,10 Bq/g in solid samples. (author)

  9. Determination of Sr-90 in rain water samples

    International Nuclear Information System (INIS)

    Lima, M.F.; Cunha, I.I.L.

    1988-01-01

    A work that aim is to establish radiochemical method for the determination of Sr-90 in rain water samples has been studied, as a step in an environmental monitoring program of radioactive elements. The analysis includes the preconcentration of strontium diluted in a large volume sample by precipitation of strontium as carbonate, separation of strontium from interfering elements (calcium, barium and rare earths), separation of strontium from ytrium, precipitation of purified strontium and ytrium respectively as carbonate and oxalate, and counting of Sr-90 and Y-90 activities in a low background anticoincidence beta counter. (author) [pt

  10. Determination of Sr90 activity in human bones

    International Nuclear Information System (INIS)

    Mendonca, Anamelia Habib

    1970-01-01

    Several studies have been published in the literature on the extent and levels of radioactive contamination of food chains caused by fallout from nuclear weapon tests. According to UNSCEAR, these studies cover a great number of:-areas of the developed world, though large, areas of Asia, Africa and South-America are left aside with only, unsatisfactory information about the levels, of radioactive contamination. In 1968, UNSCEAR recommended that a survey on the contamination of biological materials such as human - bone by fission products and particularly Sr 90 should be encouraged on those areas where only fragmentary information was available. UNSCEAR recommendations call upon the fact that many individuals on such areas of the world have been exposed to Sr 90 contamination from birth to their adult area. Therefore, that group have an Sr 90 skeletal burden very much different from people exposed only at adult age. Based on these considerations, UNSCEAR recommendations called for Sr 90 analysis on human bones from different age groups. In Brazil, studies on the of Sr 90 in human bone are practically non-existent, except for the year of 1959. Following UNSCEAR recommendations, we decided to perform such a survey on Sr 90 levels in human bones. Samples were collected from individuals that died in Rio de Janeiro from accidents. These samples were firstly classified according to social level in very poor and poor groups. Samples were then classified in three age groups ranging 0-18, 18-30 and 30-40 years of age. Results show that levels found in the Brazilian age groups are close to the ones observed in Chile (1969), Argentina and Australia (1966-1968) and slightly, higher than -those observed in Venezuela, Senegal and Jamaica (1969). If one compares the results obtained for the North and South hemispheres respectively, one sees that there was a more pronounced decrease in the levels of Sr 90 content of the of some regions of South America. Our results show no

  11. Radioactivity precipitation Sr-90 and Cs-137 in Plock

    International Nuclear Information System (INIS)

    Flakiewicz, W.; Majkowska, I.; Bonkowski, J.

    1986-01-01

    Methods of research and results for beta-activity measurements of precipitation in 1977-1983 with regard to Sr-90, Cs-137 - are presented. Climatic changes caused by big industrial plant, conditions of pollution propagation in atmosphere, parameters of pollution load of Plock region are described. 11 refs., 6 figs., 5 tabs. (author)

  12. Tritium

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

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

  13. Verification dosimetry of intravascular 90Sr/90Y source trains

    International Nuclear Information System (INIS)

    Sharma, S.D.; Shanta, A.; Tripathi, U.B.; Bhatt, B.C.

    2001-01-01

    90 Sr/ 90 Y source trains (Novoste Beta-Cath System) are currently under clinical trials in India and abroad for intracoronary brachytherapy for prevention of restenosis. Each source train of the Beta-Cath system is supplied with a source certificate giving dose rate at the reference distance of 2 mm from the centerline of the source train. It is essential that the user should check dose rates of brachytherapy sources before its application on the patients. Dose rates and depth dose measurements for 90 Sr/ 90 Y source trains of active length 40 mm using radiochromic films in a tissue equivalent phantom have been carried out. The objectives of these measurements were (1) to verify the dose rates stated in the source certificate, and (2) to obtain relative depth dose data for treatment planning. This paper presents the results of these measurements

  14. Role river flow for Sr 90 decontamination of polluted territories of Belarus

    International Nuclear Information System (INIS)

    Kudel'skij, A.V.; Smith, J.T.; Zhukova, O.M.; Rudaya, S.M.; Sasina, N.V.

    2002-01-01

    Sr 90 contamination of the water flow Dnepr, Pripyat', Sozh, Besed', Iput' rivers is considered. The dynamics of reducing the average year activities of Sr 90 and the variations of the levels of Sr 90 activities in river water during spring-autumn high water are shown. The results of investigation of Sr 90 activity of the sediments of Pripyat' and Braginka rivers are connected with the second effects of the contamination of the river flowing off Sr 90 during high water period. Sr 90 transfer in composition of the flowing off river during 1990-1995 (from Belarus to Ukraine) is being estimated. (authors)

  15. Migration of Sr 90 pine forest ecosystems in the modern technogenesis of Chernobyl exclusion zone

    International Nuclear Information System (INIS)

    Ganzha, D.D.

    2012-01-01

    The migration of Sr 90 in modern technogenesis against contamination from the Chernobyl accident was investigated. Radiometry and bioindication methods of were applied. The investigated area was divided into sections on the basis of the migration of Sr 90. (authors)

  16. Sr 90 in lake Drukshiai hydroecosystem - cooling basin of Ignalina NPP

    International Nuclear Information System (INIS)

    Dushauskene-Duzh, R.F.

    2002-01-01

    The main aim of these investigations were to investigate Sr 90 migration in lake Drukshiai hydroecosystem and determine factors stipulating the specific character of Sr 90 biological migration under the chemical and thermal pollution before and after Ignalina NPP acting

  17. Determination of Sr-90 and Cs-137 in environmental samples

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The comparison between the Ca concentration in the soil and the transfer factors involved indicates a clear dependence of Sr-90 soil-to-plant transport from the soil's Ca-content with the concentration level of 0.1 g Ca/kg soil appearing to be the limit level. Soil concentrations of Cs-137 determined to-date indicate a very even deposition density of Cs-137. The transfer factors of plant/soil in Chile differ but negligeably from those in Europe. (DG) [de

  18. Characterisation of a Sr-90 based electron monochromator

    CERN Document Server

    Arfaoui, S; CERN; Casella, C; ETH Zurich

    2015-01-01

    This note describes the characterisation of an energy filtered Sr-90 source to be used in laboratory studies that require Minimum Ionising Particles (MIP) with a kinetic energy of up to approx. 2 MeV. The energy calibration was performed with a LYSO scintillation crystal read out by a digital Silicon Photomultiplier (dSiPM). The LYSO/dSiPM set-up was pre-calibrated using a Na-22 source. After introducing the motivation behind the usage of such a device, this note presents the principle and design of the electron monochromator as well as its energy and momentum characterisation.

  19. Tritium

    International Nuclear Information System (INIS)

    Fiege, A.

    1992-07-01

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

  20. General regularities of Sr 90 distribution in system soil-plant under natural conditions

    International Nuclear Information System (INIS)

    Gudeliene, I.; Marchiulioniene, D.; Petroshius, R.

    2006-01-01

    Sr 90 distribution in system 'soil - underground part of plant - aboveground part of plant' was investigated. It was determined that Sr 90 activity concentration in underground and aboveground part of plants and in mosses was not dependent on its activity concentration in soil. There was direct dependence of Sr 90 activity concentration in aboveground on underground parts of plants. Sr 90 transfer factor from soil to underground part of plants and mosses was directly dependent on this radionuclide activity concentration in them. (authors)

  1. Study of radionuclide 90Sr-90Y on cell proliferation and apoptosis in benign prostatic hyperplasia

    International Nuclear Information System (INIS)

    Zhang Tong; Wei Wei; Zou Benjie; Liu Fang; Xu Zhishun

    2003-01-01

    Objective: To investigate the effect of 90 Sr- 90 Yon cell proliferation and apoptosis in benign prostatic hyperplasia. Methods: The apoptosis and expression of Ki-67 in benign prostatic hyperplasia (BPH) before and after irradiation 90 Sr- 90 Y were detected by transferase-mediated dUTP-biotin nick end labeling (TUNEL) method and immunohistochemical technique, respectively. Results: The proliferation index (PI) of BPH after 90 Sr- 90 Y irradiation was much lower than that before irradiation, but there was no significant change in apoptosis index (AI). Conclusion: Irradiation with 90 Sr- 90 Y could restrain cell proliferation of BPH, but could not induce apoptosis

  2. Study on the Sr-90 content in the spontaneous vegetations with fodder value

    International Nuclear Information System (INIS)

    Stoiciu, A.

    1994-01-01

    The author presents the results of a study concerning the Sr-90 activity in the spontaneous vegetation with fodder value (grass on the natural hay), during the spring - summer season of 1988, 1989 and 1990. It was possible to obtain information from 17 sampling points located in different zones of the country, at different altitudes concerning: Sr-90 accumulation in the grass (Bq/m 2 ), Sr-90 concentration of the grass in fresh state and in preserved state (Bq/kg), as well as the variation limits of the Sr-90 concentration factor in the preserved grass by drying. (author). 4 tabs., 3 refs

  3. Research And Establishment Of The Analytical Procedure For/Of Sr-90 In Milk Samples

    International Nuclear Information System (INIS)

    Tran Thi Tuyet Mai; Duong Duc Thang; Nguyen Thi Linh; Bui Thi Anh Duong

    2014-01-01

    Sr-90 is an indicator for the transfer radionuclides from environment to human. This work was setup to build a procedure for Sr-90 determination in main popular foodstuff and focus to fresh milk. The deal of this work was establish procedure for Sr-90 , assessment for chemical yield and test sample of Vietnam fresh milk, also in this work, the QA, QC for the procedure was carried out using standard sample of IAEA. The work has been completed for the procedure of determination Sr-90 in milk. The chemical yield of recovery for Y-90 and Sr-90 were at 46.76 % ±1.25% and 0.78 ± 0.086, respectively. The QA & QC program was carried out using reference material IAEA-373. The result parse is appropriate equally and well agreement with the certificate value. Three reference samples were analyses with 15 measurements. The results of Sr-90 concentration after processing statistics given a value at 3.69 Bq/kg with uncertainty of 0.23 Bq/kg. The certificate of IAEA-154 for Sr-90 (half live 28.8 year) is the 6.9 Bq/kg, with the range 95% Confidence Interval as (6.0 -8.0 ) Bq/kg at 31st August 1987. After adjusting decay, the radioactivity at this time is 3.67 Bq/kg. It means that such the result of this work was perfect matching the value of stock index IAEA. Five Vietnam fresh milk samples were analyzed for Sr-90, the specific radioactivity of Sr-90 in milk were in a range from 0.032 to 0.041 Bq/l. (author)

  4. Parameters of Cs 137 and Sr 90 transition from soil into salad crops

    International Nuclear Information System (INIS)

    Lazarevich, T.M.; Barashenko, V.V.

    2010-01-01

    Transition coefficients of Cs 137 and Sr 90 from soil into plants for potherb of bulb onion (Allium cepa), parsley (Petroselinum crispum), dill (Anethum graveolens), celery (Apium graveolens) and coriander (Coriandrum sativum) have been defined. (authors)

  5. The characteristics of intake Cs-137 and Sr-90 with food by Ozersk's inhabitants

    International Nuclear Information System (INIS)

    Dronova, M.

    2000-01-01

    This work is based on a questionnaire of production plant 'MAYAK' workers eating habits. Cs-137 and Sr-90 were chosen as experimental radionuclides. Cs-137 is distributed in organisms equally. It is considered to be a source of genetic damage after entry into the human body. This radionuclide is almost completely absorbed as it passes into the bowels after peroral entry. In the blood Cs-137 is distributed almost evenly between organs and tissues. It was established that Cs-137 accumulates in muscle, kidney, heart, spleen, lungs, and liver to a higher degree. Although the half-life of Cs-137 exceeds Sr-90, Cs-137 leaves the body more rapidly. The radiation dose of Cs-137 per unit of intake is less than for Sr-90. In addition, Cs-137 is absorbed by plants to a lesser degree than Sr-90. Taking into account that a considerable part of the daily diet consists of vegetables, it should be noted that different radionuclides get into plants in different amounts and different ways. Cs-137 concentrates in cereal, beans, oil plants, potatoes. beets and tomatoes. Meat and milk are also a main source of Cs-137 in a person's diet. Sr-90 contributes to radiation dose for bone and marrow, almost all Sr-90 is taken up by organisms through milk, greens, and cereal. (authors)

  6. Dynamics of Sr90 and its analogs accumulation by the vegetative parts of cabbage (Brassica oleracea l.) during its ontogenesis

    International Nuclear Information System (INIS)

    Nenasheva, M.N.; Timofeev, S.F.

    2003-01-01

    Field experiment demonstrated that the maximal content of Sr90 was observed in assimilative leaves of cabbage while the minimal content of Sr90 was traced in upper leaves. In conductive tissues Sr90 concentration increased insignificantly during the growth season. For assimilative plant parts the discrimination coefficient of Sr90 in relation to calcium was less than 1. It was revealed the positive correlative dependence between the contents of calcium, magnesium, stable strontium and manganese in vegetative tissues on the one hand and accumulation of Sr90 by these tissues on the other hand

  7. Dynamics of Sr 90 and its analogs accumulation by the vegetative parts of cabbage (Brassica oleracea L.) during its ontogenesis

    International Nuclear Information System (INIS)

    Nenasheva, M.N.; Timofeev, S.F.

    2004-01-01

    Field experiment demonstrated that the maximal content of Sr 90 was observed in assimilative leaves of cabbage while the minimal content of Sr 90 was traced in upper leaves. In conductive tissues Sr 90 concentration increased insignificantly during the growth season. For assimilative plant parts the discrimination coefficient of Sr 90 in relation to calcium was less than 1. The authors revealed the positive correlative dependence between the contents of calcium, magnesium, stable strontium and manganese in vegetative tissues on the one hand and accumulation of Sr 90 by these tissues on the other hand. (Authors)

  8. On the geochemistry of 'Chernobyl' Cs-137 and Sr-90 in the Black Sea

    International Nuclear Information System (INIS)

    Batrakov, G.F.; Chudinovskikh, T.V.; Zemlyanoi, A.D.; Eremeev, V.N.

    1998-01-01

    The following correlations for Cs 137 and Sr 90 were found. For Cs-137, the relation of its concentration in suspended matter to that in the dissolved component is 2.4 centre dot 10 5 in the Dnieper waters and 0.0037 centre dot 10 5 in the North-West shelf waters; for Sr-90 - 0.44 centre dot 10 3 and 2.8 centre dot 10 3 , respectively. The relation of the concentration of dissolved Cs-137 in the sea and in the Dnieper waters is 4.3, and for Sr-90 - 0.25. It is evident that these basic correlations for Cs-137 are close to those for stable cesium. These correlations for Sr-90 differ very much from those for stable strontium. So, the situation formed for the last time in the boundary area 'the Dnieper river - the North-Western Black Sea' is close to the balanced one for Cs-137, while it is very far from that for Sr-90

  9. Calibration of the 90Sr+90Y ophthalmic and dermatological applicators with an extrapolation ionization minichamber

    International Nuclear Information System (INIS)

    Antonio, Patrícia L.; Oliveira, Mércia L.; Caldas, Linda V.E.

    2014-01-01

    90 Sr+ 90 Y clinical applicators are used for brachytherapy in Brazilian clinics even though they are not manufactured anymore. Such sources must be calibrated periodically, and one of the calibration methods in use is ionometry with extrapolation ionization chambers. 90 Sr+ 90 Y clinical applicators were calibrated using an extrapolation minichamber developed at the Calibration Laboratory at IPEN. The obtained results agree satisfactorily with the data provided in calibration certificates of the sources. - Highlights: • 90 Sr+ 90 Y clinical applicators were calibrated using a mini-extrapolation chamber. • An extrapolation curve was obtained for each applicator during its calibration. • The results were compared with those provided by the calibration certificates. • All results of the dermatological applicators presented lower differences than 5%

  10. A special mini-extrapolation chamber for calibration of 90Sr+90Y sources

    International Nuclear Information System (INIS)

    Oliveira, Mercia L; Caldas, Linda V E

    2005-01-01

    90 Sr+ 90 Y applicators are commonly utilized in brachytherapy, including ophthalmic procedures. The recommended instruments for the calibration of these applicators are extrapolation chambers, which are ionization chambers that allow the variation of their sensitive volume. Using the extrapolation method, the absorbed dose rate at the applicator surface can be determined. The aim of the present work was to develop a mini-extrapolation chamber for the calibration of 90 Sr+ 90 Y beta ray applicators. The developed mini-chamber has a 3.0 cm outer diameter and is 11.3 cm in length. An aluminized polyester foil is used as the entrance window while the collecting electrode is made of graphited polymethylmethacrylate. This mini-chamber was tested in 90 Sr+ 90 Y radiation beams from a beta particle check source and with a plane ophthalmic applicator, showing adequate results

  11. Cancer risk due to Cs-137 and Sr-90 dietary intake after the Chernobyl accident

    International Nuclear Information System (INIS)

    Toader, M.; Vasilache, R.A.

    1997-01-01

    The most important radionuclides carried by the radioactive plume over Romania were I 131 , Cs 134 , Cs 137 and Sr 90 . As in many other countries, in the first days, I 131 had the main contribution to the irradiation dose released to the population. After its decay, and the decay of the other short-lived radionuclides, Cs 137 and Sr 90 remained the most important contaminants. The principal route of intake for these two radionuclides is considered to be the ingestion of contaminated foods. Assessments of radiation doses to people living in the Bucharest area have utilized data obtained from measurement of Cs 137 and Sr 90 content in dietary intake samples for a number of subjects of different ages and sexes. This paper summarizes the results of some of our measurements performed since April 1986 until March 1995. 7 refs, 8 figs

  12. In-flow technology for the determination of Sr-90 concentrations in nuclear power plants

    International Nuclear Information System (INIS)

    Dissing, E.

    1982-01-01

    Outlines for a future work concerning in-flow surveillance of Sr-90 in nuclear plant process treams have been studied. The many problems involved with the task of on-line Sr-90 determination were approached in two different ways, one applying β-counting of the 64-hour daughter Y-60 with the use of the process stream itself as a Cerenkov scintillator and the other - indirect - using simultaneous measurement of the concentrations of Sr-91 and Sr-92 for the determination of the leakage route for strontium. (Author)

  13. Seasonal variation of Sr-90 fallout in Japan through the end of 1983

    International Nuclear Information System (INIS)

    Katsuragi, Yukio; Aoyama, Michio

    1986-01-01

    Time variation of Sr-90 fallout together with that of Cs-137 is given through the end of 1983. The results at 12 stations in Japan indicate that the recent fallout in Japan was mainly derived from the 26th Chinese nuclear detonation. Relatively short stratospheric residence time of radioactive debris was obtained for the 26th Chinese detonation. The activity ratio of Cs-137 to Sr-90 in the fallout ranged from 0.8 to 6.0 with the average at 2.03. It is noted that the activity ratio of Cs-137 to Sr-90 decreased just after the detonation and then it increased. The meridional distribution of Sr-90 fallout over the Japan Islands indicates that it increased from south to north, whereas the amount of precipitation decreased in the same direction. After the thermonuclear detonation, the amount of fallout increased in the following year and the appearance of maximum fallout was delayed by one to two months from normal pattern of the spring maximum. (author)

  14. Sr 90 behaviour in the soil-plant system of moors of the Gomel' range

    International Nuclear Information System (INIS)

    Kostereva, I.V.; Kudryashov, V.P.; Shamal', N.V.; Matusevich, Zh.L.

    2005-01-01

    The wavy dependence is installed between pH of the soil of moors and rate of accumulation (RA) of Sr 90 by phytomass of plants. The direct correlation is marked between concentration of mineral materials and RA of an above - ground part of herbaceous plants. Is abnormal high values RA are observed for the green forms of semifrutexs and undershrubs. (authors)

  15. Rapid method for determining Sr-89 and Sr-90 using Cherenkov and proportional counting; Schnellmethode zur Bestimmung von SR-89 und SR-90 durch Cerenkov- und Proportionalzaehlermessungen

    Energy Technology Data Exchange (ETDEWEB)

    Lange, S.; Wende, C.; Schwokowski, R.; Alisch-Mark, M.; Abraham, A.; Heinrich, T. [Staatliche Betriebsgesellschaft fuer Umwelt und Landwirtschaft, Radebeul (Germany)

    2016-07-01

    A rapid method for determining Sr-89 and Sr-90 in water, milk and biological samples has been developed and tested. After sample preparation strontium is separated by extraction chromatography using Sr resin. Eluate is divided and transfered to LSC vial and filter paper by SrCO{sub 3} precipitation. A Hidex 300 SL TDCR liquid scintillation counter and Thermo Fisher low level proportional counter have been used. Chemical yield of Sr-85 tracer is determined by Gamma spectroscopy. Uncertainty budget, decision threshold and detection limit are calculated in accordance with GUM and ISO 11929.

  16. Accumulation of Sr90-Y90 by the developing spawn and larvae of Coregonus lavaretus L

    International Nuclear Information System (INIS)

    Kulikov, N.V.; Ozhegov, L.N.

    1976-01-01

    The results are given of laboratory experiments aimed at studying accumulation of Sr-90 and its daughter product Y-90 in the spawn and larvae of whitefish, a representative of autumn spawning fish species. The results are of interest for predicting possible consequences of radioactive contamination of water bodies and introduction of limits for permissible concentration of radioactive substances in aqueous medium. The accumulation coefficient of Sr-90 in the spawn reaches its maximum after the first twenty four hours of the experiment. Subsequently throughout the incubation period the accumulation coefficient (AC) of radionuclide does not undergo any changes being about 3. Besides it has been established that the equilibrium distribution of Sr-90 in whitefish spawn and the surrounding solution occurs during the first two hours. Unlike Sr-90 the accumulation of Y-90 occurs during the whole incubation period and results in the AC value of 80 by the end of the spawn incubation. The experiments aimed at determining the strength of radionuclide fixation by whitefish spawn and larvae revealed that both radionuclides in the spawn are rather mobile. Strontium separates fully from it during the first twenty four hours and yttrium marked by higher fixation strength transfers completely from the spawn into surrounding aqueous medium by the end of the experiment. In larvae the radionuclides are fixed stroger than in the spawn. The approximate radiation doses of embryos by the time of their hatching have been estimated. The estimates indicate that the total radiation dose of the spawn developing in aqueous solution of Sr-90 - Y-90 with 1.1 curie/0 -5 concentration is about 500 radl

  17. Measurements of Sr-90 radionuclide in Slovenian soils before and after Chernobyl accident

    International Nuclear Information System (INIS)

    Krizman, M.J.; Ambroz, S.

    2005-01-01

    Strontium-90 is a long-lived fission product (half life of 28,7 years) that is globally dispersed in the environment. It had been transported by air masses from the nuclear weapon tests sites in the period of 1951-1980 and also from Chernobyl (1986) and deposited elsewhere, especially over northern hemisphere. Contamination of surface layer (0-10 cm) of undisturbed soil in Slovenia was measured in the middle of the seventies (1973-75) and recently (2002). In parallel, long-lived radionuclide, Cs-137 was measured too, the second campaign was performed ten years after Chernobyl accident. Maps on Sr-90 and Cs-137 contents in soil were elaborated, showing different distributions of area contamination and different levels for a case of nuclear weapon tests and due to Chernobyl accident for both radionuclides. The past contamination from atmospheric nuclear tests Sr-90 and Cs-137 in Slovenian territory was characterised with the high values on western part of the country (with the exception of the coastal region) and typical values of 1,5-2 kBq/m 2 for Sr-90 and 3-4 kBq/m 2 for Cs-137. The Chernobyl accident raised the contamination with Cs-137 mostly in northwest part (Alpine region), with an average value of 20-25 kBq/m 2 for the country. Contamination with Sr-90 was much lower, the existing levels increased for about 0,2 kBq/m 2 . Recently measured levels of Sr-90 in the upper layer of soil hardly approach to 0,3 kBq/m 2 . (author)

  18. Complexation studies with 90Y from a novel 90Sr-90Y generator

    International Nuclear Information System (INIS)

    Venkatesh, M.; Pandey, U.; Banerjee, S.; Samuel, G.; Pillai, M.R.A.; Dhami, P.S.; Kannan, R.; Achuthan, P.V.; Chitnis, R.R.; Gopalakrishnan, V.; Ramanujam, A.

    2001-01-01

    Some features of a novel 90 Sr- 90 Y generator which employs supported liquid membrane (SLM) to separate carrier-free 90 Y from 90 Sr present in the high level waste of the spent fuel of reactor are described. After ascertaining the purity of 90 Y particularly with respect to 90 Sr breakthrough, its complexation was studied with a few oxo/aza donor ligands, such as DTPA, EDTMP, DOTA, TETA and a cyclic phosphonate, CTMP. These studies were primarily carried out to adjudge the quality of the 90 Y obtained from a novel 90 Sr- 90 Y generator and ascertain its usability for labelling biomolecules such as antibodies and peptides. The DOTA complexes are most stable at 37 C in human serum; they appear to be ideal bifunctional chelating agent for use in radioimmunotherapy with 90 Y. (orig.)

  19. The effect of 90Sr/90Y β-rays on benign prostatic hyperplasia

    International Nuclear Information System (INIS)

    Kong Xiangbo; Ma Qingjie; Gu Xinquan

    2004-01-01

    Objective: To investigate the effect of 90Sr/90Y β-rays on Benign prostatic hyperplasia (BPH). Methods: In order to carry out β intracavitary irradiation, the active area of the applicator was located into prostate gland section of urethra in 37 patients. The dose was controlled 30∼50Gy in each patient and the patients were observed for 3 months. Results: After two weeks 26 cases were remarkably improved, 7 cases were improved and 4 cases were ineffective. There was significant difference on MFR, PVR, I-PSS and volume of prostate gland before and after treatment (P<0.01=. Conclusion: Clinical tests indicate that the 90Sr/90Y prostatic hyperplasia applicators provide a safe, effective ,non-invasive and economical therapeutic method for BPH. It is especially applicable for old and high-risk patients. (authors)

  20. Determination of Sr-90 in environmental samples using solid phase extraction disk

    International Nuclear Information System (INIS)

    Zal U'yun Wan Mahmood

    2002-01-01

    A method is described for determination of Sr-90 in environmental samples using solid phase extraction disk (Empore TM Strontium Rad Disk) and GM counter. To determine the optimum condition for capacity of Empore TM Strontium Rad Disk, its characterization studies such as the effects Sr content, acidity (molarity) of acids, presence of Ca 2+ and other major ions (Na + , Mg 2+ etc), influence of interference (Pb and Bi) and others were carried out. An optimized the using of Empore TM Strontium Rad Disk for determination of Sr-90 was validated by application to environmental samples. Quantitative recoveries above 95%for Sr (stable) were recorded in 6M HCl condition. Typical environmental samples may contain an assortment of anionic and cationic species, but in general, Empore TM Strontium Rad Disk has enough capacity to effectively separate Sr for wide variety of aqueous solutions. Sr recovery in a matrix-free or the content of matrix less than 300 mg/sample is typically greater than 99% is reported in this research work. In particular, sample, which may contain interference such as Pb and Bi would require an addition separation step before processing to ensure an accurate measurement of Sr. In this research work, radiotracer 85 Sr was used to monitor the behavior of Sr and calculation its recovery. For analytical methods that can count Y-90, the Sr-90 activity/concentration in environmental sample was calculated. The concentration of Sr-90 in ash sample (Quality Controled Sample) of 276 ± 18 Bq/kg ash was determined from Y-90 activity. The relative percent difference of 1.1% was achievable for Empore TM Sr-Rad Disk methods when compared to the conventional method (fumed-HNO 3 method) - 279 ± 11 Bq/kg ash. (Author)

  1. The content of K-40, Sr-90 and Cs-137 in milk in Croatia

    International Nuclear Information System (INIS)

    Cesar, D.; Maracic, M.; Marovic, M.

    1996-01-01

    Milk is one of the most important foodstuffs. The children's diet is based on milk and quality of milk has always been a subject of concern and continuous control. investigation of radioactivity in milk in Croatia has been started in 1960. The samples of milk were collected daily in the towns Osijek, Zadar and Zagreb. In monthly samples specific Sr-90 activities were determined by radiochemical separation, and Cs-137 and K-40 by gammaspectrometric analysis. The values obtained are shown in Figures 1 and 2. The ratio of Sr-90, Cs-137 and K-40 specific activities was calculated as well as the ratio of their maximum prmissible levels. By the division of these two ratios the ratio of their effectiveness was calculated. On basis of the obtained data the following conclusions can be drawn: 1. The level of Sr-90 in milk had been increasing by 1964. in the period that followed Sr-90 was exponentially decreasing which indicated that its content in milk originated mostly from nuclear weapon tests carried out in the period 1945 to 1962. 2. The content of Cs-137 in milk in Croatia has been exponentially decreasing since 1965 in spite of a great increase in 1986. This points to the fact that Cs-137 content in milk originated mostly from nuclear weapon tests in the period 1945 to 1962, and to a lesser degree from Chernobyl nuclear accident in 1986. 3. The content of K-40 in milk has not changed significantly over the investigated period, but its level was not the same at all sampling locations. 4. The ratio of the impact of investigated radionuclides on man obtained from the samples of milk in the territory of Croatia is: 90 Sr : 137 Cs : 40 K = 1 : 2 : 5. (author)

  2. Enrichment and determination of small amounts of 90Sr/90Y in water samples

    International Nuclear Information System (INIS)

    Mundschenk, H.

    1979-01-01

    Small amounts of 90 Sr/ 90 Y can be concentrated from large volumes of surface water (100 l) by precipitation of the phosphates, using bentonite as adsorber matrix. In the case of samples containing no or nearly no suspended matter (tap water, ground water, sea water), the daughter 90 Y can be extracted directly by using filter beds impregnated with HDEHP. The applicability of both techniques is demonstrated under realistic conditions. (orig.) 891 HP/orig. 892 MKO [de

  3. The Sr-90 waste treatment by using sodium carbonat as a carrier

    International Nuclear Information System (INIS)

    Suroto.

    1978-01-01

    The coprecipitation processes of the Sr-90 wastes, with the lime-soda treatment, that followed by the mechanism of isomorph substitution, have shown a good performance in pH 7-8. Excess of natrium carbonate in the lime-soda treatment caused the decrease of the decontamination factor. This research shows that the coprecipitation by calcite-phosphate was better than the lime-soda treatment. (author)

  4. Dosimetric comparison of electron beam and 90Sr+90Y applicator for keloids treatment

    International Nuclear Information System (INIS)

    Coelho, Talita S.; Tada, Ariane; Antonio, Patricia L.; Yoriyaz, Helio; Fernandes, Marco A.R.

