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Sample records for accidental tritium release

  1. Method and apparatus for controlling accidental releases of tritium

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1980-01-01

    An improvement in a tritium control system based on a catalytic oxidation reactor is provided wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release

  2. Method and apparatus for controlling accidental releases of tritium

    Science.gov (United States)

    Galloway, Terry R. [Berkeley, CA

    1980-04-01

    An improvement in a tritium control system based on a catalytic oxidation reactor wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release.

  3. Results of dose calculations for NET accidental and normal operation releases of tritium and activation products

    International Nuclear Information System (INIS)

    Raskob, W.; Hasemann, I.

    1992-08-01

    This report documents conditions, data and results of dose calculations for accidental and normal operation releases of tritium and activation products, performed within the NET subtask SEP2.2 ('NET-Benchmark') of the European Fusion Technology Programme. For accidental releases, the computer codes UFOTRI and COSYMA for assessing the radiological consequences, have been applied for both deterministic and probabilistic calculations. The influence on dose estimates of different release times (2 minutes / 1 hour), two release heights (10 m / 150 m), two chemical forms of tritium (HT/HTO), and two different model approaches for the deposition velocity of HTO on soil was investigated. The dose calculations for normal operation effluents were performed using the tritium model of the German regulatory guidelines, parts of the advanced dose assessment model NORMTRI still under development, and the statistical atmospheric dispersion model ISOLA. Accidental and normal operation source terms were defined as follows: 10g (3.7 10 15 Bq) for accidental tritium releases, 10 Ci/day (3.7 10 11 Bq/day) for tritium releases during normal operation and unit releases of 10 9 Bq for accidental releases of activation products and fission products. (orig./HP) [de

  4. Modelling accidental releases of tritium in the environment: application as an excel spreadsheet

    International Nuclear Information System (INIS)

    Le Dizes, S.; Tamponnet, C.

    2004-01-01

    An application as an Excel spreadsheet of the simplified modelling approach of tritium transfer in the environment developed by Tamponnet (2002) is presented. Based on the use of growth models of biological systems (plants, animals, etc.), the two-pool model (organic tritium and tritiated water) that was developed estimates the concentration of tritium within the different compartments of the food chain and in fine the dose to man by ingestion in the case of a chronic or accidental release of tritium in a river or the atmosphere. Data and knowledge have been implemented on Excel using the object-oriented programming language VisualBasic (Microsoft Visual Basic 6.0). The structure of the conceptual model and the Excel sheet are first briefly exposed. A numerical application of the model under a scenario of an accidental release of tritium in the atmosphere is then presented. Simulation results and perspectives are discussed. (author)

  5. Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor

    International Nuclear Information System (INIS)

    Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

    2015-01-01

    As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. - Highlights: • Dynamic tritium concentration in the air and human body evaluated on the time scale. • Different intake forms and relevant radiation dose assessed to find out the potential relationships. • HT and HTO release amount limits for arbitrary accidents proposed for the fusion reactor according to dose limit

  6. Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor.

    Science.gov (United States)

    Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

    2015-10-01

    As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. Copyright © 2015 Elsevier Ltd. All rights reserved.

  7. Modelling of the tritium dispersion from postulated accidental release of nuclear power plants

    International Nuclear Information System (INIS)

    Soares, Abner Duarte; Simoes Filho, Francisco Fernando Lamego; Cunha, Tatiana Santos da; Aguiar, Andre Silva de; Lapa, Celso Marcelo Franklin

    2011-01-01

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models to simulation of tritium dispersion caused by an accident in a Candu reactor located in the ongoing Angra 3 site. The Candu reactor is one that uses heavy water (D 2 O) as moderator and coolant of the core. It was postulated, then, the LOCA accident (without fusion), where was lost 66 m3 of soda almost instantaneously. This inventory contained 35 P Bq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios (plant stopped or operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1 MBq/m 3 ) during the first 14 days after the accident. (author)

  8. The hazard to man of accidental releases of tritium

    International Nuclear Information System (INIS)

    Brearley, I.R.

    1985-03-01

    Some aspects of the atmospheric dispersion of tritium are discussed, followed by consideration of the dosimetric pathways. In order to assess the significance of a tritium release the doses from various pathways are estimated and compared with the doses estimated from a similar release of iodine-131. The major hazard from tritium is the ingestion of contaminated food products. For similar releases of tritium and I 131 the ingestion hazard can be comparable if the release occurs near and before the end of the harvest season. However, in the tritium release case the agricultural season influences the consequences markedly and, at other times during the year, the ingestion hazard from tritium may be approximately 20 times less. The dose from inhalation of tritium is sensitive to its chemical form and for similar releases of tritiated water and tritium gas then the dose from tritiated water is approximately 10 4 greater than the dose from tritium gas. For similar releases of tritiated water and iodine-131 then a comparison of the inhalation shows that the dose from the iodine is approximately 300 times greater. (author)

  9. Modelling of tritium dispersion from postulated accidental release of nuclear power plants

    International Nuclear Information System (INIS)

    Soares, Abner Duarte

    2010-01-01

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models of hydrodynamics and transport for the simulation of tritium dispersion caused by an accident in a CANDU reactor located in the ongoing Angra 3 site. This exercise was accomplished with the aid of a code system (SisBAHIA) developed in the Rio de Janeiro Federal University (COPPE/UFRJ). The CANDU reactor is one that uses heavy water (D 2 O) as moderator and coolant of the core. It was postulated, then, the LOCA (Loss of Coolant Accident) accident in the emergency cooling system of the nucleus (without fusion), where was lost 66 m 3 of soda almost instantaneously. This inventory contained 35 PBq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios (plant stopped and operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1 MBq/m 3 ) during the first 14 days after the accident. The main difference between the scenario without and with seawater recirculation (pumping and discharge) is based on the enhancement of dilution of the highest concentrations in the last one. This dilution enhancement resulting in decreasing concentrations was observed only during the first two weeks, when they ranged from 1x10 9 to 5x10 5 Bq/m 3 close to the Itaorna beach spreading just to Sandri Island. After 180 days, the plume could not be detected anymore in the bay, because their activities would be lower than the minimum detectable value ( 3 ). (author)

  10. Tritium release experiments with CATS and numerical simulation

    International Nuclear Information System (INIS)

    Munakata, Kenzo; Wajima, Takaaki; Hara, Keisuke; Wada, Kohei; Takeishi, Toshiharu; Shinozaki, Yohei; Mochizuki, Kazuhiro; Katekari, Kenichi; Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko

    2010-01-01

    In D-T fusion power plants, large amounts of tritium would be handled. Tritium is the radioisotope of protium, and is easily taken into the human body, and thus the behavior of tritium accidentally released in fusion power plants should be studied for the safety design and radioprotection of workers. Therefore, it is necessary to investigate the behavior of tritium released into large rooms with objectives, since complex flow fields should exist in such rooms and they could influence the ventilation of the air containing released tritium. Thus, tritium release experiments were conducted using Caisson Assembly for Tritium Safety Study (CATS) in TPL/JAEA. Some data were taken for tritium behavior in the ventilated area and response of tritium monitors. In the experiments, approximately 17 GBq of tritium was released into Caisson with the total volume of 12 m 3 , and the room was ventilated at the rate of 12 m 3 /h after release of tritium. It was found that placement of an objective in the vessel substantially affects decontamination efficiency. With regard to an experimental result, numerical calculation was performed and the experimental result and the result of numerical calculation were compared, which indicates that experimental results are qualitatively reproduced by numerical calculation. However, further R and D needs to be carried out for quantitative reproduction of the experimental results.

  11. Tritium release experiments with CATS and numerical simulation

    Energy Technology Data Exchange (ETDEWEB)

    Munakata, Kenzo, E-mail: kenzo@gipc.akita-u.ac.jp [Faculty of Engineering and Resource Sciences, Akita University, Tegata-gakuen-cho 1-1, Akita 010-8502 (Japan); Wajima, Takaaki; Hara, Keisuke; Wada, Kohei [Faculty of Engineering and Resource Sciences, Akita University, Tegata-gakuen-cho 1-1, Akita 010-8502 (Japan); Takeishi, Toshiharu; Shinozaki, Yohei; Mochizuki, Kazuhiro; Katekari, Kenichi [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581 (Japan); Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko [Tritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2010-12-15

    In D-T fusion power plants, large amounts of tritium would be handled. Tritium is the radioisotope of protium, and is easily taken into the human body, and thus the behavior of tritium accidentally released in fusion power plants should be studied for the safety design and radioprotection of workers. Therefore, it is necessary to investigate the behavior of tritium released into large rooms with objectives, since complex flow fields should exist in such rooms and they could influence the ventilation of the air containing released tritium. Thus, tritium release experiments were conducted using Caisson Assembly for Tritium Safety Study (CATS) in TPL/JAEA. Some data were taken for tritium behavior in the ventilated area and response of tritium monitors. In the experiments, approximately 17 GBq of tritium was released into Caisson with the total volume of 12 m{sup 3}, and the room was ventilated at the rate of 12 m{sup 3}/h after release of tritium. It was found that placement of an objective in the vessel substantially affects decontamination efficiency. With regard to an experimental result, numerical calculation was performed and the experimental result and the result of numerical calculation were compared, which indicates that experimental results are qualitatively reproduced by numerical calculation. However, further R and D needs to be carried out for quantitative reproduction of the experimental results.

  12. Modeling of the dispersion of tritium from postulated accidental releases from nuclear power plants

    International Nuclear Information System (INIS)

    Aguiar, Andre Silva de; Simoes Filho, Francisco Fernando Lamego; Lapa, Celso Marcelo Franklin; Alvim, Antonio Carlos Marques; Soares, Abner Duarte

    2013-01-01

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models of hydrodynamics and transport for the simulation of tritium dispersion caused by an accident in a CANDU reactor located in the ongoing Angra 3 site. It was postulated, then, the LOCA - Loss of Coolant Accident -, accident in the emergency cooling system of the nucleus ( without fusion), where was lost 66m 3 of soda almost instantaneously. This inventory contained 35 PBq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios ( plant stopped or operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1MBq/m 3 ), during the first 14 days after the accident. The main difference between the scenario without and with seawater recirculation (pumping and discharge) is based on the enhancement of dilution of the highest concentrations in the last one. This dilution enhancement resulting in decreasing concentrations was observed only during the first two weeks, when they ranged from 1x10 9 to 5x10 5 Bq/m 3 close to the Itaorna beach spreading just to Sandri Island. After 180 days, the plume could not be detected anymore in the bay, because their activities would be lower than the minimum detectable value ( 3 ). (author)

  13. The present status and recent applications of the accidental tritium assessment code UFOTRI

    International Nuclear Information System (INIS)

    Raskob, W.

    1999-01-01

    The computer program UFOTRI can be used for assessing the impact of accidental released tritium in the two chemical forms tritiated water vapour and tritium gas. By applying UFOTRI to potential European sites for ITER, it could be demonstrated that the main goal, the nonevacuation criteria, is fulfilled for the present release limits. Contributions in international studies together with the re-evaluation of experimental data showed that the plant sub-model as well as the soil sub-model are areas for further improvement. (author)

  14. Preliminary analysis of public dose from CFETR gaseous tritium release

    Energy Technology Data Exchange (ETDEWEB)

    Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230027 (China); Ni, Muyi, E-mail: muyi.ni@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Lian, Chao; Jiang, Jieqiong [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2015-02-15

    Highlights: • Present the amounts and limit dose of tritium release to the environment for CFETR. • Perform a preliminary simulation of radiation dose for gaseous tritium release. • Key parameters about soil types, wind speed, stability class, effective release height and age were sensitivity analyzed. • Tritium release amount is recalculated consistently with dose limit in Chinese regulation for CFETR. - Abstract: To demonstrate tritium self-sufficiency and other engineering issues, the scientific conception of Chinese Fusion Engineering Test Reactor (CFETR) has been proposed in China parallel with ITER and before DEMO reactor. Tritium environmental safety for CFETR is an important issue and must be evaluated because of the huge amounts of tritium cycling in reactor. In this work, different tritium release scenarios of CFETR and dose limit regulations in China are introduced. And the public dose is preliminarily analyzed under normal and accidental events. Furthermore, after finishing the sensitivity analysis of key input parameters, the public dose is reevaluated based on extreme parameters. Finally, tritium release amount is recalculated consistently with the dose limit in Chinese regulation for CFETR, which would provide a reference for tritium system design of CFETR.

  15. 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

  16. Results for SEAFP-subtask A 10: Assessments of individual and collective doses to the public for routine and accidental releases of tritium and activation products

    International Nuclear Information System (INIS)

    Raskob, W.

    1995-04-01

    Dose calculations have been performed for accidental releases of tritium and activation products. Unit releases of 1 GBq per nuclide have been investigated for 31 fusion relevant nuclides. Additionally, unit releases of 1g of tritium and several activated materials have been studied. Under normal operation conditions, dose calculations have been performed for real source terms of tritium and activated materials. The individual dose values at the fence of the site (1 km) as well as the collective dose to the public (from 1 km to 100 km) have been obtained. As site specific parameters are still missing, different so called ''worst case'' release conditions have been applied. To have a first guess of the influence of the release duration on the dose to the Most Exposed Individual (MEI) in the vicinity of a reactor, different release durations, ranging from 1 hour up to 168 hours have been investigated, too. Finally, dose calculations have been performed for mobilisation source terms which take account of deposition and retention in the plant. This has been done for several RPM and APM source terms. The dose values of these final source terms seem to be less than every criteria to start emergency actions, however, some problems e.g. the behaviour of tritium in the plant, remain unsolved. (orig.)

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

    International Nuclear Information System (INIS)

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

    1975-08-01

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

  18. Environmental and radiological impact of accidental tritium release

    Energy Technology Data Exchange (ETDEWEB)

    Guetat, Ph. [CEA Valduc, 21 - Is-sur-Tille (France); Patryl, L. [CEA Bruyeres-le-Chatel, 91 (France)

    2008-07-15

    Within the IAEA program EMRAS, an exercise has been performed by 7 countries to evaluate the consequences of an acute atmospheric release of tritium (10 g). This study aimed at giving practical technical information to decision-makers. Three scenarios with different meteorological conditions were modeled. The objective of this paper is to give the main information about transfer and impact, evaluate uncertainties in models/assessments and so help to set countermeasures. From the results of this exercise, reference activity values for countermeasures can be discussed. All pathways and chemical forms (HTO-HT-OBT) are considered. (authors)

  19. Environmental contamination due to release of a large amount of tritium

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    1988-01-01

    Tritium release incidents have occurred many times in the Savannah Rever Plant in the U.S. A tritium release incident also took place in the Lawrence Livermore Laboratory. The present article outlines the reports by the plant and laboratory on these incidents and makes some comments on environmental contamination that may results from release of a large amount of tritium from nuclear fusion facilities. Tritium is normally released in the form of a combination of chemical compounds such as HT, DT and T 2 and oxides such as HTO, DTO and T 2 O. The percentage of the oxides is given in the reports by the plant. Oxides, which can be absorbed through the skin, are considered to be nearly a thousand times more toxic than the other type of tritium compounds. The HT type compounds (HT, DT and T 2 ) can be oxidized by microorganisms in soil into oxides (HTO, DTO and T 2 O) and therefore, great care should also given to this type of compounds. After each accidental tritium release, the health physics group of the plant collected various environmental samples, including ground surface water, milk, leaves of plants, soil and human urine, in leeward areas. Results on the contamination of surface water, fish and underground water are outlined and discussed. (Nogami, K.)

  20. Modelling of the tritium dispersion from postulated accidental release of nuclear power plants; Modelagem da dispersao de tritio a partir de liberacoes acidentais postuladas de centrais nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Soares, Abner Duarte; Simoes Filho, Francisco Fernando Lamego; Cunha, Tatiana Santos da [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Aguiar, Andre Silva de; Lapa, Celso Marcelo Franklin, E-mail: asoares@cnen.gov.b, E-mail: flamego@ien.gov.b, E-mail: lapa@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models to simulation of tritium dispersion caused by an accident in a Candu reactor located in the ongoing Angra 3 site. The Candu reactor is one that uses heavy water (D{sub 2}O) as moderator and coolant of the core. It was postulated, then, the LOCA accident (without fusion), where was lost 66 m3 of soda almost instantaneously. This inventory contained 35 P Bq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios (plant stopped or operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1 MBq/m{sup 3}) during the first 14 days after the accident. (author)

  1. Accidental tritium release from nuclear technologies and a radiobiological survey of the impact of low dose tritium on the developing mouse brain

    International Nuclear Information System (INIS)

    Jain, Narendra; Bhatia, A.L.

    2008-01-01

    Full text: The Atomic Energy Act, 1962 provides for the development of the peaceful uses of atomic energy for the welfare of the people in India. The licensing policy adopted for nuclear power stations in India requires that the plants meet stringent requirements based on the system of dose limitation, recommended by the International Commission of Radiological Protection (ICRP). Currently, nuclear energy is contributing just 3% of the country's power generation. The share of nuclear power is proposed to be increased to 10% in the near future. With the introduction of nuclear energy, the need to assess the radioecological and radiobiological impact of radionuclides of long half- life existing in the environment for longer duration has appeared. Tritium, a radioactive by-product of power reactors is one of such major radionuclides of concern. In the world, routine releases and accidental spills of tritium from nuclear power plants pose a growing health and safety concern. Tritium has been observed in ground water in the vicinity of several nuclear stations. Exposure to tritium has been clinically proven to cause deleterious and detectable effects such as teratogenesis, cancer and life shortening in laboratory animals. There is, now, a growing emphasis on tritium in radiation protection as the challenge of nuclear fusion comes nearer. Present investigation is an attempt to elucidate the effects of low dose tritiated water exposure on developing mouse cerebellum. Pregnant Swiss albino mice (12-15 in number were given a priming injection 7.4 and 74 kBq/ml of body water) of tritiated water (HTO) on 16 th day of gestation. From the same day onward, through parturition, till the last interval studied, the pregnant females were continuously maintained respectively on 11.1 and 111 kBq/ml of tritiated drinking water provided ad libidum. After cervical dislocation the litters were autopsied on 1, 3, 5 and 6 weeks post- partum. Brains were fixed and then cerebellum from each

  2. Historical Doses from Tritiated Water and Tritiated Hydrogen Gas Released to the Atmosphere from Lawrence Livermore National Laboratory (LLNL). Part 5. Accidental Releases

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, S

    2007-08-15

    Over the course of fifty-three years, LLNL had six acute releases of tritiated hydrogen gas (HT) and one acute release of tritiated water vapor (HTO) that were too large relative to the annual releases to be included as part of the annual releases from normal operations detailed in Parts 3 and 4 of the Tritium Dose Reconstruction (TDR). Sandia National Laboratories/California (SNL/CA) had one such release of HT and one of HTO. Doses to the maximally exposed individual (MEI) for these accidents have been modeled using an equation derived from the time-dependent tritium model, UFOTRI, and parameter values based on expert judgment. All of these acute releases are described in this report. Doses that could not have been exceeded from the large HT releases of 1965 and 1970 were calculated to be 43 {micro}Sv (4.3 mrem) and 120 {micro}Sv (12 mrem) to an adult, respectively. Two published sets of dose predictions for the accidental HT release in 1970 are compared with the dose predictions of this TDR. The highest predicted dose was for an acute release of HTO in 1954. For this release, the dose that could not have been exceeded was estimated to have been 2 mSv (200 mrem), although, because of the high uncertainty about the predictions, the likely dose may have been as low as 360 {micro}Sv (36 mrem) or less. The estimated maximum exposures from the accidental releases were such that no adverse health effects would be expected. Appendix A lists all accidents and large routine puff releases that have occurred at LLNL and SNL/CA between 1953 and 2005. Appendix B describes the processes unique to tritium that must be modeled after an acute release, some of the time-dependent tritium models being used today, and the results of tests of these models.

  3. 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)

  4. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Freund, Jana, E-mail: jana.freund@kit.edu; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  7. Analysis of in-pile tritium release experiments

    International Nuclear Information System (INIS)

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

    1992-01-01

    The objective of this work is to characterize tritium release behavior from lithium ceramics and develop insight into the underlying tritium release mechanisms. Analysis of tritium release data from recent laboratory experiments with lithium aluminate has identified physical processes which were previously unaccounted for in tritium release models. A new model that incorporates the recent data and provides for release from multiple sites rather than only one site was developed. Calculations of tritium release using this model are in excellent agreement with the tritium release behavior reported for the MOZART experiment

  8. 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.)

  9. 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)

  10. 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

  11. 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)

  12. 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.))

  13. 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)

  14. Estimation of the environmental or radiological impact in the event of accidental release of radionuclides in a DCLL fusion reactor; Estimacion del impacto radiologico ambiental en caso de liberacion accidental de radionucleidos en un reactor de fusion DCLL

    Energy Technology Data Exchange (ETDEWEB)

    Palermo, I.; Gomez Ros, J. M.; Sanz, J.; Mota, F.

    2013-07-01

    Tritium production and activation in the LiPb products can pose a radiological risk in the event of accidental release in a fusion reactor. Within the research programme Consolider TECNO{sub F}US (CSD2008-079) fusion technology has developed a design for a reactor with regenerative wrap with dual refrigeration (DCLL). The purpose of this communication is to present estimates of the radiological impact derived from an accidental release of radionuclides from the circuit of LiPb provinients. (Author)

  15. 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

  16. 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

  17. Investigation of tritium release and retention in lithium aluminate

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  19. Pebble fabrication and tritium release properties of an advanced tritium breeder

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi, E-mail: hoshino.tsuyoshi@jaea.go.jp [Breeding Functional Materials Development Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Obuch, Omotedate, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan); Edao, Yuki [Tritium Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-4 Shirakata, Shirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kawamura, Yoshinori [Blanket Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ochiai, Kentaro [BA Project Coordination Group, Department of Fusion Power Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Obuch, Omotedate, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Li{sub 2}TiO{sub 3} with excess Li (Li{sub 2+x}TiO{sub 3+y}) pebble as an advanced tritium breeders was fabricated using emulsion method. • Grain size of Li{sub 2+x}TiO{sub 3+y} pebbles was controlled to be less than 5 μm. • Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties similar to that of Li{sub 2}TiO{sub 3} pebbles. - Abstract: Li{sub 2}TiO{sub 3} with excess Li (Li{sub 2+x}TiO{sub 3+y}) has been developed as an advanced tritium breeder. With respect to the tritium release characteristics of the blanket, the optimum grain size after sintering was less than 5 μm. Therefore, an emulsion method was developed to fabricate pebbles with this target grain size. The predominant factor affecting grain growth was assumed to be the presence of binder in the gel particles; this remaining binder was hypothesized to react with the excess Li, thereby generating Li{sub 2}CO{sub 3}, which promotes grain growth. To inhibit the generation of Li{sub 2}CO{sub 3}, calcined Li{sub 2+x}TiO{sub 3+y} pebbles were sintered under vacuum and subsequently under a 1% H{sub 2}–He atmosphere. The average grain size of the sintered Li{sub 2+x}TiO{sub 3+y} pebbles was less than 5 μm. Furthermore, the tritium release properties of Li{sub 2+x}TiO{sub 3+y} pebbles were evaluated, and deuterium–tritium (DT) neutron irradiation experiments were performed at the Fusion Neutronics Source facility in the Japan Atomic Energy Agency. To remove the tritium produced by neutron irradiation, 1% H{sub 2}–He purge gas was passed through the Li{sub 2+x}TiO{sub 3+y} pebbles. The Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties, similar to those of Li{sub 2}TiO{sub 3} pebbles. In particular, the released amount of tritiated hydrogen gas for easier tritium handling was greater than the released amount of tritiated water.

  20. 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)

  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. ACUTRI a computer code for assessing doses to the general public due to acute tritium releases

    CERN Document Server

    Yokoyama, S; Noguchi, H; Ryufuku, S; Sasaki, T

    2002-01-01

    Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety assessment of a nuclear fusion experimental reactor such as ITER. Thus, a computer code, ACUTRI, which calculates the radiological impact of tritium released accidentally to the atmosphere, has been developed, aiming to be of use in a discussion of licensing of a fusion experimental reactor and an environmental safety evaluation method in Japan. ACUTRI calculates an individual tritium dose based on transfer models specific to tritium in the environment and ICRP dose models. In this calculation it is also possible to analyze statistically on meteorology in the same way as a conventional dose assessment method according to the meteorological guide of the Nuclear Safety Commission of Japan. A Gaussian plume model is used for calculating the atmospheric dispersion of tritium gas (HT) and/or tritiated water (HTO). The environmental pathway model in ACUTRI considers the following internal exposures: i...

  3. 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)

  4. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Science.gov (United States)

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

    2017-06-01

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

  5. ACUTRI: a computer code for assessing doses to the general public due to acute tritium releases

    Energy Technology Data Exchange (ETDEWEB)

    Yokoyama, Sumi; Noguchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ryufuku, Susumu; Sasaki, Toshihisa; Kurosawa, Naohiro [Visible Information Center, Inc., Tokai, Ibaraki (Japan)

    2002-11-01

    Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety assessment of a nuclear fusion experimental reactor such as ITER. Thus, a computer code, ACUTRI, which calculates the radiological impact of tritium released accidentally to the atmosphere, has been developed, aiming to be of use in a discussion of licensing of a fusion experimental reactor and an environmental safety evaluation method in Japan. ACUTRI calculates an individual tritium dose based on transfer models specific to tritium in the environment and ICRP dose models. In this calculation it is also possible to analyze statistically on meteorology in the same way as a conventional dose assessment method according to the meteorological guide of the Nuclear Safety Commission of Japan. A Gaussian plume model is used for calculating the atmospheric dispersion of tritium gas (HT) and/or tritiated water (HTO). The environmental pathway model in ACUTRI considers the following internal exposures: inhalation from a primary plume (HT and/or HTO) released from the facilities and inhalation from a secondary plume (HTO) reemitted from the ground following deposition of HT and HTO. This report describes an outline of the ACUTRI code, a user guide and the results of test calculation. (author)

  6. Tritium adsorption/release behaviour of advanced EU breeder pebbles

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-15

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

  7. Analysis of accidental UF6 releases

    International Nuclear Information System (INIS)

    Fan Yumao; Tan Rui; Gao Qifa

    2012-01-01

    As interim substance in the nuclear fuel enrichment process, Uranium Hexafluoride (UF 6 ) is widely applied in nuclear processing, enrichment and fuel fabrication plants. Because of its vivid chemical characteristics and special radiological hazard and chemical toxicity, great attention must be paid to accident of UF 6 leakage. The chemical reactions involved in UF 6 release processes were introduced, therewith potential release styles, pathways and characteristics of diffusion were analyzed. The results indicated that the accidental release process of UF 6 is not a simple passive diffusion. So, specific atmospheric diffusion model related to UF 6 releases need be used in order to analyze and evaluate accurately the accidental consequences. (authors)

  8. Modelling of tritium dispersion from postulated accidental release of nuclear power plants; Modelagem da dispersao de tritio a partir de liberacoes acidentais postuladas de centrais nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Soares, Abner Duarte

    2010-07-01

    This study has the aim to assess the impact of accidental release of tritium postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models of hydrodynamics and transport for the simulation of tritium dispersion caused by an accident in a CANDU reactor located in the ongoing Angra 3 site. This exercise was accomplished with the aid of a code system (SisBAHIA) developed in the Rio de Janeiro Federal University (COPPE/UFRJ). The CANDU reactor is one that uses heavy water (D{sub 2}O) as moderator and coolant of the core. It was postulated, then, the LOCA (Loss of Coolant Accident) accident in the emergency cooling system of the nucleus (without fusion), where was lost 66 m{sup 3} of soda almost instantaneously. This inventory contained 35 PBq and was released a load of 9.7 TBq/s in liquid form near the Itaorna beach, Angra dos Reis - RJ. The models mentioned above were applied in two scenarios (plant stopped and operating) and showed a tritium plume with specific activities larger than the reference level for seawater (1.1 MBq/m{sup 3} ) during the first 14 days after the accident. The main difference between the scenario without and with seawater recirculation (pumping and discharge) is based on the enhancement of dilution of the highest concentrations in the last one. This dilution enhancement resulting in decreasing concentrations was observed only during the first two weeks, when they ranged from 1x10{sup 9} to 5x10{sup 5} Bq/m{sup 3} close to the Itaorna beach spreading just to Sandri Island. After 180 days, the plume could not be detected anymore in the bay, because their activities would be lower than the minimum detectable value (< 11 kBq/m{sup 3}). (author)

  9. Tritium release from lithium ceramics at constant temperature

    International Nuclear Information System (INIS)

    Verrall, R.A.; Miller, J.M.

    1992-02-01

    Analytic methods for post-irradiation annealing tests to measure tritium release from lithium ceramics at constant temperature are examined. Modifications to the Bertone (1) relations for distinguishing diffusion-controlled release from desorption-controlled release are shown. The methods are applied to tests on sintered LiA10 2 ; first-order desorption is shown to control tritium release for these tests

  10. Release of gaseous tritium during reprocessing

    International Nuclear Information System (INIS)

    Bruecher, H.; Hartmann, K.

    1983-01-01

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

  11. Tritium and helium release from beryllium pebbles neutron-irradiated up to 230appm tritium and 3000appm helium

    Directory of Open Access Journals (Sweden)

    Vladimir Chakin

    2016-12-01

    Full Text Available Study of tritium and helium release from beryllium pebbles with diameters of 0.5 and 1mm after high-dose neutron irradiation at temperatures of 686–968K was performed. The release rate always has a single peak, and the peak temperatures at heating rates of 0.017K/s and 0.117K/s lie in the range of 1100–1350K for both tritium and helium release. The total tritium release from 1mm pebbles decreases considerably by increasing the irradiation temperature. The total tritium release from 0.5mm pebbles is less than that from 1mm pebbles and remains constant regardless of the irradiation temperature. At high irradiation temperatures, open channels are formed which contribute to the enhanced tritium release.

  12. 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

  13. Biological effects of tritium releases from fusion power plants

    International Nuclear Information System (INIS)

    Strand, J.A.; Thompson, R.C.

    1976-09-01

    Tritium released as tritium oxide is a much more significant potential hazard to the environment than is elemental tritium. Although most biochemical reactions discriminate against the incorporation of tritium in favor of hydrogen, the possibility of some concentration should not be overlooked. A fraction of tritium accumulated as tritiated water becomes organically bound, that is, exchanges with hydrogen bound in organic molecules. The rate and extent of incorporation are dependent upon metabolic activity of the organism. On this basis, the highest concentration of organically-bound tritium would be expected in tissues and population segments which are in formative or growth stages at the time of exposure. Furthermore, as exposure duration increases from acute to chronic situations, tritium concentrations are shown to approach equilibrium levels with a single tritium-to-hydrogen ratio common to all parts of the hydrogen pool. Organic binding would not be expected to result in significant bioaccumulation of tritium from tritiated water. Tritium loss, both from tissue-free water and the tissue-bound fraction, depends upon metabolic activity. Processes that allow accumulation and incorporation of tritium also assist its elimination. Tritium which is organically bound demonstrates a longer half-time, but it would appear to constitute a small fraction of the total tritium label. The radiation exposure of all living organisms by environmentally dispersed tritium, in whatever form, is essentially a whole body exposure. Uncertainties in the individual parameters, involved in converting measured intake to estimated dose equivalent are probably no larger than a factor of three or four. If fusion reactors hold tritium releases with ICRP standards, no significant adverse impact to the environment from those releases are expected

  14. Tritium and helium retention and release from irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Longhurst, G.R.; Oates, M.A.; Pawelko, R.J. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

    1998-01-01

    This paper reports the results of an experimental effort to anneal irradiated beryllium specimens and characterize them for steam-chemical reactivity experiments. Fully-dense, consolidated powder metallurgy Be cylinders, irradiated in the EBR-II to a fast neutron (>0.1 MeV) fluence of {approx}6 x 10{sup 22} n/cm{sup 2}, were annealed at temperatures from 450degC to 1200degC. The releases of tritium and helium were measured during the heat-up phase and during the high-temperature anneals. These experiments revealed that, at 600degC and below, there was insignificant gas release. Tritium release at 700degC exhibited a delayed increase in the release rate, while the specimen was at 700degC. For anneal temperatures of 800degC and higher, tritium and helium release was concurrent and the release behavior was characterized by gas-burst peaks. Essentially all of the tritium and helium was released at temperatures of 1000degC and higher, whereas about 1/10 of the tritium was released during the anneals at 700degC and 800degC. Measurements were made to determine the bulk density, porosity and specific surface area for each specimen before and after annealing. These measurements indicated that annealing caused the irradiated Be to swell, by as much as 14% at 700degC and 56% at 1200degC. Kr gas adsorption measurements for samples annealed at 1000degC and 1200degC determined specific surface areas between 0.04 m{sup 2}/g and 0.1 m{sup 2}/g for these annealed specimens. The tritium and helium gas release measurements and the specific surface area measurements indicated that annealing of irradiated Be caused a porosity network to evolve and become surface-connected to relieve internal gas pressure. (author)

  15. Retention and release of tritium in aluminum clad, Al-Li alloys

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.

    1991-01-01

    Tritium retention in and release from aluminum clad, aluminum-lithium alloys is modeled from experimental and operational data developed during the thirty plus years of tritium production at the Savannah River Site. The model assumes that tritium atoms, formed by the 6 Li(n,α) 3 He reaction, are produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly becomes supersaturated in tritium. Newly produced tritium atoms are trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability is the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release is determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. This model is used to calculate tritium release from aluminum clad, aluminum-lithium alloys. 9 refs., 3 figs

  16. Release enhancement of tritium from graphite by addition of hydrogen

    International Nuclear Information System (INIS)

    Saeki, Masakatsu; Masaki, N.M.

    1989-01-01

    The release behavior of tritium from graphite was studied in pure He and He + H 2 atmosphere. The release from powdered graphite was significantly enhanced in hydrogen environment. Apparent diffusion coefficients of tritium in graphite also became much higher in an atmosphere containing hydrogen than values obtained in pure helium atmosphere. A careful investigation of the release processes resulted in the conclusion that the most important process of tritium behaviour in graphite was diffusion, but the desorption process of tritium from the surface played a significant role. The enhancement of the desorption process was controlled by atomic hydrogen. (orig.)

  17. An overview of organically bound tritium experiments in plants following a short atmospheric HTO exposure.

    Science.gov (United States)

    Galeriu, D; Melintescu, A; Strack, S; Atarashi-Andoh, M; Kim, S B

    2013-04-01

    The need for a less conservative, but reliable risk assessment of accidental tritium releases is emphasized in the present debate on the nuclear energy future. The development of a standard conceptual model for accidental tritium releases must be based on the process level analysis and the appropriate experimental database. Tritium transfer from atmosphere to plants and the subsequent conversion into organically bound tritium (OBT) strongly depends on the plant characteristics, seasons, and meteorological conditions, which have a large variability. The present study presents an overview of the relevant experimental data for the short term exposure, including the unpublished information, also. Plenty of experimental data is provided for wheat, rice, and soybean and some for potato, bean, cherry tomato, radish, cabbage, and tangerine as well. Tritiated water (HTO) uptake by plants during the daytime and nighttime has an important role in further OBT synthesis. OBT formation in crops depends on the development stage, length, and condition of exposure. OBT translocation to the edible plant parts differs between the crops analyzed. OBT formation during the nighttime is comparable with that during the daytime. The present study is a preliminary step for the development of a robust model of crop contamination after an HTO accidental release. Copyright © 2012 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Gutowski, H.; Bracha, M.

    1987-01-01

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

  19. Tritium releases, birth defects and infant deaths

    International Nuclear Information System (INIS)

    1991-01-01

    The AECB has published a report 'Tritium releases from the Pickering Nuclear Generating Station and Birth Defects and Infant Mortality in Nearby Communities 1971-1988' (report number INFO-0401). This presents the results of a detailed analysis of deaths and birth defects occurring in infants born to mothers living in the area (25 Km radius) of the Pickering nuclear power plant, over an 18-year period. The analysis looked at the frequency of these defects and deaths in comparison to the general rate for Ontario, and also in relation to airborne and waterborne releases of tritium from the power plant. The overall conclusion was that the rates of infant death and birth defects were generally not higher in the study population than in all of Ontario. There was no prevalent relationship between these deaths and defects and tritium releases measured either at the power plant or by ground monitoring stations t some distance from the facility

  20. Effluents and releases of tritium from Novo-Voronezh-5 reactor

    International Nuclear Information System (INIS)

    Babenko, A.G.; Mekhedov, B.N.; Podporinova, L.E.; Popov, S.V.; Shalin, A.N.

    1990-01-01

    Results of systematic measurements of tritium concentration within technological systems of reactor of Novo-Voronezh NPP conducted to evaluate tritium effluents and releases and radiation doses to population from these effluents and releases are given. It is shown that 68% concerning tritium total amount were disposed into sewerage while 17% - through vent tube and 15% - with water and steam from secondary circuit systems. Standartized tritium effluents from WWER-1000 reactor for 5 year run constitute 15±1.9 GBq/MWxyear and it corresponds to mean value of effluents for foreign NPPs. Tritium concentration in the atmosphere constituted according to calculations (4.1-20)x10 -5 Bq/l. Conclusion is made about insignificant dose to population from tritium gaseous effluents. Detail study is necessary for dose connected with tritium contained in water effluents

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

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

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

  2. The modelling of tritium behaviour in the environment

    International Nuclear Information System (INIS)

    Raskob, W.

    1991-01-01

    In view of the operation of reprocessing plants and fusion reactors the release of tritium may play a dominant role during normal operation as well as after accidents. Tritium, however, is chemically identical to hydrogen and thus interacts directly with water and organic substances, making processes like conversion of HT to HTO, re-emission after deposition, and the conversion of HTO into organically bound tritium (OBT) relevant, all of which may modify the total balance of the available HT or HTO inventories. Because of these physical and chemical properties which differ significantly from those of other radionuclides, the model UFOTRI for assessing the radiological consequences of accidental tritium releases has been developed. It describes the behaviour of tritium in the biosphere and calculates the radiological impact on individuals and the population due to inhalation and skin absorption and by ingestion pathways. The significance of the re-emission process in dose assessments - especially for HT-releases - has been clearly demonstrated in example calculations and applications of UFOTRI. The results of a comparison between an HT-release experiment in Canada 1987 and calculations of UFOTRI can be taken as a first validation of the model

  3. Development of a code to simulate dispersion of atmospheric released tritium gas in the environmental media and to evaluate doses. TRIDOSE

    International Nuclear Information System (INIS)

    Murata, Mikio; Noguchi, Hiroshi; Yokoyama, Sumi

    2000-11-01

    A computer code (TRIDOSE) was developed to assess the environmental impact of atmospheric released tritium gas (T 2 ) from nuclear fusion related facilities. The TRIDOSE simulates dispersion of T 2 and resultant HTO in the atmosphere, land, plant, water and foods in the environment, and evaluates contamination concentrations in the media and exposure doses. A part of the mathematical models in TRIDOSE were verified by comparison of the calculation with the results of the short range (400 m) dispersion experiment of HT gas performed in Canada postulating a short-time (30 minutes) accidental release. (author)

  4. Radionuclide dispersion and hydrodynamics of Ilha Grande Bay (Angra dos Reis, RJ) simulated from hypothetical accidental releases of liquid wastes

    International Nuclear Information System (INIS)

    Simoes Filho, Fernando Lamego; Lapa, Celso Marcelo Franklin; Aguiar, Andre Silva; Soares, Abner Duarte

    2011-01-01

    This study has the aim to assess the impact of accidental release of radionuclides postulate from a nuclear power reactor through environmental modeling of aquatic resources. In order to do that it was used computational models of hydrodynamic circulation and transport for the simulation of tritium dispersion caused by an accidental release in Ilha Grande Bay from the site of the future third plant in two circulation scenarios. The main difference between the scenarios is based on the enhancement of dilution of the highest concentrations in the last one. This dilution enhancement resulting in decreasing concentrations was observed only during the first two weeks, when they ranged from 1 x 10 9 to 5 x 10 5 Bq/m³ close to the Itaorna beach spreading just to Sandri Island. After 180 days, the plume could not be detected anymore in the bay, because their activities would be lower than the minimum detectable value (< 11 kBq/m³). (author)

  5. Effect of surface water on tritium release behavior from Li4SiO4

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  6. Tritium release during inspection of reactor 'RA' at 'Vinca' institute

    International Nuclear Information System (INIS)

    Sipka, V.; Miljevic, N.; Grsic, Z.; Todorovic, D.; Radenkovic, M.

