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Sample records for iter tritium plant

  1. Tritium supply assessment for ITER and DEMOnstration power plant

    International Nuclear Information System (INIS)

    Ni, Muyi; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-01-01

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource

  2. Tritium supply assessment for ITER and DEMOnstration power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ni, Muyi, E-mail: muyi.ni@fds.org.cn; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

    2013-10-15

    Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource.

  3. 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

  4. Industrial cost assessment for ITER tritium plant system (water distillation, VPCE and ISS)

    International Nuclear Information System (INIS)

    Sood, S.K.; Kalyanam, K.M.; Fong, C.

    1995-01-01

    The objective of this Industrial Cost Assessment Task for ITER Tritium Plant System consists of providing and order of magnitude cost estimate for the following major subsystems, as outlined in the Scope of Task Agreement and Work Program: water distillation (WD) system, vapour phase catalytic exchange (VPCE) system and the isotope separation system (ISS). The methodology adopted in preparing the order of magnitude cost estimate for the above three subsystems of the ITER tritium plant system is based on building the estimate from the ground up, starting with equipment cost estimates, and adding labour activities separately for engineering, fabrication, assembly, testing installation commissioning, etc. The estimate has been developed assuming that the systems are to be engineered, fabricated and constructed in Canada, (to comply with the Codes, Standards, QA and Seismic Classification applicable in Canada) since information on ITER siting is not currently available. The estimate is based on Ontario Hydro in house cost data on similar systems and equipment, such as the heavy water upgrading plants. The cost estimates are not based on quotations from suppliers for specific ITER components, since this would require completion of detailed design and specifications. 4 refs., 9 tabs., 7 figs

  5. 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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  7. ITER Dynamic Tritium Inventory Modeling Code

    International Nuclear Information System (INIS)

    Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.

    2005-01-01

    A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER

  8. 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)

  9. Optimization of tritium management within the ITER project

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. 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.)

  11. Fire risk analysis in ITER tritium building

    International Nuclear Information System (INIS)

    Lignini, Franck; Uzan-Elbez, Joelle; Girard, Jean-Philippe; Porfiri, Maria Teresa; Rodriguez-Rodrigo, Lina

    2005-01-01

    Events, such as fire, have been considered in ITER documentation of low probability and a general approach has been defined in [Technical basis for the ITER final design, EDA Documentation Series I, No. 22, IAEA, Vienna, 2001] to be developed later for the ITER specific site. It was said that 'these hazards will be treated according to the industrial safety regulations and practices of the host country'. In the framework of studies for the European ITER site in Cadarache, an assessment of fire hazard has been done in order to ensure compliance with French safety requirements. In this report, a summary of existing laws is presented and an example of the deterministic approach to be followed for the preliminary safety report (PSR) is given on the analysis of tritium building design

  12. Tritium module for ITER/Tiber system code

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Design approach for safe tritium handling in ITER

    International Nuclear Information System (INIS)

    Ohira, Shigeru

    2002-01-01

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

  14. 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)

  15. 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

  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 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

  17. Tritium turnover in succulent plants

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  18. The Tritium Systems Test Assembly applicability to ITER

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1988-01-01

    The Tritium Systems Test Assembly (TSTA), is operated by the Los Alamos National Laboratory (LANL) under the sponsorship of the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The objectives of the TSTA project are to develop, demonstrate, and evaluate the exhaust gas processing and tritium related safety systems for the magnetic fusion energy program. The applicability of these processes for the ITER Tokamak is discussed

  19. 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.

  20. Development of ITER Tritium Storage Material

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-01-15

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

  1. Tritium inventory and permeation in the ITER breeding blanket

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  2. ITER plant systems

    International Nuclear Information System (INIS)

    Kolbasov, B.; Barnes, C.; Blevins, J.

    1991-01-01

    As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs

  3. In-vessel tritium retention and removal in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

    1998-06-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  4. In-vessel tritium retention and removal in ITER

    International Nuclear Information System (INIS)

    Federici, G.; Anderl, R.A.

    1998-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world's fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

  5. 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

  6. ITER safety task NID-5a: ITER tritium environmental source terms - safety analysis basis

    International Nuclear Information System (INIS)

    Natalizio, A.; Kalyanam, K.M.