    2009-01-01

    Studies have been shown that among several methods that have been used for the treatment of keloids the surgical excision followed by the adjuvant radiotherapy presents the lowest relapsed rate of the injury. In this work a comparative dosimetric study has been performed using a 4 MeV electron beam from a Varian Clinac 2100C linear accelerator at the radiotherapy service of the Hospital das Clinicas of UNESP-HC, Botucatu-SP and an Amershan 90 Sr+ 90 Y brachytherapy applicator with 1491 MBq of activity. Percentage depth dose curves from ionization chamber measurements and through Monte Carlo simulation have been obtained and compared. Dose measurements have been obtained using parallel plates ionization chamber (Esradin A12) and extrapolation mini-chamber developed at IPEN. The dose calculations have been obtained using the well-known Monte Carlo radiation transport code MCNP-4C. Maximum dose differences obtained between measured/calculated values for 90 Sr+ 90 Y applicator and for the electron beam were, respectively: 7.8 % and 8.0%. The profiles of the depth and superficial tissue dose distribution produced by the electron beam revealed themselves flatter and more homogeneous than those produced by the 90 Sr+ 90 Y applicator, especially to wider fields, which cannot be obtained with beta therapy applicators because of their geometric limitations. In conclusion this present work has shown that 90 Sr+ 90 Y applicators could be efficient for small and very superficial lesions but in most cases electron beam sources are more adequate especially for large and deeper lesions. (author)

  5. Using Cherenkov Counting For Fast Determination of 90Sr/90Y Activity in Milk

    International Nuclear Information System (INIS)

    Tsroya, S.; Dolgin, B.; German, U.; Pelled, O.; Alfassi, Z. B.

    2014-01-01

    90Sr is one of the main long-lived fission products, and it is transferred into human body primarily by food, with milk being a substantial contributor. Due to its biochemical similarity to calcium, most strontium is efficiently incorporated into bone tissues. 90Sr is characterized by a long physical half life (28.8 y) and decays by beta particles with an Emax of 0.546 MeV to 90Y. This daughter isotope has a half life of 64 h and decays into 90Zr by beta particles with an Emax of 2.284 MeV. The milk components produce a high turbidity and light attenuation, causing a significant decrease of the counting efficiency in liquid scintillation counting (LSC) systems, mostly used for beta emitters detection. Most methods proposed in the past are time-consuming, as they are based on several stages of chemical and physical treatments, including precipitation, ashing, ion exchange and extraction (Wikins et al., 1984, Porter et al, 1961, Kimura et al., 1979). When measuring 90Sr/90Y activity by Cherenkov counting, most of the Cherenkov radiation is produced by 90Y (about 98.6%), due to the much higher energy of its beta particles relative to these from 90Sr. The counting efficiency varies strongly with color quenching, at a greater extent than in standard liquid scintillation counting (L'Annunziata, 2012), and therefore the quench correction is critical. The ‘‘external source area ratio’’ (ESAR) quench correction method was applied to measure 90Sr/90Y activities in aqueous samples with a wide range of quenching levels (Tsroya et al., 2009). This method was proved to be superior to all other quench correction methods (Tsroya et al., 2012) and is applicable also for determination of 90Sr/90Y in human urine (Tsroya et al., 2013). In the present work the applicability of the ESAR method to measurement of 90Sr/90Y activities in milk and some of its products was investigated

  6. Preliminary concentration and determination of Sr-90 in natural and waste water of Kursk region

    International Nuclear Information System (INIS)

    Basargin, N.N.; Rozovskij, Yu.G.; Grebennikova, R.V.; Salikhov, V.D.

    2001-01-01

    Synthesis and study of cheating sorbents containing functional analytical ortho-oxy-aza-ortho'-sulfonyl group are presented. Physicochemical properties of sorbents and chemisorption of Sr and Sr 90 are studied. A rapid method of preliminary concentration with subsequent atomic absorption and radiometric determination of Sr in natural and waste water is proposed. Samples of aqua-objects of Kursk region were analyzed using developed method. The results of radiometric investigations into control of strontium-90 content in cooling systems of Kursk NPP, waste waters, waters of Sejm river testifies higher values of concentration in the april - september period [ru

  7. Some analytical aspects about determination of Sr89 and Sr90 in environmental samples

    International Nuclear Information System (INIS)

    Gasco, C.; Alvarez Garcia, A.

    1988-01-01

    Some problems about determination of Sr 89 and Sr 90 in environmental samples have been studied. The main difficulties are due to the wide range in the concentration of their components and the contents of chemical and radiochemical interferent elements. The behaviour of strontium on ion exchange resin has been described by some experiments in various media: aqueous media, calcium concentration and matrix variable. The differences of alkaline-earth nitrate and carbonate solubilities have been analyzed in nitric acid. The chemical recovery in environmental samples has been determined. (Author)

  8. Determination of Cs-137, Sr-89 and Sr-90 and gamma spectroscopy of water samples from the Danube River

    International Nuclear Information System (INIS)

    Tschurlovits, M.

    1980-01-01

    Radioactivity concentration of Cs-137, Sr-90, Sr-89, Co-60 and K-40 from the Danube water in the period of 1977-1979. A few systematic changes in the radioactivity concentrations were observed and presented. (author)

  9. Depth dose curves from 90Sr+90Y clinical applicators using the thermoluminescent technique

    International Nuclear Information System (INIS)

    Antonio, Patricia L.; Caldas, Linda V.E.; Oliveira, Mercia L.

    2009-01-01

    The 90 Sr+ 90 Y beta-ray sources widely used in brachytherapy applications were developed in the 1950's. Many of these sources, called clinical applicators, are still routinely used in several Brazilian radiotherapy clinics for the treatment of superficial lesions in the skin and eyes, although they are not commercialized anymore. These applicators have to be periodically calibrated, according to international recommendations, because these sources have to be very well specified in order to reach the traceability of calibration standards. In the case of beta-ray sources, the recommended quantity is the absorbed dose rate in water at a reference distance from the source. Moreover, there are other important quantities, as the depth dose curves and the source uniformity for beta-ray plaque sources. In this work, depth dose curves were obtained and studied of five dermatological applicators, using thin thermoluminescent dosimeters of CaSO 4 :Dy and phantoms of PMMA with different thicknesses (between 1.0 mm and 5.0 mm) positioned between each applicator and the TL pellets. The depth dose curves obtained presented the expected attenuation response in PMMA, and the results were compared with data obtained for a 90 Sr+ 90 Y standard source reported by the IAEA, and they were considered satisfactory. (author)

  10. 90Sr-90Y radionuclide generator based on ionex chromatography. Part 1 - project

    International Nuclear Information System (INIS)

    Miler, V.; Budsky, F.; Malek, Z.

    2003-09-01

    This part contains a proposal for the generator column design, materials to be used (chemicals, ionexes) and technological procedures. The proposal was inspired by the 90 Sr- 90 Y generator operated by Zfk Rossendorf. The aim was to develop and launch a generator for the preparation of carrier-free 90 Y in the form of [ 90 Y] chloride solution in dilute hydrochloric acid. The separation of Y from Sr is based on ionex chromatography by sorbing the two radionuclides on a catex. While Sr remains sorbed, 90 Y is eluted with lithium citrate. During this process, 90 Y is bonded in a citrate complex which, having a negative charge, is subsequently trapped by an anex. A guard column is inserted before the anex column to trap any traces of 90 Sr. 90 Y is eluted from the anex in the yttrium chloride form by using dilute hydrochloric acid. The product from the generator can be used for the preparation of [ 90 Y] - Fe colloid injection or [ 90 Y] - yttrium citrate injection for intra-articular application or for the development of monoclonal antibodies and peptides

  11. Method for rapid screening analysis of Sr-90 in edible plant samples collected near Fukushima, Japan

    International Nuclear Information System (INIS)

    Amano, Hikaru; Sakamoto, Hideaki; Shiga, Norikatsu; Suzuki, Kaori

    2016-01-01

    A screening method for measuring 90 Sr in edible plant samples by focusing on 90 Y in equilibrium with 90 Sr is reported. 90 Y was extracted from samples with acid, co-precipitated with iron hydroxide, and precipitated with oxalic acid. The dissolved oxalate precipitate was loaded on an extraction chromatography resin, and the 90 Y-enriched eluate was analyzed by Cherenkov counting with a TDCR liquid scintillation counter. 90 Sr ( 90 Y) concentration was determined in plant samples collected near the damaged Fukushima Daiichi Nuclear Power Plants with this method. - Highlights: • A screening method for measuring 90 Sr in edible plant samples by focusing on 90 Y in equilibrium with 90 Sr is reported. • 90 Y was extracted from samples with acid, co-precipitated with iron hydroxide, and precipitated with oxalic acid. • The dissolved oxalate precipitate was loaded on an extraction chromatography resin. • 90 Y-enriched eluate was analyzed by Cherenkov counting with a TDCR liquid scintillation counter. • 90 Sr ( 90 Y) concentration was determined in edible plant samples collected near the damaged Fukushima Daiichi NPPs with this method.

  12. Thermoluminescent dosimetry of beta radiations of 90 Sr/ 90 Y using ZrO2: Eu

    International Nuclear Information System (INIS)

    Olvera T, L.; Azorin N, J.; Barrera S, M.; Soto E, A.M.; Rivera M, T.

    2005-01-01

    In this work the results of studying the thermoluminescent properties (TL) of the doped zirconium oxide with europium (ZrO 2 : Eu 3+ ) before beta radiations of 90 Sr/ 90 Y are presented. The powders of ZrO 2 : Eu 3+ were obtained by means of the sol-gel technique and they were characterized by means of thermal analysis and by X-ray diffraction. The powders of ZrO 2 : Eu 3+ , previously irradiated with beta particles of 90 Sr/ 90 Y, presented a thermoluminescent curve with two peaks at 204 and 292 C respectively. The TL response of the ZrO 2 : Eu 3+ as function of the absorbed dose was lineal from 2 Gy up to 90 Gy. The fading of the information of the ZrO 2 : Eu 3+ was of 10% the first 2 hours remaining almost constant the information by the following 30 days. The ZrO 2 doped with the (Eu 3+ ) ion it was found more sensitive to the beta radiation that the one of zirconium oxide without doping (ZrO 2 ) obtained by the same method. Those studied characteristics allow to propose to the doped zirconium oxide with europium like thermoluminescent dosemeter for the detection of the beta radiation. (Author)

  13. Development of a dosimetric system for 90Sr + 90Y betatherapy applicators

    International Nuclear Information System (INIS)

    Coelho, Talita Salles

    2010-01-01

    The 90 Sr+ 90 Y applicators, used in betatherapy for prevention of keloids and pterigium, are imported and many times their dosimetric features are shown only in an illustrated form by the manufacturers. The exhaustive routine of the medical physicists in the clinic do not make possible the accomplishment of procedures for the confirmation of these parameters. This work presents the development of a methodology for the dosimetry of 90 Sr+ 90 Y betatherapy applicators. The Monte Carlo code MCNP5 was used for the simulation of the percentage depth dose curves and dose distribution profiles produced by these applicators. The experimental measurements of the radial and axial radiation attenuation, have been done with a mini-extrapolation chamber, thermoluminescent dosimeters and radiographic films. The experimental results have been compared with the simulated values. Both percentage depth dose curves and the radial dose profiles, the theoretical and the experimental ones, have presented good agreement, which may validate the use of the MCNP5 for these simulations, confirming the viability of the usage of this method in procedures of beta emitter sources dosimetry. (author)

  14. Thermoluminescent dosimetry of beta radiations of 90 Sr/ 90 Y using amorphous ZrO2

    International Nuclear Information System (INIS)

    Rivera M, T.; Olvera T, L.; Azorin N, J.; Barrera R, M.; Soto E, A.M.

    2005-01-01

    In this work the results of studying the thermoluminescent properties (Tl) of the zirconium oxide in its amorphous state (ZrO 2 -a) before beta radiations of 90 Sr/ 90 Y are presented. The amorphous powders of the zirconium oxide were synthesized by means of the sol-gel technique. The sol-gel process using alkoxides like precursors, is an efficient method to prepare a matrix of zirconium oxide by hydrolysis - condensation of the precursor to form chains of Zr-H 3 and Zr-O 2 . One of the advantages of this technique is the obtention of gels at low temperatures with very high purity and homogeneity. The powders were characterized by means of thermal analysis and by X-ray diffraction. The powders of ZrO 2 -a, previously irradiated with beta particles of 90 Sr/ 90 Y, presented a thermoluminescent curve with two peaks at 150 and 257 C. The dissipation of the information of the one ZrO 2 -a was of 40% the first 2 hours remaining constant the information for the following 30 days. The reproducibility of the information was of ± 2.5% in standard deviation. The studied characteristics allow to propose to the amorphous zirconium oxide as thermoluminescent dosemeter for the detection of beta radiation. (Author)

  15. Transfer coefficient study of Sr-90 in the soil-grass-milk chain for Cuba

    International Nuclear Information System (INIS)

    Zerquera, J. T.; Sarria P, R.

    1996-01-01

    One of the most important problems in modern radioecology is the lack of able information about the features of radionuclide migration in tropical and subtropical environment. The development of nuclear energy and the enhancing in the applications of nuclear techniques in those latitudes indicate that studies in this area are necessary. Cuba is carrying out studies on radioecological characterization of the principal food chains in the country. One of the objectives of these studies is to define the values of the transfer coefficients to be used in the evaluation programs for the assessment of the radiological impact of practices which involve ionizing radiation. This paper shows the results obtained in the determination of Sr-90 transfer coefficients in soil-grass-milk food chain in 'La Quebrada', a place near the Havana City where an important part of the milk that the citizens consume is produced. Transfer coefficients for Sr-90 were calculated on the basis of data collected during 5 years in the region. Soil-grass transfer coefficients are in the range 0.18-5 while grass-milk coefficients are in the range of 1.2x10 -4 - 6x10 -3 day/L. These values are in accordance with values reported by other authors in the literature. (authors). 4 refs., 2 tabs

  16. A data acquisition system for indentification of 90Sr/ 90Y in environmental samples

    International Nuclear Information System (INIS)

    Medin, G.; Brajnik, D.; Starcic, M.; Stanovnik, A.

    1996-01-01

    Often, in the stage of research and development of new techniques for detection of ionizing radiation, elaborate electronic systems are required. In this paper, we describe the relatively complex detector and electronic system used for a relatively simple but nevertheless demanding measurement of the beta emitting radionuclides 90 Sr/ 90 Y in environmental samples. The detection limit of 1 Bq in a thin, disc-shaped sample, was obtained by careful elimination of background. Contribution of other radionuclides in the sample, were eliminated or at least considerably reduced by using a silica aerogel as Cherenkov radiator and thin multiwire chamber in coincidence. Cosmic ray signals were reduced by large scintillation counters in anticoincidence. Finally, persisting RF pick-up signals were eliminated by using the signal from a wire antenna and identical MWPC preamplifier and discriminator for a veto to the master coincidence. For each accepted event, both timing and pulse height information was recorded with a personal computer

  17. Preliminary study of zeolite-pva composite application in removal of SR-90

    International Nuclear Information System (INIS)

    Las, Thamzil; Zamroni, Huzen; Sugiarto; Darsono

    1998-01-01

    Zeolite-PAN composite was prepared by contacting the purified Bayah and Lampung zeolites with poly-vinyl alcohol binder and cured by using Gamma-ray of Co-60 at various doses, i.e., 10, 20, 30 dan 40 kGray with dose rates 7.5 kGy/hour. Zeolite-PAN composites were treated with solution containing Sr-90 up to 5 days and the Sr sorption was measured by Liquid Scintillation Counter for determination of their sorption efficiencies. The result obtained that, zeolite-PAN composites were shown high sorption efficiencies on the composites zeolite-PVA which was formulated from 20% zeolite, irradiated by 40 kGy and obtained the sorption efficiency of 94% with the Kd values similar to the purified zeolites. (author)

  18. Changes induced to eye lens membrane characterization after treatments with beta radiation from Sr90

    International Nuclear Information System (INIS)

    El-Refaei, F.M.; Morris, M.; Gamal, M.M.; Fadel, M.A.

    1994-12-01

    The effect of β-particles on Na + and k + content, Na + -k + ATPase and histopathological changes of cell membrane were studied in the present work. One of the two eyes of New Zealand rabbits from both sexes were irradiated with β-particles from Sr 90 source to 10, 20, 40 and 60 Gy. The effect of β-particles on lens membrane after 3 months of exposure to 20 and 60 Gy was also studied. The results indicated that the treated and untreated eyes suffered pronounced injuries which deduced from the distribution of ATPase in comparison with the normal control which showed a decrease (reached 52%). As well as uncontrolled transport of the Na + and k + through the membrane and injuries appeared in the histopathological studies. (author). 12 refs, 15 figs, 4 tabs

  19. Mobility of Cs137 and Sr90 in organic soils and its control

    International Nuclear Information System (INIS)

    Rovdan, E.

    2002-01-01

    In the management of the radionuclide contaminated areas and application of a countermeasure strategy for reduction of both the external and internal doses to the population it is extremely important to know the environmental mechanisms governing the behaviour of radionuclides in soil ecosystems. The purpose of work is to investigate by means of laboratory, field experiments and mathematical modelling the mechanisms and dynamics of radionuclide transfer in the organic soil to propose measures for control their mobility. The Chernobyl radionuclides behaviour was studied for ameliorated peat-mire soil (peat deposit Pogonyanskoye, 21 km off the ChNPP). To control the mobility of radionuclides in soils the characteristics of the migration and sorption of Cs 137 and Sr 90 in sedge peat, quartz sand, bentonite, kaolin, sapropel as well as the electrolytes impact upon the radionuclide behaviour have been experimentally investigated

  20. Characterization of an extrapolation chamber in a 90Sr/90Y beta radiation field

    International Nuclear Information System (INIS)

    Oramas Polo, I.; Tamayo Garcia, J. A.

    2015-01-01

    The extrapolation chamber is a parallel plate chamber and variable volume based on the Bragg-Gray theory. It determines in absolute mode, with high accuracy the dose absorbed by the extrapolation of the ionization current measured for a null distance between the electrodes. This camera is used for dosimetry of external beta rays for radiation protection. This paper presents the characterization of an extrapolation chamber in a 90 Sr/ 90 Y beta radiation field. The absorbed dose rate to tissue at a depth of 0.07 mm was calculated and is (0.13206±0.0028) μGy. The extrapolation chamber null depth was determined and its value is 60 μm. The influence of temperature, pressure and humidity on the value of the corrected current was also evaluated. Temperature is the parameter that has more influence on this value and the influence of pressure and the humidity is not very significant. Extrapolation curves were obtained. (Author)

  1. Development of methodology for the preparation of 90Sr-90Y generators

    International Nuclear Information System (INIS)

    Barrio, Graciela; Osso, Joao A.

    2008-01-01

    Among the more attractive radionuclides for therapeutic applications is yttrium-90. Its relatively short half-life (64.0h), maximum beta energy (2.28 MeV) make it well suited for a variety of applications, ranging from the radiolabeling of antibodies for tumor therapy to the production of radiolabeled particles for the treatment of liver malignancies. Yttrium-90 results from the decay of strontium-90 (28y) and decays to a stable daughter ( 90 Zr) and has no γ-ray, what makes this nuclide very attractive for therapeutic applications. 90 Y is produced through a generator system, from the decay of 90 Sr. Before it can safely employed in clinical applications, the 90 Y must be made essentially free of 90 Sr, an isotope known to cause bone marrow suppression. The objective of this work is to develop 90 Sr- 90 Y generator using a cation exchange method resin. This method was the most effects for the separation of 90 Y and 90 Sr. 90 Sr is strongly adsorbed in the resin and 90 Y is eluted in 0.003M EDTA. The generator efficiency and radionuclide control results were also evaluated using 85 Sr (γ-ray emitter). The generator was developed, using AG 50WX8 exchange resin converted to the Na+ form and then conditioned with 0.003M EDTA solution (pH=4.8). The quality control showed that 90 Y had no 8 '5Sr impurities, inside the detection limits of the detector. The results presented here are very promising, showing that the methodology is able to separate the two nuclides, with low quantities of 90 Sr impurity and satisfactory elution yields. (author)

  2. Dynamics of Sr-90 content in environmental objects of Ukrainian-Byelorussian Poles'e after the Chernobyl' NPP accident

    International Nuclear Information System (INIS)

    Likhtarev, I.A.; Kajro, I.A.; Shandala, N.K.; Los', I.N.; Repin, V.S.; Gul'ko, G.M.; Chepurnoj, N.I.; Berkovskij, V.B.; Tsygankov, N.Ya.; Pozhivalova, S.B.

    1990-01-01

    Ecological, dosemetric and radiation-sanitary aspects of the problem of Sr-90 contamination of environmental objects and food in Ukrainian-Byelorussian Poles'e in three-year period after the Chernobyl' NPP accident are discussed. Analysis of the materials collected shows that efficient equivalent radiation doses from Sr-90 being intaken into the region of Urkainian-Byelorussian Poles'e after the Chernobyl' NPP accident expected to be gained during 70 years of critical population group life are not exceed 1 mSv; this value is by the factor of 10-100 smaller than the dose values connected with radioactive cesium. The problem of environment contamination with Sr-90 transforms from hot radiation-sanitary situation into more quiet radioecological one

  3. Therapeutic result of radioactive nuclide 90Sr/90Y treatment in patients with benign prostatic hypertrophy (BPH)

    International Nuclear Information System (INIS)

    Chen Hanchao; Li Yuying

    2008-01-01

    Objective: To study the effect of radioactive nuclide 90 Sr/ 90 Y treatment in patients with benign prostatic hypertrophy (BPH). Methods: Sixty patients with BPH were treated with a course of transurethral radioactive nuclide 90 Sr/ 90 Y therapy. Results: The severity of BPH was assessed with four parameters: maximal flow rate (MFR), volume of residual urine (VRU), international prostatic symptom score (IPSS) and volume (size) of prostate. In this series, the total effective rate was 93.33% with no treatment- related mortality. Favorable changes of the parameters after a course of radioactive nuclide therapy were significant. Conclusion: Radioactive nuclide 90 Sr/ 90 Y therapy for patients with BPH was safe, easily performed and quite effective. This procedure is worth popularizing in appropriate patients. (authors)

  4. Study of the FeSO4 effectivity as a carrier for Sr 90 separation from liquid strontium waste

    International Nuclear Information System (INIS)

    Djoko Sardjono.

    1978-01-01

    The effectiveness of FeSO 4 as a carrier for removing Sr 90 from the strontium liquid waste is studied. This research is concerned with the study of the capacity of FeSO 4 as a carrier for removing Sr 90 from the strontium liquid waste. The method being used in the experiment is an application of the coprecipitation method to reduce the activity of the strontium liquid waste to a certain activity that is safe enough to be discharged in the environment. (author)

  5. Transfer of Cs137 and Sr90 from contaminated soil to some crops in Syria

    International Nuclear Information System (INIS)

    Yassine, T.; Al-Oudat, M.; Othman, I.; Sharaneq, A.

    1998-07-01

    Transfer factors of Cs 137 and Sr 90 from contaminated soil to some common crops were investigated under field conditions for a period of three years. The results showed large variations in transfer factor values (expressed as Bq/g of dried material to Bq/g of soil) among crops. The highest values for both radionuclides were found in green vegetables up to 0.067 for 137 Cs and 3.78 for 90 Sr, whereas cereal grains had the lowest values. The transfer factor values of 90 Sr were generally much higher than that of 137 Cs for the same crop by factors ranged up to 263. The values for both radionuclides were found to be in the lower limits of that obtained in other areas. This was attributed to the effect of several parameters such as high ph, low organic matter and high exchangeable potassium and calcium in the soil. The transfer factor values of 137 Cs were decreased from the first year to the third year for most crops by factor up to 4, while this approach was not found for 90 Sr. (author)

  6. Influence of cultivar specificity on Cs 137 and Sr 90 accumulation in green part and seeds of soya plants

    International Nuclear Information System (INIS)

    Gutseva, G.Z.; Goloveshkin, V.V.

    2008-01-01

    As a result of investigation carried out with soya sorts zoned in Belarus on soddy-podsolic sandy soils contaminated as a result of Chernobyl accident with Cs 137 and Sr 90 a high sort selectivity in accumulation of the radionuclides is established. (authors)

  7. Validation of Extraction Paper Chromatography as a Quality Control Technique for Analysis of Sr-90 in Y-90 Product

    International Nuclear Information System (INIS)

    Nipavan, Poramatikul; Jatupol, Sangsuriyan; Wiranee, Sriweing

    2009-07-01

    Full text: Yttrium-90 (Y-90) is a daughter product of strontium-90 (Sr-90). It is specified that there should be less than 2 micro-curie of Sr-90 in Y-90 radiopharmaceuticals. Since both nuclides are beta emitting and there is always a contamination of Y-90 in Sr-90 sample, validation of the analytical method is necessary. In this study, commercial Y-90 and Sr-85 (a gamma emitting isotope of strontium) were used as daughter and mother nuclides, respectively. Extraction paper chromatography technique and its efficient validation method were investigated. Bis-(2-ethylhexyl) diphosphonate was dropped at the origin of chromatography paper and air dried prior to sample drops. Validation of the separation was done by radio-chromatography scanning of the chromatography paper. Their energy spectra were identified in the spectra mode of Packard Cobra II automatic gamma counter, which can differentiate a pure gamma, a pure beta and a mixture of beta and gamma nuclides. Results showed that yttrium acetate remained fixed at the origin of the chromatography paper while strontium acetate moved to the solvent front when developed in saline. In conclusion, the extraction paper chromatography technique can effectively separate Sr-90 from Y-90

  8. Dosimetry characterization of the commercial CaF2 for beta radiation of 90Sr + 90Y

    International Nuclear Information System (INIS)

    Oliveira, Mercia L.; Caldas, Linda V.E.

    2003-01-01

    This work studies the dosimetric characteristics of the CaF 2 commercial dosimetry for detection of 90 Sr + 90 Y beta radiation for using in the calibration of flat and concave appliers. Were determined the repetitiousness and linearity of answers of the samples, and their calibration curves

  9. Methods of determination of the dynamics of mobile form of Cs 137 and Sr 90 in soils within Chernobyl catastrophe

    International Nuclear Information System (INIS)

    Osetskaya, V.V.

    2002-01-01

    In this paper the impact of the most significant mechanisms of migration of radionuclides are considered: diffusion of free and absorbed ions, migration colloid atoms transmission etc. on the dynamics of mobile forms of Cs 137 and Sr 90 in soils

  10. Depth dose distribution in the water for clinical applicators of 90Sr + 90Y, with a extrapolation mini chamber

    International Nuclear Information System (INIS)

    Antonio, Patricia de Lara; Caldas, Linda V.E.; Oliveira, Mercia L.

    2009-01-01

    This work determines the depth dose in the water for clinical applicators of 90 Sr + 90 Y, using a extrapolation mini chamber developed at the IPEN, Sao Paulo, Brazil, and different thickness acrylic plates. The obtained results were compared with the international recommendations and were considered satisfactory

  11. Interaction of Sr-90 with site candidate soil for demonstration disposal facility at Serpong

    Energy Technology Data Exchange (ETDEWEB)

    Setiawan, Budi, E-mail: bravo@batan.go.id [Radwaste Technology Center-National Nuclear Energy Agency, PUSPIPTEK, Serpong-Tangerang 15310 (Indonesia); Mila, Oktri; Safni [Dept. of Chemistry, Fac. of Math. and Nat. Sci., Andalas University, Kampus Limau Manis, Padang-West Sumatra 25163 (Indonesia)

    2014-03-24

    Interaction of radiostrontium (Sr-90) with site candidate soil for demonstration disposal facility to be constructed in the near future at Serpong has been done. This activity is to anticipate the interim storage facility at Serpong nuclear area becomes full off condition, and show to the public how radioactive waste can be well managed with the existing technology. To ensure that the location is save, a reliability study of site candidate soil becomes very importance to be conducted through some experiments consisted some affected parameters such as contact time, effect of ionic strength, and effect of Sr{sup +} ion in solution. Radiostrontium was used as a tracer on the experiments and has role as radionuclide reference in low-level radioactive waste due to its long half-live and it's easy to associate with organism in nature. So, interaction of radiostrontium and soil samples from site becomes important to be studied. Experiment was performed in batch method, and soil sample-solution containing radionuclide was mixed in a 20 ml of PE vial. Ratio of solid: liquid was 10{sup −2} g/ml. Objective of the experiment is to collect the specific characteristics data of radionuclide sorption onto soil from site candidate. Distribution coefficient value was used as indicator where the amount of initial and final activities of radiostrontium in solution was compared. Result showed that equilibrium condition was reached after contact time 10 days with Kd values ranged from 1600-2350 ml/g. Increased in ionic strength in solution made decreased of Kd value into soil sample due to competition of background salt and radiostrontium into soil samples, and increased in Sr ion in solution caused decreased of Kd value in soil sample due to limitation of sorption capacity in soil samples. Fast condition in saturated of metal ion into soil samples was reached due to a simple reaction was occurred.

  12. Studies on the radioactive contamination due to nuclear detonations VI. Theoretical analysis of the radioactive contamination due to Sr90 and Cs137

    International Nuclear Information System (INIS)

    Nishiwaki, Yasushi

    1961-01-01

    In view of the considerably different values of the fallout rate and the cumulative deposition of Sr 90 and Cs 137 reported from different parts of Japan, the author attempted to estimate the ranges of the fallout rate and of the cumulative deposition of Sr 90 and Cs 137 for different parts of Japan based upon the theoretical consideration

  13. On interrelations of Sr90 contents in the soil-forage plants-milk chain under natural conditions

    International Nuclear Information System (INIS)

    Chupka, Sh.

    1975-01-01

    Observations were made on 150 soil damples, 58 plant root samples, and 98 milk samples. Four types of soil from western Slovakia (chernozem, brown, sandy, and carbonate) and two types of plant roots (alfalfa and mixed grasses) were studied. A relation was shown between the type of soil, its physico-chemical properties, the Sr 90 accumulation in the plant roots, and the degree of contamination of milk by this radionuclide. (V.A.P.)

  14. Comparative analysis of species-based specificity in Sr 90 and Cs 137 accumulation demonstrated by ligneous plant forest communities

    International Nuclear Information System (INIS)

    Martinovich, B.S.; Vlasov, V.K.; Sak, M.M.; Golushko, R.M.; Afmogenov, A.M.; Kirykhin, O.V.

    2004-01-01

    The authors provided field-proven study of Sr 90 and Cs 137 absorption activity demonstrated by Pinus silvestris L.; Piceae abies (L.) Roth.; Quercus rubra L.; Acer platanoides L.; Betula pendula Roth.; Tilia cordata Mill, under identical habitat conditions. The above plants were examined after 5-year growth period on radionuclide-contaminated soil. To a great extent, such parameters as radionuclide accumulation in experimental plants and accumulation activity were determined by the plants' bio-ecological properties. (Authors)

  15. Fast and sensitive determination of Sr-90 and SR-89 activity in milk by ion-chromatography and liquid scintillation

    International Nuclear Information System (INIS)

    Figueiredo, V.; Herrmann, A.

    1992-01-01

    A method for fast and exact determination of both strontium isotopes in milk and other foodstuffs, combination high performance ion chromatographic separation with by liquid scintillation counting, which enables the desired results to be obtained with very satisfactory precision and reproducibility within 24 hours, has been developed. The lowest detectable activity lies by 3 Bq/liter for Sr-90 and 1 Bq/liter for Sr-89 which is satisfactory for assessing a situation in a time crisis. (author)

  16. The dynamics of Cs-137 and Sr-90 pollution of surface water systems of Belarus of Chernobyl origin

    Energy Technology Data Exchange (ETDEWEB)

    Datskevich, P.O.; Dolgov, V.M.; Golikov, Yu.N.; Zemskov, V.N.; Komissarov, F.D.; Khvaley, O.D.