    1997-01-01

    Tritium content in daily precipitation, atmospheric water vapor inside of the reactor hall and around 'Vinca' Institute as well as in soil up to 800 m distance was monitoring during the regular inspection of the fuel channels. Tritium activity in the reactor hall air moisture was in the range from 0.022 to 6.7 MBq/m 3 . Tritium content in soil moisture between 12.7 and 530.9 Bq/l indicate a certain contamination due to tritium release in the environment, depending on the depth and distance from the place of release (author) [sr

  7. Static and mobile networks design for atmospheric accidental releases monitoring

    International Nuclear Information System (INIS)

    Abida, R.

    2010-01-01

    The global context of my PhD thesis work is the optimization of air pollution monitoring networks, but more specifically it concerns the monitoring of accidental releases of radionuclides in air. The optimization problem of air quality measuring networks has been addresses in the literature. However, it has not been addresses in the context of surveillance of accidental atmospheric releases. The first part of my thesis addresses the optimization of a permanent network of monitoring of radioactive aerosols in the air, covering France. The second part concerns the problem of targeting of observations in case of an accidental release of radionuclides from a nuclear plant. (author)

  8. Organically bound tritium (OBT) for various plants in the vicinity of a continuous atmospheric tritium release

    International Nuclear Information System (INIS)

    Vichot, L.; Boyer, C.; Boissieux, T.; Losset, Y.; Pierrat, D.

    2008-01-01

    In order to quantify tritium impact on the environmental, we studied vegetation continuously exposed to a tritiated atmosphere. We chose lichens as bio-indicators, trees for determination of past tritium releases of the Valduc Centre, and lettuce as edible vegetables for dose calculation regarding neighbourhood. The Pasquill and Doury models from the literature were tested to estimate tritium concentration in the air around vegetable for distance from the release point less than 500 m. The results in tree rings show that organically bound tritium (OBT) concentration was strongly correlated with tritium releases. Using the GASCON model, the modelled variation of OBT concentration with distance was correlated with the measurements. Although lichens are recognized as bio-indicators, our experiments show that they were not convenient for environmental surveys because their age is not definitive. Thus, tritium integration time cannot be precisely determined. Furthermore, their biological metabolism is not well known and tritium concentration appears to be largely dependent on species. An average conversion rate of HTO to OBT was determined for lettuce of about 0.20-0.24% h -1 . Nevertheless, even if it is equivalent to values already published in the literature for other vegetation, we have shown that this conversion rate, established by weekly samples, varies by a factor of 10 during the different stages of lettuce development, and that its variation is linked to the biomass derivative

  9. Organically bound tritium (OBT) for various plants in the vicinity of a continuous atmospheric tritium release

    Energy Technology Data Exchange (ETDEWEB)

    Vichot, L. [Commissariat a l' Energie Atomique, CVA/DSTA/SPR/LMSE, 21120 Is-sur-Tille (France)], E-mail: laurent.vichot@cea.fr; Boyer, C.; Boissieux, T.; Losset, Y.; Pierrat, D. [Commissariat a l' Energie Atomique, CVA/DSTA/SPR/LMSE, 21120 Is-sur-Tille (France)

    2008-10-15

    In order to quantify tritium impact on the environmental, we studied vegetation continuously exposed to a tritiated atmosphere. We chose lichens as bio-indicators, trees for determination of past tritium releases of the Valduc Centre, and lettuce as edible vegetables for dose calculation regarding neighbourhood. The Pasquill and Doury models from the literature were tested to estimate tritium concentration in the air around vegetable for distance from the release point less than 500 m. The results in tree rings show that organically bound tritium (OBT) concentration was strongly correlated with tritium releases. Using the GASCON model, the modelled variation of OBT concentration with distance was correlated with the measurements. Although lichens are recognized as bio-indicators, our experiments show that they were not convenient for environmental surveys because their age is not definitive. Thus, tritium integration time cannot be precisely determined. Furthermore, their biological metabolism is not well known and tritium concentration appears to be largely dependent on species. An average conversion rate of HTO to OBT was determined for lettuce of about 0.20-0.24% h{sup -1}. Nevertheless, even if it is equivalent to values already published in the literature for other vegetation, we have shown that this conversion rate, established by weekly samples, varies by a factor of 10 during the different stages of lettuce development, and that its variation is linked to the biomass derivative.

  10. Organically bound tritium (OBT) for various plants in the vicinity of a continuous atmospheric tritium release.

    Science.gov (United States)

    Vichot, L; Boyer, C; Boissieux, T; Losset, Y; Pierrat, D

    2008-10-01

    In order to quantify tritium impact on the environmental, we studied vegetation continuously exposed to a tritiated atmosphere. We chose lichens as bio-indicators, trees for determination of past tritium releases of the Valduc Centre, and lettuce as edible vegetables for dose calculation regarding neighbourhood. The Pasquill and Doury models from the literature were tested to estimate tritium concentration in the air around vegetable for distance from the release point less than 500 m. The results in tree rings show that organically bound tritium (OBT) concentration was strongly correlated with tritium releases. Using the GASCON model, the modelled variation of OBT concentration with distance was correlated with the measurements. Although lichens are recognized as bio-indicators, our experiments show that they were not convenient for environmental surveys because their age is not definitive. Thus, tritium integration time cannot be precisely determined. Furthermore, their biological metabolism is not well known and tritium concentration appears to be largely dependent on species. An average conversion rate of HTO to OBT was determined for lettuce of about 0.20-0.24% h(-1). Nevertheless, even if it is equivalent to values already published in the literature for other vegetation, we have shown that this conversion rate, established by weekly samples, varies by a factor of 10 during the different stages of lettuce development, and that its variation is linked to the biomass derivative.

  11. Environmental health-risk assessment for tritium releases from the National Tritium Labeling Facility (NTLF) at Lawrence Berkeley Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    McKone, T.E.; Brand, K.P.

    1994-12-01

    This report is a health risk assessment that addresses continuous releases of tritium to the environment from the National Tritium Labeling Facility (NTLF) at the Lawrence Berkeley Laboratory (LBL). The NTLF contributes approximately 95% of all tritium releases from LBL. Transport and transformation models were used to determine the movement of tritium releases from the NRLF to the air, surface water, soils, and plants and to determine the subsequent doses to humans. These models were calibrated against environmental measurements of tritium levels in the vicinity of the NTLF and in the surrounding community. Risk levels were determined for human populations in each of these zones. Risk levels to both individuals and populations were calculated. In this report population risks and individual risks were calculated for three types of diseases--cancer, heritable genetic effects, and developmental and reproductive effects.

  12. Environmental health-risk assessment for tritium releases from the National Tritium Labeling Facility (NTLF) at Lawrence Berkeley Laboratory

    International Nuclear Information System (INIS)

    McKone, T.E.; Brand, K.P.

    1994-12-01

    This report is a health risk assessment that addresses continuous releases of tritium to the environment from the National Tritium Labeling Facility (NTLF) at the Lawrence Berkeley Laboratory (LBL). The NTLF contributes approximately 95% of all tritium releases from LBL. Transport and transformation models were used to determine the movement of tritium releases from the NRLF to the air, surface water, soils, and plants and to determine the subsequent doses to humans. These models were calibrated against environmental measurements of tritium levels in the vicinity of the NTLF and in the surrounding community. Risk levels were determined for human populations in each of these zones. Risk levels to both individuals and populations were calculated. In this report population risks and individual risks were calculated for three types of diseases--cancer, heritable genetic effects, and developmental and reproductive effects

  13. Evaluation of tritium release behavior from Li{sub 2}TiO{sub 3} during DT neutron irradiation by use of an improved tritium collection method

    Energy Technology Data Exchange (ETDEWEB)

    Edao, Yuki, E-mail: edao.yuki@jaea.go.jp [Tritium Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kawamura, Yoshinori [Blanket Technology Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Hoshino, Tsuyoshi [Breeding Functional Materials Development Group, Department of Blanket Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Ochiai, Kentaro [BA Project Coordination Group, Department of Fusion Power Systems Research, Rokkasho Fusion Institute, Sector of Fusion Research and Development, Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-15

    Highlights: • Behavior of tritium released from Li{sub 2}TiO{sub 3} under neutron irradiation was measured. • Tritium collection method with hydrophobic catalyst was demonstrated successfully. • Temperature of Li{sub 2}TiO{sub 3} was dominant to control the chemical form of tritium release. - Abstract: The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li{sub 2}TiO{sub 3} during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li{sub 2}TiO{sub 3}. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li{sub 2}TiO{sub 3} as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li{sub 2}TiO{sub 3} by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li{sub 2}TiO{sub 3} irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li{sub 2}TiO{sub 3} was the dominant parameter to control the chemical form of tritium released from the Li{sub 2}TiO{sub 3}.

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  15. Dose Assessment Model for Chronic Atmospheric Releases of Tritium

    International Nuclear Information System (INIS)

    Shen Huifang; Yao Rentai

    2010-01-01

    An improved dose assessment model for chronic atmospheric releases of tritium was proposed. The proposed model explicitly considered two chemical forms of tritium.It was based on conservative assumption of transfer of tritiated water (HTO) from air to concentration of HTO and organic beam tritium (OBT) in vegetable and animal products.The concentration of tritium in plant products was calculated based on considering dividedly leafy plant and not leafy plant, meanwhile the concentration contribution of tritium in the different plants from the tritium in soil was taken into account.Calculating the concentration of HTO in animal products, average water fraction of animal products and the average weighted tritium concentration of ingested water based on the fraction of water supplied by each source were considered,including skin absorption, inhalation, drinking water and food.Calculating the annual doses, the ingestion doses were considered, at the same time the contribution of inhalation and skin absorption to the dose was considered. Concentrations in foodstuffs and dose of annual adult calculated with the specific activity model, NEWTRI model and the model proposed by the paper were compared. The results indicate that the model proposed by the paper can predict accurately tritium doses through the food chain from chronic atmospheric releases. (authors)

  16. 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.

  17. 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.

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

    International Nuclear Information System (INIS)

    Quanci, J.F.

    1989-01-01

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

  19. Development of a code to simulate dispersion of atmospheric released tritium gas in the environmental media and to evaluate doses. TRIDOSE

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Mikio [Nuclear Engineering Co., Ltd., Hitachi, Ibaraki (Japan); Noguchi, Hiroshi; Yokoyama, Sumi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-11-01

    A computer code (TRIDOSE) was developed to assess the environmental impact of atmospheric released tritium gas (T{sub 2}) from nuclear fusion related facilities. The TRIDOSE simulates dispersion of T{sub 2} and resultant HTO in the atmosphere, land, plant, water and foods in the environment, and evaluates contamination concentrations in the media and exposure doses. A part of the mathematical models in TRIDOSE were verified by comparison of the calculation with the results of the short range (400 m) dispersion experiment of HT gas performed in Canada postulating a short-time (30 minutes) accidental release. (author)

  20. Individual dose due to radioactivity accidental release from fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Ni, Muyi, E-mail: muyi.ni@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wei, Shiping [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

    2017-04-05

    Highlights: • Conservative early dose of different unit fusion radioactivity release were assessed. • Data of accident level in INES for fusion reactor were proposed. • Method of environmental restoration time after fusion accident was proposed. • The maximum possible accident level for ITER like fusion reactor is 6. • We need 34–52 years to live after the fusion hypothetical accident. - Abstract: As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1 kg HTO and 1000 kg dust release) and 34–52 years for case 2 (1 kg HTO and 10kg–100 kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation.

  1. Wind tunnel investigations on tritium reemission from soil

    International Nuclear Information System (INIS)

    Taeschner, M.; Bunnenberg, C.

    1993-01-01

    Future fusion plants and tritium handling facilities will contain large amounts of tritium. Following chronical or accidental releases to the atmosphere a secondary HTO source is established in the downwind sector of the tritium release point as a result of deposition processes. To investigate HTO reemission rates, experiments were performed with a special wind tunnel, in which the air flows across the surface of soil columns under controlled conditions. In order to measure the HTO content of an air sample that was experimentally contaminated by reemission of HTO from a labeled soil column, a fast method is used. The air sample is bubbled through a flask filled with a definite volume of low-tritium water. At the end of the sampling period, the volume and the specific activity of the flask water are measured. With the help of a simple mathematical formula, that is presented in this report, the HTO activity of the air sample can be calculated. (orig.) [de

  2. The effect of helium generation and irradiation temperature on tritium release from neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Gorokhov, V.A.; Vlasov, V.V.; Kovalev, A.M.; Chakin, V.P.

    2004-01-01

    The effect of neutron irradiation condition on tritium release from beryllium is described in this paper. Beryllium samples were irradiated in the SM reactor with neutron fluence (E > 0.1 MeV) of (0.37-2.0) x 10 22 cm -2 at 70-100degC and 650-700degC. Mass-spectrometer technique was used in out of tritium release experiments during stepped-temperature anneal within a temperature range from 250 to 1300degC. The total amount of helium accumulated in irradiated beryllium samples varied from 521 appm to 3061 appm. The first signs of tritium release were detected at temperature of 406-553degC. It was shown that irradiation temperature and helium generation level significantly affect the tritium release. A fraction of 44 - 74 % of tritium content in samples irradiated at low temperature (70 - 100degC) is release from beryllium at an annealing temperature below 800degC, whereas for samples after high temperature irradiation (650 - 700 degC) tritium release did not exceed 14 %. Majority of tritium (∼68%) is released within a temperature range from 800 to 920 degC. The increase of helium generation from 521 appm to 3061 appm results in lowering the temperature of maximal tritium release rate and the upper temperature of tritium release from beryllium by 100-130degC and 200-240degC, correspondingly. On the basis of data obtained, the diffusion coefficients of tritium in beryllium were calculated. (author)

  3. Environmental transport and long-term exposure for tritium released in the biosphere

    International Nuclear Information System (INIS)

    Bergman, R.; Bergstroem, U.; Evans, S.

    1979-01-01

    Global cycling of tritium is studied with regard to long-term exposure and dose. Dose and dose commitment are calculated for releases at different latitudes to the troposphere, land and upper ocean layer, with particular regard to effects from release into recipients of intermediate size as, for example, the Baltic Sea. The global transport of tritium appears to be governed by first order kinetics. Compartment models based on linear differential equation systems, as used in this study, should therefore be adequate. The realism and applicability of ecological compartment models are analysed with respect to completeness of the systems of reservoirs and pathways as well as accuracy in assumed reservoir sizes and exchange rates. By introducing different biospheric reservoirs and transfer mechanisms, important carriers and recipients are identified for the analysis of tritium released to air, land and water. Terrestrial biota and groundwater are shown to be significant both with regard to reservoir sizes and influence on the land-troposphere and land-sea exchange of tritium. Model studies regarding the conversion of HT to HTO in different biospheric reservoirs indicate that an atmospheric release of HT may yield up to 1.7 times the dose commitment obtained after release of the same amount of tritium as HTO. The global collective dose commitment from a tropospheric release of tritium is 0.002-0.004 man.rem per Ci depending on the latitude at the release point. Release to the surface ocean layers gives a ten times lower collective dose. (author)

  4. Influence of radiation defects on tritium release parameters from Li2O

    International Nuclear Information System (INIS)

    Grishmanov, V.; Tanaka, S.; Yoneoka, T.

    1998-01-01

    The study of the influence of radiation defects on tritium release behavior from polycrystalline Li 2 O was performed by simultaneous measurements of the luminescence emission and tritium release. It was found that the radiation defects in Li 2 O introduced by electron irradiation cause the retention of tritium. It is thought that the tritium recovery is affected by the formation of a Li-T bond, which is tolerant of high temperatures. The retardation of tritium decreases with increasing absorbed dose in the dose range from 50 to 140 MGy. The aggregation of radiation defects at high irradiation doses is considered to be responsible for the decrease of the interaction of tritium with radiation defects. The mechanism of the interaction of radiation defects with tritium is discussed. (orig.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  7. Tritium recapture behavior at a nuclear power reactor due to airborne releases.

    Science.gov (United States)

    Harris, Jason T; Miller, David W; Foster, Doug W

    2008-08-01

    This paper describes the initiatives taken by Cook Nuclear Plant to study the on-site behavior of recaptured tritium released in its airborne effluents. Recapture is the process where a released radioactive effluent, in this case tritium, is brought back on-site through some mechanism. Precipitation, shifts in wind direction, or anthropogenic structures that restrict or alter effluent movement can all lead to recapture. The investigation was started after tritium was detected in the north storm drain outfall. Recent inadvertent tritium releases by several other nuclear power plants, many of which entered the groundwater, have led to increased surveillance and scrutiny by regulatory authorities and the general public. To determine the source of tritium in the outfall, an on-site surface water, well water, rainwater and air-conditioning condensate monitoring program was begun. Washout coefficients were also determined to compare with results reported by other nuclear power plants. Program monitoring revealed detectable tritium concentrations in several precipitation sample locations downwind of the two monitored containment building release vents. Tritium was found in higher concentrations in air-conditioning condensate, with a mean value of 528 Bq L(-1) (14,300 pCi L(-1)). The condensate, and to a lesser extent rainwater, were contributing to the tritium found in the north storm drain outfall. Maximum concentration values for each sample type were used to estimate the most conservative dose. A maximum dose of 1.1 x 10(-10) mSv (1.1 x 10(-8) mrem) total body was calculated to determine the health impact of the tritium detected.

  8. Lithium ceramics: sol-gel preparation and tritium release

    International Nuclear Information System (INIS)

    Renoult, O.

    1994-04-01

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

  9. Correlation of rates of tritium migration through porous concrete

    Energy Technology Data Exchange (ETDEWEB)

    Fukada, S.; Katayama, K.; Takeishi, T. [Kyushu University, Fukuoka (Japan); Edao, Y.; Kawamura, Y.; Hayashi, T.; Yamanishi, T. [JAEA-TPL, Muramatsu, Tokai-mura (Japan)

    2015-03-15

    In a nuclear facility when tritium leaks from a glovebox to room accidentally, an atmosphere detritiation system (ADS) starts operating, and HTO released is recovered by ADS. ADS starts when tritium activity in air becomes higher than its controlled level. Before ADS operates, the laboratory walls are the final enclosure facing tritium and are usually made of porous concrete coated with a hydrophobic paint. In the present study, previous data on the diffusivity and adsorption coefficient of concrete and paints are reviewed. Tritium penetrates and migrates into concrete by following 3 ways. First, gaseous HT or T{sub 2} easily penetrates into porous concrete. Its diffusivity is almost equal to that of H{sub 2}. When a gaseous molecule diffuses through pores with a smaller diameter than a mean free path, its migration rate is described by the Knudsen diffusion formula. The second mechanism is H{sub 2}O vapor diffusion in pores. Concrete holds a lot of structural water. Therefore, H{sub 2}O or HTO vapor can diffuse inside concrete pores along with adsorption-desorption and isotopic exchange with structural water, which is the third mechanism. Literature shows that the diffusivity of HTO through the epoxy-resin paint is determined as D(HTO)=1.0*10{sup -16} m{sup 2}/s. We have used this data to set a model and we have applied it to estimate residual tritium in laboratory walls. We have considered 2 accidental cases and a normal case: first, ADS starts operating 1 hour after 100 Ci HTO is released in the room, secondly, ADS starts 24 hours after 100 Ci HTO release and thirdly, when the walls are exposed to HTO for 10 years of normal operation. It appears that the immediate start up of ADS is indispensable for safety.

  10. Preliminary analysis of the safety and environmental impact of the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

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

    1980-01-01

    The Tritium Systems Test Assembly (TSTA) is a facility dedicated to the development of technologies associated with the D-T fuel cycle of future fusion reactors while demonstrating that TSTA can be operated safely with no significant losses to the environment. During the initial design stage of TSTA, a safety analysis was performed which investigated the effects of major subsystem component failure, the meteorology and seismicity of the site and their possible effect on the facility, and accident scenarios which result in tritium releases. Major releases of tritium to the environment are considered highly improbable since they require a compound failure of primary and secondary containment, along with either a breach of the building or a failure of the Emergency Tritium Cleanup system. Accidental releases caused by natural phenomena (earthquake, tornado, etc.) are considered highly improbable

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  12. Development, description and validation of a Tritium Environmental Release Model (TERM).

    Science.gov (United States)

    Jeffers, Rebecca S; Parker, Geoffrey T

    2014-01-01

    Tritium is a radioisotope of hydrogen that exists naturally in the environment and may also be released through anthropogenic activities. It bonds readily with hydrogen and oxygen atoms to form tritiated water, which then cycles through the hydrosphere. This paper seeks to model the migration of tritiated species throughout the environment - including atmospheric, river and coastal systems - more comprehensively and more consistently across release scenarios than is currently in the literature. A review of the features and underlying conceptual models of some existing tritium release models was conducted, and an underlying aggregated conceptual process model defined, which is presented. The new model, dubbed 'Tritium Environmental Release Model' (TERM), was then tested against multiple validation sets from literature, including experimental data and reference tests for tritium models. TERM has been shown to be capable of providing reasonable results which are broadly comparable with atmospheric HTO release models from the literature, spanning both continuous and discrete release conditions. TERM also performed well when compared with atmospheric data. TERM is believed to be a useful tool for examining discrete and continuous atmospheric releases or combinations thereof. TERM also includes further capabilities (e.g. river and coastal release scenarios) that may be applicable to certain scenarios that atmospheric models alone may not handle well. Copyright © 2013 Elsevier Ltd. All rights reserved.

  13. Tritium release kinetics of Li{sub 2}O with radiation defects

    Energy Technology Data Exchange (ETDEWEB)

    Grishmanov, V; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1998-03-01

    The study of an influence of radiation defects on tritium release behavior from polycrystalline Li{sub 2}O was performed by the in-pile and out-of-pile tritium release experiments. The samples were pre-irradiated by accelerated electrons to various absorbed doses up to 140 MGy and then exposed to the fluence of 10{sup 17} thermal neutrons/m{sup 2}. The radiation defects introduced by electron irradiation in Li{sub 2}O cause the retention of tritium. The linear temperature increase of the electron-irradiated samples disclosed two tritium release peaks: first starts at {approx}600 K with the maximum at {approx}800 K and second appears at {approx}950 K with the maximum at {approx}1200 K. It is thought that the tritium release at high temperatures (> 950 K) is due to the thermal decomposition of LiT. In order to further investigated the formation of lithium hydrides, the diffuse-reflectance Fourier transform infrared (FTIR) absorption spectroscopy was applied. The Li{sub 2}O powder was irradiated by electron accelerator under D{sub 2} containing atmosphere (N{sub 2} + 10% D{sub 2}). An absorption band specific to the Li{sub 2}O was observed at 668 cm{sup -1} and attributed to the Li-D stretching vibration. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  15. Transfer of Tritium in the Environment after Accidental Releases from Nuclear Facilities. Report of Working Group 7 Tritium Accidents of EMRAS II Topical Heading Approaches for Assessing Emergency Situations. Environmental Modelling for Radiation Safety (Emras II) Programme

    International Nuclear Information System (INIS)

    2014-07-01

    Environmental assessment models are used for evaluating the radiological impact of actual and potential releases of radionuclides to the environment. They are essential tools for use in the regulatory control of routine discharges to the environment and also in planning measures to be taken in the event of accidental releases. They are also used for predicting the impact of releases which may occur far into the future, for example, from underground radioactive waste repositories. It is important to verify, to the extent possible, the reliability of the predictions of such models by a comparison with measured values in the environment or with predictions of other models. The IAEA has been organizing programmes of international model testing since the 1980s. These programmes have contributed to a general improvement in models, in the transfer of data and in the capabilities of modellers in Member States. IAEA publications on this subject over the past three decades demonstrate the comprehensive nature of the programmes and record the associated advances which have been made. From 2009 to 2011, the IAEA organized a programme entitled Environmental Modelling for RAdiation Safety (EMRAS II), which concentrated on the improvement of environmental transfer models and the development of reference approaches to estimate the radiological impacts on humans, as well as on flora and fauna, arising from radionuclides in the environment. Different aspects were addressed by nine working groups covering three themes: reference approaches for human dose assessment, reference approaches for biota dose assessment and approaches for assessing emergency situations. This publication describes the work of the Tritium Accidents Working Group

  16. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    International Nuclear Information System (INIS)

    Xiao, Chengjian; Gao, Xiaoling; Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke; Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming; Wang, Xiaolin; Oya, Yasuhisa; Okuno, Kenji

    2013-01-01

    Tritium release kinetics in lithium orthosilicate (Li 4 SiO 4 ) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li 4 SiO 4 pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10 −7.0 exp (−40.3 × 10 3 /RT) cm 2 s −1

  17. Tritium Research Laboratory safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Wright, D.A.

    1979-03-01

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment. (ERB)

  18. Tritium Research Laboratory safety analysis report

    International Nuclear Information System (INIS)

    Wright, D.A.

    1979-03-01

    Design and operational philosophy has been evolved to keep radiation exposures to personnel and radiation releases to the environment as low as reasonably achievable. Each experiment will be doubly contained in a glove box and will be limited to 10 grams of tritium gas. Specially designed solid-hydride storage beds may be used to store temporarily up to 25 grams of tritium in the form of tritides. To evaluate possible risks to the public or the environment, a review of the Sandia Laboratories Livermore (SLL) site was carried out. Considered were location, population, land use, meteorology, hydrology, geology, and seismology. The risks and the extent of damage to the TRL and vital systems were evaluated for flooding, lightning, severe winds, earthquakes, explosions, and fires. All of the natural phenomena and human error accidents were considered credible, although the extent of potential damage varied. However, rather than address the myriad of specific individual consequences of each accident scenario, a worst-case tritium release caused indirectly by an unspecified natural phenomenon or human error was evaluated. The maximum credible radiological accident is postulated to result from the release of the maximum quantity of gas from one experiment. Thus 10 grams of tritium gas was used in the analysis to conservatively estimate the maximum whole-body dose of 1 rem at the site boundary and a maximum population dose of 600 man-rem. Accidental release of this amount of tritium implies simultaneous failure of two doubly contained systems, an occurrence considered not credible. Nuclear criticality is impossible in this facility. Based upon the analyses performed for this report, we conclude that the Tritium Research Laboratory can be operated without undue risk to employees, the general public, or the environment

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  20. Environmental effects of normal and off-normal releases of tritium from CTR systems

    International Nuclear Information System (INIS)

    McKone, T.E.

    1978-08-01

    Near term fusion technology will utilize the deuterium-tritium reaction. To quantify the magnitude of the hazard presented by major tritium release mechanisms, a method is presented for determining doses to the public from releases of tritium as tritiated water vapor or tritiated lithium compounds. Inclusion of this method in a computer model is described. This model uses the Gaussian dispersion method to predict distribution of tritium species in the downwind environment. Movement of tritium into biological systems is determined by treating these systems as a series of interacting water compartments. Dispersion and uptake calculations are applied to two sample sites in order to predict health effects. These effects are compared to the long range effect of introducing tritium into the world water system

  1. Is Tritium an Issue for High Temperature Reactors?

    International Nuclear Information System (INIS)

    Fütterer, Michael A.; D'Agata, Elio; Raepsaet, Xavier

    2014-01-01

    In a High Temperature Reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy. (author)

  2. Is tritium an issue for high temperature reactors?

    Energy Technology Data Exchange (ETDEWEB)

    Fütterer, Michael A., E-mail: michael.fuetterer@ec.europa.eu [European Commission – Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); D’Agata, Elio [European Commission – Joint Research Centre, Institute for Energy and Transport, P.O. Box 2, 1755 ZG Petten (Netherlands); Raepsaet, Xavier [Commissariat à l’Energie Atomique et aux Energies Alternatives, DEN/DM2S, 91191 Gif-sur-Yvette Cedex (France)

    2016-09-15

    In a high temperature reactor, tritium is produced by a number of mechanisms. Due to its high mobility, some of this tritium ends up in the primary helium cooling circuit from where it can be extracted by the coolant purification system to keep the partial pressure of tritiated compounds low. The remaining partial pressure of tritium in the coolant is the driving force for permeation across the heat exchanger from the primary cooling system into the secondary cooling system. From there the contamination may further propagate and ultimately escape into the environment. This paper summarizes a study on the different tritium control options capable of meeting possible future safety requirements. Our results indicate that compliance with plausible tritium control requirements can indeed be achieved with reasonable effort both for electricity generation using a closed steam cycle and for process steam generation with an open steam cycle. However, for new-build HTR, definite country-specific licensing requirements (e.g. chronic and accidental tritium release) are yet to be determined and will shape the required tritium control strategy.

  3. Tritium release kinetics in lithium orthosilicate ceramic pebbles irradiated with low thermal-neutron fluence

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Chengjian; Gao, Xiaoling [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Kobayashi, Makoto; Kawasaki, Kiyotaka; Uchimura, Hiromichi; Toda, Kensuke [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Kang, Chunmei; Chen, Xiaojun; Wang, Heyi; Peng, Shuming [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Box 919-214, Mian Yang 621900 (China); Wang, Xiaolin, E-mail: xlwang@caep.ac.cn [China Academy of Engineering Physics, Box 919-1, Mian Yang 621900 (China); Oya, Yasuhisa; Okuno, Kenji [Radiochemistry Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Shizuoka 422-8529 (Japan)

    2013-07-15

    Tritium release kinetics in lithium orthosilicate (Li{sub 4}SiO{sub 4}) ceramic pebbles irradiated with low thermal-neutron fluence was studied by out-of-pile annealing experiments. It was found that the tritium produced in Li{sub 4}SiO{sub 4} pebbles was mainly released as tritiated water vapor (HTO). The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from 450 K to 600 K were obtained by isothermal annealing tests, and the Arrhenius relation was determined to be D = 1 × 10{sup −7.0} exp (−40.3 × 10{sup 3}/RT) cm{sup 2} s{sup −1}.

  4. 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

  5. Tritium release from Li{sub 4}SiO{sub 4} ceramic pebbles in high magnetic field

    Energy Technology Data Exchange (ETDEWEB)

    Ran, Guangming [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Xiao, Chengjian; Chen, Xiaojun; Gong, Yu; Zhao, Linjie [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621900 (China); Wang, Xiaolin [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); China Academy of Engineering Physics, Mianyang 621900 (China)

    2015-12-15

    Highlights: • Behavior of tritium release from Li{sub 4}SiO{sub 4} pebbles in MF up to 7 T was investigated. • The MF effect on tritium release is not significant according to the TPD results. • Tritium release from the BIG sample is much easier than that from the SMALL sample. • The MF effect on tritium diffusion was probably weakened by surface desorption. - Abstract: The behavior of tritium release from Li{sub 4}SiO{sub 4} ceramic pebbles in high magnetic field (MF) was investigated by temperature programmed desorption (TPD). Two batches of Li{sub 4}SiO{sub 4} pebbles produced by wet method were used as the experimental samples, one batch with an average pebble diameter of 0.8 mm (the SMALL samples), and the other 1.2 mm (the BIG samples). A superconducting magnet was applied to generate MF up to 7 T in the sample area during annealing. For both batches of samples, the tritium release curves within and without MF showed very similar characteristics, indicating that the effect of high MF on tritium release behavior is not significant. The tritium release peaks for the BIG samples were observed at much lower temperatures than that for the SMALL samples, even though the grain sizes of the BIG samples are much bigger than that of the SMALL samples. It is considered that surface desorption process dominates the overall tritium release behavior in this work, which probably weakened the MF effect.

  6. Tritium release from neutron-irradiated Li2O: Transport in porous sintered pellets

    International Nuclear Information System (INIS)

    Tanifuji, Takaaki; Yamaki, Daiju; Jitsukawa, Shiro

    2006-01-01

    The tritium release behavior from Li 2 O sintered pellets (81-88% T.D.) is examined by isothermal heating tests. (1) For the 88% T.D. specimens, the fraction of residual tritium is found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (2) For the 81% T.D. specimens, which are annealed after irradiation at 630 K for 4 h, the fraction of residual tritium is also found to follow the square-root law of the annealing time. The rate-determining process is the migration in the connected micro-pore. (3) For the 81% T.D. specimens as irradiated, the tritium release rate is found to follow the square-root law of the annealing time. The rate-determining process is controlled by Kohlrauch stretched exponential form. Tritium trapped in irradiation defects released with recovering the defects by isothermal heating

  7. Radioecological behaviour of elementary tritium, especially dry deposition and its dependence on soil porosity

    International Nuclear Information System (INIS)

    Foerstel, H.

    1997-01-01

    The inventory of fusion reactors mainly consists of deuterium and tritium. The amount of tritium of each reactor is equal to the natural inventory of the earth's atmo- and hydrosphere. Elementary tritium (HT) itself is not dangerous to man, for it is hardly dissolved in water, that is neither taken up by human tissues nor metabolized anywhere in our body. In contrast to HT the tritiated water HTO quickly exchanges with any wet surface and with the humidity of air. After an accidental release into the atmosphere the main pathway of intake into the human body is as HTO via the lung; its surface is comparable to a soccer playground. HT released into air will be quickly oxidised within the upper centimetres of the soil when the plume touches the ground. Each soil tested by us until now had oxidized HT, that had shown hydrogenase activity. Neither the biological function nor the catalytic system wee identified yet. The hypothesis of a correlation between hydrogenase activity and soil nitrogen fixation could not be confirmed (nitrogen fixation shows a leakage of hydrogen): nitrogen fixing plants (nodules) do not oxidize HT. The presentation will summarize ten years of work in the laboratory and in the field. A concise picture of the radioecological behaviour of elementary tritium after an accidental release could be obtained. The work was partly done as cooperation within the frame of the EU or within the International Union of Radioecology

  8. Measured tritium in groundwater related to atmospheric releases from the Marcoule nuclear site

    International Nuclear Information System (INIS)

    Levy, F.; Clech, A.; Crochet, P.

    1996-01-01

    Tritium is released into the atmosphere during normal operation from the industrial facilities operated by COGEMA at Marcoule; over a 1 5-year period covered by this study (1979-1994) the quantities ranged from 4940 to 520 TBq·yr -1 . Atmospheric release in rainy weather results in tritium migration into the ground water by a series of mechanisms associated with the water cycle. COGEMA monitors the ground water by means of bore holes. Atmospheric monitoring is also routinely performed; data on the tritium activity concentration in the air and rainwater are available for the same time period. A simplified observation suggests a relation between the atmospheric tritium release and the ground water radioactivity. In 1994, the activity ranged from 100 to 200 Bq·l -1 in the boreholes located 1 km and 2 km downwind from the point of release, diminishing with the distance to less than 20 Bq·l -1 at about 3 km. The authors attempted to model two types of transfers: atmospheric transfer from the release chimney to the borehole, and transfer in the alluvial ground water. The aquifer comprises the alluvial deposits forming the Codolet plain extending to the south of Marcoule, downwind from the point of atmospheric tritium release. The hydrogeology of the entire Marcoule site has been described in previous studies by the French bureau of geological and mineralogical research (BRGM) and ANTEA. (author)

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  10. Environmental effects of a tritium gas release from the Savannah River Plant on December 31, 1975

    International Nuclear Information System (INIS)

    Jacobsen, W.R.

    1976-03-01

    At 10:00 p.m. EST on December 31, 1975, 182,000 Ci of tritium gas was released within about 1.5 min from a tritium processing facility at the Savannah River Plant. The release was caused by the failure of a vacuum gage and was exhausted to the atmosphere by way of a 200-ft-high stack. Winds averaging 20 mph carried the tritium offplant toward the east. Calculations indicate that the puff passed out to sea about 35 miles north of Charleston, South Carolina, about 7 hr after the release occurred. Samples from the facility exhaust system indicated that 99.4 percent of the tritium was in elemental form and 0.6 percent was in the more biologically active oxide (water) form. The maximum potential dose to a person (from inhalation and skin absorption) at the puff centerline on the plant boundary was calculated to be 0.014 mrem, or about 0.01 percent of the annual dose received from natural radioactivity. The integrated dose to the population under the release path was calculated to be 0.2 man-rem before the tritium passed out to sea. Over 300 environmental samples were collected and analyzed following the release. These samples included air moisture, atmospheric hydrogen, vegetation, soil, surface water, milk, and human urine. Positive results were obtained in some onplant and plant perimeter samples; these results aided in confirming the close-in puff trajectory. Tritium concentrations in nearly all samples taken beyond the plant perimeter fell within normal ranges; no urine samples indicated any tritium uptakes as a result of the release. Two milk samples did indicate a measurable tritium uptake; the maximum potential dose to an individual drinking this milk was calculated to be about 0.1 mrem. Because calculated doses from assumed exposure to the tritium are low and analyses of environmental samples indicated no significant accumulation of tritium, it is concluded that no significant environmental effects resulted from the December 31, 1975, tritium release

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

    Energy Technology Data Exchange (ETDEWEB)

    Renoult, O

    1994-04-01

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

  12. Radioactive releases into the environment under accidental conditions

    International Nuclear Information System (INIS)

    Beninson, D.

    1976-01-01

    Although accidents involving the release of radioactive materials and the unplanned exposure of people can occur at any stage of the nuclear fuel cycle, most attention has been focused on reactor accidents. Although no power reactor accidents involving exposure of the public have yet occured, it should be recognized that the probability of such accidental releases cannot be reduced to zero. Since the inventory of radioactive materials in power reactors is very large, it is usual to postulate, for safety assessments, that a release of fission products takes place in spite of all protective measures. This postulated release is of importance for reactor siting and for preparing emergency plans. (HP) [de

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

    International Nuclear Information System (INIS)

    Kosaku, Yasuo; Kuroda, Toshimasa; Enoeda, Mikio; Hatano, Toshihisa; Sato, Satoshi; Miki, Nobuharu; Akiba, Masato

    2003-06-01

    The design of the breeding blanket in ITER applies pebble bed breeder in tube (BIT) surrounded by multiplier pebble bed. It is assumed to use the same module support mechanism and coolant manifolds and coolant system as the shielding blankets. This work focuses on the verification of the design of the breeding blanket, from the viewpoints which is especially unique to the pebble bed type breeding blanket, such as, tritium breeding performance, tritium inventory and release behavior and thermo-mechanical performance of the ITER breeding blanket. With respect to the neutronics analysis, the detailed analyses of the distribution of the nuclear heating rate and TBR have been performed in 2D model using MCNP to clarify the input data for the tritium inventory and release rate analyses and thermo-mechanical analyses. With respect to the tritium inventory and release behavior analysis, the parametric analyses for selection of purge gas flow rate were carried out from the view point of pressure drop and the tritium inventory/release performance for Li 2 TiO 3 breeder. The analysis result concluded that purge gas flow rate can be set to conventional flow rate setting (88 l/min per module) to 1/10 of that to save the purge gas flow and minimize the size of purge gas pipe. However, it is necessary to note that more tritium is transformed to HTO (chemical form of water) in case of Li 2 TiO 3 compared to other breeder materials. With respect to the thermo-mechanical analyses of the pebble bed blanket structure, the analyses have been performed by ABAQUS with 2D model derived from one of eight facets of a blanket module, based on the reference design. Analyses were performed to identify the temperature distribution incorporating the pebble bed mechanical simulation and influence of mechanical behavior to the thermal behavior. The result showed that the maximum temperature in the breeding material was 617degC in the first row of breeding rods and the minimum temperature was 328

  14. Accelerator Production of Tritium Programmatic Environmental Impact Statement Input Submittal

    International Nuclear Information System (INIS)

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

    1996-02-01

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

  15. Tritium cycling in a tree spiked with tritiated water

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.; Luvall, J.C.