    1995-12-31

    The Belarus water ecosystems have been the object of investigation concerning currents and reservoirs affected by the Chernobyl APS catastrophe. The radio monitoring of samples of water systems components was implemented with the use of modern methods of radiochemistry and ionizing radiations registration. The factual material of water ecosystems sites observation presented its analysis is done and the regularities, tendencies and anomalies are revealed in the Cs-137 and Sr-90 distribution, transport and accumulation for water components.

  17. Country report: Thailand. Development of Sr-90/ Y-90 Generator and Development of Radiopharmaceuticals Using Y-90

    International Nuclear Information System (INIS)

    Nipavan, Poramatikul

    2010-01-01

    The research project has been conducted at Thailand Institute of Nuclear Technology in accordance to the 1st RCM plan during the IAEA meeting in Warsaw. The objectives of the project include the following 5 specific aims: 1. Development of Sr-90/Y-90 ion-exchange chromatography generator 2. Development of Sr-90/Y-90 extraction chromatography generator 3. Development of quality control technique 4. Development of herapeutic radiopharmaceuticals Y-90 particulates/colloids 5. Development of Re-188 DMSA–bis-phosphonates Currently we have achieved specific aims 1 to 3. The specific aims 4 and 5 are during investigation. For specific aim 4, we are during the process to extract high purity 90 Y from 90 Sr/ 90 Y generator that will yield the starting 90 Y for the production of Y-90 particulates and colloids. For the 5 th specific aim, we are on hold to receive the starting agent, bis- Phosphonates, from Dr. Blower group. Therefore, this progress report will cover our work focusing on specific aims 1 to 3

  18. Tritium determination in water

    International Nuclear Information System (INIS)

    Gavini, Ricardo M.

    2008-01-01

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

  19. WE-D-BRE-01: A Sr-90 Irradiation Device for the Study of Cutaneous Radiation Injury

    Energy Technology Data Exchange (ETDEWEB)

    Dorand, JE; Bourland, JD [Department of Radiation Oncology and Department of Physics, Wake Forest University, Winston-Salem, NC (United States); Burnett, LR [KeraNetics, LLC, Winston-Salem, NC (United States); Tytell, M [Department of Neurobiology and Anatomy, Wake Forest School of Medicine, Winston-Salem, NC (United States)

    2014-06-15

    Purpose: To determine dosimetric character for a custom-built Sr-90 beta irradiator designed for the study of Cutaneous Radiation Injury (CRI) in a porcine animal model. In the event of a radiological accident or terrorist event, Sr-90, a fission by-product, will likely be produced. CRI is a main concern due to the low energy and superficial penetration in tissue of beta particles from Sr-90. Seven 100 mCi plaque Sr-90 radiation sources within a custom-built irradiation device create a 40 mm diameter region of radiation-induced skin injury as part of a larger project to study the efficacy of a topical keratin-based product in CRI healing. Methods: A custom-built mobile irradiation device was designed and implemented for in vivo irradiations. Gafchromic™ EBT3 radiochromic film and a PTW Markus chamber type 23343 were utilized for dosimetric characterization of the beta fluence at the surface produced by this device. Films were used to assess 2-dimensional dose distribution and percent depth dose characteristics of the radiation field. Ion chamber measurements provided dose rate data within the field. Results: The radiation field produced by the irradiation device is homogeneous with high uniformity (∼5%) and symmetry (∼3%) with a steep dose fall-off with depth from the surface. Dose rates were determined to be 3.8 Gy/min and 3.3 Gy/min for film and ion chamber measurements, respectively. A dose rate of 3.4 Gy/min was used to calculate irradiation times for in vivo irradiations. Conclusion: The custom-built irradiation device enables the use of seven Sr-90 beta sources in an array to deliver a 40 mm diameter area of homogeneous skin dose with a dose rate that is useful for research purposes and clinically relevant for the induction of CRI. Doses of 36 and 42 Gy successfully produce Grade III CRI and are used in the study of the efficacy of KeraStat™. This project has been funded in whole or in part with Federal funds from the Biomedical Advanced Research and

  20. The volumes of accumulation of Cs-137 and Sr-90 per species and variety of agricultural plants

    International Nuclear Information System (INIS)

    Bogdevich, I.M.; Shmigelskaya, I.D.; Efimova, I.A.; Putyatin, Yu.V.

    2001-01-01

    The accumulation of radionuclides in various species and varieties of agricultural plants on the same conditions of soil contaminated by radionuclides and agrochemical exponents can differ hundred times. The differences in accumulation of Cs-137 and Sr-90 are less -up to 3-4 times. The article grades the basic plants cultivated on contaminated soils per volume of Cs-137 and Sr-90 as well as per crop yield. It is possible to recommend selecting the species and varieties of agricultural plants having minimal capabilities of accumulation as a simple economically justified way of reducing the contamination of agricultural produce in general. The solving of problems connected with agriculture on the contaminated territory occupies one of the leading places in the complex of actions on the consequences of Chernobyl disaster liquidation. The researches revealed that 70% of collective dose is formed by the radionuclides receipt into a human organism with food. Eventually radionuclides contents in agricultural production reduce. This process is more expressed for Cs-137 caused by protective actions realization as well as gradually fixing of Cs-137 in soils due to natural factors of decay and fixation. Sr-90 mobility and its availability to plants is not reduced, even tends to increase. Biological features of plants reveal in their different ability to absorb nutritional elements from soil. Because of that radionuclides availability and amount of their including in food chains essentially depend on the level of contra actions applied, natural conditions (soil types, granulometric structure, humidifying mode, agrochemical conditions) and features of crops. The action of Cs-137 and Sr-90 in the system soil-plant has a range of distinctive features. At same density of soil contamination Sr-90 penetration into plants much higher than Cs-137 one. The cause is in difference of radionuclides contents forms in soils. Cs-137 is strongly fixed in soil, but Sr-90 is in exchange form

  1. Transfer of Sr-90 in the environment to human bone, and radiation dose due to the atomic bomb and weapons testing

    International Nuclear Information System (INIS)

    Kawamura, H.; Shiraishi, K.; Igarashi, Y.; Sakurai, Y.

    1988-01-01

    The major source of artificial radioactivities in Japan has been the atmospheric nuclear weapons testing. Some results obtained for activities of Sr-90 in bone, particulary in Japanese, are mentioned, including trends in levels, distribution in bone, transfer from diet to bone and absorbed doses. Some litterature data on pathways of Sr-90 from environment to man are referred to, that is on contribution of different foods to the ingestion intake and transfer of Sr-90 from soil to crops. Recent topics of radioecological studies on soil-plant relationships are shortly introduced

  2. Assessment of the Mechanisms for Sr-90 and TRU Removal from Complexant-Containing Tank Wastes at Hanford

    International Nuclear Information System (INIS)

    Hallen, Richard T.; Geeting, John GH; Lilga, Michael A.; Hart, Todd R.; Hoopes, Francis V.

    2005-01-01

    Small-scale tests (∼20 mL) were conducted with samples from Hanford underground storage tanks AN-102 and AN-107 to assess the mechanisms for removing Sr-90 and transuranics (TRU) from the liquid (supernatant) portion of the waste. The Sr-90 and TRU must be removed (decontaminated), in addition to Cs-137 and the entrained solids, before the supernatant can be disposed of as low-activity waste. Experiments were conducted with various reagents and modified Sr/TRU removal process conditions to more fully understand the reaction mechanisms. The optimized treatment conditions--no added hydroxide, addition of Sr (0.02M target concentration) followed by sodium permanganate (0.02M target concentration) with mixing at ambient temperature--were used as a reference for comparison. The waste was initially two orders of magnitude undersaturated with Sr; the addition of nonradioactive Sr(NO?) ? saturated the supernatant, resulting in isotopic dilution and precipitation of Sr-90 as SrCO?. The reaction chemistry of Mn species relevant to the mechanism of TRU removal by permanganate treatment was evaluated, along with the importance of various mechanisms for decontamination, such as precipitation, absorption, ligand exchange, and oxidation of organic complexants. For TRU removal, permanganate addition generally gave the highest DF. The addition of Mn of lower oxidation states (II, IV, and VI) also resulted in good TRU removal, as did complexant oxidation with periodate and addition of Zr(IV) for ligand exchange. These results suggest that permanganate treatment leads to TRU removal by multiple routes

  3. Absorbed doses to the main parts of eyeball due to use 90Sr + 90Y ophthalmic applicator

    International Nuclear Information System (INIS)

    Chen Lishu

    1993-05-01

    The ophthalmic radiotherapy dosimetry and some affecting factors are introduced. The distributions of absorbed doses to the main parts of a fresh eyeball such as the cornea, sclera, lens and anterior chamber, during the radiotherapy by using a 90 Sr + 90 Y ophthalmic applicator are presented. An tissue-equivalent extrapolation ionization chamber was used in the dose measurement. The reasonable doses during ophthalmic radiotherapy for different depths have been obtained. Therefore, the absorbed dose to the lens, the most sensitive organ, can be given. These data are useful for radiation protection in ophthalmic radiotherapy

  4. Estimation and justification of permissible levels of Sr 90 in firewood and timber produced on the territories contaminated after the Chernobyl NPP accident

    International Nuclear Information System (INIS)

    Zabrodskij, V.N.; Bondar', Yu.I.; Sadchikov, V.I.; Kalinin, V.N.

    2014-01-01

    The permissible levels of Sr 90 in firewood and timber produced on the radioactively contaminated territory are calculated and justified. They are proposed to be used on the territories contaminated after the Chernobyl accident. (authors)

  5. Country report: Thailand. Development of Sr-90/ Y-90 Generator and Development of Radiopharmaceuticals Using Y-90

    Energy Technology Data Exchange (ETDEWEB)

    Nipavan, Poramatikul [Nuclear Research and Development Group, Thailand Institute of Nuclear Technology (Public Organization), Bangkok (Thailand)

    2010-07-01

    The research project has been conducted at Thailand Institute of Nuclear Technology in accordance to the 1st RCM plan during the IAEA meeting in Warsaw. The objectives of the project include the following 5 specific aims: 1. Development of Sr-90/Y-90 ion-exchange chromatography generator 2. Development of Sr-90/Y-90 extraction chromatography generator 3. Development of quality control technique 4. Development of herapeutic radiopharmaceuticals Y-90 particulates/colloids 5. Development of Re-188 DMSA–bis-phosphonates Currently we have achieved specific aims 1 to 3. The specific aims 4 and 5 are during investigation. For specific aim 4, we are during the process to extract high purity {sup 90}Y from {sup 90}Sr/{sup 90}Y generator that will yield the starting {sup 90}Y for the production of Y-90 particulates and colloids. For the 5{sup th} specific aim, we are on hold to receive the starting agent, bis- Phosphonates, from Dr. Blower group. Therefore, this progress report will cover our work focusing on specific aims 1 to 3.

  6. The exploration of nursing care for patients with benign prostatic hyperplasia treated using 90Sr-90Y

    International Nuclear Information System (INIS)

    Xiao Lizhen; Bai Xuemei; Zhang Bihui; Yang Xiaoling

    2004-01-01

    An exploration of nursing care for patients with benign prostatic hyperplasia (BPH) treated using 90 Sr- 90 Y through the rectum was carried out . The treatment result and nursing experience in 90 cases were reported in this paper. Before the therapy nurses explained the method and principle of this treatment to the patients for the sake of increasing their confidence and to help them complete the treatment course successfully. During the radiotherapy, nurses practiced strictly radiation protection principles and operating instructions. They assisted the patients to have a healthy life style and good diet . The result of treatment indicated that the total effectiveness rate was 96.7%. The symptoms of lower urinary obstruction were improved evidently and the life quality of the patients elevated. Observation of clinical system confirmed that 90 Sr- 90 Y may be a new treatment method of BPH with benefits of safe irradiation dos, easy operation, non-traumatization, painlessness, and remarkable curative effects. However, it should be stressed that nursing care plays a pivotal role in the treatment result. (authors)

  7. Experimental autoabsorption curve 90Sr in SrCO3. Efficiency calculation to detection of 90Sr, 90Y and 90Sr + 90Y in a beta gas proportional counter

    International Nuclear Information System (INIS)

    Gasco, C.; Alvarez, A.

    1987-01-01

    Strontium-90 has been determined by radiochemical separation techniques in environmental samples. These techniques, of course, cannot separate the two strontium radionuclides from each other of from stable strontium. Consequently the end product of the chemical separation contains all strontium isotopes in SrCO 3 . The beta particules emitted by 90 Sr are absorbed by the SrCO 3 precipitate. This is the main source of error in the activity measurement. It has been prepared sources of 90 Sr in SrCO 3 in order to determinate the counting efficiency and autoabsorption curve. Also detection efficiencies have been calibrated using known activities of 90 Y and equilibrium mixture of 90 Sr+ 90 Y in the same geometry than our samples. The activity of 90 Sr by ingrowth of 90 Y has been calculated by our computer program. (author). 2 figs., 3 refs

  8. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1990-01-01

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

  9. 90Sr- 90Y and 89Sr beta radioactivity measurement in milk samples using a proportional counter

    International Nuclear Information System (INIS)

    Mananes, A.; Perez Santos, C.; Martinez Churiaque, F.

    1987-01-01

    A thin window glas flow proportional counter is used to measure the 90 Sr- 90 Y and 89 Sr beta radioactivity in milk samples. A chemical procedure is used to separate strontium-yttrium from the other radionuclides present in milk. A calculation of the total efficiency of the system is performed which includes an empirical estimation of the backscattering factor. The calibration of the whole process allows the determination of the 90 Sr activity within 10% relative error in spite of uncertainties in the recovery yields of strontium and yttrium. No 89 Sr activity has been detected, and the mean value obtained for the 90 Sr activity in nine milk samples of Cantabria is 0.115 Bq/1 with a minimum detectable activity of 0.0105 Bq. (author) 18 refs

  10. Influence of soil fungi (basidiomycetes) on the migration of Cs 134 + 137 and Sr 90 in coniferous forest soils

    International Nuclear Information System (INIS)

    Roemmelt, R.; Hiersche, L.; Schaller, G.; Wirth, E.

    1990-01-01

    During the first three years after the Chernobyl event high Cs 134 + 137 activities in fruitbodies of basidiomycetes have been measured. A decline of activities with time has not yet been observed. The activities are considerably higher compared to agricultural products from the same area. In order to study the movement of radiocesium in coniferous forest sites, the activities in soil, fungi, and plants have been measured. Based on these results a model to describe the cesium cycling in coniferous forest ecosystems is proposed with special emphasis on the influence of soil fungi and plants on the migration of cesium. As measurements of Sr 90 in forest ecosystems are rare this nuclide has been included in the investigations. (author)

  11. To the predicted calculations of Cs-137 and Sr-90 migration in soil in the frame of the compartment model

    International Nuclear Information System (INIS)

    Knat'ko, B.A.; Skomorokhov, A.G.

    1994-01-01

    The method of calculation of migration of radionuclides in soil according to compartment model and based on analitical solution of the system of differential equations of the model is proposed. Calculations of the content of radionuclides in the soil layer for Cs-137 and Sr-90 for four types of soil for the periods of 0, 21, 34,46 and 73 months are carried out. Calculations were done for three upper soil layers 0-2.5, 2.5-5, 5.0-15.0 cm, which give the main contribution into transport of radionuclides into plants on meadows and pastures. Such method increases precision of calculations of cesium-137 and strontium-90 migration in comparison with the prosedure, which uses the system of finite difference equations. 6 refs., 1 fig

  12. Soil-plant transfer of Cs-137 and Sr-90 in digestate amended agricultural soils- a lysimeter scale experiment

    Science.gov (United States)

    Mehmood, Khalid; Berns, Anne E.; Pütz, Thomas; Burauel, Peter; Vereecken, Harry; Zoriy, Myroslav; Flucht, Reinhold; Opitz, Thorsten; Hofmann, Diana

    2014-05-01

    Radiocesium and radiostrontium are among the most problematic soil contaminants following nuclear fallout due to their long half-lives and high fission yields. Their chemical resemblance to potassium, ammonium and calcium facilitates their plant uptake and thus enhances their chance to reach humans through the food-chain dramatically. The plant uptake of both radionuclides is affected by the type of soil, the amount of organic matter and the concentration of competitive ions. In the present lysimeter scale experiment, soil-plant transfer of Cs-137 and Sr-90 was investigated in an agricultural silty soil amended with digestate, a residue from a biogas plant. The liquid fraction of the digestate, liquor, was used to have higher nutrient competition. Digestate application was done in accordance with the field practice with an application rate of 34 Mg/ha and mixing it in top 5 cm soil, yielding a final concentration of 38 g digestate/Kg soil. The top 5 cm soil of the non-amended reference soil was also submitted to the same mixing procedure to account for the physical disturbance of the top soil layer. Six months after the amendment of the soil, the soil contamination was done with water-soluble chloride salts of both radionuclides, resulting in a contamination density of 66 MBq/m2 for Cs-137 and 18 MBq/m2 for Sr-90 in separate experiments. Our results show that digestate application led to a detectable difference in soil-plant transfer of the investigated radionuclides, effect was more pronounced for Cs-137. A clear difference was observed in plant uptake of different plants. Pest plants displayed higher uptake of both radionuclides compared to wheat. Furthermore, lower activity values were recorded in ears compared to stems for both radionuclides.

  13. The radiological exposure of man from ingestion of Cs-137 and Sr-90 in seafood from the Baltic Sea. Pilot project: Marina-Balt

    Energy Technology Data Exchange (ETDEWEB)

    Nielsen, S.P. [Risoe National Lab., Roskilde (Denmark); Oehlenschlaeger, M. [National Institute of Radiation Hygiene, Broenshoej (Denmark); Karlberg, O. [Swedish Radiation Protection Institute, Stockholm (Sweden)

    1995-04-01

    This report describes a limited radiological assessment of the collective doses to man from the intake of seafood from the Baltic Sea contaminated with the radionuclides Cs-137 and Sr-90. Information on fisheries statistics is presented. The most important source terms to radioactive contamination of Cs-137 and Sr-90 in the Baltic Sea are identified and quantified. A compartment model for the dispersion of radionuclides in European coastal waters including the Baltic Sea is described and tested by comparing model predictions with observations. Collective doses are calculated with the model for each of the source-term categories. (au) (11 tabs., 28 ills., 17 refs.).

  14. A method for determination of Sr-90 Y-90 by using EDTA and ion-exchangers-Applications to the determination of those radioisotopes in milk

    International Nuclear Information System (INIS)

    Silva, C.M.; Lima, F.W. de.

    1987-05-01

    Amethod in which the complex of ethylenediaminetetraacetic acid (EDTA) with yttrium, used in conjunction with ion-exchangers for determination of Sr-90 is described. The method was to the determination of concentration of Sr-90 in milk, avoiding, in this way, protein elimination by acid precipitation or by evaporation of milk and ashing the residue. Analysis of samples of milk from various places in the state of Sao Pa ulo, Brazil, were carried out and the results are reported. Values found are much lower than maximum permissible concentration. (Author) [pt

  15. The radiological exposure of man from ingestion of Cs-137 and Sr-90 in seafood from the Baltic Sea. Pilot project: Marina-Balt

    International Nuclear Information System (INIS)

    Nielsen, S.P.; Oehlenschlaeger, M.; Karlberg, O.

    1995-04-01

    This report describes a limited radiological assessment of the collective doses to man from the intake of seafood from the Baltic Sea contaminated with the radionuclides Cs-137 and Sr-90. Information on fisheries statistics is presented. The most important source terms to radioactive contamination of Cs-137 and Sr-90 in the Baltic Sea are identified and quantified. A compartment model for the dispersion of radionuclides in European coastal waters including the Baltic Sea is described and tested by comparing model predictions with observations. Collective doses are calculated with the model for each of the source-term categories. (au) (11 tabs., 28 ills., 17 refs.)

  16. Annual deposition of Sr-90, Cs-137 and Pu-239, 240 from the 1961 - 1980 nuclear explosions: a simple model

    International Nuclear Information System (INIS)

    Hirose, K.; Aoyama, M.; Katsuragi, Y.; Sugimura, Y.

    1987-01-01

    The annual deposition of Sr-90, Cs-137, Pu-239, 240, and Pu-238 were observed from 1959 to 1984 at the Meteorological Research Institute (MRI). In order to interpret the serial trends of the annual radioactive deposition at the MRI, a semi-empirical box model of atmospheric transport was developed. The model divides the atmosphere of the Northern Hemisphere into four compartments: the atmosphere above 21 km, stratosphere below 21 km, active mixing and exchange (AME) layer near the tropopause, and the troposphere. The transfer between the compartments follows the first-order kinetics. The half residence times for transfer between upper and lower stratospheric compartments, between the lower and AME layer compartments, and between the AME layer and troposphere are 0.5, 0.7 and 0.3 yrs, respectively. It is revealed that as a long-term monitoring of the annual deposition of radioactive debris in the mid-latitude area, the model quantitatively permits the calculation of stratospheric inventories and trends of annual deposition of debris injected into the stratosphere which are characterized by apparent residence times of 0.5 to 1.7 yrs. This simple model is useful to predict the annual deposition amount of radioactive debris from the thermonuclear explosion for practical purposes. (author)

  17. Improvement of compressive strength of segmentation of zeolites as absorber of Sr-90 liquid waste using coconut fibres

    International Nuclear Information System (INIS)

    Kasmudin; Kusnanto

    2002-01-01

    The use of the coconut fibres to increase compressive strength of segmentation of zeolites as absorber of Sr-90 liquid waste was studied. The purpose of this research was to find the optimum content and length of fibres that give maximum compressive strength. This research was done with mortar-zeolites specimen of cylinder 2,2 cm diameter and 4,4 cm high, the content of zeolites was 13% volume of specimen, weight ratio of water and cement 0,3, length of fibres 1,5 cm, 2 cm, 2,5 cm, and 3 cm (aspect ratio ± 60, ± 80, ± 100 and ± 120) with the fibres content of each fibre 0%, 0,5%, 0,10%, 0,25%, 0,50%, 0,75%, and 1,00%. Addition of fibres was done with a direction of orientation longitudinal to the specimen. The specimens were tested on 28 days old test specimens. The result showed that addition of coconut fibres until certain content would increase compressive strength. The optimum size of fibres with 92,313 N/MM 2 of compressive strength or increased 119,21% of no fibres specimen were 0,50% of volume and 3 cm in length

  18. Procedure to carry out leakage test in beta radiation sealed sources emitters of 90Sr/90Y

    International Nuclear Information System (INIS)

    Alvarez R, J. T.

    2010-09-01

    In the alpha-beta room of the Secondary Laboratory of Dosimetric Calibration of the Metrology Department of Ionizing Radiations ophthalmic applicators are calibrated in absorbed dose terms in water D w ; these applicators, basically are emitter sealed sources of pure beta radiation of 90 Sr / 90 Y. Concretely, the laboratory quality system indicates to use the established procedure for the calibration of these sources, which establishes the requirement of to carry out a leakage test, before to calibrate the source. However, in the Laboratory leakage test certificates sent by specialized companies in radiological protection services have been received, in which are used gamma spectrometry equipment s for beta radiation leakage tests, since it is not reliable to detect pure beta radiation with a scintillating detector with NaI crystal, (because it could detect the braking radiation produced in the detector). Therefore the Laboratory has had to verify the results of the tests with a correct technique, with the purpose of determining the presence of sources with their altered integrity and radioactive material leakage. The objective of this work is to describe a technique for beta activity measurement - of the standard ISO 7503, part 1 (1988) - and its application with a detector Gm plane (type pankage) in the realization of leakage tests in emitter sources of pure beta radiation, inside the mark of quality assurance indicated by the report ICRU 76. (Author)

  19. Assessment of transferring Sr-90 and Cs-137 in products of processing of seeds raps for getting raps' oil and biodiesel

    International Nuclear Information System (INIS)

    Yatsino, T.S.; Mironov, V.P.

    2009-01-01

    The objects of research are soil assays, seeds of raps, straw and the stalks selected on polluted radio nuclides of territory. The work purpose is to measure specific activity of strontium and cesium in assays, to calculate factors of transition Sr 90 and Sr 137 in products of processing of seeds of raps. (authors)

  20. Evaluation of small scale laboratory and pot experiments to determine realistic transfer factors for the radionuclides Sr-90, Cs-137, Co-60 and Mn-54

    International Nuclear Information System (INIS)

    Steffens, W.; Fuhr, F.; Mittelstaedt, W.

    1980-01-01

    Lysimeter experiments were undertaken in a controlled experimental field to study the root uptake of Sr-90, Cs-137, Co-60 and Mn-54 under outdoor conditions. Parallel experiments were set up using the Kick - Brauckman experimental pots under greenhouse and the Neubauer cups under growth chamber conditions. The results obtained from the three types of experiments are compared. (H.K.)

  1. Probability distribution of dose rates in the body tissue as a function of the rhytm of Sr90 administration and the age of animals

    International Nuclear Information System (INIS)

    Rasin, I.M.; Sarapul'tsev, I.A.

    1975-01-01

    The probability distribution of tissue radiation doses in the skeleton were studied in experiments on swines and dogs. When introducing Sr-90 into the organism from the day of birth till 90 days dose rate probability distribution is characterized by one, or, for adult animals, by two independent aggregates. Each of these aggregates correspond to the normal distribution law

  2. Tritium conference days; Journees tritium

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  3. Development of an electrochemical 90Sr-90Y generator for separation of 90Y suitable for targeted therapy

    International Nuclear Information System (INIS)

    Chakravarty, Rubel; Pandey, Usha; Manolkar, Remani B.; Dash, Ashutosh; Venkatesh, Meera; Pillai, M.R. Ambikalmajan

    2008-01-01

    90 Y of high specific activity and very high radionuclidic purity (>99.998%) is essential for targeted therapy. Since the current methods used for the preparation of 90 Y from 90 Sr are not adaptable for use in a central/hospital radiopharmacy, a simple 90 Sr- 90 Y generator system based on electrochemical separation technique was developed. Methods: Two-cycle electrolysis procedure was developed for separation of 90 Y from 90 Sr in nitrate solution. The first electrolysis was performed for 90 min in 90 Sr(NO 3 ) 2 feed solution at pH 2-3 at a potential of -2.5V with 100-200 mA current using platinum electrodes. The second electrolysis was performed for 45 min in 3 mM HNO 3 at a potential of -2.5V with 100 mA current. In this step, the cathode from the first electrolysis containing 90 Y was used as anode along with a fresh circular platinum electrode as cathode. The 90 Y deposited on the circular cathode after the second electrolysis was dissolved in acetate buffer to obtain 90 Y acetate, suitable for radiolabeling. Results: >96% recovery of 90 Y could be achieved in each cycle, with an overall decay corrected yield of >90%. The recovered 90 Y had high radionuclidic purity with barely 30.2±15.2 kBq (817±411 nCi) of 90 Sr per 37 GBq (1 Ci) of 90 Y (0.817±0.411 ppm). Consistent and repeated separation could be demonstrated using up to 1.85 GBq (50 mCi) of 90 Sr. The generator was named 'Kamadhenu,' the eternally milk-yielding Indian mythological cow. Conclusions: A technique suitable for adaptation at central radiopharmacies for obtaining therapeutic quantities of pure 90 Y has been developed

  4. Tritium sources

    International Nuclear Information System (INIS)

    Glodic, S.; Boreli, F.

    1993-01-01

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

  5. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1989-01-01

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

  6. Cs 137 and Sr 90 content in the Kozloduy NPP environment; Soderzhanie 137Cs i 90Sr v obektakh okruzhayushchej sredy v rajone AEhS `Kozloduy`

    Energy Technology Data Exchange (ETDEWEB)

    Karaivanova, R; Totseva, R; Badulin, V; Zlatanova, R; Fajtondzhieva, K [National Centre of Radiobiology and Radiation Protection, Sofia (Bulgaria)

    1996-12-31

    The radioactivity status of soil and plant samples collected in the 12 km controlled zone surrounding Kozloduy has been investigated in the period 1990-1993. The average annual concentrations of Sr-90 and Cs-137 are compared to the values taken at 40 km distance. Cs-137 has in general higher concentrations in soil samples from the NPP close surroundings - e.g. 50-55 Bq/kg in 1992 versus 2-2.5 Bq/kg for Sr-90. Radionuclide concentrations in grass do not differ significantly from the data at 40 km distance. It is concluded that the average annual doses follow the radionuclide dynamics in the global fallout and the effect of Kozloduy operation can not be separated from the background radioactivity status. 4 refs., 21 figs.

  7. Radioactive contamination of the significant fish species in fish-ponds and in natural waters with particular respect to the accumulation of Sr-90 Pt. 2

    International Nuclear Information System (INIS)

    Kantor, D.; Szentjobi, O.

    1977-01-01

    The radioactive contamination of omnivorous fish species (Cyprinus carpio, Hypophtalmictys nobilis Richardson), of herbivorous fish species (Hypophtalmictys molitrix Valenciennes) and carnivorous fish species (Silurus glanis L., Esox Lucius L.) has been investigated. Orientative data have been obtained by analyzing samples originating from various sites. Though differences appeared between the analyzed samples, no incorporations of outstanding nature have been observed. In accordance with experiences described in literature, the isotope Sr-90 is accumulated in an increased degree in the skeleton of fish living in ponds. This is valid in all cases when the vital processes are regulated artificially by fishery methods. A particularly important factor is that the individual fish living in ponds are 2-3 or at most 4 years of age i.e. from the aspect of their ontogenesis their organism is in a strongly build-up phase. Sr-90 proved to be accumulated in a relatively greater amount in the organisms of herbivorous fish. (P.J.)

  8. Monte Carlo dose characterization of a new 90Sr/90Y source with balloon for intravascular brachytherapy

    International Nuclear Information System (INIS)

    Wang Ruqing; Li, X. Allen; Lobdell, John

    2003-01-01

    Beta emitting source wires or seeds have been adopted in clinical practice of intravascular brachytherapy for coronary vessels. Due to the limitation of penetration depth, this type of source is normally not applicable to treat vessels with large diameter, e.g., peripheral vessel. In the effort to extend application of its beta source for peripheral vessels, Novoste has recently developed a new catheter-based system, the Corona trade mark sign 90 Sr/ 90 Y system. It is a source train of 6 cm length and is jacketed by a balloon. The existence of the balloon increases the penetration of the beta particles and maintains the source within a location away from the vessel wall. Using the EGSnrc Monte Carlo system, we have calculated the two-dimensional (2-D) dose rate distribution of the Corona trade mark sign system in water for a balloon diameter of 5 mm. The dose rates on the transverse axis obtained in this study are in good agreement with calibration results of the National Institute of Standards and Technology for the same system for balloon diameters of 5 and 8 mm. Features of the 2-D dose field were studied in detail. The dose parameters based on AAPM TG-60 protocol were derived. For a balloon diameter of 5 mm, the dose rate at the reference point (defined as r 0 =4.5 mm, 2 mm from the balloon surface) is found to be 0.010 28 Gy min -1 mCi -1 . A new formalism for a better characterization of this long source is presented. Calculations were also performed for other balloon diameters. The dosimetry for this source is compared with a 192 Ir source, commonly used for peripheral arteries. In conclusion, we have performed a detailed dosimetric characterization for a new beta source for peripheral vessels. Our study shows that, from dosimetric point of view, the Corona trade mark sign system can be used for the treatment of an artery with a large diameter, e.g., peripheral vessel

  9. Evaluation of a pyrex glass shield for the dose reduction in extremities to manipulate a 90 Sr- 90 Y generator

    International Nuclear Information System (INIS)

    Ayra P, F.E.; Xiques C, A.; Torres B, M.B.

    2006-01-01

    The production of Y-90 of high activity it specifies (free of payee) for their use in radioimmunotherapy uses the Strontium 90 as isotope source. Depending on the method employee for the separation of both isotopes several types of generators are described in different bibliographies. The column generator used in the facilities of the Center of Isotopes requires of a frequent manipulation causing significant dose in the skin of the extremities due to the exhibition to the radiation beta of high energy. The properties of the shieldings for this radiation type have been well studied Y they consist in several publications. To be in correspondence with requirements of radiological protection in the Cuban legislation, the column was covered with a tube of glass pyrex of 5 mm of thickness and it was monitored the exposure with an ionization chamber. At the own time, the shielding using the Monte Carlo method was evaluated. It was used the MCNP 4C code to simulate the absorption of the beta particles generated in the process of disintegration of the Sr-90 and Y-90 in the glass shielding. The column generator and the fluence of beta particles were modeled in different points inside the shielding to determine if the experimentally measured values correspond to electrons that were not absorbed or to the weak stopping radiation generated in the glass due to the deceleration of these particles. A cylinder of 4 mm of diameter simulates the source (it dilutes) and a tube of walls of 6 mm of thickness simulates the shielding more the wall of the column around the generator. This it was divided in cells of 1 mm of thickness and the energy deposited in them was evaluated. The results show that all the electrons generated in the source are absorbed in the shielding and the exposure rates decrease in more of 78 times using the 5 mm of pyrex glass. The doses in extremities to the operators of the generator don't surpass the 70 mSv by year that is the dose restriction imposed in the

  10. Hanford 100N Area Apatite Emplacement: Laboratory Results of Ca-Citrate-PO4 Solution Injection and Sr-90 Immobilization in 100N Sediments

    Energy Technology Data Exchange (ETDEWEB)

    Szecsody, James E.; Burns, Carolyn A.; Moore, Robert C.; Fruchter, Jonathan S.; Vermeul, Vincent R.; Williams, Mark D.; Girvin, Donald C.; McKinley, James P.; Truex, Michael J.; Phillips, Jerry L.

    2007-10-01

    This report summarizes laboratory scale studies investigating the remediation of Sr-90 by Ca-citrate-PO4 solution injection/infiltration to support field injection activities in the Hanford 100N area. This study is focused on experimentally testing whether this remediation technology can be effective under field scale conditions to mitigate Sr-90 migration 100N area sediments into the Columbia River. Sr-90 is found primarily adsorbed to sediments by ion exchange (99% adsorbed, < 1% in groundwater) in the upper portion of the unconfined aquifer and lower vadose zone. Although primarily adsorbed, Sr-90 is still considered a high mobility risk as it is mobilized by seasonal river stage increases and by plumes of higher ionic strength relative to groundwater. This remediation technology relies upon the Ca-citrate-PO4 solution forming apatite precipitate [Ca6(PO4)10(OH)2], which incorporates some Sr-90 during initial precipitation and additionally slowly incorporates Sr-90 by solid phase substitution for Ca. Sr substitution occurs because Sr-apatite is thermodynamically more stable than Ca-apatite. Once the Sr-90 is in the apatite structure, Sr-90 will decay to Y-90 (29.1 y half-life) then Zr-90 (64.1 h half-life) without the potential for migration into the Columbia River. For this technology to be effective, sufficient apatite needs to be emplaced in sediments to incorporate Sr and Sr-90 for 300 years (~10 half-lives of Sr-90), and the rate of incorporation needs to exceed the natural groundwater flux rate of Sr in the 100N area. A primary objective of this study is to supply an injection sequence to deliver sufficient apatite into subsurface sediments that minimizes initial mobility of Sr-90, which occurs because the injection solution has a higher ionic strength compared to groundwater. This can be accomplished by sequential injections of low, then high concentration injection of Ca-citrate-PO4 solutions. Assessment of low concentration Ca-citrate-PO4, citrate-PO4

  11. Tritium trick

    Science.gov (United States)

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

    1971-01-01

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

  12. SU-G-201-12: Investigation of Beta-Emitter 90Sr-90Y Dose Distribution Using Gafchromic EBT3 Film for Application On Conformal Skin Brachytherapy Device

    International Nuclear Information System (INIS)

    Ferreira, C; Johnson, D; Ahmad, S; Rasmussen, K; Jung, J

    2016-01-01

    Purpose: To investigate 90 Sr- 90 Y as a high dose rate (HDR) source for application in a conformal skin brachytherapy (CSBT) device. The CSBT device has been previously developed to provide patient specific treatment for small inoperable lesions and irregular surfaces. Methods: A popular beta emitter, 90 Sr- 90 Y was tested for feasibility in a CSBT device. A 1 cm diameter plaque was used to deliver dose to a solid water phantom containing EBT3 Gafchromic films arranged at the surface and perpendicular to it. Additionally, a 1 cm diameter 6 MeV electron beam was used to irradiate EBT3 film at 100 cm SSD with a 0.5 cm bolus. Films were digitized with an Epson Expression 10000 XL scanner and calibrated with a 6 MeV electron specific dose curve. Normalized percent depth doses (PDD) and dose profiles for both techniques were analyzed using ImageJ. Results: Dose distributions achieved with the 90 Sr- 90 Y sources were compared with those of external electron beam radiation therapy (EBRT). Penumbra (20%- 80%) for EBRT and 90Sr-90Y were 4.3 mm and 1.6 mm, respectively. PDD values of 50% (normalized to 2 mm) were 10.1 mm and 2.8 mm for electron and 90 Sr- 90 Y, respectively. Flatness (80% of the central beam profile) was 14.1% at a 5 mm depth for EBRT and 4.0% at surface for the 90 Sr- 90 Y. Conclusion: As expected, the PDDs of 90 Sr- 90 Y in water are shallower than that of external electron beams for the same field size. 90 Sr- 90 Y can be used in CSBT to provide patient specific treatment where shallower depth doses than that provided by electron external beams may be required: e.g. eyelids, nose, lips, ears, etc. The customizability of EBRT could be replicated by using multiple adjacent 90 Sr- 90 Y plaque placements.

  13. Tritium autoradiography

    International Nuclear Information System (INIS)

    Caskey, G.R. Jr.

    1981-01-01

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

  14. Country report: Vietnam. Setting Up of a 90Sr/90Y Generator System Based on Supported Liquid Membrane (SLM) Technique and Radiolabeling of Eluted 90Y with Biomolecules

    International Nuclear Information System (INIS)

    Nguyen Thi Thu; Duong Van Dong; Bui Van Cuong; Chu Van Khoa

    2010-01-01

    In the course of participating in the IAEA-CRP during the last two years, Vietnam has achieved the goal of setting up a 90 Sr/ 90 Y generator system based on Supported Liquid Membrane (SLM) technique and also radiolabeling of the eluted 90 Y with antibody, peptides and albumin. A two stage SLM based 90 Sr- 90 Y generator was set up in-house to generate carrier-free 90 Y at different activity levels viz. 5, 20, 50 mCi. The generator system was operated in sequential mode in which 2-ethylhexyl 2-ethylhexyl phosphonic acid (PC88A) based SLM was used in the first stage for the transport 90 Y in 4.0 M nitric acid from source phase where 90 Sr- 90 Y equilibrium mixture is placed in nitric acid medium at pH to 1-2. In the second stage, octyl (phenyl)-N,N-diisobutylcarbamoylmethyl phosphine oxide (CMPO) based SLM was used for the transport of 90 Y selectively to 1.0 M acetic acid which is the best medium for radiolebeling. The eluted 90 Y from the generator was tested for the presence of any traces of 90 Sr using the Extraction Paper Chromatography (EPC) and was found suitable for radiolabeling. The generator system could be upgraded to 100 mCi level successfully due to an expert mission from India through IAEA. The 90 Y product obtained from the generator system was used for radiolabeling of antibody and peptides viz. Rituximab, DOTATATE and albumin particles under different experimental conditions. A new chromatography system could be developed for analyzing 90 Y labeled albumin using the TAE buffer as mobile phase in PC and ITLC

  15. Sequestration of Sr-90 Subsurface Contamination in the Hanford 100-N Area by Surface Infiltration of a Ca-Citrate-Phosphate Solution

    Energy Technology Data Exchange (ETDEWEB)

    Szecsody, James E.; Rockhold, Mark L.; Oostrom, Martinus; Moore, R. C.; Burns, Carolyn A.; Williams, Mark D.; Zhong, Lirong; Fruchter, Jonathan S.; McKinley, James P.; Vermeul, Vincent R.; Covert, Matthew A.; Wietsma, Thomas W.; Breshears, Andrew T.; Garcia, Ben J.

    2009-03-01

    The objective of this project is to develop a method to emplace apatite precipitate in the 100N vadose zone, which results in sorption and ultimately incorporation of Sr-90 into the apatite structure. The Ca-citrate-PO4 solution can be infiltrated into unsaturated sediments to result in apatite precipitate to provide effective treatment of Sr-90 contamination. Microbial redistribution during solution infiltration and a high rate of citrate biodegradation for river water microbes (water used for solution infiltration) results in a relatively even spatial distribution of the citrate biodegradation rate and ultimately apatite precipitate in the sediment. Manipulation of the Ca-citrate-PO4 solution infiltration strategy can be used to result in apatite precipitate in the lower half of the vadose zone (where most of the Sr-90 is located) and within low-K layers (which are hypothesized to have higher Sr-90 concentrations). The most effective infiltration strategy to precipitate apatite at depth (and with sufficient lateral spread) was to infiltrate a high concentration solution (6 mM Ca, 15 mM citrate, 60 mM PO4) at a rapid rate (near ponded conditions), followed by rapid, then slow water infiltration. Repeated infiltration events, with sufficient time between events to allow water drainage in the sediment profile can be used to buildup the mass of apatite precipitate at greater depth. Low-K heterogeneities were effectively treated, as the higher residual water content maintained in these zones resulted in higher apatite precipitate concentration. High-K zones did not receive sufficient treatment by infiltration, although an alternative strategy of air/surfactant (foam) was demonstrated effective for targeting high-K zones. The flow rate manipulation used in this study to treat specific depths and heterogeneities are not as easy to implement at field scale due to the lack of characterization of heterogeneities and difficulty tracking the wetting front over a large

  16. Development of techniques for rapid analysis of 90Sr breakthrough and 90Y activity from a 90Sr-90Y generator

    International Nuclear Information System (INIS)

    Chism, S.E.; Goodwin, D.A.; Meares, C.F.

    1986-01-01

    Recently there has been interest in labelling monoclonal antibodies with therapeutic amounts of a pure beta emitting isotope. 90 Y offers many desirable features and may be conveniently and economically obtained from a 90 Sr- 90 Y generator. A potential hazard with this system is breakthrough of the parent 90 Sr which has a half life of 28 years and is an extremely toxic bone seeking isotope. Therefore it is essential that the daughter be completely separated from its parent. We have devised two simple techniques which allow us to monitor rapidly the amount of 90 Sr breakthrough and quantitate the 90 Y activity over a 5 log range. (author)

  17. Use of MnO2 and MnO2 SiO2 for sorbing of Sr-90 from liquid rad waste

    International Nuclear Information System (INIS)

    Subiarto; Las, Thamzil; Aan BH, Martin; Utomo, Cahyo Hari

    1998-01-01

    The synthesis of MnO 2 adsorbent and MnO 2 -SiO 2 composite has been done. MnO 2 synthesis is done by the reaction of KMnO 4 , Mn(NO 3 ) 2 .4H 2 O and Na 2 S 2 O 4 ( MnO 2 -A, MnO 2 -B, and MnO 2 -T ). MnO 2 . SiO 2 is made from KMnO 4 , Na 2 SiO 3 , and H 2 O 2 . The result obtained show the best Sr-90 sorption by MnO 2 -A with Kd = 2085.63 ml/g, by MnO 2 -L with Kd = 755.09 ml/g, and by MnO 2 - SiO 2 composite with Kd = 1466.51 ml/g. From this result, we can conclude that MnO 2 -SiO 2 can be expanded for Sr-90 sorption from liquid radioactive waste. (author)

  18. Development of 90Sr/90Y Generator Systems Based on SLM Techniques for Radiolabelling of Therapeutic Biomolecules with 90Y. Chapter 14

    International Nuclear Information System (INIS)

    Thu, N.T.; Van Dong, D.; Van Cuong, B.; Van Khoa, C.; Cam Hoa, V.T.

    2015-01-01

    Yttrium-90 is one of the most useful radionuclides for radioimmunotherapeutic applications, especially for labelling peptides and antibodies. Studies were carried out to develop a 90 Sr/ 90 Y generator system based on the SLM technique. Two stages of 90 Sr/ 90 Y generator systems were developed at different activity levels of 5, 20, 50 and 100 mCi and operated with semiautomation in sequential mode. In the first stage of the system, PC88A based SLM was used, which transported 90 Y from a nitric acid medium containing 0.01–4M HNO 3 . In the second stage, the 90 Y from the first stage was transferred to the first compartment of the second stage using carbamoylmethyl phosphine oxide (CMPO) based SLM where 1M acetic acid was used as the receiving phase for 90 Y. Quality control was carried out for the products of 90 Y using EPC with paper chromatography and Tec control chromatography. Peptides and antibodies were labelled using the 90 Y product obtained from the generator developed in house. (author)

  19. Forecasting of accumulation of Cs 137 and Sr 90 in the herbage of the main types of the Belarus Palessje meadows utilizing agrochemical soil properties

    International Nuclear Information System (INIS)

    Podolyak, A.G.; Bogdevich, I.M.; Ivashkova, I.I.

    2007-01-01

    On the basis of long-term stationary experience it was established that the minimum accumulation of Cs 137 and Sr 90 in the herbage of the waterless valley, marshed and flood types of the Belarus Palessje meadows contaminated by Chernobyl radionuclides is seen when the optimum basic agrochemical soil properties are achieved with the application of the scientifically reasonable protective measures. It was demonstrated that in the remote period of the accident for the prognosis of radionuclides contents in natural and cultural meadows herbage it is advisable to use of transfer factors (TFa, Bq/kg : kBq/m2) based on the complex agrochemical parameters - basic saturation degree (V, %) and agrochemical cultivation soils index (Iac), which take into account several soil characteristics simultaneously. This article provides the equations of linear and multiple regressions that can be used to calculate the transfer factors for Cs 137 and Sr 90 uptake and the herbage contamination degree for the main types of meadows of the region, which will allow one to reduce the volume of forages production (hay, green bulk) that is not adequate to the established permissible levels: Republican allowable levels of the contents of cesium-137 and strontium-90 in agricultural raw material and forages. (authors)

  20. Tritium immobilisation

    International Nuclear Information System (INIS)

    Bridger, N.J.

    1982-01-01

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

  1. Sorption and desorption of Sr-90 and Cs-137 by sediments of the Sozh-river valley and border water collections

    International Nuclear Information System (INIS)

    Onoshko, M.P.

    2001-01-01

    From the last literature analysis it follows, that to studying of sorption and desorption soil, some rocks and minerals properties concerning radioisotopes the steadfast attention of researchers is paid nowadays. The materials of heavy particles sorption kinetics, the action of adsorption molecules and ions from solutions on leaching products are examined. Sr-90, Cs-137, Pu-239,240 diffusion is estimated. It is found out, that sorbed and desorbed amount of radioisotopes is proportionally to their concentration in soil, and sorption (S) and distributions (Cd) factors do not depend on soil contamination density, and are determined by its physical and chemical properties, parity of firm and liquid phases. It is judged, that increase of soil absorbing properties by the increase of sorbent entering are unpromising, as sorption soil capacity is filled by Cs-137 only in thousand shares of per cent from the sorbent amount, which can be absorbed by soil. With the reference to the conditions of Belarus, experiments and natural supervision on Sr-90 and Cs-137 sorption by Fe, Mn, Si, Al, Ti hydroxides were executed. At experimental researches of electrolyte influence on radioisotope sorption by peat soils Cd amount lines were established. Sediments under certain conditions, due to desorption, become a source of the secondary contamination of natural waters up to ecologically dangerous concentration. Radioisotopes desorption ambiguity is connected to many parallel proceeding processes: exchange sorption on organic and mineral components, co-sedimentation with one-and-a-half Fe, Al and Mn hydroxides and also depends on solutions structure, cationic exchange rocks and soil capacities, concentration of competing ions. At low radioisotopes contents desorption is insignificant, at high - their extraction does not depend on reagent concentration. We carried out the experiment on studying Cs-137 and Sr-90 sorption-desorption from sediments Sozh-river valley and border water

  2. Development of Technology for the Preparation of 90Sr/90Y Generators at the Radiopharmacy Directory of IPEN/CNEN-SP

    International Nuclear Information System (INIS)

    Barrio, Graciela

    2010-01-01

    90 Y (T /2 = 2,67 d; Eβmax = 2,28 MeV) is a radionuclide with efficacy established for various cancer therapies, labeling biomolecules and treating of radiosinovectomy. Due to its nuclear properties, is obtained through the decay of 90 Sr T /2 = 28 y in the form of a generator. Several types of 90 Sr/ 90 Y generators were developed, and the most employed are the cation exchange resins, where Sr and Y are adsorbed and 90 Y is selectively eluted with acetate or EDTA. The disadvantage of this type of generator is the radiolysis, which degrades its use. The electrochemical generator is a proposed solution because there is no significant effect of radiation. In this concept, the difference between the electrochemical potentials of the elements Sr and Y is used to obtain a rapid separation of 90 Y from 90 Sr. The production of 90 Y via colloid formation is the simplest method for the separation, based on the colloid formation of Y in high alkaline pH, which can be filtered and separated from Sr, and subsequently dissolved in HCl. The objective of this work was the development of technologies for the preparation of 90 Sr/ 90 Y generators, and three technologies were developed: generators using cation resins columns, generators through colloid formation and electrochemical generators. Radionuclidic quality control of 90 Y was also evaluated by liquid scintillation, radionuclide identity, extraction paper chromatography (EPC) using complexing agents for 90 Y and by Optical Emission Spectrometry with Inductively Coupled Plasma (ICP-OES). The results showed that generators using cation resins have the best results related to the elution efficiency (∼83%), the reproducibility and radionuclidic purity. The electrochemical generator showed a potential for development, having the advantage of not suffering the effects of radiolysis of the pair 90 Sr/ 90 Y as the resin. A comparison and evaluation of the methods of the radionuclidic quality control showed that the EPC is very

  3. Time and dose-related changes in the thickness of pig skin after irradiation with single doses of 90Sr/90Y β-rays

    International Nuclear Information System (INIS)

    Rezvani, M.; Hamlet, R.; Hopewell, J.W.; Sieber, V.K.

    1994-01-01

    Time-related changes in pig skin thickness have been evaluated using a non-invasive ultrasound technique after exposure to a range of single doses of 90 Sr/ 90 Yr β-rays. The reduction in relative skin thickness developed in two distinct phases: the first was between 12 and 20 weeks postirradiation. No further changes were then seen until 52 weeks postirradiation when a second phase of skin thinning was observed. This was complete after 76 weeks and no further changes in relative skin thickness were seen in the maximum follow up period of 129 weeks. The timings of these phases of damage were independent of the radiation dose, however, the severity of both phases of radiation-induced skin thinning were dose related. (Author)

  4. Results of study of Sr-90 and Cs-137 content in organism and effective doses of internal and external irradiation of Ukrainian population residing in different regions

    International Nuclear Information System (INIS)

    Kalmykov, L.; Gur, E.

    1996-01-01

    The authors have studied effective doses of internal and external radiation for 1992-1994 in the residents of Chernigov and Kharkov Regions of Ukraine, i.e. those who live in the zone of strict radioecologic control and in relatively ''clean'' zones, respectively. In 95% of the investigated residents of Chernigov Region Cs-137 activity in the organism was lower than 1500 Bq, maximum amount being 11 kBq. Conditioned Cs-137 effective dose of internal radiation did not exceed 250 micro Sv per year, in 96% of the investigated subjects it was less than 30 micro Sv per year. Mean amount of this radionuclide in the organisms of both adults and children aged 3-6 years residing in Kharkov Region was 90 and 6 Bq respectively, dose being 2 and 0.4 micro Sv per year. Sr-90 amount in the bone tissue decreases with the age and for the residents of Chernigov region it was 7-23 Bq/kg of bone, for the adult residents of Kharkov region it was about 3 Bq/kg of bone. Mean effective dose of internal radiation due to Sr-90 incorporation for the residents of both Kharkov and Chernigov Regions was 0.7 and 1.9 micro Sv per year. Effective dose of external radiation for the residents of Kharkov Region has not changed since the Chernobyl accident. Total effective dose of external and internal radiation in various professional groups for the residents of Chernigov region increased by 80 micro Sv per year which makes up 14% of mean population dose in Ukraine. (author). 11 refs, 5 tabs

  5. Magmatic tritium

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  6. Validation of an extraction paper chromatography (EPC) technique for estimation of trace levels of 90Sr in 90Y solutions obtained from 90Sr/90Y generator systems

    International Nuclear Information System (INIS)

    Usha Pandey; Yogendra Kumar; Ashutosh Dash

    2014-01-01

    While the extraction paper chromatography (EPC) technique constitutes a novel paradigm for the determination of few Becquerels of 90 Sr in MBq quantities of 90 Y obtained from 90 Sr/ 90 Y generator, validation of the technique is essential to ensure its usefulness as a real time analytical tool. With a view to explore the relevance and applicability of EPC technique as a real time quality control (QC) technique for the routine estimation of 90 Sr content in generator produced 90 Y, a systematic validation study was carried out diligently not only to establish its worthiness but also to broaden its horizon. The ability of the EPC technique to separate trace amounts of Sr 2+ in the presence of large amounts of Y 3+ was verified. The specificity of the technique for Y 3+ was demonstrated with 90 Y obtained by neutron irradiation. The method was validated under real experimental conditions and compared with a QC method described in US Pharmacopeia for detection of 90 Sr levels in 90 Y radiopharmaceuticals. (author)

  7. Tritium dosimetry and standardization

    International Nuclear Information System (INIS)

    Balonov, M.I.

    1983-01-01

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

  8. Tritium accountancy

    International Nuclear Information System (INIS)

    Avenhaus, R.; Spannagel, G.

    1995-01-01

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

  9. Study of contamination by 100 {mu}Ci of Sr 90 in the rat: clinical, hematological and osseous effects (appearance of osteosarcomas); Etude d'une contamination par 100 {mu}Ci de Sr 90 chez le rat: consequences cliniques, hematologiques et osseuses (apparition d'osteosarcomes)

    Energy Technology Data Exchange (ETDEWEB)

    Graf, B; Lafuma, J; Parmentier, C; Parmentier, N [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1968-07-01

    Clinical, hematological and osseous effects following an intramuscular injection of 100 {mu}Ci 90 Sr were studied in the rat. In spite of the magnitude of the injection and the resulting damage, the elimination of strontium could compare with what occurs after an injection at tracer doses. Comparing with the controls at the outcoming time, clinical monitoring mainly brought out a loss of body weight. Fairly early (20. - 30. day) there occurred severe hematological damage, especially on lymphocyte line, which subsided spontaneously. As foreseen, the anatomo-pathological survey of the early period showed bone and medullar lesions in the areas of enchondral ossification. In the late period, bone sarcomas occurred in nine animals out of ten. The outstanding histological type was osteogenic osteosarcomas; besides, two animals experienced bilateral tumors. (authors) [French] Les auteurs etudient les consequences cliniques, hematologiques et osseuses d'une contamination par 100 microcuries de Sr 90 injectes par voie intramusculaire chez le rat. Malgre l'importance de la contamination et les lesions consequentes, l'elimination du Sr 90 est comparable a celle que l'on observe apres injection de doses traceuses. La surveillance clinique ne met essentiellement en evidence, a la periode terminale, qu'une diminution du poids par rapport aux temoins. Les lesions hematologiques sont importantes, predominant sur la lignee lymphocytaire. Elles sont relativement precoces (20e - 30e jours) et regressent spontanement. La surveillance anatomo-pathologique de la periode precoce a montre, comme il etait previsible, des lesions osseuses et medullaires dans les zones d'ossification enchondrale. A la periode tardive, la survenue de sarcomes osseux a ete observee chez neuf animaux sur dix. Le type histologique predominant est l'osteosarcome osteogenique et il faut signaler egalement les tumeurs bilaterales observees chez deux animaux. (auteurs)

  10. Study of contamination by 100 {mu}Ci of Sr 90 in the rat: clinical, hematological and osseous effects (appearance of osteosarcomas); Etude d'une contamination par 100 {mu}Ci de Sr 90 chez le rat: consequences cliniques, hematologiques et osseuses (apparition d'osteosarcomes)

    Energy Technology Data Exchange (ETDEWEB)

    Graf, B.; Lafuma, J.; Parmentier, C.; Parmentier, N. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1968-07-01

    Clinical, hematological and osseous effects following an intramuscular injection of 100 {mu}Ci 90 Sr were studied in the rat. In spite of the magnitude of the injection and the resulting damage, the elimination of strontium could compare with what occurs after an injection at tracer doses. Comparing with the controls at the outcoming time, clinical monitoring mainly brought out a loss of body weight. Fairly early (20. - 30. day) there occurred severe hematological damage, especially on lymphocyte line, which subsided spontaneously. As foreseen, the anatomo-pathological survey of the early period showed bone and medullar lesions in the areas of enchondral ossification. In the late period, bone sarcomas occurred in nine animals out of ten. The outstanding histological type was osteogenic osteosarcomas; besides, two animals experienced bilateral tumors. (authors) [French] Les auteurs etudient les consequences cliniques, hematologiques et osseuses d'une contamination par 100 microcuries de Sr 90 injectes par voie intramusculaire chez le rat. Malgre l'importance de la contamination et les lesions consequentes, l'elimination du Sr 90 est comparable a celle que l'on observe apres injection de doses traceuses. La surveillance clinique ne met essentiellement en evidence, a la periode terminale, qu'une diminution du poids par rapport aux temoins. Les lesions hematologiques sont importantes, predominant sur la lignee lymphocytaire. Elles sont relativement precoces (20e - 30e jours) et regressent spontanement. La surveillance anatomo-pathologique de la periode precoce a montre, comme il etait previsible, des lesions osseuses et medullaires dans les zones d'ossification enchondrale. A la periode tardive, la survenue de sarcomes osseux a ete observee chez neuf animaux sur dix. Le type histologique predominant est l'osteosarcome osteogenique et il faut signaler egalement les tumeurs bilaterales observees chez deux animaux. (auteurs)

  11. STAR facility tritium accountancy

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  12. Oxidative Tritium Decontamination System

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  13. Tritium activities in Canada

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1995-01-01

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

  14. Study on the source of optical activity: Pt. 6. ESR spectroscopy on the asymmetrical radical yields in 90Sr-90Y β irradiated D-and L-alanine

    International Nuclear Information System (INIS)

    Wang Wenqing; Zhao Jian; Ding Xiang

    1993-01-01

    Free radical formation in 90 Sr- 90 Y β irradiated D- and L-alanine is studied by ESR spectroscopy. To calibrate the probable differences in the size of ESR tubes, the different densities of alanine and the incidental different impurities, samples are firstly irradiated with 60 Co γ rays. Then the D- and L- alanine samples are irradiated with 90 Sr- 90 Y source at 77 K. Soon after irradiation, ESR measurement is performed on each sample. The average ratio H β+γ /H γ of D- alanine is 0.147 higher than that of L- alanine, indicating that more free radicals are induced in D- alanine by β irradiation. Irradiation at 77 K is capable of diminishing the effect of thermal movement on the polarization of β electrons and freezing the free radical formation. The experiments show the stereoselective interaction of β electrons with D- and L- amino acids, and so supports the Vester-ulbricht hypothesis

  15. The Tritium White Paper

    International Nuclear Information System (INIS)

    2009-01-01

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

  16. ZEPHYR tritium system

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  17. TFTR tritium handling concepts

    International Nuclear Information System (INIS)

    Garber, H.J.

    1976-01-01

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

  18. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

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

  19. Tritium conference days

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. Sources of tritium

    International Nuclear Information System (INIS)

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

    1980-12-01

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

  1. Tritium pellet injector results

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  2. Environmental aspects of tritium

    International Nuclear Information System (INIS)

    Quisenberry, D.R.

    1979-01-01

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

  3. Organically bound tritium

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.

    1993-01-01

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

  4. Tritium sampling and measurement

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  5. Confinement and Tritium Stripping Systems for APT Tritium Processing

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-20

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

  6. Confinement and Tritium Stripping Systems for APT Tritium Processing

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  7. Environmental tritium in trees

    International Nuclear Information System (INIS)

    Brown, R.M.

    1979-01-01

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

  8. Tritium monitoring techniques

    International Nuclear Information System (INIS)

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

    1996-05-01

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

  9. Tritium in metals

    International Nuclear Information System (INIS)

    Schober, T.

    1990-01-01

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

  10. Tritium sources; Izvori tricijuma

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

  11. High-pressure tritium

    International Nuclear Information System (INIS)

    Coffin, D.O.

    1976-01-01

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

  12. Studies on the radioactive contamination due to nuclear detonations V. The radioactive contamination of human body by Sr90 and Cs137 and its correlation with the fallout rate and ground deposition

    International Nuclear Information System (INIS)

    Nishiwaki, Yasushi

    1961-01-01

    About 60 bone samples were analysed for Sr 90 during the period from January 1957 to July 1959. The highest average concentration was found to be about 1 S.U for the age group of 5 to 10 years old. However since most of the cases are more or less pathological ones after long period of hospitalization the results may not be considered to be the representative values for the average normal person. The highest concentration among the cases whose past history is known was 175 S.U. or roughly, about 2 S.U. During the period of observation occasionally we have found higher concentration than 2 S.U, in the bone samples of a few cases obtained from the crematory. However, since the past history of these cases is not known, it is not clear whether they depended on the rainwater or on the brown ice with much higher contamination than the white rice. In view of these findings it may not be to, much to assume that the concentration of Sr 90 in te bones of the average normal children might reach 2 S.U c corresponding to the cumulative ground deposition and the fallout rate. The concentration of Cs whose physical half-life is very much similar to Sr 90 was in the range of about 30 - 100 cesium unit in 1957 to early 1958. However, during the period from 1958 to early 1959 the concentration of Cs 137 in come of the organs of the human body has been reported to be about two to three times higher on the average in Tokyo than the above value in 1957

  13. Electrochemical separation of 90-yttrium in the electrochemical 90Sr/90Y generator and its use for radiolabelling of DOTA-conjugated somatostatin analog [DOTA0, Tyr3] octreotate

    Directory of Open Access Journals (Sweden)

    Petrović Đorđe Ž.