    1979-01-01

    Transfer and turnover rates in forests are important to compute the residence time of tritiated water in an area following an accidental release. In this study tritium was injected in the base of 7 year old, loblolly pine (Pinus taeda, L) trees to determine the rate of transfer through the trees and the turnover in the trees independent of the soil. The results indicate the flow rates depend on the rate of water movement through the tree, which is influenced by the microclimate, and exchange of tritium with hydrogen exchange sites in the tree. The initial pulse of tritium appears to move through the tree in about four days. The descending portion of the curve can be described as a two compartment model with half-lives of 1.41 and 21.7 days. There is some evidence that a longer turnover compartment is associated with metabolically fixed tritium

  16. Atmospheric Dispersion Assessment for Potential Accidental Releases at Yonggwang Nuclear Power Plants

    International Nuclear Information System (INIS)

    Na, Man Gyun; Sim, Young Rok; Jung, Chul Kee; Lee, Goung Jin; Kim, Soong Pyung; Chung, Sung Tai

    2000-01-01

    XOQ DW code is currently used to assess the atmospheric dispersion for the routine releases of radioactive gaseous effluents at Yonggwang nuclear power plants. This code was developed based on XOQDOQ code and an additional code is required to assess the atmospheric dispersion for potential accidental releases. In order to assess the atmospheric dispersion for the accidental releases, XOQAR code has been developed by using PAVAN code that is based on Reg. Guide 1.145. The terrain data of XOQ DW code inputs and the relative concentrations (X/Q) of XOQ DW code outputs are used as the inputs of the XOQAR code through the interface with XOQ DW code. By using this code, the maximum values of X/Q at exclusion area and low population zone boundaries except for sea areas were assessed as 1.33 x 10 -4 and 7.66 x 10 -6 sec/m 3 , respectively. Through the development of this code, a code system is prepared for assessing the atmospheric dispersion for the accidental releases as well as the routine releases. This developed code can be used for other domestic nuclear power plants by modifying the terrain input data

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  18. Assessment of the dose to a representative Japanese due to stationary release of tritium to the environment

    International Nuclear Information System (INIS)

    Saito, Masahiro

    2005-01-01

    The computer program TriStat was applied to estimate the dose to a representative Japanese due to a stationary release of tritium as HTO and/or HT to the atmosphere from nuclear facilities. In TriStat, the air tritium concentration is estimated by a Gaussian dispersion model. The tritium deposition to the soil was assumed to occur both by dry and wet deposition processes of atmospheric tritium. The primary process of tritium transfer to human body is assumed to take place through a local food-chain in the contaminated area. Tritium concentrations in soil, vegetables and forage were estimated as the tritium concentration per water equivalent. The food chain was modeled by assuming a vegetable compartment and an animal-food compartment. By using TriStat the annual dose to the representative Japanese was evaluated for stationary release of tritium as a function of the distance from a release point. The dose contribution from drinking water was neglected, since the drinking water is generally supplied as tap water or as commercial bottled water. In the case of HT release, the committed dose due to tritium intake through breathing and skin absorption was found to be of minor importance. (author)

  19. Computer code to assess accidental pollutant releases

    International Nuclear Information System (INIS)

    Pendergast, M.M.; Huang, J.C.

    1980-07-01

    A computer code was developed to calculate the cumulative frequency distributions of relative concentrations of an air pollutant following an accidental release from a stack or from a building penetration such as a vent. The calculations of relative concentration are based on the Gaussian plume equations. The meteorological data used for the calculation are in the form of joint frequency distributions of wind and atmospheric stability

  20. Effect of electromagnetic radiation on the release of tritium from a fusion reactor

    International Nuclear Information System (INIS)

    Causey, R.A.

    1982-01-01

    Electromagnetic radiation may play an important role in determining the amount of tritium that passes through the first wall of a fusion reactor. Photons enhance the outgassing of hydrogen from metal surfaces and therefore lower the surface concentration. The diffusion coefficient for hydrogen in metals is also increased by photon interactions. The importance of these processes on the tritium release depends on the actual conditions that will exist in the fusion reactor. Because electromagnetic radiation stimulates the release of tritium from traps, it could also affect the tritium inventory in the first wall. The effects of electromagnetic radiation on hydrogen in metals have been reviewed. Because electrons are produced by high energy photons, the effect of electrons on hydrogen has also been included

  1. Factors affecting calculations of dose resulting from a tritium release into the atmosphere

    International Nuclear Information System (INIS)

    Otaduy, P.; Easterly, C.E.; Booth, R.S.; Jacobs, D.G.

    1976-01-01

    Tritium releases in the form of HT represent a lower hazard to man than releases as HTO. However, during movement in the environment, HT is converted into HTO. The effects of the conversion rate on calcultions of dose are described, and a general method is presented for determining the dose from tritium for various conversion rates and relative HTO/HT risk factors

  2. Computer program for assessing the human dose due to stationary release of tritium

    International Nuclear Information System (INIS)

    Saito, Masahiro; Raskob, Wolfgang

    2003-01-01

    The computer program TriStat (Tritium dose assessment for stationary release) has been developed to assess the dose to humans assuming a stationary release of tritium as HTO and/or HT from nuclear facilities. A Gaussian dispersion model describes the behavior of HT gas and HTO vapor in the atmosphere. Tritium concentrations in soil, vegetables and forage were estimated on the basis of specific tritium concentrations in the free water component and the organic component. The uptake of contamination via food by humans was modeled by assuming a forage compartment, a vegetable component, and an animal compartment. A standardized vegetable and a standardized animal with the relative content of major nutrients, i.e. proteins, lipids and carbohydrates, representing a standard Japanese diet, were included. A standardized forage was defined in a similar manner by using the forage composition for typical farm animals. These standard feed- and foodstuffs are useful to simplify the tritium dosimetry and the food chain related to the tritium transfer to the human body. (author)

  3. Fabrication and tritium release property of Li2TiO3-Li4SiO4 biphasic ceramics

    Science.gov (United States)

    Yang, Mao; Ran, Guangming; Wang, Hailiang; Dang, Chen; Huang, Zhangyi; Chen, Xiaojun; Lu, Tiecheng; Xiao, Chengjian

    2018-05-01

    Li2TiO3-Li4SiO4 biphasic ceramic pebbles have been developed as an advanced tritium breeder due to the potential to combine the advantages of both Li2TiO3 and Li4SiO4. Wet method was developed for the pebble fabrication and Li2TiO3-Li4SiO4 biphasic ceramic pebbles were successfully prepared by wet method using the powders synthesized by hydrothermal method. The tritium release properties of the Li2TiO3-Li4SiO4 biphasic ceramic pebbles were evaluated. The biphasic pebbles exhibited good tritium release property at low temperatures and the tritium release temperature was around 470 °C. Because of the isotope exchange reaction between H2 and tritium, the addition of 0.1%H2 to purge gas He could significantly enhance the tritium gas release and the fraction of molecular form of tritium increased from 28% to 55%. The results indicate that the Li2TiO3-Li4SiO4 biphasic ceramic pebbles fabricated by wet method exhibit good tritium release property and hold promising potential as advanced breeder pebbles.

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. Kinetics that govern the release of tritium from neutron-irradiated lithium oxide

    International Nuclear Information System (INIS)

    Bertone, P.C.

    1986-01-01

    The Lithium Blanket Module (LBM) program being conducted at the Princeton Plasma Physics Laboratory requires that tritium concentrations as low as 0.1 nCi/g, bred in both LBM lithium oxide pellets and gram-size lithium samples, be measured with an uncertainty not exceeding +/-6%. This thesis reports two satisfactory methods of assaying LBM pellets and one satisfactory method of assaying lithium samples. Results of a fundamental kinetic investigation are also reported. The thermally driven release of tritium from neutron-irradiated lithium oxide pellets is studied between the temperatures of 300 and 400 0 C. The observed release clearly obeys first-order kinetics, and the governing activation energy appears to be 28.4 kcal/mole. Finally, a model is presented that may explain the thermally driven release of tritium from a lithium oxide crystal and assemblies thereof. It predicts that under most circumstances the release is controlled by either the diffusion of a tritiated species through the crystal, or by the desorption of tritiated water from it

  6. Release of tritium from fuel and collection for storage

    International Nuclear Information System (INIS)

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

    1976-04-01

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

  7. Consequences of tritium release to water pathways from postulated accidents in a DOE production reactor

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Olson, R.L.; Hamby, D.M.

    1991-01-01

    A full-scale PRA of a DOE production reactor has been completed that considers full release of tritium as part of the severe accident source term. Two classes of postulated reactor accidents, a loss-of-moderator pumping accident and a loss-of-coolant accident, are used to bound the expected dose consequence from liquid pathway release. Population doses from the radiological release associated with the two accidents are compared for aqueous discharge and atmospheric release modes. The expectation values of the distribution of possible values for the societal effective dose equivalent to the general public, given a tritium release to the atmosphere, is 2.8 person-Sv/PBq (9.9 x 10 -3 person-rem/Ci). The general public drinking water dose to downstream water consumers is 6.5 x 10 -2 person-Sv/Pbq (2.4 x 10 -4 person-rem/Ci) for aqueous releases to the surface streams eventually reaching the Savannah River. Negligible doses are calculated for freshwater fish and saltwater invertebrate consumption, irrigation, and recreational use of the river, given that an aqueous release is assumed to occur. Relative to the balance of fission products released in a hypothetical severe accident, the tritium-related dose is small. This study suggests that application of regional models (1610 km radius) will indicate larger dose consequences from short-term tritium release to the atmosphere than from comparable tritium source terms to water pathways. However, the water pathways assessment is clearly site-specific, and the overall aqueous dose will be dependent on downstream receptor populations and uses of the river

  8. Projected tritium releases from F ampersand H Area Seepage Basins and the Solid Waste Disposal Facilities to Fourmile Branch

    International Nuclear Information System (INIS)

    Looney, B.B.; Haselow, J.S.; Lewis, C.M.; Harris, M.K.; Wyatt, D.E.; Hetrick, C.S.

    1993-01-01

    A large percentage of the radioactivity released to the environment by operations at the Savannah River Site (SRS) is due to tritium. Because of the relative importance of the releases of tritium from SRS facilities through the groundwater to the environment, periodic evaluation and documentation of the facility operational status, proposed corrective actions, and projected changes/reductions in tritium releases are justified. Past, current, and projected tritium releases from the F and H Area Seepage Basins and the Solid Waste Disposal Facilities (SWDF) to Fourmile Branch are described. Each section provides a brief operational history along with the current status and proposed corrective actions. A conceptual model and quantitative estimates of tritium release from the facilities into the groundwater and the environment are developed. Tritium releases from the F and H Area Seepage Basins are declining and will be further reduced by the implementation of a groundwater corrective action required by the Resource Conservation and Recovery Act (RCRA). Tritium releases from the SWDF have been relatively stable over the past 10 years. It is anticipated that SWDF tritium releases to Fourmile Branch will remain approximately at current levels for at least 10--20 years. Specific characterization activities are recommended to allow an improved projection of tritium flux and to assist in developing plans for plume mitigation. SRS and the South Carolina Department of Health and Environmental Control are developing groundwater corrective action plans for the SWDF. Portions of the SWDF are also regulated under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). Reduction of tritium flux is one of the factors considered in the development of the RCRA/CERCLA groundwater corrective action. The final section of the document presents the sum of the projected tritium fluxes from these facilities to Fourmile Branch

  9. The influence of irradiation defects on tritium release from Li{sub 2}O

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Satoru; Grishmanov, V [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    During reactor irradiation of Li{sub 2}O defects are introduced by neutrons, triton and helium ions produced by {sup 6}Li(n, {alpha}){sup 3}H reactions and {gamma}-rays. Simultaneous measurements of luminescence emission and tritium release were performed under various conditions (temperature, sweep gas chemical composition) for Li{sub 2}O single crystal and polycrystal in order to elucidate possible influence of defects on tritium release. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-11-01

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

  11. Issues relating to the siting of tritium-fueled fusion experiments

    International Nuclear Information System (INIS)

    Reilly, H.J.; Holland, D.F.

    1985-01-01

    A preconceptual design study and safety analysis of the Tokamak Fusion Core Experiment (TFCX) was conducted in 1984 for the Department of Energy. This paper summarizes the calculations and comparisons related to TFCX siting and environmental issues such as radiological doses to the public living near the facility. Included are discussions of (a) routine and accidental releases of tritium, (b) routine releases of activated air, (c) direct radiation (including ''skyshine''), and (d) seismic criteria. Other potential issues are also discussed including the amount of tritium that might be retained in the graphite armor in the torus, the possible severity of magnet accidents, and the extent of damage due to plasma disruptions. The conclusions drawn from these calculations should be applicable to some of the other planned ignited core experiments that have operating parameters similar to those of TFCX

  12. Issues relating to the siting of tritium-fueled fusion experiments

    International Nuclear Information System (INIS)

    Reilly, H.J.; Holland, D.F.

    1985-01-01

    A preconceptual design study and safety analysis of the Tokamak Fusion Core Experiment (TFCX) was conducted in 1984 for the Department of Energy. This paper summarizes the calculations and comparisons related to TFCX siting and environmental issues such as radiological doses to the public living near the facility. Included are discussions of (a) routine and accidental releases of tritium, (b) routine releases of activated air, (c) direct radiation (including skyshine), and (d) seismic criteria. Other potential issues are also discussed including the amount of tritium that might be retained in the graphite armor in the torus, the possible severity of magnet accidents, and the extent of damage due to plasma disruptions. The conclusions drawn from these calculations should be applicable to some of the other planned ignited core experiments that have operating parameters similar to those of TFCX

  13. Tritium surface loading due to contamination of rainwater from atmospheric release at NAPS

    International Nuclear Information System (INIS)

    Sharma, L.N.; Dube, B.; Varakhedkar, V.K.

    2001-01-01

    Annual tritium (HTO) surface loading has been measured and calculated for the year 1998-99 within 0.8 km distance from 145m high stack of Narora Atomic Power Station (NAPS) at eight locations in different directions. The technique for measured values consists of the summation of product of tritium concentration (Bq/l) in daily rainfall samples and daily rainfall (mm) whereas that for calculated values having the use of prevailing meteorological conditions and average tritium release rate during a year. The ratios of measured and calculated values of tritium surface loading during the years 1998-99 are found to be in the range of 0.18 to 6.97. Tritium surface loading studies at NAPS reveal that a fraction 1.7E-03 of total annual tritium released through stack gets deposited on the surface due to washout / rainout of plume within 0.8 km radial distance from stack. The range of deposition velocity, V w (m.s - 1 ) i.e the ratio of annual tritium surface loading W(Bq.m - 2 . s - 1 ) and annual mean tritium concentration in air, χo(Bq.m - 3) at three locations for the years 1998-99 is found to be 5.59E-04 to 5.99E-03 ms - 1 . The average value for wet deposition velocity V bar w for NAPS site is estimated as 2.92E-03 m.s - 1. (author)

  14. Reconstruction of tritium release history from contaminated groundwater using tree ring analysis

    International Nuclear Information System (INIS)

    Kalin, R.M.; Murphy, C.E. Jr.; Hall, G.

    1995-01-01

    The history of tritium releases to the groundwater from buried waste was reconstructed through dendrochronology. Wood from dated tree rings was sectioned from a cross-section of a tree that was thought to tap the groundwater. Cellulose was chemically separated from the wood. The cellulose was combusted and the water of combustion collected for liquid scintillation counting. The tritium concentration in the rings rose rapidly after 1972 which was prior to the first measurements made in this area. Trends in the tritium concentration of water outcropping to the surface are similar to the trends in tritium concentration in tree rings. 14 refs., 3 figs

  15. 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

  16. Analyses of postulated accidental releases of UF6 inside process buildings

    International Nuclear Information System (INIS)

    Oliveira Neto, Jose Messias de; Nunes, Beatriz Guimaraes; Dias, Cristiane

    2009-01-01

    Uranium Hexafluoride is a material used in the various processes which comprise the front end of the nuclear fuel cycle (conversion, enrichment and fuel fabrication). Confinement of UF 6 is a very important safety requirement since this material is highly reactive and presents safety hazards to humans. The present paper discusses the safety relevant aspects of accidental releases of UF 6 inside process confinement buildings. Postulated accidental scenarios are analyzed and their consequences evaluated. Implant releases rates are estimated using computer code predictions. A time dependent homogeneous compartment model is used to predict concentrations of UF 6 , hydrogen fluoride and uranyl fluoride inside a confinement building, as well as to evaluate source terms released to the atmosphere. These source terms can be used as input to atmospheric dispersion models to evaluate consequences to the environment. The results can also be used to define adequate protective measures for emergency situations. (author)

  17. Methodology for estimating accidental radioactive releases in nuclear waste management

    International Nuclear Information System (INIS)

    Levy, H.B.

    1979-01-01

    Estimation of the risks of accidental radioactive releases is necessary in assessing the safety of any nuclear waste management system. The case of a radioactive waste form enclosed in a barrier system is considered. Two test calculations were carried out

  18. Estimation of release of tritium from measurements of air concentrations in reactor building of PHWR

    International Nuclear Information System (INIS)

    Purohit, R.G.; Sarkar, P.K.

    2010-01-01

    In this paper an attempt has been made to estimate the releases from measured air concentrations of tritium at various locations in Reactor Building (RB). Design data of Kaiga Generating Station and sample measurements of tritium concentrations at various locations in RB and discharges for a period of fortnight were used. A comparison has also been made with actual measurements. It has been observed that there is good matching in estimated and actual measurements of tritium release on some days while on some days there is high difference

  19. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Barry B [ORNL; Bruffey, Stephanie H [ORNL; DelCul, Guillermo Daniel [ORNL; Walker, Trenton Baird [ORNL

    2016-08-31

    Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using nonradioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  20. Capture of Tritium Released from Cladding in the Zirconium Recycle Process

    Energy Technology Data Exchange (ETDEWEB)

    Bruffey, Stephanie H [ORNL; Spencer, Barry B [ORNL; DelCul, Guillermo Daniel [ORNL

    2016-08-31

    This report is issued as the first revision to FCRD-MRWFD-2016-000297. Zirconium may be recovered from the Zircaloy® cladding of used nuclear fuel (UNF) for recycle or to reduce the quantities of high-level waste destined for a geologic repository. Recovery of zirconium using a chlorination process is currently under development at the Oak Ridge National Laboratory. The approach is to treat the cladding with chlorine gas to convert the zirconium in the alloy (~98 wt % of the alloy mass) to zirconium tetrachloride. A significant fraction of the tritium (0–96%) produced in nuclear fuel during irradiation may be found in zirconium-based cladding and could be released from the cladding when the solid matrix is destroyed by the chlorination reaction. To prevent uncontrolled release of radioactive tritium to other parts of the plant or to the environment, a method to recover the tritium may be required. The focus of this effort was to (1) identify potential methods for the recovery of tritium from the off-gas of the zirconium recycle process, (2) perform scoping tests on selected recovery methods using non-radioactive gas simulants, and (3) select a process design appropriate for testing on radioactive gas streams generated by the engineering-scale zirconium recycle demonstrations on radioactive used cladding.

  1. Tritium labeling of gonadotropin releasing hormone in its proline and histidine residues

    International Nuclear Information System (INIS)

    Klauschenz, E.; Bienert, M.; Egler, H.; Pleiss, U.; Niedrich, H.; Nikolics, K.

    1981-01-01

    3,4-dehydroproline9-GnRH prepared by solid phase peptide synthesis was tritiated catalytically under various conditions yielding 3H-GnRH with specific radioactivities in the range from 35-60 Ci/mmol and full LH releasing activity in vitro. Using palladium/alumina catalyst, the tritiation of the double bond occurs within ten minutes. Investigation of the tritium distribution between the amino acid residues showed a remarkably high incorporation of tritium into the histidine residue (11 to 37%). On the basis of this observation, the tritium labeling of GnRH and angiotensin I by direct catalytic hydrogen-tritium exchange was found to be useful for the labeling of these peptides at remarkably high specific radioactivity

  2. Numerical simulation of industrial and accidental release formation and transport

    Energy Technology Data Exchange (ETDEWEB)

    Piskunov, V.N.; Aloyan, A.A.; Gerasimov, V.M.; Pinaev, V.S.; Golubev, A.I.; Yanilkin, Yu.V.; Ivanov, N.V.; Nikonov, S.N.; Kharchenko, A.I. [and others

    1995-05-01

    Statement of work for contract 006 {open_quotes}Mathematical simulation of industrial and accidental release formation and transport{close_quotes} implies that the final result of the activity within this task will be VNIIEF developed techniques which will provide for the prediction of the post-accidental environment. Report [1] presents the description of physical models and calculation techniques which were chosen by VNIIEF to accomplish this task. These techniques were analysed for their capabilities, the areas of their application were defined, modifications within contract 006 were described, the results of test and methodical calculations were presented. Moreover, the experimental data were analysed over the source parameters and contamination measurements which can be used in the comparison with the calculation results. Based an these data this report compares the calculation results obtained with VNIIEF calculation techniques with the LANL-presented experimental results. The calculations which statements and results are given in section 1, included the following processes: explosion cloud ascent dynamics and jet release origin; aerosols kinetics in the release source including composite particle origin in the explosion cloud caused by radioactive substance sorption an the soil particles; contaminant transport in atmosphere and their in-site fallout due to the accidental explosions and fires; atmospheric flow dynamics and industrial contamination transfer over the complicated terrain. The calculated results were compared with the experimental data. Section 2 presents the parameters for a typical source in the explosion accidents based an the experimental results and calculated data from Section 1, as well as contamination patterns were calculated with basic technique {open_quotes}Prognosis{close_quotes}.

  3. Tritium release behavior from neutron-irradiated Li{sub 2}TiO{sub 3} single crystal

    Energy Technology Data Exchange (ETDEWEB)

    Tanifuji, Takaaki; Yamaki, Daiju; Noda, Kenji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nasu, Shoichi

    1998-03-01

    Li{sub 2}TiO{sub 3} single-crystals with various size (1-2mm) were used as specimens. After the irradiation up to 4 x 10{sup 18} n/cm{sup 2} with thermal neutrons in JRR-2, tritium release from the Li{sub 2}TiO{sub 3} specimens in isothermal heating tests was continuously measured with a proportional counter. The tritium release in the range from 625K to 1373K seems to be controlled by bulk diffusion. The tritium diffusion coefficient (D{sub T}) in Li{sub 2}TiO{sub 3} was evaluated to be D{sub T}(cm{sup 2}/sec) = 0.100exp(-104(kJ/mol)/RT), 625Ktritium diffusion coefficients in Li{sub 2}TiO{sub 3} is almost equal to those of Li{sub 2}O irradiated with thermal neutrons up to 2 x 10{sup 19} n/cm{sup 2}. It indicates that the tritium release performance of Li{sub 2}TiO{sub 3} is essentially good as Li{sub 2}O. (author)

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

    International Nuclear Information System (INIS)

    Wu, C.H.

    1986-01-01

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

  5. Influence of additives and impurities in sweep gas and solid tritium release behaviour from lithium ceramics (review)

    International Nuclear Information System (INIS)

    Tanaka, Satoru

    1991-01-01

    Tritium release from solid breeding material is affected by small amounts of additives or impurities in the sweep gas or solid itself. Addition of hydrogen or water vapor to the sweep gas is reported to enhance the surface reaction of tritium release. Doping to solid breeder with elements of different valence from lithium has a possibility to improve tritium diffusion in the solid. Surface reaction and migration behavior in bulk are believed to be also affected by impurities in the sweep gas and in the solid. In order to model tritium release behavior in the blanket of fusion reactor, the mechanism of interaction with these additives or impurities must be quantitatively formulated. However, the mechanism of these remains to be elucidated. In this paper effects of these additives and impurities on tritium migration are reviewed. The mechanism of surface reaction for He+H 2 sweep gas is also discussed. (orig.)

  6. ITER task title - source term data, modelling, and analysis. ITER subtask no. S81TT05/5 (SEP 1-1). Global tritium source term analysis basis document. Subtask 1: operational tritium effluents and releases. Final report (1995 TASK)

    International Nuclear Information System (INIS)

    Kalyanam, K.M.

    1996-06-01

    This document represents the final report for the global tritium source term analysis task initiated in 1995. The report presents a room-by-room map/table at the subsystem level for the ITER tritium systems, identifying the major equipment, secondary containments, tritium release sources, duration/frequency of tritium releases and the release pathways. The chronic tritium releases during normal operation, as well as tritium releases due to routine maintenance of the Water Distillation Unit, Isotope Separation System and Primary and Secondary Heat Transport Systems, have been estimated for most of the subsystems, based on the IDR design, the Design Description Documents (April - Jun 1995 issues) and the design updates up to December 1995. The report also outlines the methodology and the key assumptions that are adopted in preparing the tritium release estimates. The design parameters for the ITER Basic Performance Phase (BPP) have been used in estimating the tritium releases shown in the room-by-room map/table. The tritium release calculations and the room-by-room map/table have been prepared in EXCEL, so that the estimates can be refined easily as the design evolves and more detailed information becomes available. (author). 23 refs., tabs

  7. Description of the new version 4.0 of the tritium model UFOTRI including user guide

    International Nuclear Information System (INIS)

    Raskob, W.

    1993-08-01

    In view of the future operation of fusion reactors the release of tritium may play a dominant role during normal operation as well as after accidents. Because of its physical and chemical properties which differ significantly from those of other radionuclides, the model UFOTRI for assessing the radiological consequences of accidental tritium releases has been developed. It describes the behaviour of tritium in the biosphere and calculates the radiological impact on individuals and the population due to the direct exposure and by the ingestion pathways. Processes such as the conversion of tritium gas into tritiated water (HTO) in the soil, re-emission after deposition and the conversion of HTO into organically bound tritium, are considered. The use of UFOTRI in its probabilistic mode shows the spectrum of the radiological impact together with the associated probability of occurrence. A first model version was established in 1991. As the ongoing work on investigating the main processes of the tritium behaviour in the environment shows up new results, the model has been improved in several points. The report describes the changes incorporated into the model since 1991. Additionally provides the up-dated user guide for handling the revised UFOTRI version which will be distributed to interested organizations. (orig.) [de

  8. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

  9. The releases of krypton-85 and tritium to the environment and tritium to krypton-85 ratios as source indicators

    International Nuclear Information System (INIS)

    Schroeder, K.J.P.; Roether, W.

    1975-01-01

    More than 95% of the krypton-85 that is at present observed in the environment on a global scale originates from the combined releases following nuclear-power generation and plutonium production, while atmospheric fusion-bomb testing accounts for the same percentage of the global tritium. The global inventories at the end of 1973 were 55 MCi (estimated uncertainty +- 10%) of krypton-85, and 2900 MCi (+- 25%) of tritium. From the excess of the global krypton-85 inventory over the amounts accounted for by the sources other than plutonium production, it is concluded that 130 tonnes of plutonium-239 have been produced up to 1970. Pronounced fractionation occurs between atmosphere and hydrosphere, in the sense that the concentration of krypton-85 is lowered, and that of tritium (being overwhelmingly in the form of a water molecule, HTO) is greatly increased, in the hydrosphere. Largely as a consequence of this fractionation, the environmental tritium to krypton-85 ratios cover a wide range, with values lowest in the troposphere (typical order of magnitude 10 -1 Ci/Ci) and highest in continental surface water (of the order of 10 5 ). Preliminary oceanic krypton-85 data are reported, which indicate that the concentration decrease with depth in the ocean is very similar for both nuclides. Effluents from nuclear power reactors and reprocessing plants have tritium to krypton-85 ratios that do not differ markedly from those found at present in the troposphere; on the other hand, the very high ratios that are typical of the hydrosphere may offer opportunities of detecting such effluents in the hydrosphere on the basis of tritium to krypton-85 ratios. Considerable uncertainties in the ratios of the effluents have, however, to be anticipated, arising from varying, and fluctuating, fractionation between the two nuclides in the release processes. (author)

  10. Tritium and the environment: sources, measurement and transfer; Le tritium et l'environnement: sources, mesures et transferts

    Energy Technology Data Exchange (ETDEWEB)

    Guetat, P.; Douche, C.; Hubinois, J.C. [CEA Valduc, 21 - Is-sur-Tille (France)

    2008-10-15

    Within the framework of a seminar on environmental and health impact of tritium organized by article 31 EURATOM expert group in November 2007, it was expected to make a point on tritium knowledge and to define R and D and regulatory axes. This document presents the different sources of tritium, different methods of measurement associated to tritium processes and survey, transfer in the biosphere compartments in normal and accidental conditions. It suggests some R and D subjects and discusses some regulatory aspects. (authors)

  11. Tritium release reduction and radiolysis gas formation

    Energy Technology Data Exchange (ETDEWEB)

    Batifol, G.; Douche, Ch.; Sejournant, Ch. [CEA Valduc, 21 - Is-sur-Tille (France)

    2008-07-15

    At CEA Valduc, the usual tritiated waste container is the steel drum. It allows good release reduction performance for middle activity waste but in some cases tritium outgassing from the waste drums is too high. It was decided to over-package each drum in a tighter container called the over-drum. According to good safety practices it was also decided to measure gas composition evolution into the over-drum in order to defect hydrogen formation over time. After a few months, a significant release reduction was observed. Additionally there followed contamination reduction in the roof storage building rainwater. However hydrogen was also observed in some over-drums, in addition to other radiolysis products. Catalyst will be added to manage the hydrogen risk in the over-drums. (authors)

  12. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO/sub 2/) and lithium silicate (Li/sub 2/SiO/sub 3/) by the reaction: Li/sup 6/ + n ..-->.. /sup 4/He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100/sup 0/C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T/sub 2/), while in laboratory extractions (300-1300/sup 0/C), the tritium appeared primarily in the condensible form (HTO and T/sub 2/O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H/sub 2/O, CO/sub 2/, CO, O/sub 2/, H/sub 2/, NO, SO/sub 2/, SiF/sub 4/ and traces of hydrocarbons.

  13. Tritium release and retention properties of highly neutron-irradiated beryllium pebbles from HIDOBE-01 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A.; Klimenkov, M.; Kolb, M.; Vladimirov, P.; Kurinskiy, P.; Schneider, H.-C. [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Til, S. van; Magielsen, A.J. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-11-15

    The current helium cooled pebble bed (HCPB) tritium breeding blanket concept for fusion reactors includes a bed of 1 mm diameter beryllium pebbles to act as a neutron multiplier. Beryllium pebbles, fabricated by the rotating electrode method, were neutron irradiated in the HFR in Petten within the HIDOBE-01 experiment. This study presents tritium release and retention properties and data on microstructure evolution of beryllium pebbles irradiated at 630, 740, 873, 948 K up to a damage dose of 18 dpa, corresponding to a helium accumulation of about 3000 appm. The measured cumulative released activity from the beryllium pebbles irradiated at 948 K was found to be significantly lower than the calculated value. After irradiation at 873 and 948 K scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses revealed large pores or bubbles in the bulk and oxide films with a thickness of up to 8 μm at the surface of the beryllium pebbles. The radiation-enhanced diffusion of tritium and the formation of open porosity networks accelerate the tritium release from the beryllium pebbles during the high-flux neutron irradiation.

  14. Accidental releases of radionuclides: a preliminary study of the consequences of land contamination

    International Nuclear Information System (INIS)

    Simmonds, J.R.; Haywood, S.M.; Linsley, G.S.

    1982-10-01

    The long term consequences of land contamination from accidental releases of activity from thermal reactors are examined. The radiological consequences are assessed using an analysis of the exposure of individuals and the population to ground deposits of the radionuclides released. The contribution of the different nuclides in the release by their various exposure routes to the irradiation of man are calculated as a function of time after release and the most radiologically important are identified. A preliminary assessment is made of off-site economic and social consequences of accidental releases by estimating the areas of land which would be affected by the introduction of countermeasures to control individual radiation exposure due to external irradiation from ground deposits (relocation of populations), and the intake of radionuclides contained in locally produced foodstuffs (restrictions on food production). The areas where administrative controls would be necessary decline in size with time after the release; estimates are made of this time-dependent behaviour using dynamic environmental transfer models. Finally, the collective doses saved by the introduction of countermeasures are estimated using population and agricultural distribution data for a rural location in the United Kingdom. (author)

  15. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. Recent progress in the modelling of helium and tritium behaviour in irradiated beryllium pebbles

    International Nuclear Information System (INIS)

    Rabaglino, E.; Ronchi, C.; Cardella, A.

    2003-01-01

    One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release

  18. In-Pile Assemblies for Investigation of Tritium Release from Li2TiO3 Lithium Ceramic

    International Nuclear Information System (INIS)

    Shestakov, V.; Tazhibayeva, I.; Kawamura, H.; Kenzhin, Y.; Kulsartov, T.; Chikhray, Y.; Kolbaenkov, A.; Arinkin, F.; Gizatulin, Sh.; Chakrov, P.

    2005-01-01

    The description of algorithm to design in-pipe experimental ampoule devices (IPAD) is presented here, including description of IPAD design for irradiation tests of highly enriched lithium ceramics at WWR-K reactor. The description of the system for registration of tritium release from ceramics during irradiation is presented as well. Typical curve of tritium release from the IPAD during irradiation under various temperatures of the samples is shown here

  19. Experimental investigation on tritium release from lithium titanate pebble under high temperature of 1073 K

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Kentaro, E-mail: howartre@onid.oregonstate.edu [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Edao, Yuki; Kawamura, Yoshinori [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Ohta, Masayuki; Sato, Satoshi; Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki (Japan)

    2015-10-15

    Highlights: • We have performed the tritium recovery experiment with the DT neutron source at 1073 K. • The tritium recovery corresponded with the calculated tritium production. • The chemical form of recovered tritium is affected by the temperature and kind of sweep gas. • The recovered HT increases at higher temperature and dry hydrogen circumstance. - Abstract: The temperature of Li{sub 2}TiO{sub 3} pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li{sub 2}TiO{sub 3} pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li{sub 2}TiO{sub 3} pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li{sub 2}TiO{sub 3} layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H{sub 2}/He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H{sub 2}O/He gas (H{sub 2}O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H{sub 2}, it is indicated that the increasing tendency of HT release is similar to that of the dry H{sub 2}/He.

  20. Investigation of fire at Council, Alaska: A release of approximately 3000 curies of tritium

    International Nuclear Information System (INIS)

    Jensen, G.A.; Martin, J.B.

    1988-04-01

    On September 6, 1987, about 6:00 a.m., a fire was discovered in the community building at Council, Alaska, where 12 radioluminescent (RL) light panels containing approximately 3000 Ci were stored. All of the tritium in the panels was released as a result of the fire. This report summarizes the recovery of the remains of the panels destroyed in the fire and investigations completed to evaluate the fire site for possible exposure of community residents or contamination by tritium release in the environment. Based on the analysis of urine samples obtained from individuals in the community and from Pacific Northwest Laboratory personnel participating in the recovery operation, no evidence of exposure to individuals was found. No tritium (above normal background) was found in water and vegetation samples obtained at various locations near the site. 12 figs., 3 tabs

  1. Modeling and validating tritium transfer in a grassland ecosystem in response to {sup 3}H releases

    Energy Technology Data Exchange (ETDEWEB)

    Le Dizes, S.; Maro, D.; Rozet, M.; Hebert, D.; Solier, L.; Nicoulaud, V. [Institut de radioportection et de surete nucleaire - IRSN (France); Vermorel, F.; Aulagnier, C. [Electricite de France - EDF (France)

    2014-07-01

    Tritium ({sup 3}H) is a major radionuclide released in several forms (HTO, HT) by nuclear facilities under normal operating conditions. In terrestrial ecosystems, tritium can be found under two forms: tritium in tissue free water (TFWT) following absorption of tritiated water by leaves or roots and Organically Bound Tritium (OBT) resulting from TFWT incorporation by the plant organic matter during photosynthesis. In order to study transfers of tritium from atmospheric releases to terrestrial ecosystem such as grasslands, an in-situ laboratory has been set up by IRSN on a ryegrass field plot located 2 km downwind the AREVA NC La Hague nuclear reprocessing plant (North-West of France), as was done in the past for the assessment of transfer of radiocarbon in grasslands. The objectives of this experimental field are: (i) to better understand the OBT formation in plant by photosynthesis, (ii) to evaluate transfer processes of tritium in several forms (HT, HTO) from the atmosphere (air and rainwater) to grass and soil, (iii) to develop a modeling allowing to reproduce the dynamic response of the ecosystem to tritium atmospheric releases depending of variable environmental conditions. For this purpose, tritium activity measurements will be carried out in grass (monthly measurements of HTO, OBT), in air, rainwater, soil (daily measurements of HT, HTO) and CO{sub 2}, H{sub 2}O fluxes between soil and air compartments will be carried out. Then, the TOCATTA-c model previously developed to simulate {sup 14}C transfers to pasture on a hourly time-step basis will be adapted to take account for processes specific to tritium. The model will be tested by a comparison between simulated results and measurements. The objectives of this presentation are (1) to present the organization of the experimental design of the VATO study (Validation of TOCATTA) dedicated to transfers of tritium in a grassland ecosystem, (2) to document the major assumptions, conceptual modelling and

  2. Relocation impacts of a major release from SRTC

    International Nuclear Information System (INIS)

    Blanchard, A.; Thompson, E.A.; Thompson, J.M.

    1999-01-01

    The relocation impacts of an accidental release, scenario 1-RD-3 , are evaluated for the Savannah River Technology Center. The extent of the area potentially contaminated to a level that would result in doses exceeding the relocation protective action guide(PAG)is calculated. The maximum calculated distance downwind from the accident at which the relocation PAG is exceeded is also determined. The consequences of the particulate portion of the release are evaluated using the HOTSPOT model and an EXCEL spreadsheet. The consequences of the tritium release are evaluated using UFOTRI

  3. 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

  4. An advanced model for spreading and evaporation of accidentally released hazardous liquids on land

    NARCIS (Netherlands)

    Trijssenaar-Buhre, I.J.M.; Sterkenburg, R.P.; Wijnant-Timmerman, S.I.

    2009-01-01

    Pool evaporation modelling is an important element in consequence assessment of accidentally released hazardous liquids. The evaporation rate determines the amount of toxic or flammable gas released into the atmosphere and is an important factor for the size of a pool fire. In this paper a

  5. An advanced model for spreading and evaporation of accidentally released hazardous liquids on land

    NARCIS (Netherlands)

    Trijssenaar-Buhre, I.J.M.; Wijnant-Timmerman, S.L.

    2008-01-01

    Pool evaporation modelling is an important element in consequence assessment of accidentally released hazardous liquids. The evaporation rate determines the amount of toxic or flammable gas released into the atmosphere and is an important factor for the size of a pool fire. In this paper a

  6. A 'Puff' dispersion model for routine and accidental releases

    International Nuclear Information System (INIS)

    Grsic, Z.; Rajkovic, B.; Milutinovic, P.