    2012-01-01

    Full Text Available Radiopharmaceuticals based on 90Y are widely used in the treatment of malignant deseases. In order to meet the requirements for their future application, a 90Sr/90Y generator was developed and 90Y eluted from this locally produced generator was used for the radiolabelling of the DOTA-conjugated somatostatin analog [DOTA0,Tyr3] octreotate and the preparation of [90Y-DOTA0,Tyr3] octreotate (90Y-DOTATATE for peptide receptore radionuclide therapy. 90Sr/90Y generator was based on the electrochemical separation of 90Y from 90Sr in a two-cycle electrolysis procedure. Three electrode cells were used to perform both electrolyses. In both cycles, working electrodes were kept on constant potential. The pH of the solution was adjusted to 2.7 of the value before the electrolyses. The radionuclidic purity of the 90Y solution was analysed by ITLC and extraction paper chromatography. The labelling of peptide (100 mg DOTATATE with 90YCl3 was performed at 95°C for 30 minutes. Radiochemical purity was determined by HPLC and chromatographic separation, using a solid SepPak C-18 column. Results obtained confirmed the efficiency of our electrochemical separation technique and quality control methods for 90Y. The achieved efficiency of the 90Sr/90Y generator above 96% of the theoretical value represents a good basis for the further development of this generator. The labelling of the DOTATATE with 90Y exhibited a high efficiency, too: there was less than 1% of 90Y3+in the 90Y-DOTATATE.

  14. The ITER tritium systems

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Radionuclide Basics: Tritium

    Science.gov (United States)

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

  16. In-vessel tritium

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  17. Protection against tritium radiations

    International Nuclear Information System (INIS)

    Bal, Georges

    1964-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  19. Tritium breeding in fusion reactors

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1982-10-01

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

  20. Tritium permeation through iron

    International Nuclear Information System (INIS)

    Hagi, Hideki; Hayashi, Yasunori

    1989-01-01

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

  1. Tritium technology. A Canadian overview

    Energy Technology Data Exchange (ETDEWEB)

    Hemmings, R.L. [Canatom NPM (Canada)

    2002-10-01

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

  2. Tritium technology. A Canadian overview

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    2002-01-01

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

  3. Metabolism and dosimetry of tritium

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  4. Experience in handling concentrated tritium

    International Nuclear Information System (INIS)

    Holtslander, W.J.

    1985-12-01

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

  5. 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.

  6. Tritium permeation and recovery

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  7. Tritium protective clothing

    International Nuclear Information System (INIS)

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

    1979-06-01

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

  8. Tritium breeding blanket

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  9. Tritium in plants

    International Nuclear Information System (INIS)

    Vichot, L.; Losset, Y.

    2009-01-01

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

  10. Tritium-v. 2

    International Nuclear Information System (INIS)

    1987-01-01

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

  11. 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.

  12. Tritium protective clothing

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-06-01

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

  13. Thin film tritium dosimetry

    Science.gov (United States)

    Moran, Paul R.

    1976-01-01

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

  14. Management of tritium wastes

    International Nuclear Information System (INIS)

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

    1991-07-01

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

  15. 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)

  16. Power-law distribution of Sr-90 concentration in soil samples taken in the South part of Sverdlovsk Oblast, contaminated in the result of the Kyshtym accident at the Mayak facility in 1957

    Energy Technology Data Exchange (ETDEWEB)

    Korobitsyn, B.; Manzhurov, I.; Sergeev, A.; Subbotina, I. [Institute of Industrial Ecology, Ural Branch of Russian Academy of Sciences, Yekaterinburg (Russian Federation)

    2012-07-01

    In September 1957, a chemical explosion occurred in one of the Mayak facility's nuclear waste storage tank, spreading 740 PBq of radiation over a 23,000-km{sup 2} area. To determine a present contamination of the territory of Kamensk region of Sverdlovsk Oblast, soil samples at 315 locations were collected and analyzed in the course of the field campaign of 1993 and 1994. It was found that a distribution of Sr-90 activity in soil samples is a distribution with 'heavy tail'. Analysis of a histogram representing its frequency distribution and plotting this histogram on doubly logarithmic axes allows to conjecture that this distribution follows the power low. Testing of the power-low hypothesis provided objective evidence that the power-low is a reasonable description of the data. In this case a conclusion follows that radioactive contamination levels of the territory are not well characterized by their averages. (author)

  17. Tritium in nuclear power plants

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  18. Japanese university program on tritium radiobiology and environmental tritium

    International Nuclear Information System (INIS)

    Okada, Shigefumi

    1989-01-01

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

  19. Tritium Assay and Dispensing of KEPRI Tritium Lab

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. Longtime radionuclide monitoring in the vicinity of Salaspils nuclear reactor; Dauerhaftes Monitoring der Radionuklide in Umgebung von Salaspils Kernreaktor

    Energy Technology Data Exchange (ETDEWEB)

    Riekstina, D.; Berzins, J.; Krasta, T. [Latvia Univ. (Latvia). Inst. of Solid State Physics; Skrypnik, O.; Alksnis, J. [Latvia Univ. (Latvia). Inst. of Chemical Physics

    2016-07-01

    The research nuclear reactor in Salaspils was decommissioned in 1998. Now reactor is partially dismantled and its territory is used as a temporary storage of radioactivity contaminated materials and water. Environment radioactivity monitoring for presence of artificial radionuclides in the vicinity of Salaspils nuclear reactor is carried out since 1990. Data include Cs-137 concentration in soils, tritium concentration in ground water, as well as H-3, Cs-137, Co-60 concentration and gross beta-activity of reactors sewage and rainwater drainage. The systematic monitoring allowed to detect in December 2014 a leakage from the special wastewater basin and so to prevent a pollution of ground water outside reactors territory.

  1. Comparison of tritium production facilities

    International Nuclear Information System (INIS)

    He Kaihui; Huang Jinhua

    2002-01-01

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

  2. Tritium emissions reduction facility (TERF)

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Kidman, R.B.

    1997-11-01

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

  4. Tritium Systems Test Facility

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  5. Tritium analysis at TFTR

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  6. Tritium - is it underestimated

    International Nuclear Information System (INIS)

    Whitlock, G.D.

    1980-01-01

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

  7. Tritium in HTR systems

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1987-07-01

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

  8. Atmospheric tritium. Measurement and application

    International Nuclear Information System (INIS)

    Frejaville, Gerard

    1967-02-01

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

  9. Handling of tritium at TFTR

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  10. Tritium effluent removal system

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. 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.

  13. Tritium in rad waste management

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  14. ARIES-I tritium system

    International Nuclear Information System (INIS)

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

    1990-09-01

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

  15. Tritium breeding materials

    International Nuclear Information System (INIS)

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

    1984-03-01

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

  16. Tritium breeding materials

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  17. Properties of tritium and its compounds

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Use of tritium and sources

    International Nuclear Information System (INIS)

    Noguchi, Hiroshi

    1997-01-01

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

  19. Sr90 determination in calcareous water

    International Nuclear Information System (INIS)

    Cohen, P.; Pardo, G.; Wormser, G.

    1958-01-01

    The Straub method (ref. 2) for determination of radioactive strontium in water containing calcium is valid within very wide limits, and particularly in the case of residual water from the chemical purification treatment of the radioactive liquid effluent at the C.E.N. Saclay. (author) [fr

  20. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Evaluation algorithm for Hp (10), Hp (0.07) and response matrix in mixed fields of 137Cs+90Sr/90Y for badge TLD without use of its commercial holder

    International Nuclear Information System (INIS)

    Alvarez R, Jose T.; Tovar M, Victor M.; Castaneda P, Mario; Padilla, Ismael

    2008-01-01

    American standards HPS N 13.11 (1993, 2001) set up performance category tests for deep and shallow personal equivalent dose evaluated by dosimetry processors. These standards establish the parameters bias B, standard deviation S and tolerance level L, whose values indicate the deviations from the conventionally true value for the equivalent dose. On the other hand, for a right evaluation of the equivalent dose, all personal dosimetry systems have four basic components: badge, holder, reading protocol and dose evaluation algorithm. However, in Nuclear Plant Laguna Verde NPP dosimetry laboratory uses the badges Panasonic 802 A without Panasonic UD874AT holder, and instead employs a polyethylene holder bag to protect the badge from misuse. This situation disables the manufacturer algorithm designed for the combination: badge + holder. Therefore, the purpose of this work is to design a personal equivalent dose algorithm for this badge + bag, such this corrected algorithm let fulfill the HPS N 13.11 (1993) categories II, IV, V and VII. The algorithm starts from the readings of the elements: E1, E2, E3 and E4, only corrected by their element correction factor ECF. Additionally, we characterize the badge + bag by means of the response function for: 137 Cs, 85 Kr and 90 Sr/ 90 Y radiation fields. Finally, for validation of the algorithm one verifies that parameters B, S and L accomplish with the limits indicated by the standard. (author)

  3. Tritium in the aquatic environment

    International Nuclear Information System (INIS)

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

    1986-02-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Yamawaki, Michio; Kawamura, Hiroshi; Tsuchiya, Kunihiko

    2004-01-01

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

  6. Universal tritium transmitter

    International Nuclear Information System (INIS)

    Cordaro, J. V.; Wood, M.

    2008-01-01

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

  7. Environmental monitoring for tritium in tritium separation facility

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. Environmental monitoring for tritium at tritium separation facility

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  9. Tritium concentration monitor

    International Nuclear Information System (INIS)

    Shono, Kosuke.

    1991-01-01

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

  10. Metabolism of organically bound tritium

    International Nuclear Information System (INIS)

    Travis, C.C.

    1984-01-01

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

  11. A prototype wearable tritium monitor

    International Nuclear Information System (INIS)

    Surette, R. A.; Dubeau, J.

    2008-01-01

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

  12. Effects of tritium in elastomers

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1982-01-01

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

  13. Tritium Storage Material

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  14. Toxicity of tritium

    International Nuclear Information System (INIS)

    Dobson, R.L.

    1979-01-01

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

  15. Biological effects of tritium

    International Nuclear Information System (INIS)

    Nieto, M.

    1985-01-01

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

  16. Tritium removal and retention device

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  17. An overview of tritium production

    International Nuclear Information System (INIS)

    He Kaihui; Huang Jinghua; Feng Kaiming

    2002-01-01

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

  18. Tritium safety issues for TFCX

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  19. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

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

  20. Tritium inventory tracking and management

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  1. Tritium monitor and collection system

    Science.gov (United States)

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

    1992-01-14

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

  2. Tritium accounting for PHWR plants

    International Nuclear Information System (INIS)

    Nair, P.S.; Duraisamy, S.

    2012-01-01

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

  3. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1983-08-01

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

  4. TFTR tritium operations lessons learned

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Overview of tritium fast-fission yields

    International Nuclear Information System (INIS)

    Tanner, J.E.

    1981-03-01

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

  6. Technology developments for improved tritium management

    International Nuclear Information System (INIS)

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

    1994-06-01

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

  7. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  8. JET experiments with tritium and deuterium–tritium mixtures

    NARCIS (Netherlands)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  10. Tritium transport and control in the FED

    International Nuclear Information System (INIS)

    Rogers, M.L.

    1981-01-01

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

  11. Ontario Hydro diversifies into tritium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

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

  12. Tritium containment in fusion facilities

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1978-01-01

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

  13. Tritium accountancy in fusion systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  14. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  15. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

    von Halle, A.; Gentile, C.

    1994-01-01

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

  16. Tritium in the food chain

    International Nuclear Information System (INIS)

    Koenig, L.A.

    1988-01-01

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

  17. Tritium metabolism in rat tissues

    International Nuclear Information System (INIS)

    Takeda, H.

    1982-01-01

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

  18. Tritium behaviour in higher plants

    International Nuclear Information System (INIS)

    Guenot, J.

    1984-05-01

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

  19. Tritium practices past and present

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  20. SU-G-201-12: Investigation of Beta-Emitter 90Sr-90Y Dose Distribution Using Gafchromic EBT3 Film for Application On Conformal Skin Brachytherapy Device

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, C; Johnson, D; Ahmad, S [University of Oklahoma Health Sciences Center, Oklahoma City, OK (United States); Rasmussen, K [University of Texas HSC SA, San Antonio, TX (United States); Jung, J [East Carolina University Greenville, NC (United States)

    2016-06-15

    Purpose: To investigate {sup 90}Sr-{sup 90}Y as a high dose rate (HDR) source for application in a conformal skin brachytherapy (CSBT) device. The CSBT device has been previously developed to provide patient specific treatment for small inoperable lesions and irregular surfaces. Methods: A popular beta emitter, {sup 90}Sr-{sup 90}Y was tested for feasibility in a CSBT device. A 1 cm diameter plaque was used to deliver dose to a solid water phantom containing EBT3 Gafchromic films arranged at the surface and perpendicular to it. Additionally, a 1 cm diameter 6 MeV electron beam was used to irradiate EBT3 film at 100 cm SSD with a 0.5 cm bolus. Films were digitized with an Epson Expression 10000 XL scanner and calibrated with a 6 MeV electron specific dose curve. Normalized percent depth doses (PDD) and dose profiles for both techniques were analyzed using ImageJ. Results: Dose distributions achieved with the {sup 90}Sr-{sup 90}Y sources were compared with those of external electron beam radiation therapy (EBRT). Penumbra (20%- 80%) for EBRT and 90Sr-90Y were 4.3 mm and 1.6 mm, respectively. PDD values of 50% (normalized to 2 mm) were 10.1 mm and 2.8 mm for electron and {sup 90}Sr-{sup 90}Y, respectively. Flatness (80% of the central beam profile) was 14.1% at a 5 mm depth for EBRT and 4.0% at surface for the {sup 90}Sr-{sup 90}Y. Conclusion: As expected, the PDDs of {sup 90}Sr-{sup 90}Y in water are shallower than that of external electron beams for the same field size. {sup 90}Sr-{sup 90}Y can be used in CSBT to provide patient specific treatment where shallower depth doses than that provided by electron external beams may be required: e.g. eyelids, nose, lips, ears, etc. The customizability of EBRT could be replicated by using multiple adjacent {sup 90}Sr-{sup 90}Y plaque placements.

  1. Photoproduction of tritium

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  2. Tritium-surface interactions

    International Nuclear Information System (INIS)

    Kirkaldy, J.S.

    1983-06-01

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

  3. Torus evacuation and tritium handling on NET

    International Nuclear Information System (INIS)

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

    1986-08-01

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

  4. Contribution to the tritium continental effect

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  5. Contribution to the tritium continental effect

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. Overview of the tritium system of Ignitor

    International Nuclear Information System (INIS)

    Rizzello, C.; Tosti, S.

    2008-01-01

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

  7. Tritium, biography of an element

    International Nuclear Information System (INIS)

    Keller, C.

    1980-01-01

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

  8. Tritium monitor for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jalbert, R.A.

    1982-08-01

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

  9. The organically bound tritium: an analyst vision

    International Nuclear Information System (INIS)

    Ansoborlo, E.; Baglan, N.

    2009-01-01

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

  10. Tritium monitoring at the Sandia Tritium Research Laboratory

    International Nuclear Information System (INIS)

    Devlin, T.K.

    1978-10-01

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

  11. JET experiments with tritium and deuterium–tritium mixtures

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  12. Thought experiment with tritium

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

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

  14. Tritium transport around nuclear facilities

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  15. TFTR tritium inventory accountability system

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  16. Tritium management for fusion reactors

    International Nuclear Information System (INIS)

    Rouyer, J.L.; Djerassi, H.

    1985-01-01

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

  17. Radiotoxicity of tritium in mammals

    International Nuclear Information System (INIS)

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

    1972-12-01

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

  18. Tritium transport around nuclear faciliteis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  19. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  20. Criteria for assembly of in vivo measuring systems using high-resolution {gamma}-spectroscopy for evaluation of incorporated radionuclides; Kriterien zum Aufbau von In Vivo Messsystemen zur hochaufloesenden {gamma}-Spektrometrie inkorporierter Radionuklide

    Energy Technology Data Exchange (ETDEWEB)

    Wahl, W. [GSF Forschungszentrum fuer Umwelt und Gesundheit, Neuherberg (Germany). Inst. fuer Strahlenschutz

    1997-12-01

    The paper reviews the available, fundamental measuring methods relying on {gamma}-spectroscopy for their possible application in whole-body and partial-body counters for detection of manifold incorporation of radionuclides. Particular emphasis is placed on the response functions of various detectors, the assembly, the differentiated radioactivity distribution in the body, the various components of background activity and the corresponding suppression mechanisms, and possible ways of using the energy dependence for optimised detection of specific {gamma} energies in a given body region. Criteria and relations as well as their advantages and drawbacks are discussed. (orig./CB) [Deutsch] Diese Arbeit prueft die zur Verfuegung stehenden grundlegenden, {gamma}-spektroskopischen Messmethoden auf deren moegliche Anwendung im Spektrum der Ganz- und Teilkoerperzaehler zum Nachweis der mannigfach inkorporierten Radionuklide. Insbesondere wird eingegangen auf die Response Funktionen verschiedener Detektoren, die Anordnung, die differenzierte Aktivitaetsverteilung im Koerper, die verschiedenen Untergrundkomponenten und deren Unterdrueckungsmechanismen sowie die Beeinflussung durch die Energieabhaengigkeit zum optimalen Nachweis spezifischer {gamma}-Energien an einem bestimmten Ort am Koerper. Kriterien und Relationen sowie deren Vor- und Nachteile werden diskutiert. (orig.)

  1. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

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

    1985-08-01

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

  2. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  3. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

    Wall, W.R.

    1981-03-01

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

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

    International Nuclear Information System (INIS)

    Lei, Jiarong; Du, Yang; Yang, Yong

    2008-01-01

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

  5. Separation of Tritium from Wastewater

    International Nuclear Information System (INIS)

    JEPPSON, D.W.

    2000-01-01

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

  6. Tritium research activities in Korea

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  7. Tritium concentrations in tree ring cellulose

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Exploration for tritium-free water

    International Nuclear Information System (INIS)

    Hussain, S.D.

    1982-10-01

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

  9. Tritium problems in fusion reactor systems

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  11. Turkey Point tritium. Progress report

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  12. Tritium Management Loop Design Status

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-01

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

  13. An assembly of tritium production experiment

    International Nuclear Information System (INIS)

    Abe, Toshihiko

    1981-01-01

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

  14. Tritium in the environment. Knowledge synthesis

    International Nuclear Information System (INIS)

    2009-01-01

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

  15. Recommended radiological controls for tritium operations

    International Nuclear Information System (INIS)

    Mansfield, G.

    1992-01-01

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

  16. Tritium in metals: Techniques of preparation

    International Nuclear Information System (INIS)

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

    1982-08-01

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

  17. Tritium decontamination of machine components and walls

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  18. 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. (...

  19. Polymeric media for tritium fixation

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  20. Tritium processing using metal hydrides

    International Nuclear Information System (INIS)

    Mallett, M.W.

    1986-01-01

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

  1. Implanted-tritium permeation experiments

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  2. Tritium emissions from a detritiation facility

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. Tritium inventory prediction in a CANDU plant

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  4. Tritium oxidation and exchange: preliminary studies

    International Nuclear Information System (INIS)

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

    1978-05-01

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

  5. Bioassay guideline 2: guidelines for tritium bioassay

    International Nuclear Information System (INIS)

    1983-01-01

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

  6. 10 CFR 30.55 - Tritium reports.

    Science.gov (United States)

    2010-01-01

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

  7. Toxicity and dosimetry of tritium

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  8. The INEL Tritium Research Facility

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-01-01

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

  9. The INEL Tritium Research Facility

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  10. Tritium turnover in succulent plants

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  11. Tritium processing in JT-60U

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Masaki, Kei

    1997-01-01

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

  12. Tritium monitoring in environment at ICIT Tritium Separation Facility

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  13. Automation system for tritium contaminated surface monitoring

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. Tritium compatibility of alumina and Fosterite

    Energy Technology Data Exchange (ETDEWEB)

    Coffin, D.O.

    1979-09-01

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

  15. Management of tritium at nuclear facilities

    International Nuclear Information System (INIS)

    1984-01-01

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

  16. Study on tritium recovery from breeder materials

    International Nuclear Information System (INIS)

    Moriyama, H.; Moritani, K.

    1997-01-01

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

  17. 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.

  18. Generation of gaseous tritium standards

    International Nuclear Information System (INIS)

    Hohorst, F.A.

    1994-09-01

    The determination of aqueous and non-aqueous tritium in gaseous samples is one type of determination often requested of radioanalytical laboratories. This determination can be made by introducing the sample as a gas into a sampling train containing two silica gel beds separated by.a catalytic oxidizer bed. The first bed traps tritiated water. The sample then passes into and through the oxidizer bed where non-aqueous tritium containing species are oxidized to water and other products of combustion. The second silica gel bed then traps the newly formed tritiated water. Subsequently, silica gel is removed to plastic bottles, deionized water is added, and the mixture is permitted to equilibrate. The tritium content of the equilibrium mixture is then determined by conventional liquid scintillation counting (LSC). For many years, the moisture content of inert, gaseous samples has been determined using monitors which quantitatively electrolyze the moisture present after that moisture has been absorbed by phosphorous pentoxide or other absorbents. The electrochemical reaction is quantitative and definitive, and the energy consumed during electrolysis forms the basis of the continuous display of the moisture present. This report discusses the experimental evaluation of such a monitor as the basis for a technique for conversion of small quantities of SRMs of tritiated water ( 3 HOH) into gaseous tritium standards ( 3 HH)

  19. Tritium-labelled abscisic acid

    International Nuclear Information System (INIS)

    Pluciennik, H.; Michalski, L.

    1991-01-01

    A simple method for the preparation of biologically active abscisic acid (growth inhibiting plant hormone) labelled with tritium is described. The product obtained has a specific radioactivity of 1.12 GBq mmol -1 : the yield is about 60% as compared to the initial amount of the substance used. (author) 7 refs.; 2 figs

  20. Tritium Level in Romanian Precipitation

    Energy Technology Data Exchange (ETDEWEB)

    Varlam, C.; Stefanescu, I.; Faurescu, I.; Bogdan, D.; Soare, A. [Institute for Cryogenic and Isotope Technologies, Rm. Valcea (Romania); Duliu, O. G. [Faculty of Physics, University of Bucharest, Magurele (Romania)

    2013-07-15

    Romania is one of the countries that has no station included in GNIP (Global Network of Isotopes in Precipitation) on its territory. This paper presents results regarding the tritium concentration in precipitation for the period 1999-2009. The precipitation fell at the Institute for cryogenic and Isotope technologies (geographical coordinates: altitude 237 m, latitude 45{sup o}02'07' N, longitude 24{sup o}17'03' E) an was collected both individually and as a composite average of each month. It was individually measured and the average was calculated and compared with the tritium concentration measured in the composite sample. tritium concentration levels ranged from 9.9 {+-} 2.1 TU for 2004 and 13.7 {+-} 2.2 TU for 2009. Comparing the arithmetic mean values with the weighted mean for the period of observation, it was noticed that the higher absolute values of the weighted means were constant. It was found that for the calculated monthly average for the period of observation (1999-2009), the months with the maximum tritium concentration are the same as the months with the maximum amount of precipitation. This behaviour is typical for the monitored location. (author)

  1. Tritium transport analysis for CFETR WCSB blanket

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Pinghui, E-mail: phzhao@mail.ustc.edu.cn; Yang, Wanli; Li, Yuanjie; Ge, Zhihao; Nie, Xingchen; Gao, Zhongping

    2017-01-15

    Highlights: • A simplified tritium transport model for CFETR WCSB blanket was developed. • Tritium transport process in CFETR WCSB blanket was analyzed. • Sensitivity analyses of tritium transport parameters were carried out. - Abstract: Water Cooled Solid Breeder (WCSB) blanket was put forward as one of the breeding blanket candidate schemes for Chinese Fusion Engineering Test Reactor (CFETR). In this study, a simplified tritium transport model was developed. Based on the conceptual engineering design, neutronics and thermal-hydraulic analyses of CFETR WCSB blanket, tritium transport process was analyzed. The results show that high tritium concentration and inventory exist in primary water loop and total tritium losses exceed CFETR limits under current conditions. Conducted were sensitivity analyses of influential parameters, including tritium source, temperature, flow-rate capacity and surface condition. Tritium performance of WCSB blanket can be significantly improved under a smaller tritium impinging rate, a larger flow-rate capacity or a better surface condition. This work provides valuable reference for the enhancement of tritium transport behavior in CFETR WCSB blanket.

  2. Tritium behavior intentionally released in the room

    International Nuclear Information System (INIS)

    Kobayashi, K.; Hayashi, T.; Iwai, Y.; Yamanishi, T.; Willms, R. S.; Carlson, R. V.

    2008-01-01

    To construct a fusion reactor with high safety and acceptability, it is necessary to establish and to ensure tritium safe handling technology. Tritium should be well-controlled not to be released to the environment excessively and to prevent workers from excess exposure. It is especially important to grasp tritium behavior in the final confinement area, such as the room and/or building. In order to obtain data for actual tritium behavior in a room and/or building, a series of intentional Tritium Release Experiments (TREs) were planned and carried out within a radiologically controlled area (main cell) at Tritium System Test Assembly (TSTA) in Los Alamos National Laboratory (LANL) under US-JAPAN collaboration program. These experiments were carried out three times. In these experiments, influence of a difference in the tritium release point and the amount of hydrogen isotope for the initial tritium behavior in the room were suggested. Tritium was released into the main cell at TSTA/LANL. The released tritium reached a uniform concentration about 30 - 40 minutes in all the experiments. The influence of the release point and the amount of hydrogen isotope were not found to be important in these experiments. The experimental results for the initial tritium behavior in the room were also simulated well by the modified three-dimensional eddy flow analysis code FLOW-3D. (authors)

  3. Tritium migration in nuclear desalination plants

    International Nuclear Information System (INIS)

    Muralev, E.D.

    2003-01-01

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

  4. Storage and Assay of Tritium in STAR

    International Nuclear Information System (INIS)

    Longhurst, Glen R.; Anderl, Robert A.; Pawelko, Robert J.; Stoots, Carl J.

    2005-01-01

    The Safety and Tritium Applied Research (STAR) facility at the Idaho National Engineering and Environmental Laboratory (INEEL) is currently being commissioned to investigate tritium-related safety questions for fusion and other technologies. The tritium inventory for the STAR facility will be maintained below 1.5 g to avoid the need for STAR to be classified as a Category 3 nuclear facility. A key capability in successful operation of the STAR facility is the ability to receive, inventory, and dispense tritium to the various experiments underway there. The system central to that function is the Tritium Storage and Assay System (SAS).The SAS has four major functions: (1) receiving and holding tritium, (2) assaying, (3) dispensing, and (4) purifying hydrogen isotopes from non-hydrogen species.This paper describes the design and operation of the STAR SAS and the procedures used for tritium accountancy in the STAR facility

  5. Tritium pellet injection sequences for TFTR

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Milora, S.L.; Attenberger, S.E.; Singer, C.E.; Schmidt, G.L.

    1983-01-01

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

  6. Imaging of tritium implanted into graphite

    International Nuclear Information System (INIS)

    Malinowski, M.E.; Causey, R.A.

    1988-01-01

    The extensive use of graphite in plasma-facing surfaces of tokamaks such as the Tokamak Fusion Test Reactor, which has planned tritium discharges, makes two-dimensional tritium detection techniques important in helping to determine torus tritium inventories. We have performed experiments in which highly oriented pyrolytic graphite (HOPG) samples were first tritium implanted with fluences of ∼10 16 T/cm 2 at energies approx. 0 C resulted in no discernible motion of tritium along the basal plane, but did show that significant desorption of the implanted tritium occurred. The current results indicate that tritium in quantities of 10 12 T/cm 2 in tritiated components could be readily detected by imaging at lower magnifications

  7. Calculation of tritium release from reactor's stack

    International Nuclear Information System (INIS)

    Akhadi, M.

    1996-01-01

    Method for calculation of tritium release from nuclear to environment has been discussed. Part of gas effluent contain tritium in form of HTO vapor released from reactor's stack was sampled using silica-gel. The silica-gel was put in the water to withdraw HTO vapor absorbed by silica-gel. Tritium concentration in the water was measured by liquid scintillation counter of Aloka LSC-703. Tritium concentration in the gas effluent and total release of tritium from reactor's stack during certain interval time were calculated using simple mathematic formula. This method has examined for calculation of tritium release from JRR-3M's stack of JAERI, Japan. From the calculation it was obtained the value of tritium release as much as 4.63 x 10 11 Bq during one month. (author)

  8. Tritium systems test assembly stabilization

    International Nuclear Information System (INIS)

    Jasen, William G.; Michelotti, Roy A.; Anast, Kurt R.; Tesch, Charles

    2004-01-01

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

  9. Tritium issues in plasma wall interactions

    International Nuclear Information System (INIS)

    Tanabe, T.

    2009-01-01

    In order to establish a D-T fusion reactor as an energy source, it is not enough to have a DT burning plasma, and economical conversion of fusion energy to electricity and/or heat, a large enough margin of tritium breeding and tritium safety must be simultaneously achieved. In particular, handling of huge amount of tritium needs a significant effort to ensure that the radiation dose of radiological workers and of the public is below the limits specified by the International Commission on Radiological Protection. For the safety reasons, tritium in a reactor will be limited to only a few kg orders in weight, with radioactivity up to 10 17 Bq. Since public exposure to tritium is regulated at a level as tiny as a few Bq/cm 2 , tritium must be strictly confined in a reactor system with accountancy of an order of pg (pico-gram). Generally qualitative analysis with the accuracy of more than 3 orders of magnitude is hardly possible. We are facing to lots of safety concerns in the handling of huge amounts of radioactive tritium as a fuel and to be bred in a blanket. In addition, tritium resources are very limited. Not only for the safety reason but also for the saving of tritium resources, tritium retention in a reactor must be kept as small as possible. In the present tokamaks, however, hydrogen retention is significantly large, i.e. more than 20% of fueled hydrogen is continuously piled up in the vacuum vessel, which must not be allowed in a reactor. After the introduction of tritium as a hydrogen radioisotope, this lecture will present tritium issues in plasma wall interactions, in particular, fueling, retention and recovering, considering the handling of large amounts of tritium, i.e. confinement, leakage, contamination, permeation, regulations and tritium accountancy. Progress in overcoming such problems will be also presented. This document is made of the slides of the presentation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hirofumi; Hayashi, Takumi; Suzuki, Takumi; Nishi, Masataka [Japan Atomic Energy Research Inst., Naka Fusion Research Establishment, Department of Fusion Engineering Research, Naka, Ibaraki (Japan); Yoshida, Hiroshi [Japan Atomic Energy Research Inst., Naka Fusion Research Establishment, ITER-Joint Centeral Team, Naka, Ibaraki (Japan)

    2000-10-01

    Tritium permeation amount in a tritium storage bed with gas flowing calorimetric was evaluated under a condition of new operation mode for International Thermonuclear Experimental Reactor (ITER). As a result, tritium permeation under the new operation mode was estimated to be about twice of that under the practical operation mode. This result show that it would be regardless in a view point of material control of tritium, however, it was suggested to be required additional tritium removal or evacuate system in a view points of safety control or performance of accountability or thermal insulating of the tritium storage bed. (author)

  11. Tritium transport studies with use of the ISEP NPA during tritium trace experimental campaign on JET

    International Nuclear Information System (INIS)

    Mironov, M I; Afanasyev, V I; Murari, A; Santala, M; Beaumont, P

    2010-01-01

    The neutral particle analyzer (NPA) known as ISEP (Ion SEParator) was applied to measure the tritium neutral flux during the tritium trace experiment (TTE) on JET. The energy dependence (in the 5-28 keV energy range) of the tritium neutral flux rise time after a short ∼100 ms tritium gas puff into deuterium plasmas has been observed for the first time. The dependence has been interpreted as being due to the penetration of the tritium ions from the plasma boundary into the core and has been used for the calculation of the tritium diffusion coefficient and convective velocity values.

  12. Tritium confinement in a new tritium processing facility at the Savannah River Site

    International Nuclear Information System (INIS)

    Heung, L.K.; Owen, J.H.; Hsu, R.H.; Hashinger, R.F.; Ward, D.E.; Bandola, P.E.

    1991-01-01

    A new tritium processing facility, named the Replacement Tritium Facility (RTF), has been completed and is being prepared for startup at the Savannah River Site (SRS). The RTF has the capability to recover, purify and separate hydrogen isotopes from recycled gas containers. A multilayered confinement system is designed to reduce tritium losses to the environment. This confinement system is expected to confine and recover any tritium that might escape the process equipment, and to maintain the tritium concentration in the nitrogen glovebox atmosphere to less than 10 -2 μCi/cc tritium

  13. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

    Ono, F.; Yamawaki, M.

    1995-01-01

    In a fusion reactor or tritium-handling facilities, contamination of concrete by tritium and subsequent release from it to the reator or experimental room is a matter of problem for safe control of tritium and management of operational environment. In order to evaluate this tritium behavior, interaction of tritiated water with concrete or cement should be clarified. In the present study, HTO sorption and subsequent release from cement were experimentally studied.(1)Sorption experiments were conducted using columns packed with cement particles of different sizes. From the analysis of the breakthrough curve, tritium diffusivity in macropores and microparticles were evaluated.(2)From the short-term tritium release experiments, effective desorption rate constants were evaluated and the effects of temperature and moisture were studied.(3)In the long-term tritium release experiments to 6000h, the tritium release mechanism was found to be composed of three kinds of water: initially from capillary water, and in the second stage from gel water and from the water in the cement crystal.(4)Tritium release behavior by heat treatment to 800 C was studied. A high temperature above 600 C was required for the tritium trapped in the crystal water to be released. (orig.)

  14. The LLNL portable tritium processing system

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

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

  15. The Chalk River Tritium Extraction Plant

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Harrison, T.E.; Spagnolo, D.A.

    1990-01-01

    The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1-35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T 2 . The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: it produces pure tritium for use in fusion research and development, and it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors (author)

  16. The Chalk River Tritium Extraction Plant

    Energy Technology Data Exchange (ETDEWEB)

    Holtslander, W J; Harrison, T E; Spagnolo, D A

    1990-07-01

    The Chalk River Tritium Extraction Plant for removal of tritium from heavy water is described. Tritium is present in the heavy water from research reactors in the form of DTO at a concentration in the range of 1-35 Ci/kg. It is removed by a combination of catalytic exchange to transfer the tritium from DTO to DT, followed by cryogenic distillation to separate and concentrate the tritium to T{sub 2}. The tritium product is reacted with titanium and packaged for transportation and storage as titanium tritide. The plant processes heavy water at a rate of 25 kg/h and removes 80% of the tritium and 90% of the protium per pass. Catalytic exchange is carried out in the liquid phase using a proprietary wetproofed catalyst. The plant serves two roles in the Canadian fusion program: it produces pure tritium for use in fusion research and development, and it demonstrates on an industrial scale many of the tritium technologies that are common to the tritium systems in fusion reactors (author)

  17. Estimation of Biological Effects of Tritium.

    Science.gov (United States)

    Umata, Toshiyuki

    2017-01-01

    Nuclear fusion technology is expected to create new energy in the future. However, nuclear fusion requires a large amount of tritium as a fuel, leading to concern about the exposure of radiation workers to tritium beta radiation. Furthermore, countermeasures for tritium-polluted water produced in decommissioning of the reactor at Fukushima Daiichi Nuclear Power Station may potentially cause health problems in radiation workers. Although, internal exposure to tritium at a low dose/low dose rate can be assumed, biological effect of tritium exposure is not negligible, because tritiated water (HTO) intake to the body via the mouth/inhalation/skin would lead to homogeneous distribution throughout the whole body. Furthermore, organically-bound tritium (OBT) stays in the body as parts of the molecules that comprise living organisms resulting in long-term exposure, and the chemical form of tritium should be considered. To evaluate the biological effect of tritium, the effect should be compared with that of other radiation types. Many studies have examined the relative biological effectiveness (RBE) of tritium. Hence, we report the RBE, which was obtained with radiation carcinogenesis classified as a stochastic effect, and serves as a reference for cancer risk. We also introduce the outline of the tritium experiment and the principle of a recently developed animal experimental system using transgenic mouse to detect the biological influence of radiation exposure at a low dose/low dose rate.

  18. Tritium inventories and tritium safety design principles for the fuel cycle of ITER

    International Nuclear Information System (INIS)

    Cristescu, I.R.; Cristescu, I.; Doerr, L.; Glugla, M.; Murdoch, D.

    2007-01-01

    Within the tritium plant of ITER a total inventory of about 2-3 kg will be necessary to operate the machine in the DT phase. During plasma operation, tritium will be distributed in the different sub-systems of the fuel cycle. A tool for tritium inventory evaluation within each sub-system of the fuel cycle is important with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems; however, tritium accounting may be achieved by modelling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the sub-systems. To get reliable results, an accurate dynamic modelling of the tritium content in each sub-system is necessary. A dynamic model (TRIMO) for tritium inventory calculation reflecting the design of each fuel cycle sub-systems was developed. The amount of tritium needed for ITER operation has a direct impact on the tritium inventories within the fuel cycle sub-systems. As ITER will function in pulses, the main characteristics that influence the rapid tritium recovery from the fuel cycle as necessary for refuelling are discussed. The confinement of tritium within the respective sub-systems of the fuel cycle is one of the most important safety objectives. The design of the deuterium/tritium fuel cycle of ITER includes a multiple barrier concept for the confinement of tritium. The buildings are equipped with a vent detritiation system and re-circulation type room atmosphere detritiation systems, required for tritium confinement barrier during possible tritium spillage events. Complementarily to the atmosphere detritiation systems, in ITER a water detritiation system for tritium recovery from various sources will also be operated

  19. The tritium content of precipitation and groundwater at Yola, Nigeria ...

    African Journals Online (AJOL)

    Tritium is a radioactive isotope of hydrogen which occurs in precipitation. In groundwater studies tritium measurements give information on the time of recharge to the system; the tritium content of precipitation being used to estimate the input of tritium to the groundwater system. At Yola, the tritium ontents in precipitation and ...

  20. Tritium in Exit Signs | RadTown USA | US EPA

    Science.gov (United States)

    2017-08-07

    Many exit signs contain tritium to light the sign without batteries or electricity. Using tritium in exit signs allows the sign to remain lit if the power goes out. Tritium is most dangerous when it is inhaled or swallowed. Never tamper with a tritium exit sign. If a tritium exit sign is broken, leave the area immediately and notify the building maintenance staff.

  1. Simulation of tritium behavior after intended tritium release in ventilated room

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko; Kobayashi, Kazuhiro; Nishi, Masataka

    2001-01-01

    At the Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI), Caisson Assembly for Tritium Safety study (CATS) with 12 m 3 of large airtight vessel (Caisson) was fabricated for confirmation and enhancement of fusion reactor safety to estimate tritium behavior in the case where a tritium leak event should happen. One of the principal objectives of the present studies is the establishment of simulation method to predict the tritium behavior after the tritium leak event should happen in a ventilated room. The RNG model was found to be valid for eddy flow calculation in the 50 m 3 /h ventilated Caisson with acceptable engineering precision. The calculated initial and removal tritium concentration histories after intended tritium release were consistent with the experimental observations in the 50 m 3 /h ventilated Caisson. It is found that the flow near a wall plays an important role for the tritium transport in the ventilated room. On the other hand, tritium behavior intentionally released in the 3,000 m 3 of tritium handling room was investigated experimentally under a US-Japan collaboration. The tritium concentration history calculated with the same method was consistent with the experimental observations, which proves that the present developed method can be applied to the actual scale of tritium handling room. (author)

  2. Behaviour of tritium in the environment

    International Nuclear Information System (INIS)

    1979-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, M. [National Institute for Fusion Science, Toki, Gifu (Japan); Sugiyama, T. [Nagoya University, Fro-cho, Chikusa-ku, Nagoya (Japan)

    2015-03-15

    The detection of low level tritium is one of the key issues for tritium management in tritium handling facilities. Such a detection can be performed by tritium monitors based on proton conducting oxide technique. We tested a tritium monitoring system composed of a commercial proportional counter combined with an electrochemical hydrogen pump equipped with CaZr{sub 0.9}In{sub 0.1}O{sub 3-α} as proton conducting oxide. The hydrogen pump operated at 973 K under electrolysis conditions using tritiated water vapor (HTO). The proton conducting oxide extracts tritium molecules (HT) from HTO and tritium concentration is measured by the proportional counter. The advantage of the proposed tritium monitoring system is that it is able to convert HTO into molecular hydrogen.

  4. In-pile test of tritium release from tritium breeding materials (VOM-21H experiment)

    International Nuclear Information System (INIS)

    Kurasawa, Toshimasa; Takeshita, Hidefumi; Watanabe, Hitoshi; Yoshida, Hiroshi.

    1986-10-01

    Material development and blanket design of lithium-based ceramics such as lithium oxide, lithium aluminate, lithium silicate and lithium zirconate have been performed in Japan, United State of America and Europian Communities. Lithium oxide is a most attractive candidate for tritium breeding materials because of its high lithium density, high thermal conductivity and good tritium release performance. This work has been done to clarify the characteristics of tritium release and recovery from Li 2 O by means of in-situ tritium release measurement. The effects of temperature and sweep gas composition on the tritium release were investigated in this VOM-21H Experiment. Good measurement of tritium release was achieved but there were uncertainties in reproduciblity of data. The experimental results show that the role of surface adsorption/desorption makes a significant contribution to the tritium release and tritium inventory. Also, it is necessary to define the rate limiting process either diffusion or surface adsorption/desorption. (author)

  5. Tritium labelled steroids, preparation process and application to synthesis of tritium labelled estrane derivatives

    International Nuclear Information System (INIS)

    1978-01-01

    Process for preparing new steroids labelled with tritium in 6.7 and comprising in 3 a blocked ketonic group as ketal, thioketal or derivatives. Application of these products to the synthesis of tritium labelled estrane derivatives [fr

  6. Radiological training for tritium facilities

    International Nuclear Information System (INIS)

    1996-12-01

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

  7. 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.

  8. Tritium environmental transport studies at TFTR

    International Nuclear Information System (INIS)

    Ritter, P.D.; Dolan, T.J.; Longhurst, G.R.

    1993-01-01

    Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a weak after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER)

  9. Tritium environmental transport studies at TFTR

    Science.gov (United States)

    Ritter, P. D.; Dolan, T. J.; Longhurst, G. R.

    1993-06-01

    Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a week after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER).

  10. Purification of tritium-free water

    International Nuclear Information System (INIS)

    Hussain, S.D.

    1982-10-01

    Ground water which has been out of contact with the atmosphere for a long time as compared to the half life of tritium (12.43 years) does not contain any measureable amount of tritium. Such water is called tritium-free water. It may contain dissolved and suspended impurities and has to be purified before it can be used for the preparation of blanks and standards required in the routine measurement of low level tritium in water samples. The purification of tritium-free water by distillation in a closed system has been described. The quality of processed tritium-free water was precisely checked at International Atomic Energy Agency (IAEA) Vienna and found satisfactory. (authors)

  11. Modeling tritium transport in the environment

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.

    1986-01-01

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

  12. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

    Skinner, C.H.; Coad, J.P.; Federici, G.

    2003-01-01

    Tritium removal is a major unsolved development task for next-step devices with carbon plasma-facing components. The 2-3 order of magnitude increase in duty cycle and associated tritium accumulation rate in a next-step tokamak will place unprecedented demands on tritium removal technology. The associated technical risk can be mitigated only if suitable removal techniques are demonstrated on tokamaks before the construction of a next-step device. This article reviews the history of codeposition, the tritium experience of TFTR (Tokamak Fusion Test Reactor) and JET (Joint European Torus) and the tritium removal rate required to support ITER's planned operational schedule. The merits and shortcomings of various tritium removal techniques are discussed with particular emphasis on oxidation and laser surface heating

  13. Comparison of Tritium Component Failure Rate Data

    International Nuclear Information System (INIS)

    Lee C. Cadwallader

    2004-01-01

    Published failure rate values from the US Tritium Systems Test Assembly, the Japanese Tritium Process Laboratory, the German Tritium Laboratory Karlsruhe, and the Joint European Torus Active Gas Handling System have been compared. This comparison is on a limited set of components, but there is a good variety of data sets in the comparison. The data compared reasonably well. The most reasonable failure rate values are recommended for use on next generation tritium handling system components, such as those in the tritium plant systems for the International Thermonuclear Experimental Reactor and the tritium fuel systems of inertial fusion facilities, such as the US National Ignition Facility. These data and the comparison results are also shared with the International Energy Agency cooperative task on fusion component failure rate data

  14. History of 232-F, tritium extraction processing

    International Nuclear Information System (INIS)

    Blackburn, G.W.

    1994-08-01

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

  15. Effects of interfering constituents on tritium smears

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-02-01

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

  17. Regulating tritium in drinking water

    International Nuclear Information System (INIS)

    Fluke, R.

    1994-01-01

    This article incorporates an article by E. Koehl from an internal Ontario Hydro publication, and a letter from the Joint Committee of Health and Safety of the Royal Society of Canada and the Canadian Academy of Engineering, submitted to the Ontario Minister of the Environment and Energy. The Advisory Committee on Environmental Standards had recommended that the limit for tritium in Ontario drinking water be reduced from 40,000 to 100 Bq/L, with a further reduction to 20 in five years. Some facts and figures are adduced to show that the effect of tritium in drinking water in Ontario is negligible compared to the effect of background radiation. The risk from tritium to the people of Ontario is undetectably small, and the attempt to estimate this risk by linear extrapolation is extremely dubious. Regulation entails social and economic costs, and the government ought to ensure that the benefits exceed the costs. The costs translate into nothing less than wasted opportunity to save lives in other ways. 3 refs

  18. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

    This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

  19. Refurbishing tritium contaminated ion sources

    International Nuclear Information System (INIS)

    Wright, K.E.; Carnevale, R.H.; McCormack, B.E.; Stevenson, T.; Halle, A. von

    1995-01-01

    Extended tritium experimentation on TFTR has necessitated refurbishing Neutral Beam Long Pulse Ion Sources (LPIS) which developed operational difficulties, both in the TFTR Test Cell and later, in the NB Source Refurbishment Shop. Shipping contaminated sources off-site for repair was not permissible from a transport and safety perspective. Therefore, the NB source repair facility was upgraded by relocating fixtures, tooling, test apparatus, and three-axis coordinate measuring equipment; purchasing and fabricating fume hoods; installing exhaust vents; and providing a controlled negative pressure environment in the source degreaser/decon area. Appropriate air flow monitors, pressure indicators, tritium detectors and safety alarms were also included. The effectiveness of various decontamination methods was explored while the activation was monitored. Procedures and methods were developed to permit complete disassembly and rebuild of an ion source while continuously exhausting the internal volume to the TFTR Stack to avoid concentrations of tritium from outgassing and minimize personnel exposure. This paper presents upgrades made to the LPIS repair facility, various repair tasks performed, and discusses the effectiveness of the decontamination processes utilized

  20. Tritium production in fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.

    1981-08-01

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

  1. Tritium calorimeter setup and operation

    International Nuclear Information System (INIS)

    Rodgers, David E.

    2002-01-01

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

  2. Tritium radioluminescent devices, Health and Safety Manual

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-06-01

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

  3. Overview of light sources powered by tritium

    International Nuclear Information System (INIS)

    Wu Jian; Lei Jiarong; Liu Wenke

    2012-01-01

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

  4. Degradation of elastomers by tritium beta radiation

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  5. Tritium radioluminescent devices, Health and Safety Manual

    International Nuclear Information System (INIS)

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

    1995-06-01

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

  6. Effluent Treatment Facility tritium emissions monitoring

    International Nuclear Information System (INIS)

    Dunn, D.L.

    1991-01-01

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

  7. The movement of tritium in ecological systems

    International Nuclear Information System (INIS)

    Polevoy, Y; Laichter, Y.

    1988-11-01

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

  8. Tritium monitoring equipments for animal experiment facilities

    International Nuclear Information System (INIS)

    Sato, Hiroo

    1980-01-01

    Animal experiment facilities using tritium are described with reference to laws and regulations concerning radiological safety. Usual breeding facilities and surrounding conditions at non-radioactive animal experiments are summarized on feasible and effective designs of tritium monitors. Characteristics and desirable arrangements of various kinds of tritium monitors such as ionization chambers, proportional counters and liquid scintillation detectors are discussed from the standpoint of monitoring for room, glove-box, stack, liquid waste and personnel. (J.P.N.)

  9. Tritium in the environment. NCRP Report No. 62

    International Nuclear Information System (INIS)

    Eisenbud, M.

    1979-01-01

    The NCRP (National Council on Radiation Protection and Measurements) Report No. 62 on tritium is described. Tritium production from various sources, distribution and environmental kinetics, biological behaviour and the dosimetry of tritium are discussed. (author)

  10. National pattern for the realization of the unit of the dose speed absorbed in air for beta radiation. (Method: Ionometer, cavity of Bragg-Gray implemented in an extrapolation chamber with electrodes of variable separation, exposed to a field of beta radiation of 90Sr/90Y)

    International Nuclear Information System (INIS)

    Alvarez R, M. T.; Morales P, J. R.

    2001-01-01

    From the year of 1987 the Department of Metrology of the ININ, in their Secondary Laboratory of Calibration Dosimetric, has a patron group of sources of radiation beta and an extrapolation chamber of electrodes of variable separation.Their objective is to carry out of the unit of the dose speed absorbed in air for radiation beta. It uses the ionometric method, cavity Bragg-Gray in the extrapolation chamber with which it counts. The services that offers are: i) it Calibration : Radioactive Fuentes of radiation beta, isotopes: 90 Sr/ 90 Y; Ophthalmic applicators 9 0 S r/ 90 Y; Instruments for detection of beta radiation with to the radiological protection: Ionization chambers, Geiger-Muller, etc.; Personal Dosemeters. ii) Irradiation with beta radiation of materials to the investigation. (Author)

  11. Measurement of dose equivalent with personal dosemeters and instrumentation of radiological protection in the new operative magnitudes ICRU, for external fields of radiation beta. Part IV. Survey of the angular response of instruments used in radiological protection in secondary patron fields of beta radiation (90Sr/90Y (1850 MBq and 74 MBq), 204TI (18.5 MBq) and 147Pm (518 MBq)

    International Nuclear Information System (INIS)

    Alvarez R, J.T.

    1994-02-01

    Tests type were made (type test) in the following commercial instrumentation commonly used in radiological protection: Geiger-Mueller Counters (FH40 FE), Plastic Scintillators (NE-BP/6/4A), Ionization Chambers (RO-5) and Proportional Counters (HP-100A; gas:P-10). With object of checking the possibility that these they can carry out the new operative unit ICRU, H' (0.07; α). The tests consisted on determining the energy and angular response of the detectors in secondary patron fields of beta radiation, for isotopes of 90 Sr/ 90 Y (1850 MBq and 74 MBq and 147 Pm(518 MBq). The results show the inadequate of these commercial instruments for the realization of the H' operative unit (0.07; α) in beta external fields. Due to flaws in the design, construction and calibration of the instruments for this type of radiation fields (Author)

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

    International Nuclear Information System (INIS)

    Akulov, G.P.; Snetkova, E.V.; Kaminskij, Yu.L.; Kudelin, B.K.; Efimova, V.L.

    1991-01-01

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

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

    Science.gov (United States)

    Hebert, Detlef

    2011-09-01

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

  14. Tritium loss in molten flibe systems

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Anderl, R.A. [Idaho National Eng. and Environ. Lab., Idaho Falls, ID (United States); Scott Willms, R. [Los Alamos National Lab., NM (United States)

    2000-04-01

    An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF.BeF{sub 2}, commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons

  15. Process and system for removing tritium

    International Nuclear Information System (INIS)

    Ridgely, J.N.

    1976-01-01

    A process and system for removing tritium, particularly from high temperature gas cooled atomic reactors (HTGR), is disclosed. Portions of the reactor coolant, which is permeated with the pervasive tritium atom, are processed to remove the tritium. Under conditions of elevated temperature and pressure, the reactor coolant is combined with gaseous oxygen, resulting in the formation of tritiated water vapor from the tritium in the reactor coolant and the gaseous oxygen. The tritiated water vapor and the remaining gaseous oxygen are then successively removed by fractional liquefaction steps. The reactor coolant is then recirculated to the reactor

  16. Tritium stripping by a catalytic exchange stripper

    International Nuclear Information System (INIS)

    Heung, L.K.; Gibson, G.W.; Ortman, M.S.

    1991-01-01

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

  17. Tritium immobilization and packaging using metal hydrides

    International Nuclear Information System (INIS)

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

    1981-04-01

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

  18. 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.

  19. Tritium loss in molten flibe systems

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Scott Willms, R.

    2000-01-01

    An emerging issue relative to beryllium technology in fusion involves tritium interactions with molten beryllium-bearing salts. Innovative designs for fusion reactors, both magnetic and inertially confined, feature the molten salt mixture 2LiF.BeF 2 , commonly called Flibe, as a tritium breeder and coolant. Tritium is bred in the Flibe as neutrons from the plasma are absorbed by Li atoms, which then transmute to tritium and helium. Transmutation of tritium from Be also occurs. Among the issues to be resolved for such coolant systems is the potential loss of tritium from the Flibe coolant to the walls of the system, particularly through heat exchanger tubes, and from there into secondary coolants or working fluids and the environment. Effectively removing tritium from Flibe in clean-up units is also important. In quiescent or low Reynolds number flow, tritium movement through Flibe is governed by diffusion. For Flibe in turbulent flow, as in heat exchanger tubes, transport is by turbulent mixing, and the same flow conditions and structural design features that maximize heat transfer to the heat exchanger walls will enhance the transport of tritium to those same surfaces. Analyses have been performed to estimate the fractional loss of tritium through heat exchanger tubes and release rates from Flibe droplets in vacuum disengagers in molten Flibe systems. The calculations suggest unacceptably large losses of tritium through heat exchanger tubes. The gravity of the implications of these estimates calls for experimental verification to determine if tritium losses through molten Flibe heat exchangers or other Flibe systems can really be so high and whether vacuum disengagers will really work. There is also a need for better information on evolution of tritium from Flibe droplets in a vacuum. Several experiments are presently being planned to address these issues and are discussed. These include experiments to induce tritium in Flibe using spontaneous fission neutrons

  20. Management of Tritium in ITER Waste

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. Tritium resources available for fusion reactors

    Science.gov (United States)

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

    2018-02-01

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

  2. Tritium proof-of-principle pellet injector

    International Nuclear Information System (INIS)

    Fisher, P.W.

    1991-07-01

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

  3. Effects of tritium on electron multiplier performance

    International Nuclear Information System (INIS)

    Kerst, R.A.; Malinowski, M.E.

    1980-01-01

    In developing diagnostic instruments for fusion reactors, it is necessary to measure the effects of tritium contamination on channel electron multipliers (CEM). A CEM was exposed to T 2 pressures of up to 1.5 x 10 -1 Pa, with exposure quantities ranging up to 8800 Pa-s. The counting rate of the CEM is shown to consist of a prompt (Type I) signal caused by gas-phase tritium and a residual (Type II) signal, probably caused by near-surface tritium. The potential for using CEMs for observing the dynamics of tritium adsorption and absorption is discussed

  4. Separation of tritium from other hydrogen isotopes

    International Nuclear Information System (INIS)

    Roth, E.

    1988-01-01

    The paper describes a plant that has been operated at Marcoule for tritium production and used thermal diffusion enrichment, a facility that was built in Saclay to enrich hydrogen in tritium for low level measurements, and the Laue Langevin Institute tritium extraction plant. Details are given on the project under construction for the tritium separation facility at JET using Gas Chromatography, and on proposals for circuits for NET. Studies on catalysers for liquid phase catalytic exchange, on electrolysers, or different gas chromatography arrangements, are described. Systems designed for reprocessing plants, for detritiation of heavy water by distillation are briefly accounted for

  5. Measurement of tritium concentration in urine

    International Nuclear Information System (INIS)

    Sekiyama, Shigenobu; Deshimaru, Takehide

    1979-01-01

    Concerning the safety management of the advanced thermal reactor ''Fugen'', the internal exposure management for tritium is important, because heavy water is used as the moderator in the reactor, and tritium is produced in the heavy water. Tritium is the radioactive nuclide with the maximum β-ray energy of 18 keV, and the radiation exposure is limited to the internal exposure in human bodies, as tritium is taken in through the skin and by breathing. The tritium concentration in urine of the operators of the Fugen plant was measured. As for tritium measurement, the analysis of raw urine, the analysis after passing through mixed ion exchange resin and the analysis after distillation are applied. The scintillator, the liquid scintillation counter, the ion exchange resin and the distillator are introduced. The preliminary survey was conducted on the urine sample, the scintillator the calibration, etc. The measuring condition, the measurement of efficiency, and the limitation of detection with various background are explained, with the many experimental data and the calculating formula. Concerning the measured tritium concentration in urine, the tritium concentrations in distilled urine, raw urine and the urine refined with ion exchange resin were compared, and the correlation formulae are presented. The actual tritium concentration value in urine was less than 50 pci/ml. The measuring methods of raw urine and the urine refined with ion exchange resin are adequate as they are quick and accurate. (Nakai, Y.)

  6. Behavior of tritium in the environment. Proceedings series

    Energy Technology Data Exchange (ETDEWEB)

    1979-01-01

    Fifty papers are presented in these proceedings. Individual items are being entered onto the data base. The papers are grouped into seven sections for purposes of continuity. These sections include: distribution of tritium (7 papers); evaluation of future discharges (3 papers); measurement of tritium (3 papers); tritium in the aquatic environment (10 papers); tritium in the terrestrial environment (13 papers); tritium in man (8 papers); and monitoring of tritium (6 papers). (ERB)

  7. Determination of tritium by counting; Dosage du tritium par comptage

    Energy Technology Data Exchange (ETDEWEB)

    Schott, R; Froment, G; Pinson, J; Genty, C [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes

    1968-07-01

    Ionisation chamber assay of tritium in any gaseous mixture is a simple, fast and accurate method. We used the method of relative determination by comparison to a standard rather than the method of absolute assay in which case the constants are known with too little accuracy. The efficiency of the chamber was studied in connection to the pressure inside the chamber and its total volume. The calibration is linear in the range we are taking into account (1 to 80 millicuries). The reproducibility of the method is good: 13 runs gave a coefficient of variation of 1.6 per cent. The relative accuracy was found equal to {+-} 1.3 per cent. To end the paper, we describe in detail the apparatus and the ways of proceedings. (authors) [French] Le comptage du tritium par chambre d'ionisation est une methode simple, rapide et precise pour determiner la teneur en tritium d'un melange gazeux quelconque. Nous avons prefere utiliser la methode de determination relative par rapport a un etalon car, dans le cas d'une determination absolue, les constantes sont connues avec une trop grande incertitude. L'efficacite de la chambre a ete etudiee en fonction de la variation de la pression d'argon a l'interieur de la chambre et du volume total, de cette derniere. L'etalonnage s'est revele lineaire dans le domaine de mesures qui nous interessaient (1 a 80 millicuries). La reproductibillte de la methode est tres bonne, le coefficient de variation pour une serie de 13 essais etant de 1,6 pour cent, quant a la precision relative, elle a ete evaluee a {+-} 1,3 pour cent. Pour terminer, nous donnons une description detaillee de l'appareillage utilise et du mode operatoire suivi. (auteurs)

  8. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Evans, K.

    1983-01-01

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

  11. Tritium sorption on protective coatings for concrete

    International Nuclear Information System (INIS)

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

    1992-11-01

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

  12. Transfer and incorporation of tritium in mammals

    International Nuclear Information System (INIS)

    Hoek, J. van den; Juan, N.B.

    1979-01-01

    The metabolism of tritium in mammals has been studied in a number of laboratories which have participated in the IAEA Co-ordinated Research Programme on the Behaviour of Tritium in the Environment. The results of these studies are discussed and related to data obtained elsewhere. The animals studied are small laboratory and domestic animals. Tritium has been administered as THO, both in single and long-term dosing experiments, and also as organically bound tritium. The biological half-life of tritium in the body water pool has been determined in different species. The following values have been found: 1.1 days in mice; 13.2 days in kangaroo rats; 3.8 days in pigs; 4.1 days in lactating versus 8.3 in non-lactating goats and 3.1-4.0 days in lactating cows and steers. Much attention has been paid to the incorporation of tritium into organic constituents, both in the animal organism (organs, tissues) and in the secretions of the animal after continuous administration of tritium, mostly as THO. When compared with tritium levels in body water, and expressed as the ratio of specific activities, values of 0.25 and 0.40 have been found in mice liver and testis respectively. In cow's milk, these ratios vary from 0.30 for casein to 0.60 for lactose. The transfer of tritium into milk after continuous ingestion of THO by a lactating cow is about 1.50% of the daily ingested tritium per litre of milk. Some results of experiments, utilizing organically bound tritium, are also presented. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.

    1990-12-01

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

  17. Tritium removal using vanadium hydride

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  18. A new tritium process monitor based on scintillating fibres

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  19. Conceptual design of tritium treatment facility

    International Nuclear Information System (INIS)

    Tachikawa, Katsuhiro

    1982-01-01

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

  20. A review of tritium licensing requirements

    International Nuclear Information System (INIS)

    Meikle, A.B.

    1982-12-01

    Present Canadian regulations and anticipated changes to these regulations relevant to the utilization of tritium in fusion facilities and in commercial applications have been reviewed. It is concluded that there are no serious licensing obstacles, but there are a number of requirements which must be met. A license will be required from Atomic Energy Control Board if Ontario Hydro tritium is to be applied by other users. A license is required from the Federal Government to export or import tritium. A licensed container will be required for the storage and shipping of tritium. The containers being designed by AECL and Ontario Hydro and which are currently being tested will adequately store and ship all of the Ontario Hydro tritium but are unnecessarily large for the small quantities required by the commercial tritium users. Also, some users may prefer to receive tritium in gaseous form. An additional, smaller container should be considered. The licensing of overseas fusion facilities for the use of tritium is seen as a major undertaking offering opportunities to Canadian Fusion Fuels Technology Project to undertake health, safety and environmental analysis on behalf of these facilities

  1. Immobilization and packaging of recovered tritium

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Miller, J.M.

    1982-09-01

    The evaluation of metal hydrides as a medium for immobilization of tritium is reviewed. The work demonstrated methods of preparation and examined the properties of titanium and zirconium hydride for this application. Methods of packaging the metal hydrides for transportation and recoverable storage of tritium were also examined

  2. A study of electrolytic tritium production

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Enantiospecific tritium labeling of 28-homocastasterone

    Czech Academy of Sciences Publication Activity Database

    Elbert, Tomáš; Patil, Mahadeo Rajshekhar; Marek, Aleš

    2017-01-01

    Roč. 60, č. 3 (2017), s. 176-182 ISSN 0362-4803 R&D Projects: GA AV ČR IAA400550801 Institutional support: RVO:61388963 Keywords : 28-homocastasterone * brassinosteroids * enantiospecific reaction * tritium dehalogenation * tritium labeling Subject RIV: CC - Organic Chemistry OBOR OECD: Organic chemistry Impact factor: 1.745, year: 2016

  4. Synthesis of tritium-labeled fosfomycin

    International Nuclear Information System (INIS)

    Mertel, H.E.; Meriwether, H.T.