    1999-01-01

    A Puff dispersion model for accidental or routine releases is presented. This model was used as a constitutive part of an automatic meteorological station.All measured quantities are continuously displayed on PC monitor in a digital and graphical form, they are averaging every 10 minutes and sending to the civil information center of Belgrade. In the paper simulation of a pollutant plume dispersion from The oil refinery 'Pancevo', on April 18 th 1999 is presented. (author)

  7. Development of dose assessment code for release of tritium during normal operation of nuclear power plants

    International Nuclear Information System (INIS)

    Duran, J.; Malatova, I.

    2009-01-01

    A computer code PTM H TO has been developed to assess tritium doses to the general public. The code enables to simulate the behavior of tritium in the environment released into the atmosphere under normal operation of nuclear power plants. Code can calculate the doses for the three chemical and physical forms: tritium gas (HT), tritiated water vapor and water drops (HTO). The models in this code consist of the tritium transfer model including oxidation of HT to HTO and reemission of HTO from soil to the atmosphere, and the dose calculation model

  8. Calculation of tritium inventory and permeation in an INTOR like tokamak device and of its release after shut down

    International Nuclear Information System (INIS)

    Wienhold, P.; Waelbroeck, F.s; Winter, J.; Ali-Khan, I.

    1980-12-01

    The tritium inventory in the wall, its escape via permeation and its release in general are evaluated during the operation and later shut-down phases of a tokamak like INTOR by means of the PERI code. No real tritium problem arises after shut-down: simple outgasing techniques at wall temperatures Tsub(W) of approx. 500 0 C should decrease the tritium release rate down to a sufficient low level. Very serious problems are on the other hand expected to arise in the areas of inventory and permeation of tritium during the operation phase: one has the choice between e.g. operating at Tsub(W) = 350 0 C with a quasi-stationary tritium inventory of some 100 grams and a permeation flux phisub(p) approx. 1 gram/day or increasing Tsub(W) to 500 0 C, reducing the inventory down to some grams, but have then a leakage rate by permeation of 35 grams/day. For Tsub(W) = 100 0 C the tritium release into the torus will be intolerable. These figures are confirmed by preliminary measurements on Inconel 600. A suggestion is made as to how to alleviate the problem. (orig.) [de

  9. Transfer of tritium released by nuclear facilities to the food supply

    International Nuclear Information System (INIS)

    Bovard, P.; Delmas, J.; Belot, Y.; Camus, H.; Grauby, A.; Hoek, J. van den

    1979-01-01

    The use for agricultural purposes of river waters receiving releases or discharges of tritium results in contamination of irrigated crops and of animals given such water to drink or consuming the contaminated crops. It therefore seemed of importance to assess the part played by tritium in the contamination of the food chain, together with its possible effects on organisms. With this in mind, French, Belgian and Netherlands laboratories have joined forces to study, more especially, the relationship between environmental contamination rates and those of produce harvested in the Mediterranean region and in a humid temperate climate, the transfer process in the chain: water - fodder - bovines - dairy produce, and the role of technology in the contamination of the food chain. The present status of research undertaken jointly by organizations in the three countries is reviewed. In the Atlantic environment the experiments involved four annual crops consumed on a large scale: potatoes, sugar beet, carrots and peas, and in the Mediterranean environment several perennial species such as vine, olive, orange and apple were studied. The results obtained relate to the residence time for tritium in the various organs of each species, the part played by evapotranspiration and the physiological functions of the different parts of the plants, the uptake of tritium by tissue water and organic matter, and the distribution of tritium in the soil profile. (author)

  10. 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.)

  11. The use of consequence modelling in assessing accidental releases of activity

    International Nuclear Information System (INIS)

    Clarke, R.H.; Hemming, C.R.; Kelly, G.N.

    1982-01-01

    The radiological consequences of an accidental release can be expressed as the number of health effects in the exposed population and its descendents, and the economic and social implications that may result from restrictions placed on the consumption of foodstuffs, or on the occupation of domestic and commercial premises. For any given release of activity there will be a probability distribution of consequences which will depend on the meteorological characteristics and demographic data particular to the release location. The characteristics of this distribution should be important in reaching judgements on the acceptability of a given site or design. The relative importance of the different endpoints is analysed and the influence of the release location on the resulting distribution is considered. (author)

  12. Tritium surface loading due to contamination of rainwater from atmospheric release at NAPS (2011)

    International Nuclear Information System (INIS)

    Gautam, Y.P.; Sharma, Saivajay; Rao, K.S.; Singh, Bhikam; Kumar, Avinash; Ravi, P.M.

    2012-01-01

    Annual tritium (HTO) surface loading has been measured and calculated for the year 2011 within 0.8 km distance from 145 m high stack of Narora Atomic Power Station (NAPS) at eight locations in different directions. The technique for measured values consists of the summation of product of tritium concentration (Bq/l) in daily rainfall samples and daily rainfall (mm). Tritium surface loading studies at NAPS reveal that a fraction 1.01E-03 of total annual tritium released through stack gets deposited on the surface due to washout/rainout of plume within 0.8 km radial distance from stack. The range of deposition velocity, Vw (m.s -1 ) i.e., the ratio of annual tritium surface loading W (Bq. m -2 .s -1 ) and annual mean tritium concentration in air, c 0 (Bq.m -3 ) at three locations for the years 2011 is found to be 6.12E-04 to 2.89E-03. The average value for wet deposition velocity V w for NAPS site is estimated as 3.17E-03 m.s -1 . (author)

  13. The influence of season of the year on the predicted agricultural consequences of accidental releases of radionuclides to atmosphere

    International Nuclear Information System (INIS)

    Simmonds, J.R.

    1985-02-01

    In Europe, because of the seasonal nature of agricultural practices, the consequences for agriculture of an accidental release of radioactive materials to atmosphere are likely to vary depending upon the time of year when the release occurs. The quantification of this variation is complicated by the need to take into account the introduction of countermeasures to restrict the radiation exposure from ingestion of contaminated foods, and by the presence in accidental releases of radionuclides which persist over several seasons. In this study, the effect on agricultural consequences of accidental releases occurring at different times of the year is examined. The consequences are expressed in terms of the amount of produce affected by restrictions on food supplies and the collective radiation dose from ingestion of food. The investigation has been carried out for three hypothetical releases representing a range of releases postulated for pressurised water reactors (PWRs). The effect of season of the year was determined for accidental releases occurring both in a single, defined set of meteorological conditions and for a range of possible meteorological conditions. For the main part of the study, consideration was limited to agricultural production in the UK only, but the effect of extending the analysis beyond the UK boundary was also considered. The results of the study show that considerable variation can occur in agricultural consequences following an accidental release at different times of the year. For the larger releases considered, this variation is reduced due to the effect of the introduction of countermeasures, particularly when consideration is limited to the UK only. Seasonal variation tends to be greater for the results of a deterministic analysis, which uses a single set of constant meteorological conditions, than for the results of a full probabilistic assessment. From the results presented here it is also seen that for many applications of

  14. Accidental Continuous Releases from Coal Processing in Semi-Confined Environment

    Directory of Open Access Journals (Sweden)

    Bruno Fabiano

    2013-09-01

    Full Text Available Notwithstanding the enforcement of ATEX EU Directives (94/9/EC of 23 March 1994 and safety management system application, explosions in the coal sector still claim lives and cause huge economic losses. Even a consolidated activity like coke dry distillation allows the opportunity of preventing explosion risk connected to fugitive emissions of coke oven gas. Considering accidental releases under semi-confined conditions, a simplified mathematical approach to the maximum allowed gaseous build-up is developed on the basis of the intrinsic hazards of the released compound. The results will help identifying and assessing low rate release consequences therefore to set-up appropriate prevention and control measures. The developed methodology was tested at the real-scale and validated by numerical computational fluid dynamics (CFD simulations showing the effectiveness of the methodology to evaluate and mitigate the risk connected to confined hazardous releases.

  15. 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).

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

    International Nuclear Information System (INIS)

    Brown, D.H.

    1998-01-01

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

  17. Evaluation and mitigation of accidental releases of radioactivity

    International Nuclear Information System (INIS)

    Bruessermann, K.

    1982-12-01

    One result of the workshop was, that even in the case of severe accident sequences in modern nuclear power plants (other facilities were not discussed), there will be enough time to take active measures in order to lower the emissions and to diminish the consequences inside and outside of the plant. On the whole, new evidence from Harrisburg show that previously estimated accidental emissions, especially of radiologically relevant nuclides, have been rather conservative and that much lower emissions are possible, if the above measures are considered. Under accident conditions, models to predict radiation exposure must be applied under the event of a short-term release. (orig./DG) [de

  18. Chemical form of released tritium from molten Li{sub 2}BeF{sub 4} salt under neutron irradiation at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Akihiro; Terai, Takayuki; Yoneoka, Toshiaki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    Chemical forms of released tritium from FLIBE (the 2:1 mixture of LiF and BeF{sub 2}) by in-pile tritium release experiment were HT and TF and their proportion depended on the chemical composition of purge gas and the dehumidification time of specimen at high temperatures. The chemical form of tritium was determined by the thermodynamic equilibrium of the isotopic exchange reaction (T{sup +} + H{sub 2} {yields} H{sup +} + HT). (author)

  19. 3He release characteristics of metal tritides and scandium--tritium solid solutions

    International Nuclear Information System (INIS)

    Perkins, W.G.; Kass, W.J.; Beavis, L.C.

    1975-01-01

    Tritides of such metals as scandium, titanium, and erbium are useful materials for determining the effects of helium accumulation in metallic solids, for example, CTR first wall materials. Such effects include lattice strain and gross deformation, as reported elsewhere, which are related to 3 He retention and ultimate release. Long term gas release studies have indicated that, during the early life of a metal ditritide, a large fraction of the 3 He is retained in the solid. At more advanced ages (2 to 4 years, depending on the parent metal), the 3 He release rate becomes comparable to the generation rate. Statistical analysis of the data indicates that the acceleration in 3 He release rate depends on accumulated 3 He concentration rather than strictly on age. 3 He outgassing results are presented for thin films of ScT 2 , TiT 2 , and ErT 2 , and the critical 3 He concentrations are discussed in terms of a percolation model. Phase transformations which occur on tritide formation cast some doubt on the validity of extrapolating results obtained for metal tritides to predictions regarding the accumulation of helium in metals. Scandium is unique among the early transition and rare-earth metals in that the metal exhibits a very high room temperature tritium solubility (T/Sc = 0.4) with no phase transformation. Indeed, even the lattice parameters of the hcp scandium lattice are only minimally changed by tritium solution, and we have succeeded in obtaining single crystal ScT 0 . 3 samples in two crystallographic orientations. Using a very sensitive technique, we have measured 3 He emission from both these samples, as well as from fine-grained thin film scandium-tritium solid solution samples (ScT 0 . 3 - 0 . 4 ). The fine-grained film samples release 3 He at 2 to 3 percent of the generation rate, while the emission rate from the single-crystal samples is approximately 0.05 percent of the generation rate, indicating a strong grain size effect

  20. Individual and population dose to users of the Savannah River following K-Reactor tritium release

    International Nuclear Information System (INIS)

    Carlton, W.H.; Hamby, D.M.

    1992-01-01

    Approximately 5700 curies of tritium were released to Pen Branch between December 22, 1991 and December 25, 1991. As expected, the tritiated water traveled through the Savannah River swamp to Steel Creek and the Savannah River. Elevated tritium concentrations in the river at Becks Ferry (Beaufort-Jasper) and Abercorn Creek (Port Wentworth) has caused some concern among downstream water users as to the amount of tritium available for uptake through the domestic drinking water supplies at the Beaufort-Jasper and Port Wentworth water treatment facilities. Radiation dose to the downstream drinking water population is estimated in this report

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

    Science.gov (United States)

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

    1994-09-01

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

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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

  7. Modification and testing of the Sandia Laboratories Livermore tritium decontamination systems

    International Nuclear Information System (INIS)

    Gildea, P.D.; Birnbaum, H.G.; Wall, W.R.

    1978-08-01

    Sandia Laboratories, Livermore, has put into operation a new facility, the Tritium Research Laboratory. The laboratory incorporates containment and cleanup facilities such that any tritium accidentally released is captured rather than vented to the atmosphere. This containment is achieved with hermetically sealed glove boxes that are connected on demand by manifolds to two central decontamination systems called the Gas Purification System (GPS) and the Vacuum Effluent Recovery System (VERS). The primary function of the GPS is to remove tritium and tritiated water vapor from the glove box atmosphere. The primary function of the VERS is to decontaminate the gas exhausted from the glove box pressure control systems and vacuum pumps in the building before venting the gas to the stack. Both of these systems are designed to remove tritium to the few parts per billion range. Acceptance tests at the manufacturer's plant and preoperational testing at Livermore demonstrated that the systems met their design specifications. After preoperational testing the Gas Purification System was modified to enhance the safety of maintanance operations. Both the Gas Purification System and the Vacuum Effluent Recovery System were performance tested with tritium. Results show that concentraion reduction factors (ratio of inlet to exhaust concentrations) much in excess of 1000 per pass have been achieved for both systems at inlet concentrations of 1 ppM or less

  8. Modification and testing of the Sandia Laboratories Livermore tritium decontamination systems

    International Nuclear Information System (INIS)

    Gildea, P.D.; Birnbaum, H.G.; Wall, W.R.

    1979-01-01

    Sandia Laboratories, Livermore, has put into operation a new facility, the Tritium Research Laboratory. The laboratory incorporates containment and cleanup facilities such that any tritium accidentally released is captured rather than vented to the atmosphere. This containment is achieved with hermetically sealed glove boxes that are connected on demand by manifolds to two central decontamination systems called the Gas Purification System (GPS) and the Vacuum Effluent Recovery System (VERS). The primary function of the GPS is to remove tritium and tritiated water vapor from the glove box atmosphere. The primary function of the VERS is to decontaminate the gas exhausted from the glove box pressure control systems and vacuum pumps in the building before venting the gas to the stack. Both of these systems are designed to remove tritium to the few parts per billion range. Acceptance tests at the manufacturer's plant and preoperational testing at Livermore demonstrated that the systems met their design specifications. After preoperational testing the Gas Purification System was modified to enhance the safety of maintanance operations. Both the Gas Purification System and the Vacuum Effluent Recovery System were performance tested with tritium. Results show that concentration reduction factors (ratio of inlet to exhaust concentrations) much in excess of 1000 per pass have been achieved for both systems at inlet concentrations of 1 ppM or less

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

    International Nuclear Information System (INIS)

    Iseli, M.; Esser, B.

    1989-01-01

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

  10. 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

  11. Dose mortality relationships: implications for hypothetical accidental releases from FBRs

    International Nuclear Information System (INIS)

    Kelly, G.N.

    1979-01-01

    A summary is given of the findings detailed in the most recent of a series of radiological studies of a liquid metal-cooled fast breeder reactor (Kelly, G.N. Simmonds, J.R. Smith, H. and Stather, J.W., The radiological consequences of notional accidental releases of radioactivity from fast breeder reactors: sensitivity to the dose-effect relationships adopted for early biological effects, Harwell, National Radiological Protection Board, NRPB-R87. London, HMSO (1979)). The results have indicated those areas where effort might be most profitably directed to improve the reliability of the predicted consequences. The composition of the released activity has been shown to have a marked influence on the sensitivity of the predicted consequences, and application of simple supportive medical treatment has a potential for reducing the predicted consequences by factors of up to an order of magnitude, depending on the composition of the release. (UK)

  12. 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.)

  13. Development of CROPTRIT Model: The Dynamics of Tritium in Agricultural Crops

    Energy Technology Data Exchange (ETDEWEB)

    Galeriu, Dan; Melintescu, Anca [' Horia Hulubei' National Institute for Physics and Nuclear Engineering, Department of Environmental Physics and Life, 30 Reactorului St., POB MG-6, Bucharest-Magurele, RO-077125 (Romania); Lazar, Catalin [National Agricultural Research and Development Institute Fundulea, 915200 Fundulea, Calarasi County (Romania)

    2014-07-01

    Tritium has a complex behaviour once released into the environment. Tritium can be effectively incorporated into biological systems, including the human body, as organically bound tritium (OBT) with a larger residence time than tritiated water (HTO). In the last years robust models were developed for tritium dynamics in mammals (human included), birds and fish but all of them depend on the knowledge of intake for both terrestrial or aquatic food chain. The uncertainty of the present models for tritium in crops following an accidental atmospheric release, is very high and has impacts on the engineering actions for handling and decreasing the nuclear risk. The gaps in knowledge or the local variability of key parameters were recognised as source of uncertainty. Based on an interdisciplinary approach, CROPTRIT model was gradually developed in the last decade focusing on the detecting of the uncertainty sources. Crops of interest depends on each specific case but wheat and rice cover the majority of the practical needs for radiological risk modelling (the major food in Europe and Asia). An analysis of the processes involved in the Soil-Vegetation-Atmosphere Transfer (SVAT) of tritium was done in connection with the available experimental results. The agricultural research is focused on the improving of the yield and the crop growth models were developed in relation with the genotype, weather and management of fertilisation and water. For the radiological purposes, the interest lies in the pollutant concentration at harvest and the CROPTRIT model is focused on the influence of various processes contributing to variability and uncertainty of tritium (OBT and HTO) at harvest. The current results evidentiate the role of the stomatal conductance and difficulties at the day/night transitions, as well as the complex behaviour of the maintenance respiration. A review of the experimental results demonstrates the importance of OBT formation in night conditions and difficulties

  14. Pantex Plant Cell 12-44-1 tritium release: Re-assessment of environmental doses for 1990 to 1992

    International Nuclear Information System (INIS)

    Snyder, S.F.; Hwang, S.T.

    1994-03-01

    A release of tritium gas occurred within Cell 12-44-1 at the Pantex Plant on May 17, 1989. The release was the result of a nuclear component containment failure. This document summarizes past assessments and characterization of the release. From 1990 to 1992, the average annual dose to the offsite maximally exposed individual (MEI), re-assessed using updated methods and data, ranged from 9E-6 to 2E-4 mrem/y. Doses at this level are well below the regulatory dose limit and support the discontinuation of the distinct calculation of the MEI doses from the cell's tritium releases in future Pantex Annual Site Environmental Reports. Additional information provides guidance for the evaluation of similar releases in the future. Improved Environmental Protection Department sampling plans and assessment goals will increase the value of the data collected during future incidents

  15. MARC - the NRPB methodology for assessing radiological consequences of accidental releases of activity

    International Nuclear Information System (INIS)

    Clarke, R.H.; Kelly, G.N.

    1981-12-01

    The National Radiological Protection Board has developed a methodology for the assessment of the public health related consequences of accidental releases of radionuclides from nuclear facilities. The methodology consists of a suite of computer programs which predict the transfer of activity from the point of release to the atmosphere through to the population. The suite of programs is entitled MARC; Methodology for Assessing Radiological Consequences. This report describes the overall framework and philosophy utilised within MARC. (author)

  16. The rise of hot gases resulting from hydrogen combustion at a tritium recovery plant

    International Nuclear Information System (INIS)

    Selander, W.N.

    1981-10-01

    An accidental release of hydrogen isotopes at a proposed tritium recovery plant may result in a fire or explosion. In this report estimates are given for the initial transient rise and final height of the cloud of hot gasses which results from various modes of combustion. The radiation dose equivalent caused by the downwind passage of the tritium-bearing cloud is estimated to be less than 100 mrem in any mode of combustion or weather condition. The model used for calculating the final height of the cloud depends on an entrainment assumption, and the low-density cloud loses energy by entrainment at a slower rate than in conventional atmospheric processes. Consequently, the estimated final cloud height is conservative, and, therefore, the actual radiation dose equivalent would be lower than predicted

  17. Influence of the rate of conversion of HT and HTO on projected radiation doses from release of molecular tritium

    International Nuclear Information System (INIS)

    Jacobs, D.G.; Easterly, C.E.; Phillips, J.E.

    1979-01-01

    Releases of tritium in the past have been largely in the form of tritiated water, and the projected radiation doses could be estimated by assuming tritium behaviour to parallel that of water. There is increasing interest in potential releases of tritium in the form of HT because of significant recent advances in fusion reactor research. Several recent studies have shown that bacteria containing the enzyme hydrogenase can catalyse the conversion of HT to HTO at rates several orders of magnitude faster than the rates measured in atmospheric systems. Rates of conversion in the soil have been combined with estimates of rates of permeation of HT into the soil and with global and local models depicting tritium transport and cycling. The results suggest that for the expected conversion rates, the impact on projected radiation doses should be relatively minor. (author)

  18. Control of accidental releases of hydrogen selenide in vented storage cabinets

    Science.gov (United States)

    Fthenakis, V. M.; Moskowitz, P. D.; Sproull, R. D.

    1988-07-01

    Highly toxic hydrogen selenide and hydrogen sulfide gases are used in the production of copper-indium-diselenide photovoltaic cells by reactive sputtering. In the event of an accident, these gases may be released to the atmosphere and pose hazards to public and occupational safety and health. This paper outlines an approach for designing systems for the control of these releases given the uncertainty in release conditions and lack of data on the chemical systems involved. Accidental releases of these gases in storage cabinets can be controlled by either a venturi and packed-bed scrubber and carbon adsorption bed, or containment scrubbing equipment followed by carbon adsorption. These systems can effectively reduce toxic gas emissions to levels needed to protect public health. The costs of these controls (˜0.012/Wp) are samll in comparison with current (˜6/Wp) and projected (˜I/Wp) production costs.

  19. Thermal release of {sup 3}He from tritium aged LaNi{sub 4.25}Al{sub 0.75} hydride

    Energy Technology Data Exchange (ETDEWEB)

    Staack, G.C.; Crowder, M.L.; Klein, J.E. [Savannah River National Laboratory, Aiken, SC (United States)

    2015-03-15

    The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 (LaNi{sub 4.25}Al{sub 0.75})in the tritium process to store hydrogen isotopes. The vast majority of {sup 3}He born from the radioactive decay of tritium stored in LANA.75 is trapped in the hydride metal matrix. The SRS-TF has multiple LANA.75 tritium storage beds that have been retired from service with significant quantities of He-3 trapped in the metal. To support He-3 recovery, the Savannah River National Laboratory (SRNL) conducted thermogravimetric analysis coupled with mass spectrometry (TGA-MS) on a tritium aged LANA.75 sample. TGA-MS testing was performed in an argon environment. Prior to testing, the sample was isotopically exchanged with deuterium to reduce residual tritium and passivated with air to alleviate pyrophoric concerns associated with handling the material outside of an inert glovebox. Analyses indicated that gas release from this sample was bimodal, with peaks near 220 and 490 C. degrees. The first peak consisted of both {sup 3}He and residual hydrogen isotopes, the second was primarily {sup 3}He. The bulk of the gas was released by 600 Celsius degrees. (author)

  20. Effect of Tritium on Cracking Threshold in 7075 Aluminum

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-28

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

  1. Fact Sheet: Clean Air Act Section 112(r): Accidental Release Prevention / Risk Management Plan Rule

    Science.gov (United States)

    EPA is required to publish regulations and guidance for chemical accident prevention at facilities that pose the greatest risk of harm from accidental releases of regulated flammable and toxic substances above threshold quantities.

  2. Toxic vapor concentrations in the control room following a postulated accidental release

    International Nuclear Information System (INIS)

    Wing, J.

    1979-05-01

    An acceptable method is presented for calculating the vapor concentrations in a control room as a function of time after a postulated accidental release. Included are the mathematical formulas for computing the rates of vaporization and evaporation of liquid spills, the vapor dispersion in air, and the control room air exchange. A list of toxic chemicals and their physical properties is also given

  3. 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)

  4. 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

  5. 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)

  6. Modelling accidental releases of carbon 14 in the environment: application as an excel spreadsheet

    International Nuclear Information System (INIS)

    Le Dizes, S.; Tamponnet, C.

    2004-01-01

    An application as an Excel spreadsheet of the simplified modelling approach of carbon 14 transfer in the environment developed by Tamponnet (2002) is presented. Based on the use of growth models of biological systems (plants, animals, etc.), the one-pool model (organic carbon) that was developed estimates the concentration of carbon 14 within the different compartments of the food chain and in fine the dose to man by ingestion in the case of a chronic or accidental release of carbon 14 in a river or the atmosphere. Data and knowledge have been implemented on Excel using the object-oriented programming language VisualBasic (Microsoft Visual Basic 6.0). The structure of the conceptual model and the Excel sheet are first briefly exposed. A numerical application of the model under a scenario of an accidental release of carbon 14 in the atmosphere is then presented. Simulation results and perspectives are discussed. (author)

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

    International Nuclear Information System (INIS)

    Conrad, R.; Debarberis, L.

    1991-01-01

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

  8. 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

  9. Modeling and validating tritium transfer in a grassland ecosystem in response to {sup 3}H releases

    Energy Technology Data Exchange (ETDEWEB)

    Le Dizes, S. [Institute for Radioprotection and Nuclear Safety, IRSN/PRP-ENV/SERIS/LM2E, Centre de Cadarache, Saint-Paul-lez-Durance (France); Maro, D.; Rozet, M.; Hebert, D. [IRSN/PRP-ENV/SERIS/LRC, Cherbourg-Octeville (France)

    2015-03-15

    In this paper a radioecological model (TOCATTA) for tritium transfer in a grassland ecosystem developed on an hourly time-step basis is proposed and compared with the first data set obtained in the vicinity of the AREVA-NC reprocessing plant of La Hague (France). The TOCATTA model aims at simulating dynamics of tritium transfer in agricultural soil and plant ecosystems exposed to time-varying HTO concentrations in air water vapour and possibly in irrigation and rain water. In the present study, gaseous releases of tritium from the AREVA NC nuclear reprocessing plant in normal operation can be intense and intermittent over a period of less than 24 hours. A first comparison of the model predictions with the field data has shown that TOCATTA should be improved in terms of kinetics of tritium transfer.

  10. Environmental aspects of a tritium release from the Savannah River Plant on March 23, 1984

    International Nuclear Information System (INIS)

    Evans, A.G.; Hoel, D.D.; Kantelo, M.V.

    1985-01-01

    The environmental impact of a tritium release from the Savannah River Plant (SRP) on March 23, 1984, was assessed by using both predictive and measurement techniques. Prediction of the onsite and offsite consequences by the WIND emergency response system agreed with results determined from environmental samples. The maximum calculated radiation dose to an individual at the SRP boundary was 0.06 mrem. The maximum radiation dose to a hypothetical individual consuming milk collected at the point of highest airborne tritium concentration was calculated to be 0.17 mrem. The maximum measured (by urinalysis) dose to offsite individuals in the release trajectory was 0.02 mrem. The 0.17 mrem dose corresponds to 0.09% of the 189 mrem dose that persons who live in the vicinity of SRP receive annually from natural radiation, prorated medical radiation, and fallout. Thus the public health consequences of the release were insignificant. 11 references, 15 figures, 12 tables

  11. 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)

  12. 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

  13. 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.

  14. Tritium release in Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Linjie, E-mail: zhaolinjie1989@163.com; Long, Xinggui, E-mail: xingguil@caep.cn; Peng, Shuming, E-mail: pengshuming@caep.cn; Chen, Xiaojun; Xiao, Chengjian; Ran, Guangming; Li, Jiamao

    2016-12-15

    Li{sub 4+x}Si{sub 1−x}Al{sub x}O{sub 4} solid solution materials, which were designed as the advanced tritium breeders, were obtained by indirect solid state reactions. The behaviors of tritium release from Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} powders were investigated by temperature programmed desorption. The tritium release curves show different characteristics for the Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} ceramics. The main tritium release peak in the Li{sub 4}SiO{sub 4} and Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} powders is at approximately 600 °C after a high dose irradiation. Moreover, the temperature of the tritium release from Li{sub 4.2}Si{sub 0.8}Al{sub 0.2}O{sub 4} was lower than that of the release from Li{sub 4}SiO{sub 4}. This suggests a possible advantage to using the solid solutions as the advanced tritium breeding materials.

  15. Simulation of pulsed accidental energy release in a reactor core

    International Nuclear Information System (INIS)

    Ryshanskii, V.A.; Ivanov, A.G.; Uskov, A.A.

    1995-01-01

    At the present time the strength of the load-bearing members of VVER and fast reactors during a hypothetical accident is ordinarily investigated in model experiments [1]. A power burst during an accident is simulated by a nonnuclear exothermal reaction in water, which simulates the coolant and fills the model. The problem is to make the correct choice of the simulator of the accidental energy burst as an effective (i.e., sufficiently high working capacity) source of dangerous loads, corresponding to the conditions of an accident. What factors and parameters determine the energy release? The answers to these questions are contradictory

  16. Tritium in the Savannah River Site environment

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E. Jr.; Bauer, L.R.; Hayes, D.W.; Marter, W.L.; Zeigler, C.C.; Stephenson, D.E.; Hoel, D.D.; Hamby, D.M.

    1991-05-01

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found.

  17. Tritium in the Savannah River Site environment

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.; Bauer, L.R.; Hayes, D.W.; Marter, W.L.; Zeigler, C.C.; Stephenson, D.E.; Hoel, D.D.; Hamby, D.M.

    1991-05-01

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found

  18. Improvement of the model for surface process of tritium release from lithium oxide

    International Nuclear Information System (INIS)

    Yamaki, Daiju; Iwamoto, Akira; Jitsukawa, Shiro

    2000-01-01

    Among the various tritium transport processes in lithium ceramics, the importance and the detailed mechanism of surface reactions remain to be elucidated. The dynamic adsorption and desorption model for tritium desorption from lithium ceramics, especially Li 2 O was constructed. From the experimental results, it was considered that both H 2 and H 2 O are dissociatively adsorbed on Li 2 O and generate OH - on the surface. In the first model developed in 1994, it was assumed that either the dissociative adsorption of H 2 or H 2 O on Li 2 O generates two OH - on the surface. However, recent calculation results show that the generation of one OH - and one H - is more stable than that of two OH - s by the dissociative adsorption of H 2 . Therefore, assumption of H 2 adsorption and desorption in the first model is improved and the tritium release behavior from Li 2 O surface is evaluated again by using the improved model. The tritium residence time on the Li 2 O surface is calculated using the improved model, and the results are compared with the experimental results. The calculation results using the improved model agree well with the experimental results than those using the first model

  19. FDMH - The tritium model in RODOS

    International Nuclear Information System (INIS)

    Galeriu, D.; Mateescu, G.; Melintescu, A.; Turcanu, C.; Raskob, W.

    2000-01-01

    model for the Cernavoda area in order to obtain reliable environmental assessments on the impact of tritium under accidental release condition, with a realistic site specific data base. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  2. 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)

  3. Tritium release from EXOTIC-7 orthosilicate pebbles. Effect of burnup and contact with beryllium during irradiation

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

    EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)

  4. 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.

  5. Tritium migration studies at the Nevada Test Site

    International Nuclear Information System (INIS)

    Schulz, R.K.; Weaver, M.O.

    1993-05-01

    Emanation of tritium from waste containers is a commonly known phenomenon. Release of tritium from buried waste packages was anticipated; therefore, a research program was developed to study both the rate of tritium release from buried containers and subsequent migration of tritium through soil. Migration of tritium away from low-level radioactive wastes buried in Area 5 of the Nevada Test Site was studied. Four distinct disposal events were investigated. The oldest burial event studied was a 1976 emplacement of 3.5 million curies of tritium in a shallow land burial trench. In another event, 248 thousand curies of tritium was disposed of in an overpack emplaced 6 m below the floor of a low-level waste disposal pit. Measurement of the emanation rate of tritium out of 55 gallon drums to the overpack was studied, and an annual doubling of the emanation rate over a seven year period, ending in 1990, was found. In a third study, upward tritium migration in the soil, resulting in releases in the atmosphere were observed in a greater confinement disposal test. Releases of tritium to the atmosphere were found to be insignificant. The fourth event consisted of burial of 2.2 million curies of tritium in a greater confinement disposal operation. Emanation of tritium from the buried containers has been increasing since disposal, but no significant migration was found four years following backfilling of the disposal hole

  6. The radiological consequences of notional accidental releases of radioactivity from fast breeder reactors: the influence of the meteorological conditions

    International Nuclear Information System (INIS)

    Hemming, C.R.; Hallam, J.; Kelly, G.N.

    1980-04-01

    The radiological consequences of a wide range of notional accidental releases from a 1300 MW(e) LMFBR (liquid Metal-cooled Fast Breeder Reactor) were assessed in a previous study by the National Radiological Protection Board. In the present study the sensitivity of the predicted consequences to the meteorological conditions was investigated. The influence of the wind direction, and hence the distribution of the exposed population, and of atmospheric stability at the time of the release are considered separately. Conclusions are reached on the precision required in specifying atmospheric stability conditions to estimate reliably the distribution of possible consequences following a given release of activity. The probability distributions of consequences for selected releases from two locations are evaluated taking account of the frequency distribution of wind direction and atmospheric stability at each location. The variation is large and demonstrates the importance of taking account of the whole range of meteorological conditions when assessing the risk presented by an accidental release of activity. (author)

  7. Modeling of tritium behavior in ceramic breeder materials

    International Nuclear Information System (INIS)

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

    1988-11-01

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

  8. Tritium dynamics in soils and plants at a tritium processing facility in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Mihok, S.; St-Amanat, N.; Kwamena, N.O. [Canadian Nuclear Safety Commission (Canada); Clark, I.; Wilk, M.; Lapp, A. [University of Ottawa (Canada)

    2014-07-01

    The dynamics of tritium released as tritiated water (HTO) have been studied extensively with results incorporated into environmental models such as CSA N288.1 used for regulatory purposes in Canada. The dispersion of tritiated gas (HT) and rates of oxidation to HTO have been studied under controlled conditions, but there are few studies under natural conditions. HT is a major component of the tritium released from a gaseous tritium light manufacturing facility in Canada (CNSC INFO-0798). To support the improvement of models, a garden was set up in one summer near this facility in a spot with tritium in air averaging ∼ 5 Bq/m{sup 3} HTO (passive diffusion monitors). Atmospheric stack releases (575 GBq/week) were recorded weekly. HT releases occur mainly during working hours with an HT:HTO ratio of 2.6 as measured at the stack. Soils and plants (leaves/stems and roots/tubers) were sampled for HTO and organically-bound tritium (OBT) weekly. Active day-night monitoring of air was conducted to interpret tritium dynamics relative to weather and solar radiation. The experimental design included a plot of natural grass/soil, contrasted with grass (sod) and Swiss chard, pole beans and potatoes grown in barrels under different irrigation regimes (in local topsoil at 29 Bq/L HTO, 105 Bq/L OBT). All treatments were exposed to rain (80 Bq/L) and atmospheric releases of tritium (weekdays), and reflux of tritium from soils (initial conditions of 284 Bq/L HTO, 3,644 Bq/L OBT) from 20 years of operations. Three irrigation regimes were used for barrel plants to mimic home garden management: rain only, low tritium tap water (5 Bq/L), and high tritium well water (mean 10,013 Bq/L). This design provided a range of plants and starting conditions with contrasts in initial HTO/OBT activity in soils, and major tritium inputs from air versus water. Controls were two home gardens far from any tritium sources. Active air monitoring indicated that the plume was only occasionally present for

  9. A test to enhance the excretion of tritium in man by excessive water intake

    International Nuclear Information System (INIS)

    Takada, K.; Fukuda, H.; Hattori, T.; Akaishi, J.

    1981-01-01

    A worker who accidentally inhaled tritium water vapour at a heavy-water-moderated reactor was recommended to drink an excessive amount of water to enhance the excretion of incorporated tritium. The change in biological half-life of tritium in the body was studied by measuring the tritium in urine, saliva and exhaled water. The results obtained in the present test showed that if the intake of excessive water is performed under no special control the expected shortening of tritium half-life is not necessarily achieved. When severe incorporation of tritium occurs, greater amounts of water than those in the present test must be taken. (U.K.)

  10. Two investigations concerning the release of tritium. I. Tritium leakage from 3H(Sc) EC-detectors

    International Nuclear Information System (INIS)

    Bergman, C.; Wesslen, E.

    1977-01-01

    Recently the manufacturers of EC-detectors for gas chromatographs introduced a new type of 3 H EC-detector where the tritium is bound to scandium instead of to titanium and has an activity up to 1 Ci. It is expected that the scandium-based detector will take a great part of the Swedish EC-detector market. The Swedish National Institute of Radiation Protection is anxious to make sure that the introduction of the new detector, which will be used at higher temperature, will not give rise to any increased risk of tritium intake to the personnel handling the chromatographs. The leakage of tritium from commercially available 3 H(Sc) EC-detectors containing 1 Ci of tritium was measured as a function of the detector temperature. Tritium appears both in the form of tritium gas dissolved in the scandium and in the form of tritide. The gas evaporates rather easily with increasing temperature while the dissociation of the tritide is a slower process. The evaporation of tritium due to the dissociation of the tritide was found to be negligible, less than 0.2 μCi/h at temperatures less than 100 0 C, but rises rapidly with temperature. The study also showed that even when the detector is stored at room temperature, a re-distribution of the tritium occures, from the tritide to the dissolved tritium gas, which then easily evaporates even at moderately elevated temperatures

  11. 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

  12. Uncertainties in modeling of consequences of tritium release from fusion reactors. Plasma Fusion Center No. PFC/TR-79-5

    International Nuclear Information System (INIS)

    Piet, S.J.; Kazimi, M.S.

    1979-07-01

    The bases for various models concerned with all phases of estimating doses from routine tritium releases from fusion reactors have been examined. The implications of uncertainties in parameters and assumptions for the uncertainty of the calculated doses and resulting maximum permissible releases are presented. Global dispersion models are most affected by the assumptions made concerning movement, such as the role of the ocean as a sink. Dose models were generally found to agree within a factor of two, with the largest variation due to agricultural data. Plant tritium flow studies are the least developed and require substantial improvement in the data base. Based on two possible arbitrary global standards, the maximum allowable releases were found to range from 1.6 to 20,000 Ci/day. The local criteria imply releases between 5 and 20 Ci/day

  13. 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

  14. Tritium in the DIII-D carbon tiles

    International Nuclear Information System (INIS)

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

    1993-06-01

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

  15. Tritium release from fast neutron irradiated boron carbide

    International Nuclear Information System (INIS)

    Hollenberg, G.W.

    1977-01-01

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

  16. Analysis of tritium production in the vicinity of Linac and LEB tunnels at the Superconducting Super Collider Laboratory

    International Nuclear Information System (INIS)

    Nabelssi, B.K.

    1994-01-01

    Monte Carlo calculations were performed to estimate the tritium production in groundwater around the Linear Accelerator (Linac) and the Low Energy Booster (LEB) tunnels at the Superconducting Super Collider Laboratory (SSCL). The calculations were performed using the new version of the Los Alamos High Energy Transport (LAHET) code system (SUPERHET). Most of the tritium activity was found to occur in a zone extending 2 m from the tunnel wall. The calculated tritium production rate was used to derive the. maximum allowable beam losses that would result in an average groundwater concentration in the activation zone of 20 pCi/cm 3 , the federal maximum contaminant level (MCL) for tritium in drinking water. The maximum allowable beam losses were found to be about 4% and 2% of the maximum operating be.-un for the Linac at 1 GeV and the LEB at 11 GeV, resnectively. These percentages are well in excess of typical operational losses at existing highenergy accelerators. The results are in good agreement with previously reported calculations. Tritium saturation activity in water pipes resultina, from the derived maximum allowable beam loss was found to be 355 pCi/cm 3 in the Linac operating at 600 MeV and 363 pCi/cm 3 in the LEB operating at 11 GeV. Accidental tritium releases from water pipes were found to cause an inhalation dose rate of less than 0.013 (Linac at 600 MeV) and 0.009 mrem/hr (LEB at 11 Gev) in the tunnels. These dose rates are well within the laboratory's design limit of 0.1 mrem/hr for controlled areas. Accidental beam losses were found to cause activation in excess of the MCL only after an irradiation time of more than 557 hours in the Linac at 600 MeV and 69 hours in the LEB at 11 GeV. A full-beam accident lasting more than one hour is considered unlikely

  17. 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)

  18. 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).

  19. 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

  20. 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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  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. SUBDOSA: a computer program for calculating external doses from accidental atmospheric releases of radionuclides

    International Nuclear Information System (INIS)

    Strenge, D.L.; Watson, E.C.; Houston, J.R.

    1975-06-01

    A computer program, SUBDOSA, was developed for calculating external γ and β doses to individuals from the accidental release of radionuclides to the atmosphere. Characteristics of SUBDOSA are: doses from both γ and β radiation are calculated as a function of depth in tissue, summed and reported as skin, eye, gonadal, and total body dose; doses are calculated for releases within each of several release time intervals and nuclide inventories and atmospheric dispersion conditions are considered for each time interval; radioactive decay is considered during the release and/or transit using a chain decay scheme with branching to account for transitions to and from isomeric states; the dose from gamma radiation is calculated using a numerical integration technique to account for the finite size of the plume; and the program computes and lists the normalized air concentrations at ground level as a function of distance from the point of release. (auth)

  4. Desorption of tritium and helium from high dose neutron irradiated beryllium

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Vlasov, V. V.; Kovalev, A. M.; Chakin, V. P.

    2007-08-01

    The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5-16) × 10 22 cm -2 at 70-100 °C and 380-420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250-1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312-445 °C and 500-740 °C, respectively. It is shown that most tritium (˜82%) from sample irradiated at 70-100 °C releases in temperature range of 312-700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380-420 °C, tritium release starts at a higher temperature ( Ts > Tann = 445 °C) and most of the tritium (˜99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at Tann > 600 °C and in helium release - at Tann > 750 °C.

  5. Isothermal release of tritium from neutron-irradiated Li/sub 2/O pellets

    Energy Technology Data Exchange (ETDEWEB)

    O' Hira, Shigeru; Nagao, Hiroshi; Fujii, Yasuhiko; Okamoto, Makoto

    1986-04-01

    Li/sub 2/O pellets irradiated with thermal neutrons were isothermally annealed to release tritium in a helium atmosphere at temperatures ranging from 673 to 1073 K. The release rates were found to significantly increase at elevated temperatures and to depend on the density of the Li/sub 2/O pellet. The overall diffusion coefficients of the release process were calculated using the cylindrical geometry model for the pellets as D(cm/sup 2/ s/sup -1/)=1.02 x 10/sup -3/ exp(-51.0 kJ mol/sup -1//RT)(90% theoretical density pellets), and D (cm/sup 2/ s/sup -1/)=2.64 x 10/sup -3/ exp(-46.5 kJ mol/sup -1//RT)(ca. 80% T.D. pellets) over the region 773 <= T <= 1073/sup 0/K. The result of the release experiment at 673/sup 0/K sugested that the diffusion rate was controlled by the decomposition of lithium hydroxide on the surface of Li/sub 2/O grains.

  6. 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.)

  7. Tritium in the Savannah River Site environment. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, C.E. Jr.; Bauer, L.R.; Hayes, D.W.; Marter, W.L.; Zeigler, C.C.; Stephenson, D.E.; Hoel, D.D.; Hamby, D.M.

    1991-05-01

    Tritium is released to the environment from many of the operations at the Savannah River Site. The releases from each facility to the atmosphere and to the soil and streams, both from normal operations and inadvertent releases, over the period of operation from the early 1950s through 1988 are presented. The fate of the tritium released is evaluated through environmental monitoring, special studies, and modeling. It is concluded that approximately 91% of the tritium remaining after decay is now in the oceans. A dose and risk assessment to the population around the site is presented. It is concluded that about 0.6 fatal cancers may be associated with the tritium released during all the years of operation to the population of about 625,000. This same population (based on the overall US cancer statistics) is expected to experience about 105,000 cancer fatalities from all types of cancer. Therefore, it is considered unlikely that a relationship between any of the cancer deaths occurring in this population and releases of tritium from the SRS will be found.

  8. 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

  9. 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)

  10. Nuclear power: Accidental releases - principles of public health action

    International Nuclear Information System (INIS)

    1984-01-01

    This report is based on the collective knowledge and experience of the members of a Working Group, convened by WHO in collaboration with the Government of Belgium in Brussels on 23-27 November 1981, to discuss and appraise the different actions that might be taken following accidental radioactive releases from nuclear plants. It does not provide detailed technical data, but broadly surveys the rational basis for decision-making, indicating the present position as assessed by members of the Working Group. Four major disciplines (radiological protection, health physics, environmental science and technology, and human biology) and three main professional categories (physicians, engineers and physicists) were represented, providing a comprehensive multidisciplinary approach to the topic. The purpose of this report is to give guidance to national authorities on how to develop the capacity to take action in a nuclear emergency

  11. Production sources and management of tritium generated by nuclear facilities. Works reporting from the reflection group 'tritium defense-in-depth'

    International Nuclear Information System (INIS)

    2001-01-01

    This report synthesizes current knowledge of tritium, its sources of tritium and its behavior in reactors and nuclear facilities. In addition, liquid and gaseous tritiated releases and the processes of separation and possible trapping of this radionuclide are displayed. This report also provides elements of reflection on the management of tritium produced today by these facilities, both in France and abroad. It presents insights on the relevance of current management modes of tritiated liquid and solid waste and also on the associated measures to reduce releases of tritium into the environment. (author)

  12. Tritium trapping states induced by lithium-depletion in Li{sub 2}TiO{sub 3}

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Makoto, E-mail: kobayashi.makoto@LHD.nifs.ac.jp [National Institute for Fusion Science (Japan); Oya, Yasuhisa; Okuno, Kenji [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan)

    2017-04-15

    Identifications of tritium trapping states in neutron-irradiated Li{sub 1.8}TiO{sub 2.9} (lithium-depleted Li{sub 2}TiO{sub 3}) were carried out by the out-of-pile tritium release behavior. Tritium release behaviors for neutron-irradiated Li{sub 2}TiO{sub 3} and tritium gas-exposed TiO{sub 2} were also measured for comparison. Among the tritium release spectra for these samples, three tritium release peaks were appeared. By the kinetic analyses of tritium release behaviors, the Arrhenius parameters for three peaks were evaluated. Especially for Li{sub 1.8}TiO{sub 2.9}, there were two tritium release peaks, and the peak in lower temperature region was assigned to the tritium release controlled by the diffusion process in Li{sub 2}TiO{sub 3} structure. The other tritium release peak, which was hardly appeared for Li{sub 2}TiO{sub 3}, was assigned to the release of tritium trapped as hydroxyl groups in Li{sub 1.8}TiO{sub 2.9}, indicating that lithium-depletion would result in the formation of hydroxyl groups in Li{sub 2}TiO{sub 3}. Lithium vacancies existed in Li{sub 2}TiO{sub 3} crystal structure would promote the tritium trapping as hydroxyl groups due to the decreased charge repulsion between lithium ions and tritium ion, resulting in the difficulty of recovering tritium from Li{sub 2}TiO{sub 3} effectively.

  13. 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.)

  14. Dispersion modeling of accidental releases of toxic gases - Comparison of the models and their utility for the fire brigades.

    Science.gov (United States)

    Stenzel, S.; Baumann-Stanzer, K.

    2009-04-01

    Dispersion modeling of accidental releases of toxic gases - Comparison of the models and their utility for the fire brigades. Sirma Stenzel, Kathrin Baumann-Stanzer In the case of accidental release of hazardous gases in the atmosphere, the emergency responders need a reliable and fast tool to assess the possible consequences and apply the optimal countermeasures. For hazard prediction and simulation of the hazard zones a number of air dispersion models are available. The most model packages (commercial or free of charge) include a chemical database, an intuitive graphical user interface (GUI) and automated graphical output for display the results, they are easy to use and can operate fast and effective during stress situations. The models are designed especially for analyzing different accidental toxic release scenarios ("worst-case scenarios"), preparing emergency response plans and optimal countermeasures as well as for real-time risk assessment and management. There are also possibilities for model direct coupling to automatic meteorological stations, in order to avoid uncertainties in the model output due to insufficient or incorrect meteorological data. Another key problem in coping with accidental toxic release is the relative width spectrum of regulations and values, like IDLH, ERPG, AEGL, MAK etc. and the different criteria for their application. Since the particulate emergency responders and organizations require for their purposes unequal regulations and values, it is quite difficult to predict the individual hazard areas. There are a quite number of research studies and investigations coping with the problem, anyway the end decision is up to the authorities. The research project RETOMOD (reference scenarios calculations for toxic gas releases - model systems and their utility for the fire brigade) was conducted by the Central Institute for Meteorology and Geodynamics (ZAMG) in cooperation with the Vienna fire brigade, OMV Refining & Marketing GmbH and

  15. 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

  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. Forecasting consequences of accidental release: how reliable are current assessment models

    International Nuclear Information System (INIS)

    Rohwer, P.S.; Hoffman, F.O.; Miller, C.W.

    1983-01-01

    This paper focuses on uncertainties in model output used to assess accidents. We begin by reviewing the historical development of assessment models and the associated interest in uncertainties as these evolutionary processes occurred in the United States. This is followed by a description of the sources of uncertainties in assessment calculations. Types of models appropriate for assessment of accidents are identified. A summary of results from our analysis of uncertainty is provided in results obtained with current methodology for assessing routine and accidental radionuclide releases to the environment. We conclude with discussion of preferred procedures and suggested future directions to improve the state-of-the-art of radiological assessments

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  19. Dispersion modeling of accidental releases of toxic gases - utility for the fire brigades.

    Science.gov (United States)

    Stenzel, S.; Baumann-Stanzer, K.

    2009-09-01

    Several air dispersion models are available for prediction and simulation of the hazard areas associated with accidental releases of toxic gases. The most model packages (commercial or free of charge) include a chemical database, an intuitive graphical user interface (GUI) and automated graphical output for effective presentation of results. The models are designed especially for analyzing different accidental toxic release scenarios ("worst-case scenarios”), preparing emergency response plans and optimal countermeasures as well as for real-time risk assessment and management. The research project RETOMOD (reference scenarios calculations for toxic gas releases - model systems and their utility for the fire brigade) was conducted by the Central Institute for Meteorology and Geodynamics (ZAMG) in cooperation with the Viennese fire brigade, OMV Refining & Marketing GmbH and Synex Ries & Greßlehner GmbH. RETOMOD was funded by the KIRAS safety research program of the Austrian Ministry of Transport, Innovation and Technology (www.kiras.at). The main tasks of this project were 1. Sensitivity study and optimization of the meteorological input for modeling of the hazard areas (human exposure) during the accidental toxic releases. 2. Comparison of several model packages (based on reference scenarios) in order to estimate the utility for the fire brigades. For the purpose of our study the following models were tested and compared: ALOHA (Areal Location of Hazardous atmosphere, EPA), MEMPLEX (Keudel av-Technik GmbH), Trace (Safer System), Breeze (Trinity Consulting), SAM (Engineering office Lohmeyer). A set of reference scenarios for Chlorine, Ammoniac, Butane and Petrol were proceed, with the models above, in order to predict and estimate the human exposure during the event. Furthermore, the application of the observation-based analysis and forecasting system INCA, developed in the Central Institute for Meteorology and Geodynamics (ZAMG) in case of toxic release was

  20. Summary of accidental releases of radioactivity detected off the Nevada Test Site, 1963--1986

    International Nuclear Information System (INIS)

    Patzer, R.G.; Phillips, W.G.; Grossman, R.F.; Black, S.C.; Costa, C.F.

    1988-08-01

    Of the more than 450 underground nuclear explosives tests conducted at the Nevada Test Site from August 1963 (signing of the Limited Test Ban Treaty) through the end of 1986, only 23 accidentally released radioactivity that was detectable beyond the boundary of the NTS. Of these 23, 4 were detectable off the NTS only by aircraft while the remainder were detectable by ground monitoring instruments. Since the Baneberry venting of December 1970, only two tests released radioactivity that was detectable off the NTS, and this was a seepage of radioactive noble gases. None of these releases from underground tests designed for complete containment caused exposure of the population living in the area that exceeded standards recommended by national and international radiation protection agencies. This report summarizes the releases from each of the tests, describes the monitoring that was conducted, and lists the location of the maximum exposure

  1. Modeling of tritium behavior in Li2O

    International Nuclear Information System (INIS)

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

    1992-08-01

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

  2. 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

  3. Modelling the Environmental Transfer of Tritium and Carbon-14 to Biota and Man. Report of the Tritium and Carbon-14 Working Group of EMRAS Theme 1

    International Nuclear Information System (INIS)

    2012-01-01

    -specific information as possible on land use, local soil properties and predominant plant cultivars and animal breeds; - implementing realistic growth curves for the plant cultivars of interest; - basing all sub-models on the physical approaches available for the disciplines in question. For example, knowledge from the agricultural sciences should be used to improve models for crop growth, photosynthesis and translocation; - recognizing and accounting for any unusual conditions (water stress, an uncommon cultivar or breed) in the model application. Further work in the following areas would help to improve tritium and 14 C dose assessments: - testing and improving models for the following processes: plant uptake of HTO at night and when it is raining; OBT formation in plants at night; translocation of OBT to fruit and roots; isotopic discrimination; tritium behaviour in soils following deposition from the atmosphere; and tritium behaviour in winter; - modifying the steady-state models for chronic releases to account for the fact that fluctuations in release rates and meteorological conditions result in a state of quasi- equilibrium in the environment, rather than the complete equilibrium assumed by the models; - developing a standard conceptual model for accidental tritium releases; - investigating and understanding the large OBT/HTO ratios that have been observed in soils, plants and fish under conditions that are ostensibly at equilibrium. The ten scenarios developed by the Tritium/C14 WG provide a valuable source of test data for validating environmental tritium and 14 C models.

  4. Modelling the Environmental Transfer of Tritium and Carbon-14 to Biota and Man. Report of the Tritium and Carbon-14 Working Group of EMRAS Theme 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-06-15

    of interest; - incorporating as much site-specific information as possible on land use, local soil properties and predominant plant cultivars and animal breeds; - implementing realistic growth curves for the plant cultivars of interest; - basing all sub-models on the physical approaches available for the disciplines in question. For example, knowledge from the agricultural sciences should be used to improve models for crop growth, photosynthesis and translocation; - recognizing and accounting for any unusual conditions (water stress, an uncommon cultivar or breed) in the model application. Further work in the following areas would help to improve tritium and {sup 14}C dose assessments: - testing and improving models for the following processes: plant uptake of HTO at night and when it is raining; OBT formation in plants at night; translocation of OBT to fruit and roots; isotopic discrimination; tritium behaviour in soils following deposition from the atmosphere; and tritium behaviour in winter; - modifying the steady-state models for chronic releases to account for the fact that fluctuations in release rates and meteorological conditions result in a state of quasi- equilibrium in the environment, rather than the complete equilibrium assumed by the models; - developing a standard conceptual model for accidental tritium releases; - investigating and understanding the large OBT/HTO ratios that have been observed in soils, plants and fish under conditions that are ostensibly at equilibrium. The ten scenarios developed by the Tritium/C14 WG provide a valuable source of test data for validating environmental tritium and {sup 14}C models.

  5. Improving tritium exposure reconstructions using accelerator mass spectrometry

    International Nuclear Information System (INIS)

    Love, A.H.; Hunt, J.R.; Vogel, J.S.; Knezovich, J.P.

    2004-01-01

    Direct measurement of tritium atoms by accelerator mass spectrometry (AMS) enables rapid low-activity tritium measurements from milligram-sized samples and permits greater ease of sample collection, faster throughput, and increased spatial and/or temporal resolution. Because existing methodologies for quantifying tritium have some significant limitations, the development of tritium AMS has allowed improvements in reconstructing tritium exposure concentrations from environmental measurements and provides an important additional tool in assessing the temporal and spatial distribution of chronic exposure. Tritium exposure reconstructions using AMS were previously demonstrated for a tree growing on known levels of tritiated water and for trees exposed to atmospheric releases of tritiated water vapor. In these analyses, tritium levels were measured from milligram-sized samples with sample preparation times of a few days. Hundreds of samples were analyzed within a few months of sample collection and resulted in the reconstruction of spatial and temporal exposure from tritium releases. Although the current quantification limit of tritium AMS is not adequate to determine natural environmental variations in tritium concentrations, it is expected to be sufficient for studies assessing possible health effects from chronic environmental tritium exposure. (orig.)

  6. Improving tritium exposure reconstructions using accelerator mass spectrometry

    Science.gov (United States)

    Hunt, J. R.; Vogel, J. S.; Knezovich, J. P.

    2010-01-01

    Direct measurement of tritium atoms by accelerator mass spectrometry (AMS) enables rapid low-activity tritium measurements from milligram-sized samples and permits greater ease of sample collection, faster throughput, and increased spatial and/or temporal resolution. Because existing methodologies for quantifying tritium have some significant limitations, the development of tritium AMS has allowed improvements in reconstructing tritium exposure concentrations from environmental measurements and provides an important additional tool in assessing the temporal and spatial distribution of chronic exposure. Tritium exposure reconstructions using AMS were previously demonstrated for a tree growing on known levels of tritiated water and for trees exposed to atmospheric releases of tritiated water vapor. In these analyses, tritium levels were measured from milligram-sized samples with sample preparation times of a few days. Hundreds of samples were analyzed within a few months of sample collection and resulted in the reconstruction of spatial and temporal exposure from tritium releases. Although the current quantification limit of tritium AMS is not adequate to determine natural environmental variations in tritium concentrations, it is expected to be sufficient for studies assessing possible health effects from chronic environmental tritium exposure. PMID:14735274

  7. Demonstration of an Emergency Containment System

    International Nuclear Information System (INIS)

    Flanagan, T.M.; Rogers, M.L.; Wilkes, W.R.

    1978-01-01

    A system called an Emergency Containment System (ECS) to be used for tertiary containment of tritium was reported at the 13th Air Cleaning Conference. This system was part of the Tritium Effluent Control Laboratory then under construction at Mound Facility. A series of experiments has recently been conducted to evaluate the performance of an ECS in capturing tritium accidentally released into an operating laboratory. The ECS is an automatically actuated laboratory air detritiation system utilizing a catalytic oxidation reactor and presaturated oxide adsorption/exchange columns. In the event of an accidental release of tritium into the laboratory, the ECS is automatically activated, and quick-acting pneumatic dampers divert the laboratory air supply and exhaust through the ECS until room concentrations are returned to safe operating levels. The results of the experiments have shown that a tertiary containment of tritium is feasible. In the event of a catastrophic accident, the ECS is capable of preventing the release of a large quantity of tritium to the environment

  8. The radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR

    CERN Document Server

    Kelly, G N; Charles, D; Hemming, C R

    1983-01-01

    This report contains an assessment of the radiological impact on the Greater London population of postulated accidental releases from the Sizewell PWR. Three of the degraded core accident releases postulated by the CEGB are analysed. The consequences, conditional upon each release, are evaluated in terms of the health impact on the exposed population and the impact of countermeasures taken to limit the exposure. Consideration is given to the risk to the Greater London population as a whole and to individuals within it. The consequences are evaluated using the NRPB code MARC (Methodology for Assessing Radiological Consequences). The results presented in this report are all conditional upon the occurrence of each release. In assessing the significance of the results, due account must be taken of the frequency with which such releases may be predicted to occur.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  10. Comparison of numerical models for calculating dispersion from accidental releases of pollutants

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, D W [Savannah River Lab., Aiken, SC; Cooper, R E; Baker, A J

    1982-01-01

    A modular, data-based system approach has been developed to facilitate computational simulation of multi-dimensional pollutant dispersion in atmospheric, steam, estuary, and groundwater applications. This system is used to assess effects of accidental releases of pollutants to the environment. Model sophistication ranges from simple statistical to complex three-dimensional numerical methods. The system used specifies desired degree of model sophistication from a terminal. The model used depends on the particular type of problem being solved, and on a basis of merit related to computer cost. The results of prediction for several model problems are presented.

  11. 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

  12. A study of wet deposition of atmospheric tritium releases at the Ontario Power Generation, Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Crooks, G.; DeWilde, J.; Yu, L.

    2001-01-01

    The Ontario Power Generation,Pickering Nuclear Generating Station (PNGS) has been investigating deposition of atmospheric releases of tritium on their site. This study has included numerical dispersion modelling studies conducted over the past three years, as well as an ongoing field monitoring study. The following paper will present results of the field monitoring study and make comparisons to the numerical modelling. The results of this study could be of potential use to nuclear stations in quantifying tritium deposition in near field regions where building wake effects dominate pollutant dispersion

  13. Environmental aspects of a tritium oxide release from the Savannah River Site on September 2 and 3, 1984

    International Nuclear Information System (INIS)

    Hoel, D.D.; Kurzeja, R.J.; Evans, A.G.

    1990-01-01

    Tritium was released to the atmosphere from the Savannah River Plant during an incident on September 2 and 3, 1984 between 10 PM and 3 AM. During this five hour period, 43,800 Ci of tritium, principally in the form of the oxide (HTO), was released. An additional 14,000, Ci was released during subsequent cleanup operations between September 3 and 7. The total amount released from the incident was 57,800 Ci. The HTO cloud initially moved northward and passed near the towns of New Ellenton and Aiken, SC. Two hours after the release began, the wind shifted and carried the cloud toward Columbia, SC. The cloud moved northeast during the daytime on September 3 over the east-central portion of North Carolina. Environmental sampling teams were dispatched by SRL, SRP, and SCDHEC (South Carolina Department of Health and Environmental Control). SRL collected air and vegetation samples and SRP collected vegetation, water, milk and bioassay samples. SCDHEC collected vegetation, milk, and water samples. The highest activity of HTO measured in vegetation was 501 pCi/mL onsite, 2522 pCi/mL at the plant boundary, and 9859 pCi/mL offsite. These concentrations were approximately 100 times larger than normal values. 13 refs., 7 figs., 10 tabs

  14. Effectiveness of cleanup criteria relative to an accidental nuclear release

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.

    1988-01-01

    In the event of an accidental nuclear release, the associated long-term radiological risks would result primarily from ground contamination pathways. Cleanup of the contaminated ground surfaces is a necessary step toward reducing the radiological risk to the general population. Ideally, the radiological risk decreases as the level of cleanup effort increases; however, as the cleanup criterion (i.e., the required contaminant concentration after cleanup) becomes more stringent, the cleanup effort may become prohibitively costly. This study examines several factors that are important in determining the effectiveness of the cleanup criteria for selected radionuclides: (a) annual individual dose commitment (mrem/yr), (b) total population environmental dose commitment (person-rem), and (c) total area (km 2 ) requiring cleanup following an accident. To effectively protect the general population, the benefits of cleanup should be weighed against the potentially large increase in cleanup area (and the associated costs) as the cleanup criterion becomes more stringent. The effectiveness of cleanup will vary, depending largely on site-specific parameters such as population density and agricultural productivity as well as on the amount and type of radionuclide released. Determination of an optimum cleanup criterion should account for all factors, including a comprehensive cost/benefit analysis

  15. Preliminary investigations of the significance of the ingestion pathway following accidental releases with actinides

    International Nuclear Information System (INIS)

    Steinhauer, C.

    1985-10-01

    Preliminary accident consequence assessments have been performed with the computer code UFOMOD to study the significance of the ingestion pathway in accidental releases with actinides. The investigation was based on the release category K1 of the 'Risk Oriented Analysis of the SNR 300', in which a higher fraction of actinides is released than in the worst release category for an LWR. The analysis was carried out using the currently implemented food chain transport model WASH-1400/BSU and data from the dynamic model from the MARC methodology. To study the influence of the time of the accident on the food chain-related results, releases in January and July were considered by means of the MARC data. In this report the differences are presented between both food chain transport models for transuranium elements and those which are observed in the potential doses due to ingestion, the areas affected by food-bans and the late health effects when using both models and taking the influence of the season into account. (orig./HP) [de

  16. Modelling for radiological and radioecological consequences of an accidental radionuclide release at Sea

    International Nuclear Information System (INIS)

    Paschoa, A.S.

    1997-01-01

    Full text: Scenarios concerning accidental releases of radionuclides into water bodies can be found in the open literature, mostly in connection with nuclear power plants located either onshore or inland. However, meager attention has been given to nuclear reactors used as energy sources for propulsion at sea, which are also subject to accidents. Such potential accidents may involve the loss of part of the reactor core to the surrounding water body. In addition of the initial instantaneous releases, one can estimate delayed source terms based on the rate at which radionuclides are dissolved or leached from the solidified material, like part of the core or structural materials in contact with water. Most of such solidified material might be a mixture of uranium, zirconium, iron, calcium, silica, fission and activation products, and transuranium elements as oxides, forming a glassy type solid. Transport models were used to calculate radionuclide concentrations in water resulting from short and delayed source terms. Oceanographic data used in the calculations were taken either from the open literature or from unclassified reports of the Brazilian Navy, being, however, as generic as possible. Time-dependent concentration functions for radionuclides in aquatic food following an accidental release reflect the net result of intake and elimination processes. However, to avoid the complexities of multiple parameters involved in such processes, the model accounts only for trophic transfer of radionuclides, and yet avoids the necessity of analyzing the details of each transfer step used to determine fish, crustacea, molluscs and seaweed accumulation. Swimming and other aquatic sports are not included in the model used for dose calculations because of theirs relatively low importance in comparison with the pathways concerning ingestion of aquatic food

  17. Contaminative influence of beef due to the inhalation of air and the ingestion of soil of cattle in an accidental release of radioactive materials

    International Nuclear Information System (INIS)

    Hwang, W. T.; Kim, E. H.; Seo, K. S.; Jung, H. J.; Lee, S. M.; Hang, M. H.

    2004-01-01

    The contaminative influence of beef due to the inhalation of air and the ingestion of soil of cattle, both of which are dealt with as minor contaminative pathways in most radioecological models but may not be neglected, was comprehensively investigated with the improvement of the Korean dynamic food chain model DYNACON. As the results, it was found that both pathways can not be neglected at all in the contamination of beef in the case of an accidental release during the non-grazing period of cattle. The ingestion of soil was more influential in the contamination of beef than the inhalation of air over most time following an release. If precipitation is encountered during an accidental release, contaminative influence due to the ingestion of soil was far greater compared with the cases of no precipitation. This fact was more distinct for a long-lived radionuclide 137 Cs than a short-lived radionuclide 131 I (elemental iodine). Compared with the results for milk performed prior to this study, the contaminative pathways due to the inhalation of air and the ingestion of soil were more important in beef because of longer biological half-lives. In the meantime, in the case of an accidental release during the grazing period of cattle, radioactive contamination due to the ingestion of pasture was dominant irrespective of the existence of precipitation during an accidental release. It means that contaminative influence due to the inhalation of air and the ingestion of soil is negligible like the cases of milk

  18. 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....

  19. The estimation of areas of ground that may be contaminated after an accidental release of pollutant to the atmosphere

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1979-01-01

    A method is developed for calculating the area of ground contaminated above a prescribed level after an accidental release of radioactivity or any other pollutant to the atmosphere. Numerical calculations are made for a wide range of releases, atmospheric conditions and rates of wet and dry deposition. It is shown that high atmospheric stability and rain both tend to maximize the area of significant contamination for most of the plausible range of releases. However, for very large hypothetical releases, dry conditions with an unstable atmosphere spread significant contamination furthest afield. (author)

  20. Modelization of tritium transfer into the organic compartments of algae

    International Nuclear Information System (INIS)

    Bonotto, S.; Gerber, G.B.; Arapis, G.; Kirchmann, R.

    1982-01-01

    Uptake of tritium oxide and its conversion into organic tritium was studied in four different types of algae with widely varying size and growth characteristics (Acetabularia acetabulum, Boergesenia forbesii, two strains of Chlamydomonas and Dunaliella bioculata). Water in the cell and the vacuales equilibrates rapidly with external tritium water. Tritium is actively incorporated into organically bound form as the organisms grow. During the stationary phase, incorporation of tritium is slow. There exists a discrimination against the incorporation of tritium into organically bound form. A model has been elaborated taking in account these different factors. It appears that transfer of organic tritium by algae growing near the sites of release would be significant only for actively growing algae. Algae growing slowly may, however, be useful as cumulative indicators of discontinuous tritium release. (author)

  1. Partitioning of perfluorooctanesulfonate and perfluorohexanesulfonate in the aquatic environment after an accidental release of aqueous film forming foam at Schiphol Amsterdam Airport

    NARCIS (Netherlands)

    Kwadijk, C.J.A.F.; Kotterman, M.J.J.; Koelmans, A.A.

    2014-01-01

    In the summer of 2008, an accidental release of Aqueous Film Forming Foam (AFFF) took place at Schiphol Amsterdam Airport (The Netherlands). After the release, water, fish and sediment samples were collected and analyzed for perfluoroalkyl sulfonates (PFSA). In situ perfluorooctane sulfonate (PFOS)

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

    International Nuclear Information System (INIS)

    Muller, M.E.

    1978-01-01

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

  3. International conference and workshop on modeling and mitigating the consequences of accidental releases of hazardous materials

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    This conference was held September 26--29, 1995 in New Orleans, Louisiana. The purpose of this conference was to provide a multidisciplinary forum for exchange of state-of-the-art information on the consequences of accidental releases of hazardous materials. Attention is focused on air dispersion of vapors. Individual papers have been processed separately for inclusion in the appropriate data bases

  4. Biokinetic aspects of tissue-bound tritium in algae

    International Nuclear Information System (INIS)

    Strack, S.; Kistner, G.

    1978-01-01

    For the estimate of the radiation exposure of man and for the calculation of the risk of artificial tritium from nuclear power plants, organic tissue-bound tritium is of decisive importance. In model experiments, a tritium incorporation of 61 to 71% was found from tritiated water (HTO) into organic matter of planctonic algae under reproducible conditions and this was related to the theoretical value. In further experiments the tritium release from these high tritiated algae was of interest. Kept in darkness in tritium-free, non-sterile river water, so that autolytic processes and bacterial decomposition could occur, the concentration of HTO was measured over a period of three weeks. A relatively long half-life of tissue-bound tritium was found under various temperature conditions. Therefore it must be considered that a significant retention of tritium in biological matter has to be taken into account in a natural ecosystem. In streams into which the cooling water of a nuclear reactor is released all conditions are found already for a long turnover and cycling of artificial tritium in living organisms as well as the conditions for a favourable transport of tritium by food chains to man. (Auth.)

  5. Distribution of tritium in a chronically contaminated lake

    International Nuclear Information System (INIS)

    Blaylock, B.G.; Frank, M.L.

    1978-01-01

    White Oak Lake located on the U.S. Department of Energy's Oak Ridge Reservation receives a continuous input of tritium from operating facilities and waste disposal operations at the Oak Ridge National Laboratory. The purpose of this paper was (1) to determine the distribution and concentration of tritium in an aquatic environment which has received releases of tritium significantly greater than expected releases from nuclear power plants, and (2) to determine the effect of fluctuating tritium concentrations in ambient water on the concentration of tritium in fish. Aquatic biota from White Oak Lake were analyzed for tissue water tritium and tissue bound tritium. Except for one plant species, the ratio of tissue water tritium to lake water tritium ranged from 0.80 to 1.02. The tissue water tritium in Gambusia affinis, the mosquito fish, followed closely the significant changes in tritium concentration in lake water. The turnover of tissue water tritium was very rapid; Gambusia from White Oak Lake eliminated 50% of their tissue water tritium in 14 min. The ratio of the specific activity of the tissue bound tritium to the specific activity of the lake water was greatest for the larger species of fish but never exceeded unity. The radiation dose to man from tritium which could be acquired through the aquatic food chain was relatively small when compared to other pathways. The whole body dose to a hypothetical individual taking in concentrations of tritium measured in White Oak Lake was 1.8 mrem/yr from eating fish and 10.0 mrem/yr from drinking water

  6. Dependency of irradiation damage density on tritium migration behaviors in Li2TiO3

    International Nuclear Information System (INIS)

    Kobayashi, Makoto; Toda, Kensuke; Oya, Yasuhisa; Okuno, Kenji

    2014-01-01

    Tritium migration behaviors in Li 2 TiO 3 with the increase of irradiation damage density were investigated by means of electron spin resonance and thermal desorption spectroscopy. The irradiation damages of F + -centers and O − -centers were formed by neutron irradiation, and their damage densities were increased with increasing neutron fluence. Tritium release temperature was clearly shifted toward higher temperature side with increasing neutron fluence, i.e. increasing damage density. The rate determining process for tritium release was also clearly changed depending on the damage density. Tritium release was mainly controlled by tritium diffusion process in crystalline grain of Li 2 TiO 3 at lower neutron fluence. The apparent tritium diffusivity was reduced as the damage density in Li 2 TiO 3 increased due to the introduction of tritium trapping/detrapping sites for diffusing tritium. Then, tritium trapping/detrapping processes began to control the overall tritium release with further damage introductions as the amount of tritium trapping sites increased enough to trap most of tritium in Li 2 TiO 3 . The effects of water vapor in purge gas on tritium release behaviors were also investigated. It was considered that hydrogen isotopes in purge gas would be dissociated and adsorbed on the surface of Li 2 TiO 3 . Then, hydrogen isotopes diffused inward Li 2 TiO 3 would occupy the tritium trapping sites before diffusing tritium reaches to these sites, promoting apparent tritium diffusion consequently. Kinetics analysis of tritium release for highly damaged Li 2 TiO 3 showed that the rate determining process of tritium release was the detrapping process of tritium formed as hydroxyl groups. The rate of tritium detrapping as hydroxyl groups was determined by the kinetic analysis, and was comparable to tritium release kinetics for Li 2 O, LiOH and Li 4 TiO 4 . The dangling oxygen atoms (O − -centers) formed by neutron irradiation would contribute strongly on the

  7. Estimate of the absorbed dose in the mouse organs and tissues after tritium administration

    International Nuclear Information System (INIS)

    Saito, Masahiro

    2000-01-01

    Chronic and accidental release of tritium from future fusion facilities may cause some extent of hazardous effect to the public health. Various experiments using small animals such as mice have been performed to mimic the dose accumulation due to tritium intake by the human body. An difficulty in such animal experiments using small animals is that it is rather difficult to administer tritium orally and estimate the dose to small organs or tissues. In the course of our study, a simple method to administer THO and T-labeled amino acids orally to the mouse was dictated and dose accumulation in various organs and tissues was determined. The tritium retention in the bone marrow was also determined using the micro-centrifuge method. Throughout our experiment, colony-bred DDY mice were used. The 8-10 week old male mice were orally and intraperitoneally administered THO water or T-amino acids mixture solution. For the purpose of oral administration, a 10 μl aliquot of T-containing saline solution was placed on the tongue of the mice using an automatic micropipette. At various times after tritium administration, the animals were sacrificed and the amount of tritium in various tissues and organs including bone marrow was examined. Dose accumulation pattern after THO intake and T-amino acids was compared between intraperitoneal injection and oral administration. The accumulated dose after oral administration of THO exhibited a tendency to be 10-20% higher than after intraperitoneal injection. The bone marrow dose after oral intake of THO was found to be lower than the doses to urine, blood, liver and testis. In contrast, the blood dose gave a conservative estimate for the dose to the other tissues and organs. (author)

  8. Tritium waste control: April-September 1982

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  9. Safety analysis of tritium processing system based on PHA

    International Nuclear Information System (INIS)

    Fu Wanfa; Luo Deli; Tang Tao

    2012-01-01

    Safety analysis on primary confinement of tritium processing system for TBM was carried out with Preliminary Hazard Analysis. Firstly, the basic PHA process was given. Then the function and safe measures with multiple confinements about tritium system were described and analyzed briefly, dividing the two kinds of boundaries of tritium transferring through, that are multiple confinement systems division and fluid loops division. Analysis on tritium releasing is the key of PHA. Besides, PHA table about tritium releasing was put forward, the causes and harmful results being analyzed, and the safety measures were put forward also. On the basis of PHA, several kinds of typical accidents were supposed to be further analyzed. And 8 factors influencing the tritium safety were analyzed, laying the foundation of evaluating quantitatively the safety grade of various nuclear facilities. (authors)

  10. Dose estimation models for environmental tritium released from fusion facilities

    International Nuclear Information System (INIS)

    Murata, Mikio

    1993-01-01

    Various mathematical models are being developed to predict the behavior of HT released to the natural environment and their consequent impact. This report outlines models and the major findings of HT field release studies in France and Canada. The models are constructed to incorporate the key processes thought to be responsible for the formation of atmospheric HTO from a release of HT. It has been established from the experiments that HT oxidized in surface soil is incorporated almost entirely into soil water as HTO. This tritium may be reemitted to the atmosphere in the form of HTO through exchange of soil and atmospheric moisture as well as through the bulk water mass flux from the soil the atmosphere due to evaporation and transpiration. The direct conversion of HT to HTO in air and direct uptake of HT by vegetation are expected to be negligible for the time and space scales of interest in considering short duration releases. HTO emitted to the atmosphere is can further exchange with soil and vegetation water. Validation of these models against experimental data is conducted to demonstrate their credibility. It may be concluded that further laboratory and field works are needed in order to develop a sufficiently good understanding of the dependence of the key processes on environmental factors (including diurnal cycling and seasonality) to allow the rates of the processes to be predicted from a knowledge of environmental conditions. (author)

  11. A database on tritium behavior in the chronic HT release experiment. 1. Meteorological data and tritium concentrations in air and soil

    Energy Technology Data Exchange (ETDEWEB)

    Noguchi, Hiroshi; Yokoyama, Sumi; Kinouchi, Nobuyuki; Murata, Mikio; Amano, Hikaru; Ando, Mariko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Fukutani, Satoshi

    1999-03-01

    This report comprises a database that can be used to develop and validate tritium models to assess doses to the general public due to HT continuously released from fusion facilities into the atmosphere. The data was collected in the 1994 chronic HT release experiment carried out at the Chalk River Laboratories in Canada. The data set include meteorological conditions such as solar radiation, net solar radiation, wind speed, air temperature and humidity, soil temperature and soil heat flux; soil conditions such as bulk density, water content and free pore volume fraction; HT and HTO concentrations in air, HTO concentrations in soil moisture and HTO deposition to water surface. Evapo-transpiration rates and turbulent diffusivity are estimated and tabulated. The report also contains experimental methods to observe meteorological conditions and take air and soil samples. (author)

  12. 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

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  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. Impacts of accidental radioactive releases to the hydrosphere from floating and land-based nuclear power plants

    International Nuclear Information System (INIS)

    Vollmer, R.H.

    1978-01-01

    To assess the impacts of accidental radioactive releases to the hydrosphere from floating and land-based nuclear power plants, the Liquid Pathway Generic Study was initiated. The objective of this study was to compare the risks associated with releases through the liquid pathway from accidents that may reasonably be expected and those that are very improbable. Consequences were estimated in terms of radiation dose to man from drinking water ingestion, shell-fish and fish flesh consumption, and direct exposure (swimming, beach) as well as long-term effects, such as genetic effects or aquatic species degradation. (author)

  16. 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.