    1982-01-01

    Tritium gas was used as a labeling agent for the preparation of [1,2- 3 H]fosfomycin. Introduction of tritium into a precursor, the synthesis including resolution of the intermediate racemic 1,2-epoxypropylphosphonic acid, and preparation of both amine and calcium salts of the labeled antibiotic are described. (author)

  5. Tritium Systems Test Facility. Volume II. Appendixes

    International Nuclear Information System (INIS)

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

    1976-10-01

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

  6. Tritium Systems Test Facility. Volume I

    International Nuclear Information System (INIS)

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

    1976-10-01

    Sandia Laboratories proposes to build and operate a Tritium Systems Test Facility (TSTF) in its newly completed Tritium Research Laboratory at Livermore, California (see frontispiece). The facility will demonstrate at a scale factor of 1:200 the tritium fuel cycle systems for an Experimental Power Reactor (EPR). This scale for each of the TSTF subsystems--torus, pumping system, fuel purifier, isotope separator, and tritium store--will allow confident extrapolation to EPR dimensions. Coolant loop and reactor hall cleanup facilities are also reproduced, but to different scales. It is believed that all critical details of an EPR tritium system will be simulated correctly in the facility. Tritium systems necessary for interim devices such as the Ignition Test Reactor (ITR) or The Next Step (TNS) can also be simulated in TSTF at other scale values. The active tritium system will be completely enclosed in an inert atmosphere glove box which will be connected to the existing Gas Purification System (GPS) of the Tritium Research Laboratory. In effect, the GPS will become the scaled environmental control system which otherwise would have to be built especially for the TSTF

  7. Tritium waste disposal technology in the US

    International Nuclear Information System (INIS)

    Albenesius, E.L.; Towler, O.A.

    1983-01-01

    Tritium waste disposal methods in the US range from disposal of low specific activity waste along with other low-level waste in shallow land burial facilities, to disposal of kilocurie amounts in specially designed triple containers in 65' deep augered holes located in an aird region of the US. Total estimated curies disposed of are 500,000 in commercial burial sites and 10 million curies in defense related sites. At three disposal sites in humid areas, tritium has migrated into the ground water, and at one arid site tritium vapor has been detected emerging from the soil above the disposal area. Leaching tests on tritium containing waste show that tritium in the form of HTO leaches readily from most waste forms, but that leaching rates of tritiated water into polymer impregnated concrete are reduced by as much as a factor of ten. Tests on improved tritium containment are ongoing. Disposal costs for tritium waste are 7 to 10 dollars per cubic foot for shallow land burial of low specific activity tritium waste, and 10 to 20 dollars per cubic foot for disposal of high specific activity waste. The cost of packaging the high specific activity waste is 150 to 300 dollars per cubic foot. 18 references

  8. Tritium labeling of detonation nanodiamonds.

    Science.gov (United States)

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

    2014-03-18

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

  9. HYLIFE-II tritium management system

    International Nuclear Information System (INIS)

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

    1993-06-01

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

  10. Experiences with decontaminating tritium-handling apparatus

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  11. Uptake of atmospheric tritium by market foods

    International Nuclear Information System (INIS)

    Inoue, Y.; Tanaka-Miyamoto, K.; Iwakura, T.

    1992-01-01

    In this paper uptake of tritium by market foods from tritiated water vapor in the air is investigated using cereals and beans purchased in Deep River, Canada. The concentrations of tissue free water tritium (TFWT) and organically bound tritium (OBT) range from 12 to 79% and from 10 to 38% respectively, of that estimated for atmospheric water vapor of the sampling month. The specific activity ratios of OBT to TFWT were constant for cereals, but variable for beans. The elevated OBT was shown to be the result of isotopic exchange of labile hydrogen by the fact that washing the foods with tritium free-water reduced their tritium contents to levels characteristic of their production sites

  12. Disposal of tritium-exposed metal hydrides

    International Nuclear Information System (INIS)

    Nobile, A.; Motyka, T.

    1991-01-01

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

  13. DOE handbook: Tritium handling and safe storage

    International Nuclear Information System (INIS)

    1999-03-01

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

  14. Diurnal variations of tritium uptake by plants

    International Nuclear Information System (INIS)

    Hettinger, M.; Diabate, S.; Strack, S.

    1991-02-01

    The influence of the diurnal cycle is important for the behaviour of environmental tritium in the vegetation. A mathematical model has been used to calculate the deposition of tritium in plants as a function of diurnal variations of climatic parameters. The necessary physiological parameters (relationship of net photosynthesis and growth) were derived from growth experiments for tomatoes and maize. In chamber experiments, tomato and maize plants were exposed to tritium with natural diurnal variations of the climatic conditions. Within the range of standard deviations the measured concentrations of tritium in tissue free water of tomatoes correspond well to the estimated values. Furthermore, the incorporation into non-exchangeable organically bound tritium (OBT nx) can be sufficiently modelled and explained. There are deviations from the estimated concentrations in some parts of maize leaves. (orig.) [de

  15. DOE handbook: Tritium handling and safe storage

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

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

  16. Methane generated from graphite--tritium interaction

    International Nuclear Information System (INIS)

    Coffin, D.O.; Walthers, C.R.

    1979-01-01

    When hydrogen isotopes are separated by cryogenic distillation, as little as 1 ppM of methane will eventually plug the still as frost accumulates on the column packings. Elemental carbon exposed to tritium generates methane spontaneously, and yet some dry transfer pumps, otherwise compatible with tritium, convey the gas with graphite rotors. This study was to determine the methane production rate for graphite in tritium. A pump manufacturer supplied graphite samples that we exposed to tritium gas at 0.8 atm. After 137 days we measured a methane synthesis rate of 6 ng/h per cm 2 of graphite exposed. At this rate methane might grow to a concentration of 0.01 ppM when pure tritium is transferred once through a typical graphite--rotor transfer pump. Such a low methane level will not cause column blockage, even if the cryogenic still is operated continuously for many years

  17. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  18. The organic tritium in the environment

    International Nuclear Information System (INIS)

    Kirchmann, R.

    1979-01-01

    Sources, organization process, and biological availability of organic tritium released in the environment, transfer of organic tritium in the environment from methane or soil to plants and from food to mammals, transfer of tritium in aquatic ecosystems, and dose to man resulting of the ingestion of tritiated food were reviewed and discussed. Some data about transfer of organic tritium in terrestrial and aquatic ecosystems reported by literatures were summarized and were supplied with recent data on biological accumulation of organic tritium in the food chain. It was stressed that more research must be done in future because data available were still insufficient. Last, some research programs in progress or planned were stated. (Tsunoda, M.)

  19. Handling of tritium-bearing wastes

    International Nuclear Information System (INIS)

    1981-01-01

    The generation of nuclear power and reprocessing of nuclear fuel results in the production of tritium and the possible need to control the release of tritium-contaminated effluents. In assessing the need for controls, it is necessary to know the production rates of tritium at different nuclear facilities, the technologies available for separating tritium from different gaseous and liquid streams, and the methods that are satisfactory for storage and disposal of tritiated wastes. The intention in applying such control technologies and methods is to avoid undesirable effects on the environment, and to reduce the radiation burden on operational personnel and the general population. This technical report is a result of the IAEA Technical Committee Meeting on Handling of Tritium-bearing Effluents and Wastes, which was held in Vienna, 4 - 8 December 1978. It summarizes the main topics discussed at the meeting and appends the more detailed reports on particular aspects that were prepared for the meeting by individual participants

  20. Recent environmental tritium levels in Japan

    International Nuclear Information System (INIS)

    Iwakura, T.; Inoue, Y.; Tanaka, K.; Kasida, Y.

    1982-01-01

    Data of the tritium surveillance program are summarized for the period of 1967 through 1980. Samples of surface water, tap water, coastal sea water and ground water were collected from environs of commercial nuclear power plants and nuclear facilities, and were analyzed by liquid scintillation counting. Although the results show some differences in tritium concentrations in water samples from various part of the country, there is a general tendency of the concentration in surface waters to decline as a function of time. This implies that environmental waters in Japan generally have not been influenced by the discharged effluents of the facilities or the stations with regard to tritium contamination and that the tritium content of precipitation still plays the dominant role in reflecting annual variation of tritium concentration in surface waters. (J.P.N.)

  1. Tritium decay helium-3 effects in tungsten

    Directory of Open Access Journals (Sweden)

    M. Shimada

    2017-08-01

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

  2. Tritium behavior in an aquatic ecosystem

    International Nuclear Information System (INIS)

    Komatsu, K.

    1982-01-01

    Tritium behavior in aquatic organisms through a model food chain was investigated. In this model food chain, tritium in water reaches bacteria or Japanese killifish via diatoms and brine shrimps. Tritium accumulation in these organisms as organic bound form was expressed as the R value which is defined as the ratio of tritium specific activity in lyophilized organisms (μCi/gH) to that in water (μCi/gH). The maximum R values were 0.5 in diatoms: Chaetoceros gracilis, 0.2 in bacteria: Escherichia coli, 0.5 in brine shrimps: Artemia salina, and 0.32 in Japanese killifish: Oryzias latipes under the growing condition in which tritium accumulation was due to tritium in tritiated water and not tritiated foods. Brine shrimps and Japanese killifish were grown from larve to adult in tritiated sea water and were fed on tritiated foods (model food chain). Their R values were 0.70 and 0.67, respectively. Bacteria, which grew in tritiated water by adding the hydrolysate of tritiated brine shrimps, showed a maximum R value at 0.32. The R values of each organ of Japanese killifish and of DNA and the nucleotides purified from brine shrimps growing in tritiated water with or without tritiated food were measured to estimate the tritium distribution in the body or various molecules of the organisms. These results did not indicate concentration of tritium in specific organs or compounds. The changes of specific activity of tritium in these organisms were measured when they were transferred to non-tritiated water. These retention of tritium was not only different among the tissues but also depended on whether or not the organisms were reared with tritiated foods. (author)

  3. Issues Associated with Tritium Legacy Materials

    International Nuclear Information System (INIS)

    Mills, Michael

    2008-01-01

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

  4. PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Smith, P.; Sheetz, S.

    2013-09-30

    Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case of Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL

  5. Tritium-related materials problems in fusion reactors

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1976-01-01

    Pressing materials problems that must be solved before tritium can be used to produce energy economically in fusion reactors are discussed. The following topics are discussed: (1) breeding tritium, (2) recovering bred tritium, (3) containing tritium, (4) fuel recycling, and (5) laser-fusion fueling

  6. Optimization of tritium management within the ITER project

    International Nuclear Information System (INIS)

    Cortes, P.; Elbez-Uzan, J.; Glugla, M.; Rosanvallon, S.; Ciattaglia, S.; Iseli, M.; Rodriguez-Rodrigo, L.

    2009-01-01

    The authors describe the tritium cycle existing within the ITER project and which has been considered since its beginning. They indicate how confinement systems ensure tritium confinement, how tritium is recovered and processed. They indicate the different tritium management optimization ways which have been identified and integrated into the ITER design

  7. 10 CFR 39.55 - Tritium neutron generator target sources.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Tritium neutron generator target sources. 39.55 Section 39... Equipment § 39.55 Tritium neutron generator target sources. (a) Use of a tritium neutron generator target....77. (b) Use of a tritium neutron generator target source, containing quantities exceeding 1,110 GBg...

  8. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  9. Measurement of dose speed absorbed in depth imparted by sources external secondary patterns of beta radiation. Part 1 Measurement of dose speed absorbed in the surface of soft fabric for isotopes of 90Sr/90Y, 147Pm and 204TI

    International Nuclear Information System (INIS)

    Alvarez R, J.T.

    1993-01-01

    The dose speed was measured absorbed for depth zero, (superficial) in soft equivalent fabric, for the secondary pattern s four sources of beta radiation, (Nr. 86): 90 Sr/ 90 Y, (1850 MBq and 74 MBq respectively); 147 Pm, (518 MBq) and 204 TI, (18.5 MBq). The measurement is carried out to different distances of source-detecting separation, (11.0, 30.0 and 50.0 cm for the source of 1850 MBq, 30.0 cm for that of 74 MBq; 11.00 cm for the source of 147 Pmand to contact for all the sources); maintaining the radiation sheaf aligned the one axis of symmetry of the detector, (α 0 degrees). The detector employed was a extrapolation chambers of variable electrodes and electrode fixed collector, (30 mm of diameter). In accordance with the principle of Bragg-Gray the volume of the chambers is varied and they register the variations of the current of collected ionization, correcting until for a maximum of thirteen correction factors that take into account the deviation to the suppositions that it establishes this principle. The certain values of the speed of superficial absorbed dose are in the following intervals: 90 Sr/ 90 Y, (1850 MBq, 0.0, 11.0, 30.0 and 50.0 cm): 43.164 mGy S-t, 0.544 mGy s-1 ,0.075 mGy s -1 and 0.027 mGy s -1 , respectively, with a Global Analysis of the order of 1.17%, 1.17%, 1.14% and 1.66%, K J; 90 Sr / 90 Y, (74 MBq, 0.0 and 30 cm): 1.536 mGy s -1 and 0.002 mGy s -1 , with Global Analysis of 1.19.0% and 5.22%, (K = 1) respectively, for the 147 Pm, (0.0 and 11.0 in the interval of: 0.36 μGy s -1 and 0.43 μGy s -1 , with one Global Analysis of 1 .42% and 4.28%, (K = 1), respectively; and finally for the 204 TI, (0.0 cm) in the interval of 0.10 μGy s -1 with a Global Analysis of 1.27%. He calculates of the Global Analysis one carries out of agreement with those recommendations of the BIPM. In all the cases of source-detecting arrangement with separations different from zero, models of simple lineal regression were used. However for the case of the

  10. Tritium Decay Helium-3 Effects in Tungsten

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. Transfer of fallout tritium from environment to human body

    International Nuclear Information System (INIS)

    Hisamatsu, Shun-ichi; Takizawa, Yukio

    1989-01-01

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

  13. Tritium distributing in stainless steel determined by chemical etchin

    International Nuclear Information System (INIS)

    Xiong Yifu; Luo Deli; Chen Changan; Chen Shicun; Jing Wenyong

    2009-01-01

    The depth distribution of tritium in stainless steel was measured by chemical etching. The results show that the method can more quantitatively evaluate the tritium distributing in stainless steel. The maximum amount of tritium which distributed in crystal lattice of stainless steel is limitted by its solubility at room temperature. The other form of tritium in stainless steel is gaseous tritium that are trapped by defects, impurities, fractures, etc. within it. The gaseous tritium is several times more than the solid-dissolved tritium. (authors)

  14. Tritium production and processing in a Tokamak reactor

    International Nuclear Information System (INIS)

    Leger, D.

    1986-09-01

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

  15. Tritium monitor with improved gamma-ray discrimination

    Science.gov (United States)

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

    1985-01-01

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

  16. Introduction to Wolsong Tritium Removal Facility (WTRF)

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  17. Environmental monitoring for tritium separation facility

    International Nuclear Information System (INIS)

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

    2001-01-01

    The Cryogenic Pilot is an experimental project within 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 almost all the neighbors of the Experimental Cryogenic Pilot are chemical plants. It is necessary to emphasize this aspect because the sewage system is connected with the other tree chemical plants from the 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 and waste water. The tritium level was between 10 TU and 27 TU what indicates that there is no sources of tritium contamination in the neighborhood of Cryogenic Pilot. In order to determine baseline levels we decided to monitor monthly each location. In this paper it is presented the standard method used for tritium determination in water samples, the precautions needed in order to achieve reliable results, and the evolution of tritium level in different location near the Tritium and Deuterium Cryogenic Separation Experimental Pilot. (authors)

  18. Analysis of the organically bound tritium

    International Nuclear Information System (INIS)

    Baglan, N.; Alanic, G.

    2011-01-01

    In environmental samples, tritium is very often combined with the fraction of bulk water accumulated in the sample but also in the form of organically bound tritium. When the tritium is organically bound, 2 forms can coexist: the exchangeable fraction and the non-exchangeable fraction. The analysis of the different forms of tritium present in the sample is necessary to assess the sanitary hazards due to tritium. The total tritium is obtained from the analysis of the water released when the fresh sample is burnt while the organically bound tritium is obtained from the analysis of the water released when the dry extract of the sample is burnt. The measurement of the exchangeable fraction and the non-exchangeable fraction requires an additional stage of labile exchange. The exchangeable fraction is determined from the analysis of the water released during the labile exchange and the non-exchangeable fraction is determined from the water released during the combustion of the dry extract of the labile exchange

  19. Little tritium goes a long way

    International Nuclear Information System (INIS)

    Albright, D.; Taylor, T.B.

    1988-01-01

    Faced with mounting safety problems in its military production reactors, the Energy Department will soon ask Congress to fund the construction of at least one new multibillion dollar tritium production reactor. Energy estimates that building such a reactor could take ten years, and it says that in the interim it needs to continue producing tritium at the Savannah River reactors. In fact, it plans to resume operating its Savannah River reactors at full power as soon as possible. The United States must keep producing tritium if the US-Soviet nuclear arms race continues its present course. If the arms race continues, the Energy Department has two basic options: it could run the Savannah River reactors for several more decades or it could use these reactors until it has built a new one. Operating the Savannah River reactors at full or low power may be risky, even if they undergo extensive safety modifications, since no one knows at what power these reactors can be operated safely. Despite these pressing issues, most of the substantive debate about the role of tritium in nuclear weapons and the requirement for more tritium production is taking place in secret. The public debate largely ignores the broader questions of whether the United States needs to produce tritium and what impact possible agreements reducing nuclear arsenals might have on US tritium requirements

  20. Tritium Room Air Monitor Operating Experience Review

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader; B. J. Denny

    2008-09-01

    Monitoring the breathing air in tritium facility rooms for airborne tritium is a radiological safety requirement and a best practice for personnel safety. Besides audible alarms for room evacuation, these monitors often send signals for process shutdown, ventilation isolation, and cleanup system actuation to mitigate releases and prevent tritium spread to the environment. Therefore, these monitors are important not only to personnel safety but also to public safety and environmental protection. This paper presents an operating experience review of tritium monitor performance on demand during small (1 mCi to 1 Ci) operational releases, and intentional airborne inroom tritium release tests. The tritium tests provide monitor operation data to allow calculation of a statistical estimate for the reliability of monitors annunciating in actual tritium gas airborne release situations. The data show a failure to operate rate of 3.5E-06/monitor-hr with an upper bound of 4.7E-06, a failure to alarm on demand rate of 1.4E-02/demand with an upper bound of 4.4E-02, and a spurious alarm rate of 0.1 to 0.2/monitor-yr.

  1. Quick management of accidental tritium exposure cases

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  2. Quick management of accidental tritium exposure cases.

    Science.gov (United States)

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

    2012-07-01

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

  3. Isotopic fractionation of tritium in biological systems.

    Science.gov (United States)

    Le Goff, Pierre; Fromm, Michel; Vichot, Laurent; Badot, Pierre-Marie; Guétat, Philippe

    2014-04-01

    Isotopic fractionation of tritium is a highly relevant issue in radiation protection and requires certain radioecological considerations. Sound evaluation of this factor is indeed necessary to determine whether environmental compartments are enriched/depleted in tritium or if tritium is, on the contrary, isotopically well-distributed in a given system. The ubiquity of tritium and the standard analytical methods used to assay it may induce biases in both the measurement and the signification that is accorded to the so-called fractionation: based on an exhaustive review of the literature, we show how, sometimes large deviations may appear. It is shown that when comparing the non-exchangeable fraction of organically bound tritium (neOBT) to another fraction of tritium (e.g. tritiated water) the preparation of samples and the measurement of neOBT reported frequently led to underestimation of the ratio of tritium to hydrogen (T/H) in the non-exchangeable compartment by a factor of 5% to 50%. In the present study, corrections are proposed for most of the biological matrices studied so far. Nevertheless, the values of isotopic fractionation reported in the literature remain difficult to compare with each other, especially since the physical quantities and units often vary between authors. Some improvements are proposed to better define what should encompass the concepts of exchangeable and non-exchangeable fractions. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. Method of extracting tritium from heavy water

    International Nuclear Information System (INIS)

    Tsuchiya, Hiroyuki; Kikuchi, Makoto; Asakura, Yamato; Yusa, Hideo.

    1979-01-01

    Purpose: To extract tritium in heavy water by combining isotope exchange reaction with liquefaction distillation to increase the concentration of recovered tritium, thereby reducing the quantity of radioactive wastes recovered. Constitution: Heavy water containing tritium from a reactor is introduced into a tritium separator through a conduit pipe. On the other hand, a D 2 gas is introduced through the conduit pipe in the lower part of a tritium separator to transfer tritium into D 2 gas by isotope exchange. The D 2 gas containing DT is introduced into a liquefaction distillation tower together with an outlet gas of a converter supplied through a pipeline. The converter is filled with net-like metals of platinum group such as Pt, Ni, Pd and the like, and the D 2 gas affluent in DT, extracted from the distillation tower is converted into D 2 and T 2 . The gas which has been introduced into the liquefaction distillation tower is liquefied. The D 2 gas of low boiling point components reaches the tower top, and the T 2 gas of high boiling point components is concentrated at the tower bottom, and is rendered into tritium water in a recoupler and stored in a water storage apparatus. (Yoshino, Y.)

  5. Risks of tritium and their mitigation

    International Nuclear Information System (INIS)

    Ichimasa, Y.; Shiba, H.; Ichimasa, M.; Chikuuti, M.; Akita, Y.

    1992-01-01

    In this study, the effects of an antibacterial drug, norfloxacin, and an antibiotic, clindamycin, on in vivo oxidation of tritium gas in rats were investigated. Wistar strain male rats were used. They were provided with a standard diet, water ad libitum, and maintained in glass metabolic cages of approximately 20 liters capacity. The air flow and temperature were controlled. To investigate the availability of norfloxacin and clindamycin on the inhibition effects of the oxidation of tritium gas, two types of the experiments were conducted one was that, before the exposure to tritium gas for 2 hours, norfloxacin or clindamycin was administrated to rats three times a day for 4 days, and the other was administration of a drug after tritium gas exposure. After the exposure to tritium gas, blood, the liver, urine and feces samples were collected from rats and the radioactivity of them was determined after combustion using a sample oxidizer. In the case of norfloxacin, tritium concentration in rat body decreased one fifth of that in non-treated rats. On the other hand, administration of clindamycin shortened the biological half-life of tritium in urine to three fifth of that of non-treated rats. (author)

  6. ITER Dynamic Tritium Inventory Modeling Code

    International Nuclear Information System (INIS)

    Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.

    2005-01-01

    A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER

  7. Measurement of tritium in environment, (2)

    International Nuclear Information System (INIS)

    Chaya, Ikuo; Kagami, Tadaaki; Hamamura, Norikatsu

    1975-01-01

    In order to know the amount of natural tritium in environmental water and also to know the tendency of tritium concentration in surface water which is necessary for the measurement of ground water age, the tritium concentration in rain, river, and sea water in Aichi Prefecture were measured. In order to make the appropriate utilization of ground water such as city water and hot springs and to elucidate the effect of ground water utilization on ground subsidence, it is desirable to clarify the state of underground water-bearing strata, the flow direction and flow speed of ground water, and the change of ground water quality owing to the flow. As the means of knowing the flow speed of ground water, the age determination with tritium was carried out. The amount of tritium was determined by measuring the concentrated samples with a liquid scintillation counter. The tritium concentration in river was 1.7 times as much as that in rain water, and it is attributed to the time difference from raining to flowing in rivers. The tritium concentration in sea water was high at the estuary of Kiso River, and about a half of it in the other places. The water of the hot spring source in Nobi Plain is the old ground water soaked more than 20 years ago. The city water sources utilizing ground water shallower than 300 m use both new and old ground water. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  9. Conceptual design of tritium accountancy system for LLCB TBM

    International Nuclear Information System (INIS)

    Patel, Rudreksh; Sircar, Amit

    2017-01-01

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

  10. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua

    2003-01-01

    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)

  11. Tritium

    Science.gov (United States)

    2011-11-01

    Fraction The probability that a triton injected into a reactor is burned in the reaction (1) before it escapes the confinement region in the case of MFE , or...vσ〉 = 3.68× 10−18 T 2/3 i exp ( − 19.94 T 1/3 i ) . (12) For a representative MFE ion temperature Ti = 20 keV, (13) we find from (12) that 〈vσ〉 = 3.22

  12. Tritium extraction technologies and DEMO requirements

    Energy Technology Data Exchange (ETDEWEB)

    Demange, D., E-mail: david.demange@kit.edu [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Antunes, R.; Borisevich, O.; Frances, L. [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rapisarda, D. [Laboratorio Nacional de Fusión, EURATOM-CIEMAT, 28040 Madrid (Spain); Santucci, A. [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy)

    2016-11-01

    Highlights: • We detail the R&D plan for tritium technology of the European DEMO breeding blanket. • We study advanced and efficient extraction techniques to improve tritium management. • We consider inorganic membranes and catalytic membrane reactor for solid blankets. • We consider permeator against vacuum and vacuum sieve tray for liquid blankets. - Abstract: The conceptual design of the tritium extraction system (TES) for the European DEMO reactor is worked out in parallel for four different breeding blankets (BB) retained by EUROfusion. The TES design has to be tackled in an integrated manner optimizing the synergy with the directly interfacing inner fuel cycle, while minimizing the tritium permeation into the coolant. Considering DEMO requirements, it is most likely that only advanced technologies will be suitable for the tritium extraction systems of the BB. This paper overviews the European work programme for R&D on tritium technology for the DEMO BB, summaries the general first outcomes, and details the specific and comprehensive R&D program to study experimentally immature but promising technologies such as vacuum sieve tray or permeator against vacuum for tritium extraction from PbLi, and advanced inorganic membranes and catalytic membrane reactor for tritium extraction from He. These techniques are simple, fully continuous, likely compact with contained energy consumption. Several European Laboratories are joining their efforts to deploy several new experimental setups to accommodate the tests campaigns that will cover small scale experiments with tritium and inactive medium scale tests so as to improve the technology readiness level of these advanced processes.

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  15. Tritium kinetics in a freshwater marsh ecosystem

    International Nuclear Information System (INIS)

    Adams, L.W.

    1976-01-01

    Ten curies of tritium (as tritiated water, HTO) were applied to a 2-ha enclosed Lake Erie marsh in northwestern Ohio on 29 October 1973. Tritium kinetics in the marsh water, bottom sediment, and selected aquatic plants and animals were determined. Following HTO application, peak tritium levels in the sediment were observed on day 13 in the top 1-cm layer, on day 27 at the 5-cm depth, and on day 64 at the 10-cm depth. Peak levels at 15 and 20 cm were not discernible, although there was some movement of HTO to the 20-cm depth. A model based on diffusion theory described tritium movement through the sediment. Unbound and bound tritium levels in curly-leaf pondweed (Potamogeton crispus), pickerelweed (Pontederia cordata), and smartweed (Polygonum lapathifolium) generally tended to follow tritium levels in marsh water. The unbound tritium:marsh water tritium ratio was significantly larger (P < 0.001) in curly-leaf pondweed than in either of the two emergents. Tritium uptake into the unbound compartments of crayfish (Procambarus blandingi), carp (Cyprinus carpio), and bluegills (Lepomis macrochirus) was rapid. For crayfish, maximum HTO levels were observed on days 3 and 2 for viscera and muscle, respectively. Unbound HTO in carp viscera peaked on day 2, and levels in carp muscle reached a maximum in 4 hours. Maximum levels of unbound HTO in bluegill viscera and muscle were observed on day 1. After peak levels were obtained, unbound HTO paralleled marsh water HTO activity in all species. Tritium uptake into the bound compartments was not as rapid nor were the levels as high as for unbound HTO in any of the species. Peak bound levels in crayfish viscera were observed on day 20 and maximum levels in muscle were noted on day 10. Bound tritium in carp viscera and muscle reached maximum levels on day 20. In bluegills, peaks were reached on days 7 and 5 for viscera and muscle, respectively. Bound tritium in all species decreased following maximum levels

  16. Review of general tritium accountancy techniques

    International Nuclear Information System (INIS)

    Vassallo, G.; Engelmann, U.

    1995-01-01

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

  17. Preparation of honey sample for tritium monitoring

    International Nuclear Information System (INIS)

    Chen Bingru; Wang Chenlian; Wang Weihua

    1989-01-01

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

  18. Conversion of tritium gas to tritiated water

    International Nuclear Information System (INIS)

    Papagiannakopoulos, P.J.; Easterly, C.E.

    1979-05-01

    The mechanisms of conversion of tritium gas to tritiated water (HTO) have been examined for several tritium gaseous mixtures. The physical and chemical processes involved in the self-radiolysis of such mixtures have been analyzed and the kinetics involved in the formation of HTO has been presented. It has been determined that the formation of the H and/or OH free radicals, as intermediate species, are of significance in the formation of HTO. Therefore, the problem of reducing the rate of formation of tritiated water in a mixture of gaseous tritium with atmospheric components is one of finding an effective scavenger for the H and/or OH free radicals

  19. Electrolytic gettering of tritium from air

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  20. The tritium and the controlled fusion reactors

    International Nuclear Information System (INIS)

    Leger, D.; Rouyer, J.L.

    1986-04-01

    It is shown how tritium is used how it is circulating in a fusion reactor. The great functions of tritium circuits are detailed: reprocessing of burnt gases, reprocessing of gases coming from neutral injectors, reprocessing from gaseous wastes, detritiation of cooling fluids. Current technologic developments are quoted. Then tritium confinement and containment, in normal or accidental situations, are displayed. Limitation devices of effluents and release for normal operating (noticeably the reprocessing systems of atmosphere) and safety and protection systems in case of accident are described [fr

  1. Management of Tritium in European Spallation Source

    DEFF Research Database (Denmark)

    Ene, Daniela; Andersson, Kasper Grann; Jensen, Mikael

    2015-01-01

    with the country regulation criteria. The aim of this paper is to give an overview of the different aspects of the tritium management in ESS facility. Besides the design parameter study of the helium coolant purification system of the target the consequences of the tritium releasing into the environment were also...... of the results on soil examinations. With the assumption of 100% release of tritium to the atmosphere during the occurring of the extreme accidents, it was found as well that the total dose complies with the constraint....

  2. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

  4. Tritium containing polymers having a polymer backbone substantially void of tritium

    Science.gov (United States)

    Jensen, G.A.; Nelson, D.A.; Molton, P.M.

    1992-03-31

    A radioluminescent light source comprises a solid mixture of a phosphorescent substance and a tritiated polymer. The solid mixture forms a solid mass having length, width, and thickness dimensions, and is capable of self-support. In one aspect of the invention, the phosphorescent substance comprises solid phosphor particles supported or surrounded within a solid matrix by a tritium containing polymer. The tritium containing polymer comprises a polymer backbone which is essentially void of tritium. 2 figs.

  5. Tritium inventory measurements using calorimetry

    International Nuclear Information System (INIS)

    Kapulla, H.; Kraemer, R.; Heine, R.