  17. 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.

  18. Tritium liquid effluents from the Krsko NPP

    International Nuclear Information System (INIS)

    Savli, S.; Krizman, M.; Nemec, T.; Cindro, M.; Stritar, A.; Vokal Nemec, B.; Janzekovic, H.

    2007-01-01

    In the past, 12-months' fuel cycles in the Krsko NPP had not caused any problems regarding compliance with its Technical Specifications and license limits on liquid tritium releases (20 TBq/year, 8 TBq/three months). The first 18-months' fuel cycle, which was introduced in 2004, required fuel with higher enrichment, higher boron concentration in the primary coolant and more fuel rods with burnable poisons. In 2005, the NPP operated without refueling outage for the whole year and produced the highest amount of energy so far. Due to these facts and a few unplanned shutdowns and power reductions, production of tritium and releases increased strongly in 2005. As a result, the Krsko NPP hardly succeeded to stay within regulatory limits on tritium releases. However, the three-months' limit was exceeded in the first quarter of 2006. On the basis of conclusions acquired from the SNSA's study and practice of other European countries the SNSA considerably increased the annual limit of permitted liquid tritium releases (from 20 TBq to 45 TBq) and abolished the three-months' limit. At the same time, the SNSA reduced the limit of fission and activation products by halves. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  20. 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)

  1. Tritium and radon risks for humans

    International Nuclear Information System (INIS)

    Mauna, Traian; Mauna, Andriesica

    2008-01-01

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

  2. An updated review on tritium in the environment.

    Science.gov (United States)

    Eyrolle, Frédérique; Ducros, Loïc; Le Dizès, Séverine; Beaugelin-Seiller, Karine; Charmasson, Sabine; Boyer, Patrick; Cossonnet, Catherine

    2018-01-01

    Various studies indicated more or less recently that organically bound tritium (OBT) formed from gaseous or liquid tritium releases into the environment potentially accumulates in organisms contradicting hypotheses associated to methods used to assess the biological impact of tritium on humans (ASN, 2010). Increasing research works were then performed during the last decade in order to gain knowledge on this radionuclide expected to be increasingly released by nuclear installations in the near future within the environment. This review focusses on publications of the last decade. New unpublished observations revealing the presence of technogenic tritium in a sedimentary archive collected in the upper reaches of the Rhône river and findings from the Northwestern Mediterranean revealing in all likelihood the impact of terrigenous tritium inputs on OBT levels recorded in living organisms are also presented. Identifying and understanding the physicochemical forms of tritium and the processes leading to its persistence in environmental compartments would explain most observations regarding OBT concentrations in organisms and definitively excludes that tritium would "bio accumulate" within living organisms. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. 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)

  4. Tritium: An analysis of key environmental and dosimetric questions

    International Nuclear Information System (INIS)

    Till, J.E.; Meyer, H.R.; Etnier, E.L.; Bomar, E.S.; Gentry, R.D.; Killough, G.G.; Rohwer, P.S.; Tennery, V.J.; Travis, C.C.

    1980-05-01

    This document summarizes new theoretical and experimental data that may affect the assessment of environmental releases of tritium and analyzes the significance of this information in terms of the dose to man. Calculated doses resulting from tritium releases to the environment are linearly dependent upon the quality factor chosen for tritium beta radiation. A reevaluation of the tritium quality factor by the ICRP is needed; a value of 1.7 would seem to be more justifiable than the old 1.0 value. A new exposure model is proposed, based primarily upon the approach recommended by the National Council on Radiation Protection and Measurements. Employing a /open quotes/typical/close quotes/ LMFBR reprocessing facility source term, a /open quotes/base case/close quotes/ dose commitment to total body (for a maximally exposed individual) was calculated to be 4.0 /times/ 10/sup /minus/2/ mSv, with 3.2 /times/ 10/sup /minus// mSv of the dose due to intake of tritium. The study analyzes models which exist for evaluating the buildup of global releases of tritium from man-made sources. Scenarios for the release of man-made tritium to the environment and prediction of collective dose commitment to future generations suggest that the dose from nuclear weapons testing will be less than that from nuclear energy even though the weapons source term is greater than that for any of our energy scenarios

  5. Tritium waste control: April--September 1977

    International Nuclear Information System (INIS)

    1978-01-01

    A pilot scale system was used in an initial experiment to investigate the combined-electrolysis-catalytic-exchange process (CECE) for the detritiation of water. Data taken during the experiment indicate the process does indeed strip tritium from gaseous hydrogen at the top and concentrate it in water at the bottom of the catalyst-filled column. A high activity tritiated water electrolysis system was designed and built using a solid polymer electrolyte (SPE) cell. The system was successfully operated at currents up to 50 A using deionized tap water. Triplicate samples of cement, cement-plaster (1:1 ratio by weight), and cement-plaster (1:1 ratio by volume) were injected with 386 Ci of tritium. Preliminary results indicate Type III Portland cement retains the tritium slightly better than the cement-plaster mixtures. The tritium release study of actual waste burial packages is continuing. The fractional release is 1 x 10 -5 on a 4-y old package, only 4 x 10 -7 on 3-y old packages, and 1 x 10 -9 on a 1-y old package. Pressure increase and gas composition determinations were repeated for octane (activity = 1000 Ci/liter) with and without tritium fixation using argon as the initial overgas. Pressure buildup measurements for octane without fixation were repeated a third time using hydrogen gas. The rate of pressure increase did not change significantly from previously determined values. Four elemental tritium samples were released into a laboratory to determine the efficiency of the Emergency Containment System. The ventilation system was modified during the fourth experiment to minimize leakage

  6. Nuclear graphite waste's behaviour under disposal conditions: Study of the release and repartition of organic and inorganic forms of carbon 14 and tritium in alkaline media

    International Nuclear Information System (INIS)

    Vende, L.

    2012-01-01

    23000 tons of graphite wastes will be generated during dismantling of the first generation of French reactors (9 gas cooled reactors). These wastes are classified as Long Lived Low Level wastes (LLW-LL). As requested by the law, the French National Radioactive Waste Management Agency (Andra) is studying concepts of low-depth disposals.In this work we focus on carbon 14, the main long-lived radionuclide in graphite waste (5730 y), but also on tritium, which is the main contributor to the radioactivity in the short term. Carbon 14 and tritium may be released from graphite waste in many forms in gaseous phase ( 14 CO 2 , HT...) or in solution ( 14 CO 3 2- , HTO...). Their speciation will strongly affect their migration from the disposal site to the environment. Leaching experiments, in alkaline solution (0.1 M NaOH simulating repository conditions) have been performed on irradiated graphite, from Saint-Laurent A2 and G2 reactors, in order to quantify their release and characterize their speciation. The studies show that carbon 14 exists in both gaseous and aqueous phases. In the gaseous phase, release is weak (≤0.1%) and corresponds to oxidizable species. Carbon 14 is mainly released into liquid phase, as both inorganic and organic species. 65% of released fraction is inorganic and 35% organic carbon. Two tritiated species have been identified in gaseous phase: HTO and HT/Organically Bond Tritium. More than 90% of tritium in that phase corresponds to HT/OBT. But release is weak (≤0.1%). HTO is mainly in the liquid phase. (author)

  7. Comparison of the regulatory models assessing off-site radiological dose due to the routine releases of tritium

    International Nuclear Information System (INIS)

    Hwang, W. T.; Kim, E. H.; Han, M. H.; Choi, Y. H.; Lee, H. S.; Lee, C. W.

    2005-01-01

    Methodologies of NEWTRIT model, NRC model and AIRDOS-EPA model, which are off-site dose assessment models for regulatory compliance from routine releases of tritium into the environment, were investigated. Using the domestic data, if available, the predictive results of the models were compared. Among them, recently developed NEWTRIT model considers only doses from organically bounded tritium (OBT) due to environmental releases of tritiated water (HTO). A total dose from all exposure pathways predicted from AIRDOS-EPA model was 1.03 and 2.46 times higher than that from NEWTRIT model and NRC model, respectively. From above result, readers should not have an understanding that a predictive dose from NRC model may be underestimated compared with a realistic dose. It is because of that both mathematical models and corresponding parameter values for regulatory compliance are based on the conservative assumptions. For a dose by food consumption predicted from NEWTRIT model, the contribution of OBT was nearly equivalent to that of HTO due to relatively high consumption of grains in Korean. Although a total dose predicted from NEWTRIT model is similar to that from AIRDOS-EPA model, NEWTRIT model may be have a meaning in the understanding of phenomena for the behavior of HTO released into the environment

  8. Review of decontamination and clean-up techniques for use in the UK following accidental releases of radioactivity to the environment

    International Nuclear Information System (INIS)

    Brown, J.; Cooper, J.R.; Jones, J.A.; Flaws, L.; McGeary, R.; Spooner, J.

    1996-10-01

    This review examines decontamination and clean-up methodologies that may be applied following an accidental release of radionuclides to the atmosphere in the UK. An accidental release of radionuclides to the environment could cause contamination of land and property, and may necessitate movement of the resident population to reduce their exposure. Decontamination procedures may reduce the level of contamination and, in determining whether to implement such procedures in an affected area, it is necessary to weigh the effectiveness of the techniques against the associated economic costs and other disadvantages. The report gives a review of available methods of decontamination and clean-up in terms of the effectiveness of the various methods to remove activity from building or land surfaces and dose saved. It also considers associated disadvantages, costs, clean-up rates for land and buildings, waste disposal implications, and personnel and resource requirements. On the basis of these factors, the techniques are prioritised and overall strategies for decontamination following releases of both fission products and alpha emitters are presented. (author)

  9. Tritium handling facilities at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  10. Uptake of tritium through foliage in capsicum fruitescens, L

    International Nuclear Information System (INIS)

    Iyengar, T.S.; Sadarangani, S.H.; Vaze, P.K.; Soman, S.D.

    1977-01-01

    Tritium uptake and release patterns throuogh foliage in Capsicum fruitescens, L. were investigated using twelve potted plants, under different conditions of exposure and release. The plants studied belonged to two age groups, 3 months and 5 months. The average half residence time for the species was found to be 42.6 min, on the basis of treating the entire group of plants as a single cluster. The individual release rates showed a variation of up to a factor of two, for half residence time values (Tsub(1/2)). The second component was not easily resolvable in most of the cases. Tissue bound tritium showed interesting uptake patterns. The ratios between tissue bound tritium and tissue free water tritium concentrations indicated regular mode of uptake with well defined rate constants in the case of long exposure periods. (author)

  11. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-15

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

  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. Tritium monitoring in the environment of the French territory

    Energy Technology Data Exchange (ETDEWEB)

    Leprieur, F.; Roussel-Debet, S.; Pierrard, O.; Tournieux, D.; Boissieux, T.; Caldera-Ideias, P. [Institut de radioprotection et de surete nucleaire (France)

    2014-07-01

    Introduction: Radioactive releases in the environment from civilian and military nuclear facilities have significantly decreased over the last few decades, except for discharges of tritium which are forecast to increase due to changes in the fuel management in power plants and in the longer term by new tritium-emitting units (fusion reactors). In the aim to perform its radiological monitoring mission throughout the French territory, IRSN uses and develops advanced technology equipment to sample and to analyze tritium in the different environmental compartments. Methodology: IRSN uses bubblers to collect both tritium vapour (HTO) and gaseous tritium (mainly HT) in the air. Another method, developed by IRSN, consists in directly sampling the water vapour in the air by condensing in a cold trap and more recently with passive sampler. In continental and marine surface water, samples are usually collected by automatic water samplers. Instantaneous surface water samples are also collected by grab sample devices. In addition, IRSN conducts animal and plant samples near French nuclear facilities. Natural origin and tritium remaining from testing of nuclear weapons In the atmosphere, the background levels of tritium of 1 to 2 Bq/L measured in water vapour, equivalent to an activity of 0.01 to 0.02 Bq/m{sup 3} of air. In fresh waters, the tritium activity currently ranges between 1 and 3 Bq/L of water. In the marine environment, tritium emitted during nuclear weapon tests has been totally 'diluted' in cosmogenic tritium and concentration levels at the surface have remained around 0.1 to 0.2 Bq/L. In biological matrices, total tritium concentration range from 1 to 3 Bq/kg f.w. with a variable proportion of free and organically bounded forms. Tritium around nuclear facilities: Close to facilities releasing more than 2x10{sup 13} Bq/year of gaseous tritium, higher activity levels, ranging from a few tens to a few hundred Bq/L, are observed in the atmospheric and

  15. Research of CITP-II tritium production irradiation device design

    International Nuclear Information System (INIS)

    Zhang Zhihua; Deng Yongjun; Mi Xiangmiao; Li Rundong; Liu Zhiyong

    2012-01-01

    As the core component of CITP-II, the online tritium production irradiation device is the pivotal equipment in the research on tritium production and release of tritium breeders. The design of CITP-II online tritium production irradiation device creatively makes replacing the breeders online come true; as tritium production capacity, the self-shielding factor of device, and neutron flux were studied. The influence of different load models and load thicknesses of breeders to tritium production capacity was calculated. The hydrodynamics parameters of device in solid-gas phase were computed. Thermal parameters, such as the heat power of breeders, hotspot, temperature grads distributions, utmost temperature, uneven factors, were analyzed. Creatively designed nonlinear electric heater equalized breeders' even heat power. The influence laws of the components, pressure of gap gas and carrier gas to the balance temperature were got. And the key thermal parameters were ascertained. The key thermal parameters and the changing laws were got and provide the basis for structural optimization and safety analysis. They can also be referenced for the study of breeders' tritium production and release. (authors)

  16. Temporal sealing material of tritium-contaminated stainless steel

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  17. 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.)

  18. Tritium in the burial ground of the Savannah River Site

    International Nuclear Information System (INIS)

    Hyder, M.L.

    1993-06-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  20. Tritium dynamics in soils and plants grown under three irrigation regimes at a tritium processing facility in Canada

    International Nuclear Information System (INIS)

    Mihok, S.; Wilk, M.; Lapp, A.; St-Amant, N.; Kwamena, N.-O.A.; Clark, I.D.

    2016-01-01

    The dynamics of tritium released from nuclear facilities as tritiated water (HTO) have been studied extensively with results incorporated into regulatory assessment models. These models typically estimate organically bound tritium (OBT) for calculating public dose as OBT itself is rarely measured. Higher than expected OBT/HTO ratios in plants and soils are an emerging issue that is not well understood. To support the improvement of models, an experimental garden was set up in 2012 at a tritium processing facility in Pembroke, Ontario to characterize the circumstances under which high OBT/HTO ratios may arise. Soils and plants were sampled weekly to coincide with detailed air and stack monitoring. The design included a plot of native grass/soil, contrasted with sod and vegetables grown in barrels with commercial topsoil under natural rain and either low or high tritium irrigation water. Air monitoring indicated that the plume was present infrequently at concentrations of up to about 100 Bq/m"3 (the garden was not in a major wind sector). Mean air concentrations during the day on workdays (HTO 10.3 Bq/m"3, HT 5.8 Bq/m"3) were higher than at other times (0.7–2.6 Bq/m"3). Mean Tissue Free Water Tritium (TFWT) in plants and soils and OBT/HTO ratios were only very weakly or not at all correlated with releases on a weekly basis. TFWT was equal in soils and plants and in above and below ground parts of vegetables. OBT/HTO ratios in above ground parts of vegetables were above one when the main source of tritium was from high tritium irrigation water (1.5–1.8). Ratios were below one in below ground parts of vegetables when irrigated with high tritium water (0.4–0.6) and above one in vegetables rain-fed or irrigated with low tritium water (1.3–2.8). In contrast, OBT/HTO ratios were very high (9.0–13.5) when the source of tritium was mainly from the atmosphere. TFWT varied considerably through time as a result of SRBT's operations; OBT/HTO ratios showed no clear

  1. 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

  2. Surface desorption and bulk diffusion models of tritium release from Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Avila, R.E., E-mail: ravila@cchen.c [Departamento de Materiales Nucleares, Comision Chilena de Energia Nuclear, Cas. 188-D, Santiago (Chile); Pena, L.A.; Jimenez, J.C. [Departamento de Produccion y Servicios, Comision Chilena de Energia Nuclear, Cas. 188-D, Santiago (Chile)

    2010-10-30

    The release of tritium from Li{sub 2}TiO{sub 3} and Li{sub 2}ZrO{sub 3} pebbles, in batch experiments, is studied by means of temperature programmed desorption. Data reduction focuses on the analysis of the non-oxidized and oxidized tritium components in terms of release limited by diffusion from the bulk of ceramic grains, or by first or second order surface desorption. By analytical and numerical methods the in-furnace tritium release is deconvoluted from the ionization chamber transfer functions, for which a semi-empirical form is established. The release from Li{sub 2}TiO{sub 3} follows second order desorption kinetics, requiring a temperature for a residence time of 1 day (T{sub 1dRes}) of 620 K, and 603 K, of the non-oxidized, and the oxidized components, respectively. The release from Li{sub 2}ZrO{sub 3} appears as limited by either diffusion from the bulk of the ceramic grains, or by first order surface desorption, the first possibility being the more probable. The respective values of T{sub 1dRes} for the non-oxidized component are 661 K, according to the first order surface desorption model, and 735 K within the bulk diffusion limited model.

  3. Assessment of the Public Health impact from the accidental release of UF6 at the Sequoyah Fuels Corporation Facility at Gore, Oklahoma (Docket No. 40-8027). Volume 2

    International Nuclear Information System (INIS)

    1986-03-01

    Following the accidental release of UF 6 from the Sequoyah Fuels Facility on January 4, 1986, an Ad Hoc Interagency Public Health Assessment Task Force was established. The Task Force consists of technical staff members from various agencies who have prepared this assessment of the public health impact associated with the accidental release. Volume 2 of the report contains Appendices which provide more detailed information used in the assessment and support the discussion in Volume 1

  4. 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

  5. Use of passive sampling for atmospheric tritium monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Caldeira Ideias, P.; Pierrard, O.; Tournieux, D. [Institut de Radioprotection et de Surete Nucleaire - IRSN (France); Tenailleau, L. [Marine nationale (France)

    2014-07-01

    Tritium is one of the most important radionuclide in environmental radiological monitoring. In French civil and military nuclear facilities, the releases levels are between 100 to 100 000 higher than any other radionuclide (rare gas excluded). Moreover these levels will probably increase in the next decades. With an average energy of 6 keV, the beta particle from tritium radioactive decay is difficult to detect and quantify within the environmental levels. To monitor the tritium in the air, French actors (authorities, operator, and experts) commonly use atmospheric bubblers and water vapour condensers. This type of sampling approach is time-consuming and very costly. To simplify and complete these methods, the Institute for Radiological Protection and Nuclear Safety (IRSN), had developed an atmospheric tritium monitoring device based on passive sampling. The passive sampler developed consists in a small container designed with a patented specific geometry and filled with 13X molecular sieve. This system is based on free diffusion flow principle (Fick's law). The driving force is the partial pressure gradient existing between the environmental atmosphere and the passive sampler. The constancy of the sampling rate for different moisture conditions assures the representativeness of the proposed device. The desorption bench developed specifically allows the recovery of 99% of the water vapour sampled in the molecular sieve. More than 99% of the sampled tritium (HTO) activity is recovered in the range between 0 and 100 Bq.L{sup -1}. Above 100 Bq.L{sup -1} to 25 k Bq.L{sup -1} (max tested activity), it was verified that no more than 3% of the tritium remains in the molecular sieve.. Thus, the use of passive sampler provides: - a representative sampling method, - a good detection limit (0,01 Bq.m{sup -3}), - no electric power supply needs, - a wide range of sampling duration (1 day to 1 month), - a low-cost method for monitoring. Different performance tests were

  6. Preparation of Tritium from irradiated lithium compounds (study)

    International Nuclear Information System (INIS)

    1989-01-01

    The aim of the present study is the selection of a certain scheme for release, collection, measurement and analysis of Tritium as produced in Lithium compounds (Li 2 O, LiOH, Li 2 CO 3 LiF and some Lithium alloys) in accordance with the 6 Li(n, α)T reaction. Tritium technology is of vital concern to power reactor programmes as well as to fusion technology. Meanwhile the fields of activity include, tritium generation and mangement in fission and fusion reactors; enviromental studies, release modeling HT/HTO conversion and dose assessments, absorption/ deabsorption, monitoring and plant design; research and development, labeling, compatibility and physical chemical properties

  7. Preparation of Tritium from irradiated lithium compounds (study)

    Energy Technology Data Exchange (ETDEWEB)

    1989-01-01

    The aim of the present study is the selection of a certain scheme for release, collection, measurement and analysis of Tritium as produced in Lithium compounds (Li[sub 2]O, LiOH, Li[sub 2]CO[sub 3] LiF and some Lithium alloys) in accordance with the [sup 6]Li(n, [alpha])T reaction. Tritium technology is of vital concern to power reactor programmes as well as to fusion technology. Meanwhile the fields of activity include, tritium generation and mangement in fission and fusion reactors; enviromental studies, release modeling HT/HTO conversion and dose assessments, absorption/ deabsorption, monitoring and plant design; research and development, labeling, compatibility and physical chemical properties.

  8. Non-labile tritium in Savannah River Plant pine trees

    International Nuclear Information System (INIS)

    Sanders, S.M. Jr.

    1976-06-01

    Non-labile tritium bound in cellulose of pine trees was measured to learn about the effects and fate of tritium contributed to the environment by the Savannah River Plant (SRP). An estimation of the regional inventory and the distance tritium can be observed from SRP was desired because tritium is a major component of the radioactivity released by SRP, and as the oxide, it readily disperses in the environment

  9. EXOTIC: Development of ceramic tritium breeding materials

    International Nuclear Information System (INIS)

    Kwast, H.; Conrad, R.

    1989-09-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  12. A uranium bed with ceramic body for tritium storage

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  13. 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.)

  14. 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

  15. 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

  16. Consideration of disposal alternatives for tritium-contaminated wastewater streams at Hanford

    International Nuclear Information System (INIS)

    Waters, E.D.

    1988-03-01

    Small quantities of tritium are produced as an undesirable by-product of the operation of light-water reactors. At the US Department of Energy Hanford Site in Washington State, some tritium has been discharged to the environment in low-level liquid and gaseous wastes from the N Reactor plant, but more than 97% of the tritium stays typically within the irradiated fuel as it is delivered for reprocessing. During fuel reprocessing, the tritium is distributed in the process streams, and most of the tritium is presently released to the soil column with excess process condensates from the Plutonium-Uranium Extraction (PUREX) Plant. On an annual basis, approximately 1 g of tritium is discharged in more than 1 x 10 6 L of process condensate water. Principal tritium release points and quantities are presented in section 4.0. The present study is intended to identify and evaluate alternate methods of tritium control and disposal that might merit additional study or development for potential application to Hanford Site effluents. 30 refs., 15 figs., 5 tabs

  17. A new blanket tritium recovery experiment with intense DT neutron source at JAEA/FNS

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Kentaro, E-mail: ochiai.kentaro@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Edao, Yuki [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Kawamura, Yoshinori [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ohta, Masayuki; Kwon, Saerom; Konno, Chikara [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan)

    2016-11-01

    Highlights: • For detail investigation of the tritium recovery performance on the fusion reactor blanket, we have started a new blanket tritium recovery experiment with ionization chamber at JAEA/FNS. • A new improved container was provided for the appropriate tritium measurement by IC and also utilized for the enhancement of TPR in the new container. The TPR was calculated with a calculation code MCNP5 and some typical nuclear data libraries and then the radioactivity of the tritium recovery with LSC corresponded with that of calculation. • The tritium release curves by the IC outputs are similar to those by the LSC output. However, it was indicated that the quantitative measurement by IC needed further improvement for the tritium recovery. - Abstract: We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

  18. MISTRAL: A comprehensive model for tritium transport in lithium-base ceramics. Pt. 2

    International Nuclear Information System (INIS)

    Federici, G.; Raffray, A.R.; Abdou, M.A.

    1990-01-01

    A new tritium transport model called MISTRAL (Model for Investigative Studies of Tritium Release in Lithium Ceramics) has been developed to describe and predict the kinetics of tritium release in lithium ceramic materials for tritium breeding applications in fusion blankets. The model has transient capabilities and has been developed to analyze the full range of transient conditions produced in in-pile tritium recovery experiments and expected in fusion blankets. Calibration of the model against experiments has been done in parallel with its development in order to assess its predictive capabilities and to identify the ranges of potential applicability. The comparisons of the results available for lithium metasilicate and aluminate samples irradiated respectively in the two in-pile tritium recovery experiments LISA1 and MOZART are presented and discussed in this paper. They have been selected for the calibration of the codes as being good examples of various features relevant for tritium release analysis in ceramic breeders under different transient conditions such as change in temperature, purge gas composition and reactor power. (orig.)

  19. Tritium accident containment within a large fusion enclosure: cost, benefit, and risk considerations

    International Nuclear Information System (INIS)

    Sherwood, A.E.

    1983-01-01

    Containment of a tritium accident within a large fusion device building will be difficult and costly. Complete containment is impossible, and with this fact in mind, the global dispersion and health effects of tritium are reviewed. Atmospheric tritium emissions lead to an estimated population dose to the Northern Hemisphere of 5.6 x 10 - 3 man-rem/Ci, which may also be interpreted as 1 cancer fatality per MCi. Updating the NRC $1000 per man-rem criterion to 1982 costs gives 9.5 $/y per Ci/y as the unit annual health benefit rate from averting tritium release at a continuous rate. Present worth considerations lead to an estimate of $100 per Ci/y for the maximum capital investment justified per expected curie per year of tritium release averted. A simplified enclosure model is used to explore the trade-off between processing capacity and recycle time with the health cost of residual tritium release included in the analysis

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

    International Nuclear Information System (INIS)

    Powell, J.R.

    1975-01-01

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

  1. TRIF - an intermediate approach to environmental tritium modelling

    International Nuclear Information System (INIS)

    Higgins, N.A.

    1997-01-01

    The movement of tritium through the environment, from an initial atmospheric release to selected end points in the food chain, involves a series of closely coupled and complex processes which are, consequently, difficult to model. TRIF (tritium transfer into food) provides a semi-empirical approach to this transport problem, which can be adjusted to bridge the gap between simple steady state approximations and a fully coupled model of tritium dispersion and migration (Higgins et al., 1996). TRIF provides a time-dependent description of the behaviour of tritium in the form of tritium gas (HT) and tritiated water (HTO) as it enters and moves through the food chain into pasture, crops and animals. This includes a representation of the production and movement of organically bound tritium (OBT). (Author)

  2. Tritium storage metal-bed pyrophoricity measurements

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  3. Plutonium in the environment: key factors related to impact assessment in case of an accidental atmospheric release

    Energy Technology Data Exchange (ETDEWEB)

    Guetat, P. [CEA Valduc, 21 - Is-sur-Tille (France); Moulin, V.; Reiller, P. [CEA Saclay, 91 (FR)] (and others)

    2009-07-01

    This paper deals with plutonium and key factors related to impact assessment. It is based on recent work performed by CEA which summarize the main features of plutonium behaviour from sources inside installations to the environment and man, and to report current knowledge on the different parameters used in models for environmental and radiological impact assessment. These key factors are illustrated through a case study based on an accidental atmospheric release of Pu in a nuclear facility. (orig.)

  4. Tritium in organic matter around Krsko Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kristof, Romana; Zorko, Benjamin; Kozar Logar, Jasmina; Kosenina, Suzana

    2017-01-01

    The aim of the research was to obtain first results of tritium in the organic matter of environmental samples in the vicinity of Krsko NPP. The emphasis was on the layout of suitable sampling network of crops and fruits in nearby agricultural area. Method for determination of tritium in organic matter in the form of Tissue Free Water Tritium (TFWT) and Organically Bound Tritium (OBT) has been implemented. Capabilities of the methods were tested on real environmental samples and its findings were compared to modeled activities of tritium from atmospheric releases and literature based results of TFWT and OBT. (author)

  5. 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.

  6. Migration of tritium from a nuclear waste burial site

    International Nuclear Information System (INIS)

    Hawkins, R.H.

    1975-09-01

    The Savannah River Plant (SRP) has routinely and continuously monitored the local environment (land, water, air, flora, and fauna) since 1951. As part of this intensive program, a three-part study was made to assess the tritium migration from an onsite burial ground for solid nuclear wastes and the resulting dose-to-man. A major source of tritium is buried, massive, Li-Al residues (referred to as melts) from the thermal extraction step in the SRP tritium production process. A melt with its extraction crucible and lid were immersed in water to measure the amounts of tritium released as HTO and HT to the water and to air. The result was a rapid release of 23 curies, of which approximately 99 percent was HTO that remained in the immersion water, and 1 percent was HT that passed into the air. (auth)

  7. Tritium Removal by Laser Heating and Its Application to Tokamaks

    International Nuclear Information System (INIS)

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

    2001-01-01

    A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm 2 , and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  9. Computer programs at SRL to evaluate environmental effects SRP operations and postulated accidental releases

    International Nuclear Information System (INIS)

    Cooper, R.E.

    1975-09-01

    Savannah River Plant operations unavoidably result in the release of some chemical and radioactive effluents to the environs. The most environmentally significant releases are gaseous effluents to the atmosphere; computer codes dealing with these atmospheric releases are discussed in this report. There is a wide variety of effluents, both chemical and radioactive, to be considered, and each must be correlated with meteorological dispersion data as a function of time to estimate the environmental effects. In addition, large inventories of toxic and radioactive materials in some facilities represent a potential for accidental releases. Accidents are postulated for these facilities, and the environmental effects of resulting releases are again evaluated by correlating with meteorological dispersion data. In accordance with AEC Regulatory Guide 23, a 2-year meteorological data base is used in performing all analyses. Due to the diversity of possible releases and the large meteorological data base, the environmental analyses are necessarily performed with the aid of a large computer facility. Several computer programs have been written to facilitate these analyses according to the type of analysis desired. The computer programs described in this report are basically of three categories: probability distributions of estimated concentrations or doses as a function of distance from a point of origin, estimates of average concentrations or doses over a specified time period such as annual averages, and some miscellaneous programs in support of the first two categories to optimize the use of the computing facility. A complete documentation of each program is included with a program listing and sample input-output

  10. 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)

  11. 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%

  12. 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

  13. Tritium Management In HCLL-PPCS Model AB Blanket

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. Tritium release from beryllium pebbles after high temperature irradiation up to 3000 appm He in the HIDOBE-01 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Til, S. van, E-mail: vantil@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Fedorov, A.V.; Stijkel, M.P.; Cobussen, H.L.; Mutnuru, R.K.; Idsert, P. van der [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and The Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

    2013-11-15

    In the HIDOBE (HIgh DOse irradiation of BEryllium) irradiation program, various grades of constrained and unconstrained beryllium pebbles, beryllium pellets and titanium-beryllide samples are irradiated in the High Flux Reactor (HFR) in Petten at four different temperatures (between 698 K and 1023 K) for 649 days [1]. The first of two HIDOBE irradiation experiments, HIDOBE-01, was completed after achieving a DEMO relevant helium production level of 3000 appm and the samples are retrieved for postirradiation examination (PIE). This work shows preliminary results of the out-of-pile tritium release analysis performed on different grades of irradiated beryllium pebbles (different in size). Relationships between irradiation temperature, tritium inventory and microstructural evolution have been observed by light microscopy and scanning electron microscopy.

  15. Implantation measurements to determine tritium permeation in first wall structures

    International Nuclear Information System (INIS)

    Holland, D.F.; Causey, R.A.

    1983-01-01

    A principal safety concern for a D-T burning fusion reactor is release of tritium during routine operation. Tritium implantation into first wall structures, and subsequent permeation into coolants, is potentially an important source of tritium loss. This paper reports on an experiment in which an ion accelerator was used to implant deuterium atoms in a stainless steel disk to simulate tritium implantation in first wall structures. The permeation rate was measured under various operating conditions. These results were used in the TMAP computer code to determine potential tritium loss rates for fusion reactors

  16. Implantation measurements to determine tritium permeation in first-wall structures

    International Nuclear Information System (INIS)

    Holland, D.F.; Causey, R.A.; Sattler, M.L.

    1983-01-01

    A principal safety concern for a D-T burning fusion reactor is release of tritium during routine operation. Tritium implantation into first-wall structures, and subsequent permeation into coolants, is potentially an important source of tritium loss. This paper reports on an experiment in which an ion accelerator was used to implant deuterium atoms in a stainless steel disk to simulate tritium implantation in first-wall structures. The permeation rate was measured under various operating conditions. These results were used in the TMAP computer code to determine potential tritium loss rates for fusion reactors

  17. 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)

  18. An application of cost-effectiveness analysis to restrict the damage caused by an accidental release of radioactive material to the environment

    International Nuclear Information System (INIS)

    Frittelli, L.; Tamburrano, A.

    1980-01-01

    When an accidental release of radioactive material occurs health effects in the exposed population can be mitigated by remedial actions applied to individuals or their environment. A cost-effectiveness analysis is performed by comparing the cost of remedial action with the monetary value of the collective dose avoided. (H.K.)

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  20. Analysis of tritium migration and deposition in fusion-reactor systems

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-08-01

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

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

    International Nuclear Information System (INIS)

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

    1983-10-01

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

  3. Experimental and modelling studies on the exposure of wall surfaces to tritium gas in ambient room conditions

    International Nuclear Information System (INIS)

    Housiadas, C.; Douglas, K.

    1995-01-01

    An experimental set-up is used to carry out static exposure tests to study the uptake and oxidation of tritium released in ambient room air, routinely or accidentally, in the presence of selected surface materials. Tritium, in its elemental form at concentrations of the order of approx.0.4 GBq/m 3 (10 -2 Ci/m 3 ), is injected into the glass exposure chamber containing the selected surface material and air at atmospheric pressure. Periodically, samples of the chamber atmosphere are analysed, using liquid scintillation counting, to obtain the concentrations of HTO and T 2 . The exposures have been performed using aluminium, 316L stainless steel and painted stainless steel plates, as the selected surface materials. Results are compared with predictions using the ITER approved TMAP4 code. The results indicate practically the same conversion rate, of about 0.02% per day, for both the aluminium and stainless steel samples and give reasonable agreement with modelling predictions. Strong absorption of both T 2 and HTO by the painted surface is observed, suggesting the use of high values for the solubility constant to correctly predict this behaviour. 19 refs., 5 figs., 1 tab

  4. Synthesis of tritium measurement results performed by IRSN and CEA from November 5, 2010, to February 1, 2012, in the environment of the 2M Process company building at Saint-Maur-des-Fosses; Synthese des resultats des mesures de tritium effectuees par l'IRSN et le CEA du 5 novembre 2010 au 1er fevrier 2012 dans l'environnement de l'entreprise 2M Process a Saint-Maur-des-Fosses

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-02-22

    A building of 2M Process company at Saint-Maur-des-Fosses (France) has been accidentally contaminated after the handling of a tritium contaminated device. The consequence has been a tritium release inside and in the close environment of the building. Since November 5, on request of the French nuclear safety authority (ASN), the French institute of radiation protection and nuclear safety (IRSN) has performed tritium measurements in water, air and plants sampled in the vicinity of the building in order to delimit the contaminated area and to follow its decay during and after the cleansing of the building. From March 16, 2011 onward, the IRSN's monitoring has been lightened as an answer to the complementary monitoring implemented by the CEA in the framework of the environment monitoring plan ordered by the Prefecture. The results published in previous reports have shown a strong contamination of the building and the presence of significant amounts of tritium in plant leaves and superficial waters in a 50 m radius around the 2M Process building. This note makes a synthesis of the tritium concentrations measured in plant leaves, surface waters, air, and rainwater sampled near the 2M Process building since November 5, 2010, and up to February 1, 2012. (J.S.)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

    tritium release into the environment. The contribution of the OBT to population ingestion dose is important and must be considered in models for dose assessment. (authors)

  6. Tritium and helium-3 in metals

    International Nuclear Information System (INIS)

    Lasser, R.

    1989-01-01

    The book surveys recent results on the behaviour of tritium and its decay product helium-3 metals. In contrast to many earlier books which discuss the properties of the stable hydrogen isotopes without mentioning tritium, this book reviews mainly the results on tritium in metals. Due to the difficulties in preparing metal tritide samples, very important quantities such as diffusivity, superconductivity, solubility, etc. have only been determined very recently. The book not only presents the measured tritium data, but also the isotopic dependency of the different physical properties by comparing H, D and T results. A chapter is devoted to helium-3 in metals. Aspects such as helium release, generation of helium bubbles, swelling, and change of the lattice parameter upon aging are discussed. The book provides the reader with up-to-date information and deep insight into the behaviour of H, D, T and He-3 in metals. Further important topics such a tritium production, its risks, handling and discharge to the environment are also addressed

  7. Tritium releases from the Pickering Nuclear Generating Station and birth defects and infant mortality in nearby communities 1971-1988

    International Nuclear Information System (INIS)

    Johnson, K.C.; Rouleau, J.

    1991-10-01

    This study was commissioned to examine whether there were elevated rates of stillbirth, birth defects, or death in the first year of life between 1971 and 1988 among offspring of residents of communities within a 25-kilometre radius of the Pickering Nuclear Generating Station. The study was also to investigate whether there were any statistical associations between the monthly airborne or waterborne tritium emissions from the Pickering Nuclear Generating Station and the rates of these reproductive outcomes. Overall analysis did not support a hypothesis of increased rates of stillbirths, neonatal mortality or infant mortality near the Pickering Nuclear Generating Station, or a hypothesis of increased birth prevalence of birth defects for 21 of 22 diagnostic categories. The prevalence of Down Syndrome was elevated in both Pickering and Ajax; however, there was no consistent pattern between tritium release levels and Down Syndrome prevalence, chance could not be ruled out for the associations between Down Syndrome and tritium releases or ground-monitored concentrations, the association was detected in an analysis where multiple testing was done which may turn up significant associations by change, and maternal residence at birth and early in pregnancy needs to be verified. The association between Down Syndrome and low-level radiation remains indeterminate when existing evidence from epidemiological studies is summed. The estimated radiation exposure from the nuclear plant for residents of Pickering and Ajax is lower by a factor of 100 than the normal natural background radiation. Further study is recommended. (21 tabs., 29 figs., 5 maps, 37 refs.)

  8. 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.)

  9. 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.

  10. 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

  11. Effect of hydrophobic paints coating for tritium reduction in concrete materials

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  12. Spatial distribution of tritium in the Rawatbhata Rajasthan site environment

    International Nuclear Information System (INIS)

    GilI, Rajpal; Tiwari, S.N.; Gocher, A.K.; Ravi, P.M.; Tripathi, R.M.

    2014-01-01

    Tritium is one of the most environmentally mobile radionuclides and hence has high potential for migration into the different compartments of environment. Tritium from nuclear facilities at RAPS site is released into the environment through 93 m and 100 m high stack mainly as tritiated water (HTO). The released tritium undergoes dilution and dispersion and then follows the ecological pathway of water molecule. Environmental Survey Laboratory of Health Physics Division, Bhabha Atomic Research Centre (BARC), located at Rajasthan Atomic Power Station (RAPS) site is continuously monitoring the concentration of tritium in the environment to ensure the public safety. Atmospheric tritium activity during the period (2009-2013) was measured regularly around Rajasthan Atomic Power Station (RAPS). Data collected showed a large variation of H-3 concentration in air fluctuating in the range of 0.43 - 5.80 Bq.m -3 at site boundary of 1.6 km. This paper presents the result of analyses of tritium in atmospheric environment covering an area up to 20 km radius around RAPS site. Large number of air moisture samples were collected around the RAPS site, for estimating tritium in atmospheric environment to ascertain the atmospheric dispersion and computation of radiation dose to the public

  13. Aquatic dispersion modelling of a tritium plume in Lake Ontario

    International Nuclear Information System (INIS)

    Klukas, M.H.; Moltyaner, G.L.

    1996-05-01

    Approximately 2900 kg of tritiated water, containing 2.3E+15 Bq of tritium, were released to Lake Ontario via the cooling water discharge when a leak developed in a moderator heat exchanger in Unit 1 at the Pickering Nuclear Generating Station (PNGS) on 1992 August 2. The release provided the opportunity to study the dispersion of a tritium plume in the coastal zone of Lake Ontario. Current direction over the two-week period following the release was predominantly parallel to the shore, and elevated tritium concentrations were observed up to 20 km east and 85 km west of the PNGS. Predictions of the tritium plume movement were made using current velocity measurements taken at 8-m depth, 2.5 km offshore from Darlington and using a empirical relationship where alongshore current speed is assumed to be proportional to the alongshore component of the wind speed. The tritium migration was best described using current velocity measurements. The tritium plume dispersion is modelled using the one-dimensional advection-dispersion equation. Transport parameters are the alongshore current speed and longitudinal dispersion coefficient. Longitudinal dispersion coefficients, estimated by fitting the solution of the advection-dispersion equation to measured concentration distance profiles ranged from 3.75 to 10.57 m 2 s -1 . Simulations using the fitted values of the dispersion coefficient were able to describe maximum tritium concentrations measured at water supply plants located within 25 km of Pickering to within a factor of 3. The dispersion coefficient is a function of spatial and temporal variability in current velocity and the fitted dispersion coefficients estimated here may not be suitable for predicting tritium plume dispersion under different current conditions. The sensitivity of the dispersion coefficient to variability in current conditions should be evaluated in further field experiments. (author). 13 refs., 7 tabs., 12 figs

  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. 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

  16. Parametric study of the Ignalina reactor building capability as barrier against accidental releases of radioactivity

    International Nuclear Information System (INIS)

    Blomquist, R.; Johansson, Kjell; Nilsson, Lars.

    1993-01-01

    The results of a parametric study are offered to the Ignalina plant management staff and to the Lithuanian and Swedish nuclear inspectorates as a basis for a decision whether there is mutual interest in a project for the purpose of strengthening the Ignalina reactor buildings inherent capabilities to provide a barrier against accidental releases of radioactivity. Practical measures to consider are: * establish natural convection of warm air from the steam drums to the tall stack of 150 m height. * reduce the resulting draught of air through the reactor hall floor between the fuel channel shield blocks into the steam drum compartments. * apply filtration to the stack air flow. 18 refs

  17. ANCLI's conclusions and recommendations made after the ANCLI colloquium 'Tritium, discrete, but present everywhere'

    International Nuclear Information System (INIS)

    Sene, M.

    2009-01-01

    The authors briefly state the conclusions of the colloquium about the presence of tritium in the environment, its sanitary impact, the re-examination of a management based on release, the need to reduce tritium production. The recommendations are also indicated: to continue researches on organically bound tritium, not to allow any release increase as long as effects are not better known. The role of the ANCLI is outlined

  18. Description of NORMTRI: a computer program for assessing the off-site consequences from air-borne releases of tritium during normal operation of nuclear facilities

    International Nuclear Information System (INIS)

    Raskob, W.

    1994-10-01

    The computer program NORMTRI has been developed to calculate the behaviour of tritium in the environment released into the atmosphere under normal operation of nuclear facilities. It is possible to investigate the two chemical forms tritium gas and tritiated water vapour. The conversion of tritium gas into tritiated water followed by its reemission back to the atmosphere as well as the conversion into organically bound tritium is considered. NORMTRI is based on the statistical Gaussian dispersion model ISOLA, which calculates the activity concentration in air near the ground contamination due to dry and wet deposition at specified locations in a polar grid system. ISOLA requires a four-parametric meteorological statistics derived from one or more years synoptic recordings of 1-hour-averages of wind speed, wind direction, stability class and precipitation intensity. Additional features of NORMTRI are the possibility to choose several dose calculation procedures, ranging from the equations of the German regulatory guidelines to a pure specific equilibrium approach. (orig.)

  19. Tritium dynamics in soils and plants grown under three irrigation regimes at a tritium processing facility in Canada.

    Science.gov (United States)

    Mihok, S; Wilk, M; Lapp, A; St-Amant, N; Kwamena, N-O A; Clark, I D

    2016-03-01

    The dynamics of tritium released from nuclear facilities as tritiated water (HTO) have been studied extensively with results incorporated into regulatory assessment models. These models typically estimate organically bound tritium (OBT) for calculating public dose as OBT itself is rarely measured. Higher than expected OBT/HTO ratios in plants and soils are an emerging issue that is not well understood. To support the improvement of models, an experimental garden was set up in 2012 at a tritium processing facility in Pembroke, Ontario to characterize the circumstances under which high OBT/HTO ratios may arise. Soils and plants were sampled weekly to coincide with detailed air and stack monitoring. The design included a plot of native grass/soil, contrasted with sod and vegetables grown in barrels with commercial topsoil under natural rain and either low or high tritium irrigation water. Air monitoring indicated that the plume was present infrequently at concentrations of up to about 100 Bq/m(3) (the garden was not in a major wind sector). Mean air concentrations during the day on workdays (HTO 10.3 Bq/m(3), HT 5.8 Bq/m(3)) were higher than at other times (0.7-2.6 Bq/m(3)). Mean Tissue Free Water Tritium (TFWT) in plants and soils and OBT/HTO ratios were only very weakly or not at all correlated with releases on a weekly basis. TFWT was equal in soils and plants and in above and below ground parts of vegetables. OBT/HTO ratios in above ground parts of vegetables were above one when the main source of tritium was from high tritium irrigation water (1.5-1.8). Ratios were below one in below ground parts of vegetables when irrigated with high tritium water (0.4-0.6) and above one in vegetables rain-fed or irrigated with low tritium water (1.3-2.8). In contrast, OBT/HTO ratios were very high (9.0-13.5) when the source of tritium was mainly from the atmosphere. TFWT varied considerably through time as a result of SRBT's operations; OBT/HTO ratios showed no clear temporal

  20. Fate of gaseous tritium and carbon-14 released from buried low-level radioactive waste

    International Nuclear Information System (INIS)

    Striegl, R.G.

    1988-01-01

    Microbial decomposition, chemical degradation, and volatilization of buried low-level radioactive waste results in the release of gases containing tritium ( 3 H) and carbon-14 ( 14 C) to the surrounding environment. Water vapor, carbon dioxide, and methane that contain 3 H or 14 C are primary products of microbial decomposition of the waste. Depending on the composition of the waste source, chemical degradation and volatilization of waste also may result in the production of a variety of radioactive gases and organic vapors. Movement of the gases in materials that surround waste trenches is affected by physical, geochemical, and biological mechanisms including sorption, gas-water-mineral reactions, isotopic dilution, microbial consumption, and bioaccumulation. These mechanisms either may transfer 3 H and 14 C to solids and infiltrating water or may result in the accumulation of the radionuclides in plant or animal tissue. Gaseous 3 H or 14 C that is not transferred to other forms is ultimately released to the atmosphere

  1. Tritium clouds environmental impact in air into the Western Mediterranean Basin evaluation

    International Nuclear Information System (INIS)

    Castro, P.; Velarde, M.; Ardao, J.; Perlado, J.M.; Sedano, L.

    2012-01-01

    The paper considers short-term releases of tritium (mainly but not only tritium hydride (HT)) to the atmosphere from a potential ITER-like fusion reactor located in the Mediterranean Basin and explores if the short range legal exposure limits are exceeded (both locally and downwind). For this, a coupled Lagrangian ECMWF/FLEXPART model has been used to follow real time releases of tritium. This tool was analyzed for nominal tritium operational conditions under selected incidental conditions to determine resultant local and Western Mediterranean effects, together with hourly observations of wind, to provide a short-range approximation of tritium cloud behavior. Since our results cannot be compared with radiological station measurements of tritium in air, we use the NORMTRI Gaussian model. We demonstrate an overestimation of the sequence of tritium concentrations in the atmosphere, close to the reactor, estimated with this model when compared with ECMWF/FLEXPART results. A Gaussian “mesoscale” qualification tool has been used to validate the ECMWF/FLEXPART for winter 2010/spring 2011 with a database of the HT plumes. It is considered that NORMTRI allows evaluation of tritium-in-air-plume patterns and its contribution to doses.

  2. UFOING: A program for assessing the off-site consequences from ingestion of accidentally released radionuclides

    International Nuclear Information System (INIS)

    Steinhauer, C.

    1988-12-01

    The program UFOING estimates foodchain-related consequences following accidental releases of radionuclides to the atmosphere. It was developed as a stand-alone supplement to the accident consequence assessment program system UFOMOD to allow faster and more detailed investigations of the consequences arising from the foodchain pathways than possible with the version of UFOING which is implemented in UFOMOD. For assumed releases at different times of the year, age dependent individual doses, collective doses, individual risks for fatal stochastic somatic health effects as a function of time, the total numbers of the effects, and the areas affected by foodbans together with the estimated duration of the bans are calculated. In addition, percentage contributions of radionuclides and foodstuffs to the doses and risks can be evaluated. In the first part of this report, an overview over the program is given. The other parts contain a user's guide, a program guide, and descriptions of the data employed and of the version of UFOING which is implemented in UFOMOD. (orig.) [de

  3. Control of tritium permeation through fusion reactor strucural materials

    International Nuclear Information System (INIS)

    Maroni, V.A.

    1978-01-01

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

  4. 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)

  5. Development of a dynamic compartment model for the prediction of tritium behavior around NPPs

    International Nuclear Information System (INIS)

    Choi, Heui-Joo; Lee, Hansoo; Kang, Hee Suk; Choi, Yong Ho

    2003-01-01

    KAERI has developed a new model to find the relationship between the tritium release rate and tritium concentration in the environment. The model was based upon a dynamic compartment model. In this paper three kinds of global tritium cycling model were compared to estimate the natural background concentration of tritium in Korea. The dry and wet deposition rates were calculated using a computer program called DEPOS to derive a source term. The mechanisms considered for the transfer of tritium between the compartments were evaporation, groundwater flow, infiltration, runoff, and hydrodynamic dispersion. Also, transfer coefficients between the compartments were obtained using realistic geographical data. In order to illustrate the model various release scenarios were developed, and the change of tritium concentration in groundwater and surface water around the nuclear power plants was estimated. (author)

  6. Basic study of influence of radiation defects on tritium release processes from lithium silicates

    Energy Technology Data Exchange (ETDEWEB)

    Abramenkovs, A.; Tiliks, J.; Kizane, G.; Supe, A. [Latvia Univ., Riga (Latvia). Dept. of Chem.; Grishmanovs, V. [Department of Quantum Engineering and System Science, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113 (Japan)

    1997-09-01

    The radiolysis of Li{sub 2}SiO{sub 3} and Li{sub 4}SiO{sub 4} was studied using the chemical scavengers method (CSM), thermoluminescence, lyoluminescence, electron spin resonance and spectrometric methods. The influence of the absorbed dose and many another parameters such as: irradiation conditions, sample preparation conditions and concentration of impurities on the accumulation rate of each type RD and RP were studied. Several possibilities for reducing the radiolysis of silicates were discussed. It has been found that tritium localization on the surface and in grains proceed by two different mechanisms. Tritium thermoextraction from the surface proceeds as chemidesorption of tritiated water, but from the bulk as diffusion. The tritium retention processes were studied. It has been found that tritium retention depends on irradiation conditions. Tritium retention is due to the formation of chemical bonds Li-T and thermal stable {identical_to}Si-T bonds. The accumulation of colloidal silicon and lithium can increase the tritium retention up to 25-35%. (orig.).

  7. Tritium Mitigation/Control for Advanced Reactor System

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-03-31

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

  8. Tritium waste management on the La Hague AREVA NC site: associated impact and monitoring

    International Nuclear Information System (INIS)

    Devin, P.; Deguette, H.

    2009-01-01

    The authors propose an analysis of tritium behaviour in the nuclear fuel processed in the AREVA NC plant in La Hague, of its presence in the plant and in its wastes, and of the impact of these wastes and the tritium monitoring in the environment. First, they present the AREVA NC plant and evoke the legal context concerning the waste management. They report and discuss the analysis of the presence and behaviour of tritium in irradiated fuel, of its behaviour during spent fuel processing, the evolution of tritium releases (legal limitations, evolutions since 1992), of measurement of activity in effluents, and discuss a study of possible reductions of tritium releases by La Hague plants (mainly in sea waters). They also report the computational assessment of the dosimetric impact of tritium on neighbouring population. They describe how the presence of tritium in the environment is monitored within the annual radioactivity monitoring programme

  9. Investigation of tritium in groundwater at Pickering NGS

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  10. Chemical forms and discharge ratios to stack and sea of tritium from Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Mikami, Satoshi; Akiyama, Kiyomitsu; Miyabe, Kenjiro

    2002-03-01

    Chemical forms and discharge ratios to stack and sea of tritium form Tokai Reprocessing Plant of Japan Nuclear Cycle Development Institute (JNC) were investigated by analyzing monitoring data. It was ascertained that approximately 70-80% of tritium discharged from the main stack was tritiated water vapor (HTO) and approximately 20-30% was tritiated hydrogen (HT) as a result of analyzing the data taken from reprocessing campaign's in 1994, 1995, 1996, 1997, 2000 and 2001, and also that the amount of tritium released from the stack was less than 1% of tritium inventory in spent fuel and the amount of tritium released into sea was approximately 20-40% of inventory. (author)

  11. Tritium stripping in a nitrogen glove box using palladium/zeolite and SAES St 198 trademark

    International Nuclear Information System (INIS)

    Klien, J.E.; Wermer, J.R.

    1995-01-01

    Glove box clean-up experiments were conducted in a nitrogen glove box using palladium deposited on zeolite (Pd/z) and a SAES St 198 trademark getter as tritium stripping materials. Protium/deuterium samples spiked with tritium were released into a 620 liter glove box to simulate tritium releases in a 10,500 liter glove box. The Pd/z and the SAES St 198 trademark stripper beds produced a reduction in tritium activity of approximately two to three orders of magnitude and glove box clean-up was limited by a persistent background tritium activity level. Attempts to significantly reduce the glove box activity to lower levels without purging were unsuccessful

  12. Tritium extraction mechanisms from lithium aluminates during in pile irradiation experiments

    International Nuclear Information System (INIS)

    Briec, M.; Roth, E.

    1987-04-01

    The principal aim was to determine ranges of parameters governing tritium release from γ lithium aluminates within which acceptable rates for their contemplated usage as tritium breeder material in a fusion reactor blanket could be obtained. in the first place values of every quantity involved should be known as well as possible. Reproducible results should be a criterium of validity of the selected parameters. It is shown from a description of a series of experiments that processes limiting tritium release rates are not the same in different temperature ranges. By varying the composition of purge gases used for tritium extraction, the level of irradiation fluxes, and by studying simultaneously samples of different textures, results were obtained and an assignment of the respective role of defect formation, texture, surface effect is attempted to interpret them

  13. Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bourham, Mohamed A.; Gilligan, John G.

    1999-08-14

    Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

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

    International Nuclear Information System (INIS)

    Caldwell-Nichols, C.J.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. Removal of tritium from gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    Muller, M.E.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-04-01

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

  19. Evaluation of Tritium Behavior in the Epoxy Painted Concrete Wall of ITER Hot Cell

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

    2005-01-01

    Tritium behavior released in the ITER hot cell has been investigated numerically using a combined analytical methods of a tritium transport analysis in the multi-layer wall (concrete and epoxy paint) with the one dimensional diffusion model and a tritium concentration analysis in the hot cell with the complete mixing model by the ventilation. As the results, it is revealed that tritium concentration decay and permeation issues are not serious problem in a viewpoint of safety, since it is expected that tritium concentration in the hot cell decrease rapidly within several days just after removing the tritium release source, and tritium permeation through the epoxy painted concrete wall will be negligible as long as the averaged realistic diffusion coefficient is ensured in the concrete wall. It is also revealed that the epoxy paint on the concrete wall prevents the tritium inventory increase in the concrete wall greatly (two orders of magnitudes), but still, the inventory in the wall is estimated to reach about 0.1 PBq for 20 years operation

  20. Environmental Release Prevention and Control Plan

    International Nuclear Information System (INIS)

    Mamatey, A.; Arnett, M.

    1997-01-01

    During the history of SRS, continual improvements in facilities, process, and operations, and changes in the site''s mission have reduced the amount of radioactive liquid releases. In the early years of SRS (1958 to 1965), the amount of tritium discharged to the Savannah River averaged approximately 61,000 curies a year. During the mid-1980''s (1983 to 1988), liquid releases of tritium averaged 27,000 curies a year. By 1996, liquid releases of tritium are projected to be just 3000 curies for the year. This large projected decrease is the result of the planned shut-down of all reactors and the anticipated significant decline in the amount of tritium migrating from the site seepage basins and the Solid Waste Disposal Facility

  1. Human Health and the Biological Effects of Tritium in Drinking Water: Prudent Policy Through Science – Addressing the ODWAC New Recommendation

    Science.gov (United States)

    Dingwall, S.; Mills, C.E.; Phan, N.; Taylor, K.; Boreham, D.R.

    2011-01-01

    Tritium is a radioactive form of hydrogen and is a by-product of energy production in Canadian Deuterium Uranium (CANDU) reactors. The release of this radioisotope into the environment is carefully managed at CANDU facilities in order to minimize radiation exposure to the public. However, under some circumstances, small accidental releases to the environment can occur. The radiation doses to humans and non-human biota from these releases are low and orders of magnitude less than doses received from naturally occurring radioisotopes or from manmade activities, such as medical imaging and air travel. There is however a renewed interest in the biological consequences of low dose tritium exposures and a new limit for tritium levels in Ontario drinking water has been proposed. The Ontario Drinking Water Advisory Council (ODWAC) issued a formal report in May 2009 in response to a request by the Minister of the Environment, concluding that the Ontario Drinking Water Quality Standard for tritium should be revised from the current 7,000 Bq/L level to a new, lower 20 Bq/L level. In response to this recommendation, an international scientific symposium was held at McMaster University to address the issues surrounding this change in direction and the validity of a new policy. Scientists, regulators, government officials, and industrial stakeholders were present to discuss the potential health risks associated with low level radiation exposure from tritium. The regulatory, economic, and social implications of the new proposed limit were also considered. The new recommendation assumed a linear-no-threshold model to calculate carcinogenic risk associated with tritium exposure, and considered tritium as a non-threshold chemical carcinogen. Both of these assumptions are highly controversial given that recent research suggests that low dose exposures have thresholds below which there are no observable detrimental effects. Furthermore, mutagenic and carcinogenic risk calculated from

  2. Tritium persistence pattern in some terrestrial plants-field investigations

    International Nuclear Information System (INIS)

    Soman, S.D.; Iyengar, T.S.; Krishnamoorthy, T.M.; Sadarangani, S.H.; Vaze, P.K.; Gogate, S.S.; Deo, J.V.

    1977-01-01

    The uptake and release pattern of tritium in certain trees in their natural conditions of growth were investigated by artificial simulation of active conditions by incorporating tritium in the system through stem or roots. These trees are grown in some of the nuclear sites wherein a number of nuclear facilities are located. The species studied include palms, casuarinas and banana trees. In most of the cases a single component corresponding to the tree compartment tritium was obtained. The second component of the tissue free water tritium and the tissue bound compartment of tritium were not easily resolvable due to tremendous variation caused by the environmental conditions such as rain, humidity etc. Repeated humps were observed in certain cases of root uptake studies due to the variation in the meteorological factors. In most of the cases the half residence times for tritium (Tsub(1/2)) (tissue free water tritium) were found to be below two days. (author)

  3. Use of Tritium Accelerator Mass Spectrometry for Tree Ring Analysis

    Science.gov (United States)

    LOVE, ADAM H.; HUNT, JAMES R.; ROBERTS, MARK L.; SOUTHON, JOHN R.; CHIARAPPA - ZUCCA, MARINA L.; DINGLEY, KAREN H.

    2010-01-01

    Public concerns over the health effects associated with low-level and long-term exposure to tritium released from industrial point sources have generated the demand for better methods to evaluate historical tritium exposure levels for these communities. The cellulose of trees accurately reflects the tritium concentration in the source water and may contain the only historical record of tritium exposure. The tritium activity in the annual rings of a tree was measured using accelerator mass spectrometry to reconstruct historical annual averages of tritium exposure. Milligram-sized samples of the annual tree rings from a Tamarix located at the Nevada Test Site are used for validation of this methodology. The salt cedar was chosen since it had a single source of tritiated water that was well-characterized as it varied over time. The decay-corrected tritium activity of the water in which the salt cedar grew closely agrees with the organically bound tritium activity in its annual rings. This demonstrates that the milligram-sized samples used in tritium accelerator mass spectrometry are suited for reconstructing anthropogenic tritium levels in the environment. PMID:12144257

  4. Spherical diffusion of tritium from a point of release in a uniform unsaturated soil

    International Nuclear Information System (INIS)

    Smiles, D.E.; Gardner, W.R.; Schulz, R.K.

    1993-10-01

    Tritium (Tr), when released as tritiated water at a point in a uniform and relatively dry soil, redistributes in both the liquid and vapor phases. The flux density of Tr in the liquid will exceed that in the vapor phase provided the water content is greater than approximately 15% of the total soil porosity. Thus Tr redistribution must be modeled recognizing transfer ''in parallel'' in both phases. The authors use the diffusion equation cast in spherical coordinates to analyze this problem in order to provide a basis for design of field experiments, and to offer observations on the long term behavior of such systems. The solution of the diffusion equation permits calculation of the evolution of profiles of Tr concentration, within and external to the sphere of released solution, assuming the initial concentration within this sphere to be uniform. The authors also predict the rate of advance of the maximum of Tr as it advances, and attenuates, in the soil. Calculations for the case of 1 million Curies of Tr diluted in 1 liter of water and released at a depth of 20 meters, and 200 meters above the water table, are demonstrated. If the soil has an initial water volume fraction of 0.06 and total porosity of 0.3 they show, for example, that at 5 meters from the point of discharge, the Tr concentration increases to a maximum in 24 years and then slowly declines. That maximum is 1 Curie/liter. The concentration in the gas phase will be 5 orders-of-magnitude less than this. At 60 meters the maximum ever reached in the liquid phase is ca 10 -21 Ci/liter; that maximum will be achieved after 408 years. The authors discuss the effects of variation in the volume fractions of water and air originally present in the soil on the effective diffusion coefficient of Tr in soil, consider the effects of a net flux of water in the system, and identify questions to be answered to achieve safe systematic disposal of tritium in the deep unsaturated zone of desert soil

  5. Development of dose assessment code for accidental releases of activation products

    International Nuclear Information System (INIS)

    Noguchi, H.; Yokoyama, S.

    2000-01-01

    It is expected that activation products will be important radionuclides as well as tritium in the assessment of the public doses necessary for licensing of a future fusion reactor. In order to calculate the public doses due to the activation products released in cases of accidents, a code named ACUTAP (dose assessment code for ACUTe Activation Product releases) has been developed. Major characteristics of the code are as follows: (1) the transfer model reflects specific behavior of the activation products in the environment, (2) the doses are assessed based on ICRP dose models, (3) it is possible to calculate individual doses using annual meteorological data statistically according to the guide of the Nuclear Safety Commission of Japan, and (4) the code can calculate collective doses as well as individual doses. Individual doses are calculated for the following pathways: internal exposure by inhalation of activation products in a plume and those resuspended from the ground, external exposure from a plume (cloudshine), and external exposure from activation products deposited on the ground (groundshine). The inhalation in a plume and cloudshine pathways are included in the model for calculating collective doses. In addition to parent nuclides released from the facilities, progeny nuclides produced during the atmospheric dispersion are considered in calculating inhalation doses, and those during the deposition period in calculating groundshine doses. External doses from the cloudshine are calculated for 18 energy groups instead of individual energy of emitted gamma rays in order to save the computation time. Atmospheric concentrations are calculated using a Gaussian plume model with atmospheric dispersion parameters prescribed in the guide of the Nuclear Safety Commission of Japan. Data sets of parameters necessary for the dose assessment, such as internal dos coefficients, external dose rate conversion factors and half lives, are prepared for about 100 radionuclides

  6. Elemental tritium deposition and conversion in the terrestrial environment

    International Nuclear Information System (INIS)

    Dunstall, T.G.; Ogram, G.L.; Spencer, F.S.

    1985-01-01

    Studies were undertaken to determine the deposition and conversion of atmospheric elemental tritium in soils and vegetation. In the field tritium deposition velocities ranged between 0.007 and 0.07 cm s -1 during the summer and autumn and were less than 0.0005 cm s -1 during the winter. Deposition velocity was found to depend significantly on soil water content, total pore space and organic content in controlled laboratory experiments. In contrast to soils, exposure of vegetation to atmospheric elemental tritium resulted in negligible uptake and conversion in foliage. These studies are of significance to the assessment of behaviour and impact of elemental tritium releases

  7. Report of the consultant's meeting on monitoring accidentally released radionuclides in the environment

    International Nuclear Information System (INIS)

    2001-01-01

    At an earlier Research Co-ordination Meeting (12-16 August 1991, VIC, Vienna) for the IAEA Programme on 'Rapid Instrumental and Separation Methods for Monitoring Radionuclides in Food and Environmental Samples', it was recognized that a network of analytical laboratories with specially qualified scientists is needed which will provide a timely coordinated response to any future requests for assistance from Member States. The experience gained from the International Chernobyl Project revealed that there is a need for such a closely linked group when the Member State lacks sufficient technical resource to independently perform an assessment of environmental contamination following an accidental release of radioactivity. The existence of such a network of laboratories would help to assure responsible government representatives and the general public that the Agency has additional available resources including specialized experts to perform assessments on which to base required actions for many accident scenarios and their radiological consequences. The purpose of the present Consultants' Meeting is to discuss the organization and scope of the proposed network of analytical laboratories. The Consultants should discuss the matrices and nuclides of interest at various time frames of an accidental release of radioactivity. It is also intended that the Consultants should discuss the laboratories from their experience which would have the specialized expertise needed, in both terrestrial and aquatic sample collection and analysis. Finally, it is intended that the Consultants should discuss whether it would be desirable for the Agency to produce a technical document based on the achievements of the Agency programme on 'Rapid Instrumental and Separation Methods for Monitoring Radionuclides in Food and Environmental Samples'. A Guide Book was considered essential by the participants of this Co-ordinated Research Programme because it would be a succinct form and on a common

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

    International Nuclear Information System (INIS)

    Dickson, R.S.

    1990-11-01

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

  9. Measurement of tritium in the Sava and Danube Rivers.

    Science.gov (United States)

    Grahek, Željko; Breznik, Borut; Stojković, Ivana; Coha, Ivana; Nikolov, Jovana; Todorović, Nataša

    2016-10-01

    Two nuclear power plants (NPP), the KrškoNPP (Slovenia) on the Sava River and the Paks NPP (Hungary) on the Danube River, are located in the immediate vicinity of Croatia and Serbia. Some of the radioactivity monitoring around the NPPs involves measuring tritium activity in the waters of rivers and wells. The authors present the tritium measurement results taken over several years from the Sava and Danube Rivers, and groundwater. The measurements were carried out in two laboratories including an impact assessment of the tritium released into the rivers and groundwater. The routine methods for determining tritium (with/without electrolytic enrichment) were tested in two laboratories using two different instruments, a Tri-Carb 3180 and Quantulus 1220. Detection limits for routine measurements were calculated in compliance with ISO 11929 and Currie relations, and subsequently the results were compared with those determined experimentally. This has shown that tritium can be reliably determined within a reasonable period of time when its activity is close to the calculated detection limit. The Krško NPP discharged 62 TBq of tritium into the River Sava over a period of 6 years (23% of permitted activity, 45 TBq per year). The natural level of tritium in the Sava River and groundwater is 0.3-1 Bq/l and increases when discharges exceed 1 TBq per month. Usually, the average monthly activity in the Sava River and groundwater is maintained at a natural level. The maximum measured activity was 16 Bq/l in the Sava River and 9.5 Bq/l in groundwater directly linked to the river. In the majority of water samples from the Danube River, measured tritium activity ranged between 1 and 2 Bq/l. The increased tritium levels in the Danube River are more evident than in the Sava River because tritium activity above 1.5 Bq/l appears more frequently on the Danube River. All measured values were far below the allowed tritium limit in drinking water. Dose assessment has shown that

  10. FDNH - the tritium module in RODOS

    International Nuclear Information System (INIS)

    Galeriu, D.; Melintescu, A.; Turcanu, C. O.; Raskob, W.

    2001-01-01

    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 1997 it was realised that tritium should be included in the list of radionuclides, as large tritium sources exists 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 reactors in Romania (with potential tritium accidents). At present, the Food and Dose Module Hydrogen -(FDMH) - for tritium applications - is integrated and documented in the RODOS system. It calculates the time dependent tritium concentration (as tritiated water or organically bound tritium) in crops (as much as 22 different species) and up to 12 animal products, inhalation doses and ingestion dose from up to 34 diet items for various groups of the population and for up to 2520 locations around the source, following an accidental emission of tritiated water. FDMH incorporates many improved techniques in radiological assessment and makes intensively use of interdisciplinary research. It is developed in a modular structure with a variable time grid according to the physical processes. Differing from other models, using generic transfer parameters or parameters fitted on individual experiments, FDMH derives tritium transfer rates based on physical and physiological process analysis, using scientifically accepted results from interdisciplinary research on, among others, land-atmosphere interaction, water cycle in the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  12. Mesoscale atmospheric modeling of the July 12, 1992 tritium release from the Savannah River Site

    International Nuclear Information System (INIS)

    Fast, J.D.; O'Steen, B.L.; Addis, R.P.

    1992-01-01

    In August of 1991, the Environmental Transport Group (ETG) began the development of an advanced Emergency Response (ER) system based upon the Colorado State University Regional Atmospheric Modeling System (RAMS). This model simulates the three-dimensional, time-dependent, flow field and thermodynamic structure of the planetary boundary layer (PBL). A companion Lagrangian Particle Dispersion Model (LPDM) simulates contaminant transport based on the flow and turbulence fields generated by RAMS. This paper describes the performance of the advanced ER system in predicting transport and diffusion near the SRS when compared to meteorological and sampling data taken during the July 12, 1992 tritium release. Since PUFF/PLUME and 2DPUF are two Weather INformation and Display (WIND) System atmospheric models that were used to predict the transport and diffusion of the plume at the time of the release, the results from the advanced ER system are also compared to those produced by PUFF/PLUME and 2DPUF

  13. Development of a tritium dispersion code

    International Nuclear Information System (INIS)

    Bell, R.P.; Davis, M.W.; Joseph, S.; Wong, K.Y.

    1985-01-01

    This paper describes the development and verification of a computer code designed to calculate the radiation dose to man following acute or chronic atmospheric releases of tritium gas and oxide from a point source. The Ontario Hydro Tritium Dispersion Code calculates tritium concentrations in air, soil, and vegetation and doses to man resulting from inhalation/immersion and ingestion of food, milk meat and water. The deposition of HT to soil, conversion of HT to HTO by soil enzymes and resuspension of HTO to air have been incorporated into the terrestrial compartment model and are unique features of the code. Sensitivity analysis has identified the HT deposition velocity and the equivalent water depth of the vegetation compartment as two parameters which have a strong influence on dose calculations. Tritium concentrations in vegetation and soil calculated by the code were in reasonable agreement with experimental results. The radiological significance of including the mechanisms of HT to HTO conversion and resuspension of HTO to air is illustrated

  14. Safety aspects of tritium in ICF reactors with internally-breeding targets

    International Nuclear Information System (INIS)

    Ragheb, M.; Miley, G.H.; University of Illinois, Urbana, IL)

    1985-01-01

    The LOTRIT inertial confinement reactor concept employs a deuterium burning target with a DT spark trigger core. This eliminates the need for tritium breeding in a blanket, and leads to a minimization of the tritium inventory and of the possibility of metal fire hazards if lead is used instead of lithium for first wall protection. The active fuel inventory in the fuel cycle and blanket per MJ of energy produced is only 5 percent of the DT case. The most significant reduction in the total tritium inventory is in the target manufacture and storage areas, and is about 1.8% of the DT case per unit of fusion energy produced. If the goal is to reduce the risk from tritium releases from fusion reactors to below that of fission reactors, it is estimated that the tritium releases must be maintained at 0.13-5.0 Ci/day. Attaining these values will be costly, technologically difficult and will constrain the design options in DTbased systems, but may be within the realm of systems using the LOTRIT concept

  15. Large-scale distribution of tritium in a commercial product

    International Nuclear Information System (INIS)

    Combs, F.; Doda, R.J.

    1979-01-01

    Tritium enters the environment from various sources including nuclear reactor operations, weapons testing, natural production, and from the manufacture, use and ultimate disposal of commercial products containing tritium. A recent commercial application of tritium in the United States of America involves the backlighting of liquid crystal displays (LCD) in digital electronic watches. These watches are distributed through normal commercial channels to the general public. One million curies (MCi) of tritium were distributed in 1977 in this product. This is a significant quantity of tritium compared with power reactor-produced tritium (3MCi yearly) or with naturally produced tritium (6MCi yearly). This is the single largest commercial application involving tritium to date. The final disposition of tritium from large quantities of this product, after its useful life, must be estimated by considering the means of disposal and the possibility of dispersal of tritium concurrent with disposal. The most likely method of final disposition of this product will be disposal in solid refuse; this includes burial in land fills and incineration. Burial in land fills will probably contain the tritium for its effective lifetime, whereas incineration will release all the tritium gas (as the oxide) to the atmosphere. The use and disposal of this product will be studied as part of an environmental study that is at present being prepared for the U.S. Nuclear Regulatory Commission. (author)

  16. Tritium burning in inertial electrostatic confinement fusion facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  17. Tritium burning in inertial electrostatic confinement fusion facility

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  18. 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

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

    Science.gov (United States)

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

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

  20. Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    International Nuclear Information System (INIS)

    Bourham, Mohamed A.; Gilligan, John G.

    1999-01-01

    Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m 2 over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER

  1. 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

  2. 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

  3. Tritium-management requirements for D-T fusion reactors (ETF, INTOR, FED)

    International Nuclear Information System (INIS)

    Finn, P.A.; Clemmer, R.G.; Misra, B.

    1981-10-01

    The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design

  4. The influence of the physico-chemical form of the aerosol on the radiological consequences of notional accidental releases of radioactivity from a fast breeder reactor

    International Nuclear Information System (INIS)

    Kelly, G.N.; Jones, J.A.; Simmonds, J.R.

    1979-01-01

    The radiological consequences of a wide range of notional accidental releases from a 1300 MW(e) LMFBR (Liquid Metal-cooled Fast Breeder Reactor) were assessed in a study published by the National Radiological Protection Board (NRPB) in 1977. In that study representative values were in general adopted for each of the important parameters while recognising that in reality they could vary considerably. The present study is concerned with the sensitivity of the predicted consequences to the physico-chemical form of the released aerosol. Of particular interest is the importance of a mixed sodium-transuranium element aerosol which may be formed in accidental releases of activity from sodium cooled FBRs. Two significant findings emerge from the study. First the predicted consequences in general are relatively insensitive to the range of physico-chemical forms analysed. For generic assessments therefore it is sufficient to assume the properties of the aerosol adopted in the initial study (1 μm AMAD and each element in the oxide form); the exception concerns the estimation of the incidence of early morbidity, and to a lesser extent early mortality, but only for a limited range of release composition. The second finding is that the radiological consequences are not, contrary to what might have been expected, significantly increased for the release of a mixed sodium-element aerosol

  5. Prediction of tritium behavior in rice plant after a short-term exposure of HTO

    International Nuclear Information System (INIS)

    Yook, Dae Sik; Lee, Kun Jai; Choi, Heui Joo; Lee, Chang Min

    2001-01-01

    In many Asian countries including Korea, rice is a very important food crop. Its grain is consumed by humans and its straw is used to feed animals. Because four CANDU reactors are in operation in Korea, relatively large amounts of tritium are released into the environment and the dose by these tritium in the rice plant must be estimated. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out experimental studies to obtain domestic data on various parameters related to the direct tritium contamination of plant. But the analysis of the tritium behavior in the rice plant has been insufficient. In this study, the behavior of the tritium in the rice plant is predicted and compared with the measurement performed at KAERI. Using the conceptual model of the soil-plant-atmosphere tritiated water transport system which was suggested by Charles E. Murphy, transient tritium concentrations in soil and leaves were predicted. If the effect of tritium concentration in the soil is taken into account, the tritium concentration in leaves can be described by a double exponential model, however if the tritium concentration in the soil is disregarded, the tritium concentration in leaves can be described by a single exponential term like other relevant models e.g. UFOTRI or STAR-H3 model. The results can be used to predict the tritium concentration in the rice plant near the plant site and to estimate the ingestion dose after the release of tritium to the environment

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    Science.gov (United States)

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

    1997-02-01

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

  8. 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.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  10. Enhanced activities of organically bound tritium in biota samples.

    Science.gov (United States)

    Svetlik, I; Fejgl, M; Malátová, I; Tomaskova, L

    2014-11-01

    A pilot study aimed on possible occurrence of elevated activity of non-exchangable organically bound tritium (NE-OBT) in biota was performed. The first results showed a significant surplus of NE-OBT activity in biota of the valley of Mohelno reservoir and Jihlava river. The liquid releases of HTO from the nuclear power plant Dukovany is the source of tritium in this area. This area can be a source of various types of natural samples for future studies of tritium pathways. Copyright © 2014 Elsevier Ltd. All rights reserved.

  11. Assessment of radiological releases to the environment from a fusion reactor power plant

    International Nuclear Information System (INIS)

    Shank, K.E.; Oakes, T.W.; Easterly, C.E.

    1978-05-01

    This report summarizes the expected tritium and activation-product inventories and presents an assessment of the potential radiological releases from a fusion reactor power plant, hypothetically located at the Oak Ridge National Laboratory. Routine tritium releases and the resulting dose assessment are discussed. Uncertainties associated with the conversion of tritium gas to tritium oxide and the global tritium cycling are evaluated. The difficulties of estimating releases of activated materials and the subsequent dose commitment are reviewed

  12. Tritium: an underestimated health risk- 'ACROnic du nucleaire' nr 85, June 2009

    International Nuclear Information System (INIS)

    Barbey, Pierre

    2009-06-01

    After having indicated how tritium released in the environment (under the form of tritiated water or gas) is absorbed by living species, the author describes the different biological effects of ionizing radiations and the risk associated with tritium. He evokes how the radiation protection system is designed with respect to standards, and outlines how the risk related to tritium is underestimated by different existing models and standards. The author discusses the consequences of tritium transmutation and of the isotopic effect

  13. 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.)