    1992-01-01

    In the past calorimetry has been developed as a powerful tool in radiometrology. Calorimetric methods have been applied for the determination of activities, half lives and mean energies released during the disintegration of radioactive isotopes. The fundamental factors and relations which determine the power output of radioactive samples are presented and some basic calorimeter principles are discussed in this paper. At the Kernforschungszentrum Karlsruhe (KfK) a family of 3 calorimeters has been developed to measure the energy release from radiative waste products arising from reprocessing operations. With these calorimeters, radiative samples with sizes from a few cm 3 to 2 ·10 5 cm 3 and heat ratings ranging from a few nW to kW can be measured. After modifications of tits inner part the most sensitive calorimeter among the three calorimeters mentioned above would be best suited for measuring the tritium inventory in T-getters of the Amersham-type

  6. Environmental monitoring of molecular tritium

    Energy Technology Data Exchange (ETDEWEB)

    Ichimasa, M.; Ichimasa, Y.; Akita, Y. (Ibaraki Univ., Mito (Japan). Faculty of Science); Suzuki, M.; Obayashi, H.; Sakuma, Y.

    1992-01-01

    The oxidation of atmospheric molecular tritium (HT) in vegetation was determined by in vitro experiments for various kinds of woody and herbaceous plant leaves, mosses and lichens taken from a forest and a garden in Ibaraki prefecture and a forest in Gifu prefecture, and comparison of the HT oxidation activity in vegetation was made with those in its neighboring surface soil (0-5cm in depth). The oxidation of HT in woody plant leaves was extremely low, only about 1/10000-1/1000 that in the surface soil as well as herbaceous plant leaves with some exception, whereas HT oxidation in mosses and lichens was 50-500 times that in pine needles. These results suggest the usefulness of mosses and lichens as monitor vegetation for accidental release of HT into the environment. (author).

  7. Tritium metabolism in cow's milk after administration of tritiated water and of organically bound tritium

    International Nuclear Information System (INIS)

    Hoek, J. van den

    1982-01-01

    Tritium was administered as THO and as organically bound tritium (OBT) to lactating cows. Urine and milk samples were collected and analyzed for tritium content. Plateau levels in milk water and in milk fat, lactose and casein were reached in about 20 days after feeding either THO or OBT. Comparison of the specific activity (pCi 3 H/g H) of the various milk constituents with the specific activity of the body water showed that, after administration of THO, the highest tritium incorporation occurred in lactose (0.58), followed by milk fat (0.36) and casein (0.27). Tritium incorporation in milk dry matter (0.45) is considerably higher than in most tissue components of several mammalian species after continuous ingestion of THO as reported in the literature. After feeding OBT, the highest tritium incorporation occurred in milk fat and to a lesser extent in casein. Tritium levels in lactose were surprisingly low and the reason for this is not clear. They were similar to those in milk water. Tritium levels in milk and urine water showed systematic differences during administration of OBT and after this was stopped. There was more tritium in milk water until the last day of OBT feeding and this situation was reversed after this. (author)

  8. Overview of tritium processing development at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  10. Tritium metabolism in cow's milk after administration of tritiated water and of organically bound tritium

    Energy Technology Data Exchange (ETDEWEB)

    van den Hoek, J [Landbouwhogeschool Wageningen (Netherlands). Lab. voor Fysiologie der Dieren; Gerber, G; Kirchmann, R [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

    1982-01-01

    Tritium was administered as THO and as organically bound tritium (OBT) to lactating cows. Urine and milk samples were collected and analyzed for tritium content. Plateau levels in milk water and in milk fat, lactose and casein were reached in about 20 days after feeding either THO or OBT. Comparison of the specific activity (pCi/sup 3/H/g H) of the various milk constituents with the specific activity of the body water showed that, after administration of THO, the highest tritium incorporation occurred in lactose (0.58), followed by milk fat (0.36) and casein (0.27). Tritium incorporation in milk dry matter (0.45) is considerably higher than in most tissue components of several mammalian species after continuous ingestion of THO as reported in the literature. After feeding OBT, the highest tritium incorporation occurred in milk fat and to a lesser extent in casein. Tritium levels in lactose were surprisingly low and the reason for this is not clear. They were similar to those in milk water. Tritium levels in milk and urine water showed systematic differences during administration of OBT and after this was stopped. There was more tritium in milk water until the last day of OBT feeding and this situation was reversed after this.

  11. Assessment of Sr-90 in water samples: precision and accuracy

    Energy Technology Data Exchange (ETDEWEB)

    Nisti, Marcelo B.; Saueia, Cátia H.R.; Castilho, Bruna; Mazzilli, Barbara P., E-mail: mbnisti@ipen.br, E-mail: chsaueia@ipen.br, E-mail: bcastilho@ipen.br, E-mail: mazzilli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The study of artificial radionuclides dispersion into the environment is very important to control the nuclear waste discharges, nuclear accidents and nuclear weapons testing. The accidents in Fukushima Daiichi Nuclear Power Plant and Chernobyl Nuclear Power Plant, released several radionuclides in the environment by aerial deposition and liquid discharge, with various level of radioactivity. The {sup 90}Sr was one of the elements released into the environment. The {sup 90}Sr is produced by nuclear fission with a physical half-life of 28.79 years with decay energy of 0.546 MeV. The aims of this study are to evaluate the precision and accuracy of three methodologies for the determination of {sup 90}Sr in water samples: Cerenkov, LSC direct method and with radiochemical separation. The performance of the methodologies was evaluated by using two scintillation counters (Quantulus and Hidex). The parameters Minimum Detectable Activity (MDA) and Figure Of Merit (FOM) were determined for each method, the precision and accuracy were checked using {sup 90}Sr standard solutions. (author)

  12. Assessment of Sr-90 in water samples: precision and accuracy

    International Nuclear Information System (INIS)

    Nisti, Marcelo B.; Saueia, Cátia H.R.; Castilho, Bruna; Mazzilli, Barbara P.

    2017-01-01

    The study of artificial radionuclides dispersion into the environment is very important to control the nuclear waste discharges, nuclear accidents and nuclear weapons testing. The accidents in Fukushima Daiichi Nuclear Power Plant and Chernobyl Nuclear Power Plant, released several radionuclides in the environment by aerial deposition and liquid discharge, with various level of radioactivity. The 90 Sr was one of the elements released into the environment. The 90 Sr is produced by nuclear fission with a physical half-life of 28.79 years with decay energy of 0.546 MeV. The aims of this study are to evaluate the precision and accuracy of three methodologies for the determination of 90 Sr in water samples: Cerenkov, LSC direct method and with radiochemical separation. The performance of the methodologies was evaluated by using two scintillation counters (Quantulus and Hidex). The parameters Minimum Detectable Activity (MDA) and Figure Of Merit (FOM) were determined for each method, the precision and accuracy were checked using 90 Sr standard solutions. (author)

  13. Beta dosimetry in teeth from SR-90 exposed subjects

    International Nuclear Information System (INIS)

    Fattibene, P.; De Coste, V.; Onori, S.; Veronese, I.; Giussani, A.; Cantone, M.C.; Shishkina, E.

    2006-01-01

    Tooth enamel is a well recognized dosimeter for retrospective dose reconstruction of individuals accidentally exposed to ionizing radiation. The measurements of the absorbed dose in tooth enamel is conventionally carried out with the Electron Paramagnetic Resonance technique. Tooth enamel is sensitive to all kind of ionizing radiation. Its response to photons has been widely investigated. For application to contaminated teeth with 90 Sr, one of the most common osteo tropic radionuclides, the effectiveness of tooth enamel response to the β spectrum needs be evaluated. The response function to 90 Sr of the EPR/tooth enamel systems, its linearity and reproducibility have been investigated under a controlled geometry, and the results will be presented and compared to those obtained with photons. When the subject has been exposed to both external and internal radiation, a combined EPR/T.L. method can be used to distinguish the internal from the external contribution to the cumulative dose in tooth (Gosku et al., 2002;=Veronese et al. 2004, Shishkina et al. 2005). The T.L. measurement, performed putting thin ± Al 2 O 3 :C dosimeters at contact with the tooth surfaces, enables to estimate the beta dose rate due to the radionuclides present in tooth. The combination of this information with that coming from EPR allows, under specific assumptions, to evaluate separately the internal and external contribution to the tooth dose. In a previous work (Veronese et al., 2004) the dose in enamel measured by EPR in a tooth contaminated with 90 Sr of a Techa River resident was compared to the dose rate measured by TLDs. The test has been extended to a larger number of 90 Sr contaminated teeth. EPR measurements have been also performed in other portions of the teeth, i.e. tooth dentin and root. The correlation between the results, obtained from EPR and TLD measurements, and the evaluation of the relative proportion of internal and external dose are presented and discussed. (authors)

  14. Review of tritium behavior in HTGR systems

    International Nuclear Information System (INIS)

    Gainey, B.W.

    1976-01-01

    The available experimental evidence from laboratory and reactor studies pertaining to tritium production, capture, release, and transport within an HTGR leading to release to the environment is reviewed. Possible mechanisms for release, capture, and transport are considered and a simple model was used to calculate the expected tritium release from HTGRs. Comparison with Federal regulations governing tritium release confirm that expected HTGR releases will be well within the allowable release limits. Releases from HTGRs are expected to be somewhat less than from LWRs based on the published LWR operating data. Areas of research deserving further study are defined but it is concluded that a tritium surveillance at Fort St. Vrain is the most immediate need

  15. Development of tritium-handling technique

    International Nuclear Information System (INIS)

    Ohmura, Hiroshi; Hosaka, Akio; Okamoto, Takahumi

    1988-01-01

    The overview of developing activities for tritium-handling techniques in IHI are presented. To establish a fusion power plant, tritium handling is one of the key technologies. Recently in JAERI, conceptual design of FER (Fusion Experimental Reactor) has been carried out, and the FER system requires a processing system for a large amount of tritium. IHI concentrate on investigation of fuel gas purification, isotope separation and storage systems under contract with Toshiba Corporation. Design results of the systems and each components are reviewed. IHI has been developing fundamental handling techniques which are the ZrNi bed for hydrogen isotope storage and isotope separation by laser. The ZrNi bed with a tritium storage capacity of 1000 Ci has been constructed and recovery capability of the hydrogen isotope until 10 -4 Torr {0.013 Pa} was confirmed. In laser isotope separation, the optimum laser wave length has been determined. (author)

  16. Tritium-management survey of Wolsong 1

    International Nuclear Information System (INIS)

    Allsop, P.J.; Boss, C.R.; Song, M-J.; Son, S-H.; Choi, J-K.

    1996-10-01

    Commissioned in 1983, Wolsong 1 has had one of the best lifetime capacity factors in the world. It has also maintained tritium emissions and heavy-water losses at or below those of similar CANDU 6 reactors. To further ensure that emissions remain as low as reasonably achievable (ALARA), Wolsong 1, AECL and the KEPC0 (Korean Electric Power Company)Research Center collaborated on a survey of tritium management at Wolsong 1 during the spring of 1995. This survey identified similarities and differences between Wolsong 1 and the Canadian CANDU 6 stations. It also corroborated several of Wolsong 1's plans to further refine and upgrade tritium management. This report summarizes those aspects of the Wolsong 1 tritium survey. (author)

  17. Tritium proof-of-principle injector experiment

    International Nuclear Information System (INIS)

    Fisher, P.W.; Milora, S.L.; Combs, S.K.; Carlson, R.V.; Coffin, D.O.

    1988-01-01

    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 fision 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 Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). 18 refs., 3 figs

  18. Tritium Measurements in Slovenia - Chronology Till 2004

    International Nuclear Information System (INIS)

    Logar, Jasmina Kozar; Vaupotic, Janja; Kobal, Ivan

    2005-01-01

    Almost all the analyses of tritium in Slovenia have been performed by the tritium laboratory at the Jozef Stefan Institute. Nearly 90 % of its measurements have been covered by two national programs, both approved by the Slovenian Nuclear Safety Administration: the radioactive monitoring program in the environs of Krsko Nuclear Power Plant (KNPP) and the program of global radioactive contamination monitoring in the environment. These programs include samples of groundwaters, surface waters, precipitation and drinking waters, as well as liquid and gaseous effluents from KNPP. Tritium was determined in some research projects and in hydrological studies of thermal waters, groundwater and coalmine waters. Tritium in the Karst region was mapped as well as the springs of entire territory of Slovenia. Around 5500 samples have been analyzed up to 2004

  19. Breeding blanket development. Tritium release from breeder

    International Nuclear Information System (INIS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Nagao, Yoshiharu

    2006-01-01

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

  20. Technologies for immobilization and disposal of tritium

    International Nuclear Information System (INIS)

    Coppari, N.R.

    1996-01-01

    This study was done within a program one of whose objectives was to know the state of the technology development for tritium separation in the moderator circuit at HWR and to define the possible technologies to be applied to the Argentine nuclear power plants. Within this framework the strategies adopted by each country and the available technologies for a safe disposal of tritium, not only in its gaseous state tritium but also as tritiated water were analyzed. It is considered that if the selected separation method is such that the tritium is in its gaseous state, the hydride formation for long periods of immobilization should be studied. whereas if it were triated water immobilization should be studied to choose the technology between cementation and drying agents, in both cases the final disposal site will have to be selected. (author). 8 refs

  1. Survey of pumps for tritium gas

    International Nuclear Information System (INIS)

    Dowell, T.M.

    1983-05-01

    This report considers many different types of pumps for their possible use in pumping tritium gas in the low, intermediate and high vacuum ranges. No one type of pump is suitable for use over the wide range of pumping pressure required in a typical pumping system. The favoured components for such a system are: bellows pump (low vacuum); orbiting scroll pump (intermediate vacuum); magnetically suspended turbomolecular pump (high vacuum); cryopump (high vacuum). Other pumps which should be considered for possible future development are: mound modified vane pump; SRTI wobble pump; roots pump with canned motor. It is proposed that a study be made of a future tritium pumping system in a Canadian tritium facility, e.g. a tritium laboratory

  2. Calibrations of a tritium extraction facility

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. Tritium detection in installations and in environment

    International Nuclear Information System (INIS)

    Calando, J.P.

    1986-04-01

    The different tritium detection devices in the atmosphere are reviewed: ionization chamber, proportional counters, Peltier effect, bubble-through device. Characteristics of those allowing to quantify the HTO form (more ''radiotoxic'' than T 2 form) are emphasized [fr

  4. Tritium Inventory in ARIES-AT

    International Nuclear Information System (INIS)

    Longhurst, Glen R.

    2001-01-01

    This report documents an investigation into the tritium inventory expected in the ARIES-AT fusion reactor. ARIES-AT features silicon carbide fibers in a silicon carbide matrix as its primary construction. It uses the same fusion power core as the previous ARIES-RS. Based on experimental results of several researchers, consideration was given to swelling, sputtering, film coatings, erosion, and implantation. Estimates were made of tritium inventory using the TMAP4 code. About 700 g of tritium may be expected in the machine, two thirds of which would reside in the first wall. Under assumed accident conditions that involve first wall temperatures up to 1000 C, evolution of retained tritium may be expected to vary from 0.8 to nearly 40 percent depending on the temperature of the first wall

  5. Methods of tritium recovery from molten lithium

    International Nuclear Information System (INIS)

    Farookhi, R.; Rogers, J.E.

    1968-01-01

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

  6. Investigation of tritium in the aquatic environment

    International Nuclear Information System (INIS)

    Cohen, L.K.

    1977-01-01

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

  7. FDMH - The tritium model in RODOS

    International Nuclear Information System (INIS)

    Galeriu, D.; Mateescu, G.; Melintescu, A.; Turcanu, C.; Raskob, W.

    2000-01-01

    Under the auspices of its RTD (Research and Technological Development) Framework Programmes, the European Commission has supported the development of the RODOS (Real-time On-line DecisiOn Support) system for off-site emergency management. The project started in 1989 focusing on PWR/LWR type accidents and using experience from the Chernobyl accident. In 1996 it was realised that tritium should be included in the list of radionuclides, as large tritium sources exist in Europe and to allow a potential expansion of the RODOS system for application on future fusion reactor accidents. The National Institute for Physics and Nuclear Engineering (IFIN-HH) in Romania - in close co-operation with the Research Centre Karlsruhe (FZK) - was charged to develop the tritium module, based on previous experience in environmental tritium modelling and the operation of CANDU reactor-based NPP in Romania (with potential tritium accidents). Tritium, being an isotope of hydrogen, is incorporated immediately in the life cycle and its transport into the biosphere differs considerably from other radionuclides treated by the RODOS system. Concentrations in the individual compartments may change very rapidly (hours) under varying environmental conditions and conversion to organic forms by biochemical and metabolic processes takes place in plants and animals. Consequently, the tritium code in RODOS was developed as a separate module and harmonisation in data sets and interfaces with other food chain modules integrated in RODOS was ensured. Presently, the tritium module - FDMH- is integrated and documented in the RODOS system, delivering time dependent tritium concentration (as tritiated water or organically bound tritium) in plant and animal products, inhalation dose and ingestion dose for various groups of population, after an accident emitting tritiated water and for up to 2520 locations around the source. FDMH incorporates many improved techniques in radiological assessment and makes

  8. Leaching of tritium from a cement composite

    International Nuclear Information System (INIS)

    Matsuzuru, Hideo; Ito, Akihiko

    1978-10-01

    Leaching of tritium from cement composites into an aqueous phase has been studied to evaluate the safety of incorporation of the tritiated liquid waste into cement. Leaching tests were performed by the method recommended by the International Atomic Energy Agency. The Leaching fraction was measured as functions of waste-cement ratio (Wa/C), temperature of leachant and curing time. The tritium leachability of cement in the long term test follows the order: alumina cement portland cement slag cement. The fraction of tritium leached increases with increasing Wa/C and temperature and decreasing curing period. A deionized water as a leachant gives a slightly higher leachability than synthetic sea water. The amount leached of tritium from a 200 l drum size specimen was estimated on the basis of the above results. (author)

  9. Tritium target manufacturing for use in accelerators

    Science.gov (United States)

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

    2001-07-01

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

  10. Tritium means of detection and of protection

    International Nuclear Information System (INIS)

    Sutra-Fourcade, Y.

    1967-01-01

    The report is an attempt to correlate present data concerning tritium, especially from the health physics points of view. The various detection and measurement methods are reviewed in turn: measurement of tritium in the atmosphere, in liquids and on surfaces. The operation of various types of apparatus is analyzed and the sensitivity limits deduced from laboratory tests are given. Otter sections are devoted to the means of protection which can be used against inhalation of tritium (ventilation, protective clothing) and to calculations of the changes in atmospheric pollution in a given place and of the time spent in a contaminated zone. The last part deals with the decontamination of equipment contaminated with tritium. (author) [fr

  11. Tritium handling facility at KMS Fusion Inc

    International Nuclear Information System (INIS)

    Bowman, C.C.; Vis, V.A.

    1990-01-01

    The tritium facility at KMS Fusion, Inc. supports the inertial confinement fusion research program. The main function of the facility is to fill glass and polymer Microshell (TM) capsules (small fuel containers) to a maximum pressure of 100 atm with tritium (T 2 ) or deuterium--tritium (DT). The recent upgrade of the facility allows us to fill Microshell capsules to a maximum pressure of 200 atm. A second fill port allows us to run long term fills of Macroshell (TM) capsules (large fuel containers) concurrently. The principle processes of the system are: (1) storage of the tritium as a uranium hydride; (2) pressure intensification using cryogenics; and (3) filling of the shells by permeation at elevated temperatures. The design of the facility was centered around a NRC license limit of 6000 Ci

  12. Tritium Issues in Next Step Devices

    International Nuclear Information System (INIS)

    C.H. Skinner; G. Federici

    2001-01-01

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

  13. Tritium Issues in Next Step Devices

    Energy Technology Data Exchange (ETDEWEB)

    C.H. Skinner; G. Federici

    2001-09-05

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

  14. Observation of tritium in cold fusion experiments

    International Nuclear Information System (INIS)

    Krishnan, M.S.; Malhotra, S.K.; Gaonkar, D.G.; Sadhukhan, H.K.

    1990-01-01

    This paper describes the results of tritium measurements carried out during the electrolysis of heavy water in different electrolysers employing palladium and titanium as cathodes. The tritium level of electrolytes have been found to be many orders of magnitudes higher than what can be explained on the basis of isotope enrichment and evaporation during electrolysis. The neutron measurement results have also been included and these observations have been attributed to the phenomenon of cold fusion. (author). 6 refs., 1 tab

  15. Desactivation of tritium waters by rectification methods

    International Nuclear Information System (INIS)

    Egorov, A.I.; Tyunis, V.M.

    2002-01-01

    Results of experiments into the basic rectification processes dedicated to tritium separation from reactor, technological and waste waters are presented. Coefficients of separation for rectification of water (1.028), ammonia (1.05), azeotrope H 2 O - HTO - HNO 3 (1.098) and D 2 O - DTO - DNO 3 (1.039) are performed. Operating schemes of tritium separating units are reviewed [ru

  16. Tritium-containment systems: a tradeoff study

    International Nuclear Information System (INIS)

    Folkers, C.L.; Cena, R.J.

    1978-01-01

    Various design parameters are evaluated that affect the performance of tritium-containment systems for fusion reactors. Our study included a review of such parameters as tritium forms, impurities, catalysts, adsorbents, getters, and as low as reasonably achievable principles. We organized these schemes, which can be considered for treating either air or inert atmospheres, so one could easily make orderly choices and tradeoffs for optimum performance. The relationships examined involved purification-system decontamination factors, flow rates, recycling and leakage, and environmental losses

  17. Energy Metabolism and Human Dosimetry of Tritium

    International Nuclear Information System (INIS)

    Galeriu, D.; Takeda, H.; Melintescu, A.; Trivedi, A.

    2005-01-01

    In the frame of current revision of human dosimetry of 14 C and tritium, undertaken by the International Commission of Radiological Protection, we propose a novel approach based on energy metabolism and a simple biokinetic model for the dynamics of dietary intake (organic 14 C, tritiated water and Organically Bound Tritium-OBT). The model predicts increased doses for HTO and OBT comparing to ICRP recommendations, supporting recent findings

  18. Preparation of pyronaridine labelled with tritium

    International Nuclear Information System (INIS)

    Jiang Shangen; Zhang Liufang; Zheng Dongzhu; Feng Zheng; Wu Zufan

    1987-01-01

    Pyronaridine is a high efficient and low toxic new antimalarial drug. 3 H-pyronaridine was prepared by catalytic isotopic exchange in solution with tritium gas using PdO/BaSO 4 as catalyst. That crude product was purified by extraction. 3 H-NMR spectra of pyronaridine showed that tritium was labelled at the 6-position. Specific activity of 3 H-pyronaridine was 5.5 Ci/mmol and radiochemical purity over 95%

  19. Preparation of pyronaridine labelled with tritium

    Energy Technology Data Exchange (ETDEWEB)

    Shangen, Jiang; Liufang, Zhang; Dongzhu, Zheng; Zheng, Feng; Zufan, Wu

    1987-12-01

    Pyronaridine is a high efficient and low toxic new antimalarial drug. /sup 3/H-pyronaridine was prepared by catalytic isotopic exchange in solution with tritium gas using PdO/BaSO/sub 4/ as catalyst. That crude product was purified by extraction. /sup 3/H-NMR spectra of pyronaridine showed that tritium was labelled at the 6-position. Specific activity of /sup 3/H-pyronaridine was 5.5 Ci/mmol and radiochemical purity over 95%.

  20. Enantiospecific tritium labeling of 28-homocastasterone.

    Science.gov (United States)

    Elbert, Tomáš; Patil, Mahadeo R; Marek, Aleš

    2017-03-01

    A regiospecific and enantiospecific synthesis of tritium-labeled 28-homocastasterone is reported. Appropriate chlorocarbonate, efficiently synthesized from the starting 28-homocastasterone in an overall yield of 46%, undergoes catalytic tritium dechlorination by the T 2 /Pd[0]/Et 3 N system, providing 28-[3β- 3 H]homocastasterone, in a good yield, radiochemical purity (>97%), and with a high specific activity (5.8 Ci/mmol). Copyright © 2016 John Wiley & Sons, Ltd.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  2. Tritium evolution from various morphologies of palladium

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  3. Tritium and helium behavior in irradiated beryllium

    International Nuclear Information System (INIS)

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

    1990-11-01

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

  4. Tritium control: October 1982-March 1983

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  5. 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)

  6. A model for global cycling of tritium

    International Nuclear Information System (INIS)

    Killough, G.G.; Kocher, D.C.

    1988-01-01

    Dynamic compartment models are widely used to describe global cycling of radionuclides for purposes of dose estimation. In this paper the authors present a new global tritium model that reproduces environmental time-series data on concentrations in precipitation, ocean surface waters, and surface fresh waters in the northern hemisphere, concentrations of atmospheric tritium in the southern hemisphere, and the latitude dependence of tritium in both hemispheres. Names TRICYCLE (for TRItium CYCLE) the model is based on the global hydrologic cycle and includes hemispheric stratospheric compartments, disaggregation of the troposphere and ocean surface waters into eight latitude zones, consideration of the different concentrations of atmospheric tritium over land and over the ocean, and a diffusive model for transport in the ocean. TRICYCLE reproduces the environmental data if it is assumed that about 50% of the tritium from atmospheric weapons testing was injected directly into the northern stratosphere as HTO. The model's latitudinal disaggregation permits taking into account the distribution of population. For a uniformly distributed release of HTO into the worldwide troposphere, TRICYCLE predicts a collective dose commitment to the world population that exceeds the NCRP model's corresponding prediction by a factor of three

  7. A tritium vessel cleanup experiment in TFTR

    International Nuclear Information System (INIS)

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

    1995-03-01

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

  8. A model for global cycling of tritium

    International Nuclear Information System (INIS)

    Killough, G.G.; Kocher, D.C.

    1988-01-01

    Dynamic compartment models are widely used to describe global cycling of radionuclides for purposes of dose estimation. In this paper, we present a new global tritium model that reproduces environmental time-series data on concentrations in precipitation, ocean surface waters, and surface fresh waters in the northern hemisphere, concentrations of atmospheric tritium in the soutehrn hemisphere, and the latitude dependence of tritium in both hemispheres. Named TRICYCLE for Tritium CYCLE, the model is based on the global hydrologic cycle and includes hemisphereic stratospheric compartments, disaggregation of the troposphere and ocean surface waters into eight latitudezones, consideration of the different concentrations of atmospheric tritium over land and over the ocean, and a diffusive model for transport in the ocean. TRICYCLE reproduces the environmental data if we assume that about 50% of the tritium from atmospheric weapons testing was injected directly into the northern stratosphere as HTO. The models latitudinal disaggregation permits taking into account the distribution of population. For a unfiormaly distributed release of HTO into the worldwide troposphere, TRICYCLE predicts a collective dose commitment to the world population that exceeds the corresponding prediction by the NCRP model by about a factor of 3. 11 refs., 5 figs., 1 tab

  9. Helium effects on tritium storage materials

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  10. Helium effects on tritium storage materials

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-15

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

  11. Separation of tritium from aqueous effluents

    International Nuclear Information System (INIS)

    Bruggeman, A.; Leysen, R.; Meynendonckx, L.; Parmentier, C.; Bellien, H.; Smets, D.; Stevens, J.

    1984-01-01

    This report describes the further development of the so-called ELEX process, carried out from 1 July 1980 until 31 December 1982. The ELEX process is the combination of electrolysis with the catalytic tritium exchange between hydrogen and water in order to accumulate the tritium in the liquid phase. The experimental study of the catalytic tritium exchange between hydrogen and liquid water was continued and the overall exchange rate could be substantially increased. An alternative process based on bithermal exchange of tritium has been evaluated. In the 10 mol h -1 mini-pilot bench scale detritiation unit the ELEX process was successfully demonstrated by detritiating up to now more than 1m 3 of water containing up to 100 mCi tritium per dm 3 , which is the feed concentration to be expected for application of the process in a reprocessing plant. A 280 mol h -1 pilot detritiation installation now being constructed is described. This installation will realize a volume reduction factor of 100 and a process decontamination factor of 100. The maximum total tritium inventory will be about 1000 Ci. The plant consists mainly of a 80 kW electrolyser and a 10 cm diameter exchange column and can be considered as the ultimate step before industrial application of the ELEX process

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

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  13. A Survey of Tritium in Irish Seawater

    International Nuclear Information System (INIS)

    Currivan, L.; Kelleher, K.; McGinnity, P.; Wong, J.; McMahon, C.

    2013-07-01

    This report provides a comprehensive record of the study and measurements of tritium in Irish seawater undertaken by the Radiological Protection Institute of Ireland RPII. The majority of the samples analysed were found to have tritium concentrations below the limit of detection and a conservative assessment of radiation dose arising showed a negligible impact to the public. Tritium is discharged in large quantities from various nuclear facilities, and mostly in liquid form. For this reason it is included in the list of radioactive substances of interest to the OSPAR (Oslo-Paris) Convention to protect the marine environment of the North-East Atlantic. To fulfil its role within OSPAR, to provide technical support to the Irish Government, RPII carried out a project to determine the levels of tritium in seawater from around the Irish coast to supplement its routine marine monitoring programme. A total of 85 seawater samples were collected over a three year period and analysed at the RPII's laboratory. Given that the operational discharges for tritium from the nuclear fuel reprocessing plant at Sellafield, UK, are expected to increase due to current and planned decommissioning activities RPII will continue to monitor tritium levels in seawater around the Irish coast, including the Irish Sea, as part of its routine marine monitoring programme

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  15. Design options to minimize tritium inventories at Savannah River

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  16. Design options to minimize tritium inventories at Savannah River

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  17. Studies on steps affecting tritium residence time in solid blanket

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  19. Tritium means of detection and of protection; Le tritium moyens de detection et de protection

    Energy Technology Data Exchange (ETDEWEB)

    Sutra-Fourcade, Y [Commissariat a l' Energie Atomique, Marcoule (France). Centre d' Etudes Nucleaires

    1967-07-01

    The report is an attempt to correlate present data concerning tritium, especially from the health physics points of view. The various detection and measurement methods are reviewed in turn: measurement of tritium in the atmosphere, in liquids and on surfaces. The operation of various types of apparatus is analyzed and the sensitivity limits deduced from laboratory tests are given. Otter sections are devoted to the means of protection which can be used against inhalation of tritium (ventilation, protective clothing) and to calculations of the changes in atmospheric pollution in a given place and of the time spent in a contaminated zone. The last part deals with the decontamination of equipment contaminated with tritium. (author) [French] Le rapport represente un essai de synthese des connaissances actuelles sur le tritium, essentiellement du point de vue de la radioprotection. Les differents moyens de detection et de mesure sont successivement passes en revue: mesure du tritium dans l'atmosphere, dans les liquides, sur les surfaces. Le fonctionnement de differents types d'appareils est analyse et les limites de sensibilite sont donnees d'apres les essais effectues en laboratoire. D'autres paragraphes sont consacres aux moyens de protection contre l'inhalation du tritium (ventilation, vetements de protection), a des calculs d'evolution de pollution atmospherique dans les locaux et de temps de presence en atmosphere contaminee. La derniere partie se rapporte a la de contamination de materiel contamine par du tritium. (auteur)

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

    Science.gov (United States)

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

    2016-01-01

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