  14. 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

  15. Tritium Issues in Next Step Devices

    International Nuclear Information System (INIS)

    C.H. Skinner; G. Federici

    2001-01-01

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

  16. Using the Tritium Plasma Experiment to evaluate ITER PFC safety

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  17. Estimation of accumulated individual doses from waterborne pulse releases of tritium and activation products into the Baltic Sea from a hypothetical fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Edlund, O

    1995-04-01

    The committed effective dose to a critical group during 50 years attributed to pulse unit releases of tritium and activation products (in total 66 nuclides) from a thought postulated fusion power plant into Tvaeren Bay, a part of the Baltic Sea with brackish water, outside Studsvik, Sweden, is calculated. The purpose of this work is to obtain (simple) relationships between release and dose for relevant nuclides for use in the future when realistic source terms are known. 26 refs, 8 tabs, 4 figs.

  18. Estimation of accumulated individual doses from waterborne pulse releases of tritium and activation products into the Baltic Sea from a hypothetical fusion reactor

    International Nuclear Information System (INIS)

    Edlund, O.

    1995-04-01

    The committed effective dose to a critical group during 50 years attributed to pulse unit releases of tritium and activation products (in total 66 nuclides) from a thought postulated fusion power plant into Tvaeren Bay, a part of the Baltic Sea with brackish water, outside Studsvik, Sweden, is calculated. The purpose of this work is to obtain (simple) relationships between release and dose for relevant nuclides for use in the future when realistic source terms are known. 26 refs, 8 tabs, 4 figs

  19. Updating the tritium quality factor: the argument for conservatism

    International Nuclear Information System (INIS)

    Till, J.E.; Etnier, E.L.; Meyer, H.R.

    1980-01-01

    Estimated doses resulting from tritium releases to the environment are linearly dependent upon the quality factor (Q) chosen for tritium beta radiation. In 1969 the International Commission on Radiological Protection (ICRP) recommended using 1 as the Q for all low energy beta radiation. Considerable improvements have been made in evaluating exposures to tritium at very low dose rates and in refining physiological and biological endpoints since the 1969 ICRP recommendations. This study summarizes recent experiments to determine the relative biological effectiveness of tritium. Based upon our study of published data related to quality factor, its importance in the calculation of dose, and the currently accepted conservative philosophy in radiation protection, it is concluded that a value of 2 would seem to be more defensible for environmental assessments and that a reevaluation of the tritium quality factor by the ICRP is needed

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  1. Transfer of tritium released into the marine environment by French nuclear facilities bordering the English Channel.

    Science.gov (United States)

    Fiévet, Bruno; Pommier, Julien; Voiseux, Claire; Bailly du Bois, Pascal; Laguionie, Philippe; Cossonnet, Catherine; Solier, Luc

    2013-06-18

    Controlled amounts of liquid tritium are discharged as tritiated water (HTO) by the nuclear industry into the English Channel. Because the isotopic discrimination between 3H and H is small, organically bound tritium (OBT) and HTO should show the same T/H ratio under steady-state conditions. We report data collected from the environment in the English Channel. Tritium concentrations measured in seawater HTO, as well as in biota HTO and OBT, confirm that tritium transfers from HTO to OBT result in conservation of the T/H ratio (ca. 1 × 10(-16)). The kinetics of the turnover of tritium between seawater HTO, biota HTO, and OBT was investigated. HTO in two algae and a mollusk is shown to exchange rapidly with seawater HTO. However, the overall tritium turnover between HTO and the whole-organism OBT is a slow process with a tritium biological half-life on the order of months. Nonsteady-state conditions exist where there are sharp changes in seawater HTO. As a consequence, for kinetic reasons, the T/H ratio in OBT may deviate transiently from that observed in HTO of samples from the marine ecosystem. Dynamic modeling is thus more realistic for predicting tritium transfers to biota OBT under nonsteady-state conditions.

  2. Wind field forecast for accidental release of radiative materials

    International Nuclear Information System (INIS)

    Kang Ling; Chen Jiayi; Cai Xuhui

    2003-01-01

    A meso-scale wind field forecast model was designed for emergency environmental assessment in case of accidental release of radiative materials from a nuclear power station. Actual practice of the model showed that it runs fast, has wind field prediction function, and the result given is accurate. With meteorological data collected from weather stations, and pre-treated by a wind field diagnostic model, the initial wind fields at different times were inputted as initial values and assimilation fields for the forecasting model. The model, in turn, worked out to forecast meso-scale wind field of 24 hours in a horizontal domain of 205 km x 205 km. And then, the diagnostic model was employed again with the forecasting data to obtain more detail information of disturbed wind field by local terrain in a smaller domain of 20.5 km x 20.5 km, of which the nuclear power station is at the center. Using observation data in January, April, July and October of 1996 over the area of Hangzhou Bay, wind fields in these 4 months were simulated by different assimilation time and number of the weather stations for a sensitive test. Results indicated that the method used here has increased accuracy of the forecasted wind fields. And incorporating diagnostic method with the wind field forecast model has greatly increased efficiency of the wind field forecast for the smaller domain. This model and scheme have been used in Environmental Consequence Assessment System of Nuclear Accident in Qinshan Area

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

    International Nuclear Information System (INIS)

    Garber, H.J.

    1975-10-01

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

  4. Levels of tritium in soils and vegetation near Canadian nuclear facilities releasing tritium to the atmosphere: implications for environmental models

    International Nuclear Information System (INIS)

    Thompson, P.A.; Kwamena, N.-O.A.; Ilin, M.; Wilk, M.; Clark, I.D.

    2015-01-01

    Concentrations of organically bound tritium (OBT) and tritiated water (HTO) were measured over two growing seasons in vegetation and soil samples obtained in the vicinity of four nuclear facilities and two background locations in Canada. At the background locations, with few exceptions, OBT concentrations were higher than HTO concentrations: OBT/HTO ratios in vegetation varied between 0.3 and 20 and values in soil varied between 2.7 and 15. In the vicinity of the four nuclear facilities OBT/HTO ratios in vegetation and soils deviated from the expected mean value of 0.7, which is used as a default value in environmental transfer models. Ratios of the OBT activity concentration in plants ([OBT] plant ) to the OBT activity concentration in soils ([OBT] soil ) appear to be a good indicator of the long-term behaviour of tritium in soil and vegetation. In general, OBT activity concentrations in soils were nearly equal to OBT activity concentrations in plants in the vicinity of the two nuclear power plants. [OBT] plant /[OBT] soil ratios considerably below unity observed at one nuclear processing facility represents historically higher levels of tritium in the environment. The results of our study reflect the dynamic nature of HTO retention and OBT formation in vegetation and soil during the growing season. Our data support the mounting evidence suggesting that some parameters used in environmental transfer models approved for regulatory assessments should be revisited to better account for the behavior of HTO and OBT in the environment and to ensure that modelled estimates (e.g., plant OBT) are appropriately conservative. - Highlights: • We measured tritium in soils and plants near four nuclear facilities in Canada. • OBT/HTO ratios in plants are higher than default value in environmental models. • OBT/HTO ratios in background soils reflect historically higher atmospheric tritium. • Implications for environmental transfer models are discussed

  5. Tritium in groundwater investigation at the Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  6. Tritium pellet injector design for tokamak fusion test reactor

    International Nuclear Information System (INIS)

    Fisher, P.W.; Baylor, L.R.; Bryan, W.E.

    1985-01-01

    A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper

  7. Determination of low-level tritium concentrations in surface water and precipitation in the Czech Republic

    International Nuclear Information System (INIS)

    Maresova, Diana; Hanslik, Eduard; Sedlarova, Barbora; Juranova, Eva; Charles University, Prague

    2017-01-01

    Past tests of nuclear weapons in the atmosphere, nuclear energy facilities and tritium of natural origin are main sources of tritium in the environment. Thanks to its presence in environment and its favourable properties, tritium is used as a radiotracer. Since stopping of atmospheric nuclear tests, tritium in precipitation has been decreasing towards natural levels below 1 Bq l -1 and precise analyses of low level tritium activities are necessary. This paper focuses on tritium development at sites not influenced by any technogenic release of tritium in Elbe River basin (Bohemia) in the Czech Republic using liquid scintillation measurement with electrolytic enrichment. (author)

  8. Tritium saturation in plasma-facing materials surfaces

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  9. Tritium saturation in plasma-facing materials surfaces

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

  10. Accidental Release of Chlorine from a Storage Facility and an On-Site Emergency Mock Drill: A Case Study

    Directory of Open Access Journals (Sweden)

    Ambalathumpara Raman Soman

    2015-01-01

    Full Text Available In the current industrial scenario there is a serious need for formulating strategies to handle hazardous substances in the safest way. Manufacture, storage, and use of hazardous substances pose a serious risk to industry, people, and the environment. Accidental release of toxic chemicals can lead to emergencies. An emergency response plan (ERP is inevitable to minimize the adverse effects of such releases. The on-site emergency plan is an integral component of any process safety and risk management system. This paper deals with an on-site emergency response plan for a chlorine manufacturing industry. It was developed on the basis of a previous study on chlorine release and a full scale mock drill has been conducted for testing the plan. Results indicated that properly trained personnel can effectively handle each level of incidents occurring in the process plant. As an extensive guideline to the district level government authorities for off-site emergency planning, risk zone has also been estimated with reference to a chlorine exposure threshold of 3 ppm.

  11. The European Community dimension to the problem of accidental radioactive releases to the environment

    International Nuclear Information System (INIS)

    Bennett, E.

    1989-01-01

    The European Community has to comply with the provisions of the Euratom Treaty, 1957. As well as providing the conditions necessary for the growth of nuclear industries it lays down the task to establish uniform Safety Standards to protect the health of workers and of the general public. This includes emergency planning in case of nuclear accidents. As a result the Community followed the guide, published in 1982 by the International Commission on Radiological Protection, on the radiological protection criteria for controlling doses to the public in the event of accidental releases of radioactive materials. However when the Chernobyl accident occured the Community and International arrangements proved to be inadequate to respond to the consequent widespread contamination over Europe. Measures have been taken since then to improve the preparedness of the Community. They include better exchange of information, control of contaminated foodstuffs and better information to the public. (U.K.)

  12. Wet scrubber technology for tritium confinement at ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

  13. Tritium retention properties of tungsten, graphite and co-deposited carbon film

    International Nuclear Information System (INIS)

    Nobuta, Y.; Hatano, Y.; Matsuyama, M.; Abe, S.; Akamaru, S.; Yamauchi, Y.; Hino, T.; Suzuki, S.; Akiba, M.

    2014-01-01

    DT + ion irradiation was performed on polycrystalline tungsten, graphite and carbon film and both the amount of retained tritium and the reduction of retained tritium after preservation in vacuum were investigated using an IP technique and BIXS. In addition, the relationship between the retention properties of tritium and the microstructure of graphite and carbon film were studied with Raman spectroscopy. The amount of retained tritium in tungsten was smaller than in both graphite and carbon film. After 1 keV of DT + irradiation, graphite showed no reduction of the amount of retained tritium after six months preservation while that of carbon film decreased by approximately 20% after 40 days preservation. It was suggested that this difference might be associated with differences in the microstructure between graphite and carbon film. In tungsten, the amount of retained tritium decreased to approximately half after 18 days preservation. As the incident energy of implanted tritium to tungsten increased, the decrease in tritium retention during preservation became slower. Tungsten's properties of releasing tritium while preserved in vacuum would be a useful tool for the reduction/removal of retained tritium

  14. Monsanto Mound Laboratory tritium waste control technology development program

    International Nuclear Information System (INIS)

    Bixel, J.C.; Kershner, C.J.; Rhinehammer, T.B.

    1975-01-01

    Over the past four years, implementation of tritium waste control programs has resulted in a 30-fold reduction in the gaseous tritium effluents from Mound Laboratory. However, to reduce tritium waste levels to the ''as low as practicable'' guideline poses problems that are beyond ready solution with state-of-the-art tritium control technology. To meet this advanced technology need, a tritium waste control technology program was initiated. Although the initial thrust of the work under this program was oriented toward development of gaseous effluent treatment systems, its natural evolution has been toward the liquid waste problem. It is thought that, of all the possible approaches to disposal of tritiated liquid wastes, recovery offers the greatest advantages. End products of the recovery processes would be water detritiated to a level below the Radioactivity Concentration Guide (RCG) or detritiated to a level that would permit safe recycle in a closed loop operation and enriched tritium. The detritiated water effluent could be either recycled in a closed loop operation such as in a fuel reprocessing plant or safely released to the biosphere, and the recovered tritium could be recycled for use in fusion reactor studies or other applications

  15. 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)

  16. Environmental Aspects of Tritium Around the Vinca Institute of Nuclear Sciences

    International Nuclear Information System (INIS)

    Miljevic, N.; Sipka, V.

    1995-01-01

    An overview of environmental distribution of tritium around the Institute of Nuclear Sciences Vinca during the period 1988-1994 is presented. Temporal and local variations of the specific tritium variations in precipitation (Usek, Zeleno Brdo), river waters (the Danube, the Sava and Mlaka Creek) as well as atmospheric water vapor are given. Estimates based on precipitation measurements have shown that 6.3 TBq of tritium activity should be released annually into the atmosphere from the Vinca Institute of Nuclear Sciences. (author)

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  18. Experimental determination of the kinetic conversion rate of gaseous tritium into HTO

    International Nuclear Information System (INIS)

    Gulden, W.; Guenot, J.; Djerassi, H.; Clerc, H.

    1986-09-01

    To assess the impact of atmospheric tritium releases from a fusion plant to the environment, the dose to man usually is determined under the conservative assumption that all tritium is released in the more dangerous oxide form. To quantify this overprediction, experiments are presently being performed by CEA. Oxidation of tritium gas by soil and subsequent resuspension of HTO from soil to atmosphere is one of them. First results have been obtained by CEA on the kinetics of HTO resuspension from contaminated soils. Immediately after contamination, the fraction of the deposited activity which is resuspended to atmosphere is in the range of 1% to 5% per hour. This resuspension rate then decreases more or less slowly with time, depending on specific conditions

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  20. Assessment of tritium in the Savannah River Site environment

    Energy Technology Data Exchange (ETDEWEB)

    Carlton, W.H.; Murphy, C.E. Jr.; Bauer, L.R. [and others

    1993-10-01

    This report is the first revision to a series of reports on radionuclides inn the SRS environment. Tritium was chosen as the first radionuclide in the series because the calculations used to assess the dose to the offsite population from SRS releases indicate that the dose due to tritium, through of small consequence, is one of the most important the radionuclides. This was recognized early in the site operation, and extensive measurements of tritium in the atmosphere, surface water, and ground water exist due to the effort of the Environmental Monitoring Section. In addition, research into the transport and fate of tritium in the environment has been supported at the SRS by both the local Department of Energy (DOE) Office and DOE`s Office of Health and Environmental Research.

  1. Assessment of tritium in the Savannah River Site environment

    International Nuclear Information System (INIS)

    Carlton, W.H.; Murphy, C.E. Jr.; Bauer, L.R.

    1993-10-01

    This report is the first revision to a series of reports on radionuclides inn the SRS environment. Tritium was chosen as the first radionuclide in the series because the calculations used to assess the dose to the offsite population from SRS releases indicate that the dose due to tritium, through of small consequence, is one of the most important the radionuclides. This was recognized early in the site operation, and extensive measurements of tritium in the atmosphere, surface water, and ground water exist due to the effort of the Environmental Monitoring Section. In addition, research into the transport and fate of tritium in the environment has been supported at the SRS by both the local Department of Energy (DOE) Office and DOE's Office of Health and Environmental Research

  2. Modeling tritium processes in plasma-facing beryllium

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  3. Measurements of tritium for radiological protection purposes in dial painting industry

    International Nuclear Information System (INIS)

    Sawant, J.V.; Rudran, Kamala; Pillai, K.C.

    1990-01-01

    Tritium is used as the active component in self-luminous paint. During dial painting process luminous paints releases tritium in air, causing air contamination. In the present paper results of a preliminary study on air monitoring and estimation of air samples in a local watch industry are given. (author). 5 refs., 2 t abs

  4. 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

  5. Environmental considerations for alternative fusion reactor blankets

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.; Young, J.R.

    1975-01-01

    Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

  6. Implementation of the Clean Air Act, Title III, Section 112(r) Prevention of Accidental Release Rule requirements at U.S. DOE Oak Ridge Reservation facilities

    International Nuclear Information System (INIS)

    Humphreys, M.P.

    1997-01-01

    Title III, Section 112(r) of the Clean Air Act (CAA) Amendments of 1990 requires the Environmental Protection Agency (EPA) to promulgate regulations to prevent accidental releases of regulated substances and to reduce the severity of those releases that do occur. The final EPA rule for Risk Management Programs under Section 112(r)(7) of the CAA, promulgated June 20, 1996, applies to all stationary sources with processes that contain more than a threshold quantity of any of 139 regulated substances listed under 40 CFR 68.130. All affected sources will be required to prepare a risk management plan which must be submitted to EPA and be made available to state and local governments and to the public. This paper will provide details of initiatives underway at US Department of Energy (DOE) Oak Ridge Reservation (ORR) Facilities for implementation of the Prevention of Accidental Release Rule. The ORR encompasses three DOE Facilities: the Y-12 Plant, Oak Ridge National Laboratory (ORNL), and the K-25 Site. The Y-12 Plant manufactures component parts for the national nuclear weapons program; the ORNL is responsible for research and development activities including nuclear engineering, engineering technologies, and the environmental sciences; and the K-25 Site conducts a variety of research and development activities and is the home of a mixed waste incinerator. ORR activities underway and soon to be undertaken toward implementation of the Prevention of Accidental Release Rule include: compilation of inventories of regulated substances at all processes at each of the three ORR Facilities for determination of affected processes and facilities; plans for inventory reduction to levels below threshold quantities, where necessary and feasible; determination of the overlap of processes subject to the OSHA PSM Standard and determination of parallel requirements; preparation of Risk Management Plans and Programs for affected processes and facilities including detailed requirements

  7. Deposition of radionuclides and their subsequent relocation in the environment following an accidental release to the atmosphere

    International Nuclear Information System (INIS)

    Underwood, B.Y.; Roed, J.; Paretzke, H.G.

    1993-01-01

    The objective of the project is to improve, as necessary, the models and parameterizations used in estimating the intensity and spatial distribution of deposited activity, and the total health/economic impact of such deposits in assessments of the consequences of accidental releases of radioactivity. The study comprises the influence of various weather conditions on deposition; the resuspension of deposited 137 Cs activity; the weathering of deposits in urban and rural environments; the ultimate fate and dosimetric impact of radionuclides carried by urban run-off water; the impact of the atmosphere's dispersion capabilities. Objectives and results of the four contributions to the project for the reporting period are presented. (R.P.) 5 refs., 4 figs., 1 tab

  8. Tritium concentration reducing method in atmosphere in nuclear reactor containment facility

    International Nuclear Information System (INIS)

    Hirasawa, Yoshiya; Kigoshi, Yasutane; Yonenaga, Haruo.

    1992-01-01

    A portion of water content in an atmosphere is condensed by a condensation/evaporation device disposed in a nuclear reactor containment building and then a portion of the condensed water is evaporated in the atmosphere. A portion of hydrogen nuclides constituting the evaporated water content is subjected to isotopic exchange with tritium nuclides in the atmosphere. A portion of water content in the atmosphere applied with the isotopic exchange is condensed in the condensation/evaporation device. That is, the hydrogen nuclides in steams are applied with isotopic exchange with tritium nuclides, and steams incorporating tritium nuclides are condensed again in the condensation/evaporation device, to transfer the tritium nuclides in the atmosphere to condensed water. The condensed water is recovered without releasing the tritium nuclides to the outside of the reactor containment building, thereby enabling to reduce the tritium concentration in the atmosphere. (N.H.)

  9. Relation between the tritium in continuous atmospheric release and the tritium contents of fruits and tubers.

    Science.gov (United States)

    Korolevych, V Y; Kim, S B

    2013-04-01

    Concentrations of organically bound tritium (OBT) and tissue-free water tritium (TFWT, also referred to as HTO) in fruits and tubers were measured at a garden plot in the vicinity of the source of chronic airborne tritium emissions during the 2008, 2010, and 2011 growing seasons. A continuous record of HTO concentration in the air moisture was reconstructed from the continuous record of Ar-41 ambient gamma radiation, as well as from frequent measurements of air HTO by active samplers at the garden plot and Ar-41 and air HTO monitoring data from the same sector. Performed measurements were used for testing the modified Specific Activity (SA) model based on the assumption that the average air HTO during the pod-filling period provides an appropriate basis for estimating the levels of OBT present in pods, fruits and tubers. It is established that the relationship between the OBT of fruits and tubers and the average air HTO from a 15-20 day wide window centred at the peak of the pod-filling period is consistent throughout the three analysed years, and could be expressed by the fruit or tuber's OBT to air-HTO ratio of 0.93 ± 0.21. For all three years, the concentration of HTO in fruits and tubers was found to be related to levels of HTO in the air, as averaged within a 3-day pre-harvest window. The variability in the ratio of plant HTO to air HTO appears to be three times greater than that for the OBT of the fruits and tubers. It is concluded that the OBT of fruits and tubers adequately follows an empirical relationship based on the average level of air HTO from the pod-filling window, and therefore is clearly in line with the modified SA approach. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

  10. Report of the first research co-ordination meeting on the co-ordinated research programme: Development and selection of analytical techniques for measuring accidentally released radionuclides in environment

    International Nuclear Information System (INIS)

    2001-01-01

    The participants at the second Research Co-ordination Meeting (Vienna, 12-16 August 1991) of the CRP on 'Rapid Instrumental and Separation Methods for Monitoring Radionuclides in Food and Environmental Samples', recommended that a new CRP be established. The current CRP on 'Development and Selection of Analytical Techniques and Procedures for Measuring Accidentally Released Radionuclides in Environment' was established based on this recommendation. The objectives of this CRP are to conduct research and development on applicable methodologies for response to accidental releases, and to improve and maintain the capabilities of the network of laboratories and provide training of individuals within member states. Thus, the CRP serves as a vehicle to maintain contact within the network of laboratories, while developing and transferring analytical techniques and procedures for measuring accidentally released radioactivity. The purpose of the Research Co-ordination Meeting is to discuss the proposed research programs, the status to date and the work planned for the duration of the CRP. The meeting also provides the opportunity for the CRP participants to exchange ideas and possibly develop collaborations in their research. The members of the CRP also need to discuss issues related to the previous CRP on Rapid methods. These include: the preparation of the final report of the previous CRP, the preparation of an addendum to TRS-295 and the ultimate revision or updating of TRS-295. Finally, it is intended that the members of the CRP should discuss the mandate and scope of work of the network of analytical laboratories and the steps needed to firmly establish this network

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. Distribution of tritium in water vapour and precipitation around Wolsung nuclear power plant.

    Science.gov (United States)

    Chae, Jung-Seok; Lee, Sang-Kuk; Kim, Yongjae; Lee, Jung-Min; Cho, Heung-Joon; Cho, Yong-Woo; Yun, Ju-Yong

    2011-07-01

    The distribution of tritium in water vapour and precipitation with discharge of tritiated water vapour and meteorological factors was studied around the Wolsung nuclear power plant (NPP) site during the period 2004-2008. The tritium concentrations in atmospheric water vapour and precipitation had a temporal variation with relatively high values in the early summer. Spatial distribution of tritium concentrations was affected by various factors such as distance from the NPP site, wind direction, tritium discharge into the atmosphere and atmospheric dispersion factor. The annual mean concentrations of atmospheric HTO and precipitation were correlated with the amount of gaseous tritium released from the Wolsung NPP. The tritium concentrations in precipitation decrease exponentially with an increase of the distance from the Wolsung NPP site.

  13. 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.

  14. 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

  15. Radiological safety system based on real-time tritium-in-air monitoring indoors and in effluents

    International Nuclear Information System (INIS)

    Bidica, N.; Sofalca, N; Balteanu, O.; Srefan, I.

    2006-01-01

    Exposure to tritium is an important health hazard in any tritium processing facility so that implementing a real-time tritium monitoring system is necessary for its operation in safety conditions. The tritium processing facility operators need to be informed at any time about the in-air tritium concentration indoors or in the stack effluents, in order to detect immediately any leaks in tritium containments, or any releases inside the buildings or to the environment. This information is very important for adopting if necessary protection measures and correcting actions as quickly as possible. In this paper we describe an improved real-time tritium monitoring system designed for the Heavy Water Detritiation Pilot Plant of National Institute for Cryogenics and Isotopes Separation, Rm. Valcea, Romania. The design of the Radiological Safety System implemented for the ICIT Water Detritiation Pilot Plant is intended to provide the maximum safety level based on the ALARA concept. The main functions of tritium monitoring system are: - monitoring the working areas and gaseous effluents by determination of the tritium-in-air activity concentration; - local and remote data display; - assessing of environment dose equivalent rates and dose equivalents in the working environment (for personnel exposure control and work planning); - assessing the total tritium activity released to the environment through ventilation exhaust stack; - safety functions, i.e., local and remote, locking/unlocking personnel access, process shut-down in emergency conditions and start of the air cleaning systems. With all these features our tritium monitoring system is really a safety system adequate for personnel and environmental protection. (authors)

  16. ITER vent filter conceptual study. Final report ITER Task S71 TD 70 (NID-9e) EU Task SEA 3-4

    Energy Technology Data Exchange (ETDEWEB)

    Blomquist, R; Johansson, Kjell; Natalizio, A; Nilsson, Lars; Shen, K; Sjoeberg, A

    1995-02-01

    The aim of this study is to investigate vent filter technologies that could be applied to ITER in accidental situations and to discuss how effective they could be in reducing accidental releases of gases, vapors and aerosols to the atmosphere. The study`s main contents are: A description of vent filter technologies as they are used for fission reactors; A review of appropriate tritium removal technologies. The vent filter part deals mainly with aerosol removal. The tritium part includes removal of both elemental and oxidized tritium from the vent flow. An important conclusion is that current fission reactor vent filter systems do not remove elemental tritium from the vent stream. A third part of the study is scoping thermal-hydraulic and aerosol transport analyses in order to find realistic vent flow rates and aerosol transport characteristics. Theses analyses, which are based on ITER design data assumed at the time of analysis, have been performed by means of the codes RELAP5 and CONTAIN. Release estimates are also included. Suggestions and recommendations are given regarding how to design a reliable vent filter system. Suggested vent systems include: Vent ducts from vacuum vessel and heat exchanger departments, a water pool for condensation of steam, and equipment for removal of aerosols, elemental tritium and oxidized tritium. 42 refs, 6 tabs, 34 figs.

  17. ITER vent filter conceptual study. Final report ITER Task S71 TD 70 (NID-9e) EU Task SEA 3-4

    International Nuclear Information System (INIS)

    Blomquist, R.; Johansson, Kjell; Natalizio, A.; Nilsson, Lars; Shen, K.; Sjoeberg, A.

    1995-02-01

    The aim of this study is to investigate vent filter technologies that could be applied to ITER in accidental situations and to discuss how effective they could be in reducing accidental releases of gases, vapors and aerosols to the atmosphere. The study's main contents are: A description of vent filter technologies as they are used for fission reactors; A review of appropriate tritium removal technologies. The vent filter part deals mainly with aerosol removal. The tritium part includes removal of both elemental and oxidized tritium from the vent flow. An important conclusion is that current fission reactor vent filter systems do not remove elemental tritium from the vent stream. A third part of the study is scoping thermal-hydraulic and aerosol transport analyses in order to find realistic vent flow rates and aerosol transport characteristics. Theses analyses, which are based on ITER design data assumed at the time of analysis, have been performed by means of the codes RELAP5 and CONTAIN. Release estimates are also included. Suggestions and recommendations are given regarding how to design a reliable vent filter system. Suggested vent systems include: Vent ducts from vacuum vessel and heat exchanger departments, a water pool for condensation of steam, and equipment for removal of aerosols, elemental tritium and oxidized tritium. 42 refs, 6 tabs, 34 figs

  18. Secondary containment system for a high tritium research cryostat

    International Nuclear Information System (INIS)

    Tsugawa, R.T.; Fearon, D.; Souers, P.C.; Hickman, R.G.; Roberts, P.E.

    1976-01-01

    A 4.2- to 300-K liquid helium cryostat has been constructed for cryogenic samples of D--T containing up to 4 x 10 14 dis/s (10,000 Ci) of tritium radioactivity. The cryostat is enclosed in a secondary box, which acts as the ultimate container in case of a tritium release. Dry argon is flushed through the box, and the box atmosphere is monitored for tritium, oxygen, and water vapor. A rupture disk and abort tank protect the box atmosphere in case the sample cell breaks. If tritium breaks into the box, a powdered uranium getter trap reduces the 4 x 10 14 dis/s (10,000 Ci) to 4 x 10 9 dis/s (0.1 Ci) in 24 h. A backup palladium--zeolite getter system goes into operation if an overabundance of oxygen contaminates the uranium getter

  19. Distribution of tritium in a nuclear process heat plant with HTR

    International Nuclear Information System (INIS)

    Steinwarz, W.; Stoever, D.; Hecker, R.; Thiele, W.

    1984-01-01

    The application of HTR-process heat in chemical processes involves low contamination of the product by tritium permeation through the heat exchanger walls. According to conservative assumptions for the tritium release rate and based on experimental permeation data of the German R und D-program a tritium concentration in the PNP-product gas of about 10 pCi/g was calculated. The domestic use of the product gas in unvented kitchen ranges as the most important direct radiation exposure pathway then leads to an effective equivalent radiation dose of only 20 μrem/a. (orig.)

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. Preliminary Tritium Management Design Activities at ORNL

    International Nuclear Information System (INIS)

    Harrison, Thomas J.; Felde, David K.; Logsdon, Randall J.; McFarlane, Joanna; Qualls, A. L.

    2016-01-01

    Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritium mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year's efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.

  4. Preliminary Tritium Management Design Activities at ORNL

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Felde, David K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Logsdon, Randall J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); McFarlane, Joanna [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-01

    Interest in salt-cooled and salt-fueled reactors has increased over the last decade (Forsberg et al. 2016). Several private companies and universities in the United States, as well as governments in other countries, are developing salt reactor designs and/or technology. Two primary issues for the development and deployment of many salt reactor concepts are (1) the prevention of tritium generation and (2) the management of tritium to prevent release to the environment. In 2016, the US Department of Energy (DOE) initiated a research project under the Advanced Reactor Technology Program to (1) experimentally assess the feasibility of proposed methods for tritium mitigation and (2) to perform an engineering demonstration of the most promising methods. This document describes results from the first year’s efforts to define, design, and build an experimental apparatus to test potential methods for tritium management. These efforts are focused on producing a final design document as the basis for the apparatus and its scheduled completion consistent with available budget and approvals for facility use.

  5. Risk assessment of a fusion-reactor fuel-processing system

    International Nuclear Information System (INIS)

    Bruske, S.Z.; Holland, D.F.

    1983-07-01

    The probabilistic risk assessment (PRA) methodology provides a means to systematically examine the potential for accidents that may result in a release of hazardous materials. This report presents the PRA for a typical fusion reactor fuel processing system. The system used in the analysis is based on the Tritium Systems Test Assembly, which is being operated at the Los Alamos National Laboratory. The results of the evaluation are presented in a probability-consequence plot that describes the probability of various accidental tritium release magnitudes

  6. Unclassified information on tritium extraction and purification technology: attachment 1

    International Nuclear Information System (INIS)

    McNorrill, P.L.

    1976-01-01

    Several tritium recovery and purification techniques developed at non-production sites are described in the unclassified and declassified literature. Heating of irradiated Li-Al alloy under vacuum to release tritium is described in declassified reports of Argonne National Laboratory. Use of palladium membranes to separate hydrogen isotopes from other gases is described by Argonne, KAPL, and others. Declassified KAPL reports describe tritium sorption on palladium beds and suggest fractional absorption as a means of isotope separation. A thermal diffusion column for tritium enrichment is described in a Canadian report. Mound Laboratory reports describe theoretical and experimental studies of thermal diffusion columns. Oak Ridge reports tabulate ''shape factors'' for thermal diffusion columns. Unclassified journals contain many articles on thermal diffusion theory, experiments, and separation of gas mixtures by thermal diffusion columns; much of these data can be readily extended to the separation of hydrogen-tritium mixtures. Cryogenic distillation for tritium recovery is described in the Mound Laboratory reports. Process equipment such as pumps, valves, Hopcalite beds, and uranium beds are described in reports by ANL, KAPL, and MLM, and in WASH-1269, Tritium Control Technology

  7. In-situ tritium recovery from Li2O irradiated in fast neutron flux: BEATRIX-II initial results

    International Nuclear Information System (INIS)

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

    1990-10-01

    The BEATRIX-II experiment in FFTF is an in-situ tritium recovery experiment to evaluate the tritium release characteristics of Li 2 O and its stability under fast neutron irradiation to extended burnups. This experiment includes two specimens: a thin annular specimen capable of temperature transients and a larger temperature gradient specimen. During the first 85 days of the operating cycle of the reactor, the tritium recovery rate of a temperature transient capsule was examined as a function of temperature, gas flow rate, gas composition and burnup. Temperature changes in the range from 525 to 625 degree C resulted in decreasing tritium inventory with increasing temperature. Lower gas flow rates resulted in slightly lower tritium release rates while gas composition changes affected the tritium release rate significantly, more than either flow rate or temperature changes. Three different sweep gases were used: He with 0.1% H 2 , He with 0.01% H 2 , and pure He. Decreasing the amount of hydrogen in the sweep gas decreased the steady-state release rate by as much as a factor of two. A temperature gradient capsule is more prototypic of the conditions expected in a fusion blanket and was designed to provide data that can be used in evaluating the operational parameters of a solid breeder in a blanket environment. The operation of this canister during the first 85 EFPD cycle suggests that Li 2 O is a viable solid breeder material. 9 refs., 5 figs., 3 tabs

  8. Data on the bound tritium level in fish from the great French rivers

    International Nuclear Information System (INIS)

    Foulquier, L.; Pally, M.

    1982-01-01

    The sampling stations were chosen as a function of the French nuclear program. Considering the sampling periods (1977-1982) and the operating nuclear plants, three areas were determined: 1) affected by fallout, 2) affected by power plant releases and 3) also affected by the releases from a fuel reprocessing plant. Tritium levels in water in the first, second and third areas were estimated at 210, 330 and 490 pCil -1 respectively. In the first two areas, differences in bound tritium levels in fish were not significant (average value: 1230+-520 pci.kg -1 dry). Downstream from the reprocessing plant, the average value reached 5360+-1330. Tritium ratios in fish vs water varied according to the sampling points and were always above 1. (Data) (author)

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

    International Nuclear Information System (INIS)

    Maroni, V.A.

    1976-01-01

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

  10. Using the Tritium Plasma Experiment to evaluate ITER PFC safety

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  11. Design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

    International Nuclear Information System (INIS)

    Finn, P.A.; Reedy, G.T.; Homa, M.I.; Clemmer, R.G.; Pappas, G.; Slawecki, M.A.; Graczyk, D.G.; Bowers, D.L.; Clemmer, E.D.

    1983-01-01

    The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

  12. An interdisciplinary approach to modeling tritium transfer into the environment

    International Nuclear Information System (INIS)

    Galeriu, D; Melintescu, A.

    2005-01-01

    More robust radiological assessment models are required to support the safety case for the nuclear industry. Heavy water reactors, fuel processing plants, radiopharmaceutical factories, and the future fusion reactor, all have large tritium loads. While of low probability, large accidental tritium releases cannot be ignored. For Romania that uses CANDU600 for nuclear energy, tritium is the national radionuclide. Tritium enters directly into the life cycle in many physicochemical forms. Tritiated water (HTO) is leaked from most nuclear installations but is partially converted into organically bound tritium (OBT) through plant and animal metabolic processes. Hydrogen and carbon are elemental components of major nutrients and animal tissues and their radioisotopes must be modeled differently from those of most other radionuclides. Tritium transfer from atmosphere to plant and conversion into organically bound tritium strongly depend on plant characteristics, season, and weather conditions. In order to cope with this large variability and avoid expensive calibration experiments, we developed a model using knowledge of plant physiology, agrometeorology, soil sciences, hydrology, and climatology. The transfer of tritiated water to plant was modeled with resistance approach including sparse canopy. The canopy resistance was modeled using the Jarvis-Calvet approach modified in order to make direct use of the canopy photosynthesis rate. The crop growth model WOFOST was used for photosynthesis rate both for canopy resistance and formation of organically bound tritium. Using this formalism, the tritium transfer parameters were directly linked to processes and parameters known from agricultural sciences. Model predictions for tritium in wheat were close to a factor two, according to experimental data without any calibration. The model was also tested on rice and soybean and can be applied for various plants and environmental conditions. For sparse canopy, the model used coupled

  13. Tritium pollution in the Swiss luminous compound industry

    International Nuclear Information System (INIS)

    Krejci, K.; Zeller, Jr.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Varghese, C.; Khan, Z.A.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-15

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Hassanein, A.

    1999-01-01

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

  18. Water - The radiological health of rivers: releases are very much controlled downstream power plants. What do hospital releases represent? The Seine reserves a surprise

    International Nuclear Information System (INIS)

    Anon.

    2015-01-01

    After a brief presentation of the role of the IRSN in the control of the radioactivity present in waters and in the control and follow-up of all sources of radioactivity, a first article briefly present the hydro-collector network, indicates that some point samplings of sediment and aquatic species are performed, that a national network of beacons for a continuous radioactivity measurement is installed in the main French rivers, downstream nuclear installations, and that advanced measurement techniques are used to detect very small level of tritium. Maps giving a brief indication of the radiological condition of the Loire and Rhone are provided. A second article addresses the control of releases downstream power plants, and evokes the legal context and the associated objectives and produced documents. The third article discusses the risk associated with hospital wastes and releases (liquid and solid effluents), how radioactivity is controlled between the hospital and tap water distribution. The last article reports and comments the results obtained by an analysis of historical pollutions trapped in the sediments of the Seine: 40 year-old traces of plutonium have been discovered, due to an accidental release from a CEA installation in Fontenay-aux-Roses, with no detrimental impact on population or on sewer workers

  19. Tritium formation and elimination in light-water electronuclear plants

    International Nuclear Information System (INIS)

    Dolle, L.; Bazin, J.

    1977-01-01

    In light-water reactors, the tritium balance should be considered from both the working constraint and environmental pollution aspects. In light-water electronuclear stations with pressurized reactors using boric acid in solution for reactivity control, the amounts of tritium formed in the primary circuit are worthy of note. The estimations concerning the tritium production in a hypothetical 1000 MWe reactor are discussed. In the tritium build-up, the part which takes the tritium formed by fission in the fuel, owing to diffusion through cladding, is still difficult to estimate. The tritium balance in different working nuclear power stations are consequently of interest. But the tritium produced by ternary fission in the fuel is always much more abundant, and remains almost entirely confined in the uranium oxide if the fuel is clad with zircaloy. The annual quantity stored in the fuel elements is more than 20 times larger than that of the built up free tritium in the primary circuit water of a reactor. It reaches about 12,400 Ci in the hypothetical reactor. In the presently operated reprocessing plants, tritium is all going over in the effluents, and is almost entirely released in the environment. Taking into account the increasing quantities of high irradiated fuel to be reprocessed, it seems necessary to develop separation processes. Development work and tests have been achieved jointly by CEA and SAINT-GOBAIN TECHNIQUES NOUVELLES in order to: contain the tritium in the high activity part of the plant; and keep small the tritiated effluent volume, about 300 liters per ton of reprocessed uranium. It is then possible to envisage a storage for decay of isotopic separation processes. Such separation processes have been estimated by CEA assuming a daily output of 1500 liters of water containing 2,3 Ci.1 -1 of tritium, the desired decontamination factor being 100 [fr

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

    International Nuclear Information System (INIS)

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

    1990-09-01

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