    1994-09-01

    The Canadian Fusion Fuels Technology Project's (CFFTP) is part of the contribution to ITER task NID-5a, Initial Tritium Source Term. This safety analysis basis constitutes the first part of the work for establishing tritium source terms and is intended to solicit comments and obtain agreement. The analysis objective is to provide an early estimate of tritium environmental source terms for the events to be analyzed. Events that would result in the loss of tritium are: a Loss of Coolant Accident (LOCA), a vacuum vessel boundary breach. a torus exhaust line failure, a fuelling machine process boundary failure, a fuel processing system process boundary failure, a water detritiation system process boundary failure and an isotope separation system process boundary failure. 9 figs

  7. In-vessel dust and tritium control strategy in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M., E-mail: michiya.shimada@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France); Pitts, R.A.; Ciattaglia, S.; Carpentier, S.; Choi, C.H.; Dell Orco, G.; Hirai, T.; Kukushkin, A.; Lisgo, S.; Palmer, J.; Shu, W.; Veshchev, E. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

    2013-07-15

    A baseline strategy for dust and tritium-inventory control and recovery in ITER has been established and preparations are underway for its implementation. Limits on dust and tritium-inventory are an integral part of the ITER safety case and are fixed at 1 kg for tritium, 1000 kg for mobilisable dust and 11 kg (beryllium)/76 kg (tungsten) for dust on hot surfaces. Maximum average T-retention rates of ∼1 g/shot are estimated for baseline inductive operation at Q{sub DT} = 10, suggesting that the in-vessel T-retention could reach the administrative limit of 640 g in as little as ∼2 months of operation. Baking is expected to remove a significant fraction of the T co-deposited on the divertor targets. Despite large uncertainties, dust quantities are expected to remain well below safety limits over the divertor cassette lifetime. In situ aspiration during divertor cassette exchange is foreseen as the main dust removal technique.

  8. Tritium in nuclear power plants

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  9. The Chalk River Tritium Extraction Plant

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  10. The Chalk River Tritium Extraction Plant

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-07-01

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

  11. 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)

  12. Tritium control in reprocessing plants

    International Nuclear Information System (INIS)

    Goumondy, J.P.; Miquel, P.

    1977-01-01

    There is a danger that the T which is formed in water reactors will prove detrimental to the environment over the next few years, and studies have been undertaken to develop techniques to contain and process it where possible. In order to retain T, which is present largely in the fuel and on the possible to adapt for use in the conventional design of reprocessing plant. In this process T is maintained in the form of an aqueous solution in the high-active area of the plant. Control is achieved by restricting as far as possible the ingress of non-tritiated water into this area, and by setting up a tritiated water barrier at the first U and Pu extraction stage, stripping the tritium-containing solvent at that point with ordinary water. In this way the T can be extracted in a small volume of water with a view to intermediate storage, disposal at sea additional processing to remove the T from the water. Experiments carried out so far have demonstrated the effectiveness of the T barrier and have shown what equipment would be required for the application of the process in new reprocessing plants. (orig.) [de

  13. Diurnal variations of tritium uptake by plants

    International Nuclear Information System (INIS)

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

    1991-02-01

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

  14. Tritium in water and plants

    International Nuclear Information System (INIS)

    Koenig, L.A.; Winter, M.

    1977-10-01

    A summary is presented of the results obtained in programs on the measurement of the T concentration in the immediate and more distant environment of the Karlsruhe Nuclear Research Center (KFK). The amount of T released with the waste water and the exhaust air from the KFK in the years 1969 until 1976 is indicated. The total releases ranged from 2000 Ci/a to 5200 Ci/a in the period of reporting. The fractions contained in the exhaust air and in the waste water were subjected to considerable variations over the years. The results of measurements for water samples are presented as annual mean values. The annual values of measured results obtained for precipitations on the site clearly exceed the values measured at sampling stations outside the KFK. Of the surface waters monitored the rivers Rhine and Neckar showed the highest T values. In smaller flowing waters the T concentration is also influenced by the percentage of waste water from communities. Obviously the tritium content of the ground and drinking water depends on the depth of sampling. Drinking water raised from a small depth in the vicinity of the Rhine is subjected to the same variations as the water from the river Rhine. To find out relations to the T offer of the relevant media close to the plants, the T concentrations in tissue water of plants and in air humidity, ground water and precipitations were investigated. Variations of the T concentration in air humidity correlate with the variations of the T concentration in the tissue water of plants. The T concentration level in the tissue water of plants is close to the T concentration in air humidity. The following time constants and half-lives, respectively, are found: for oak and hornbeam leaves 2+-1 days, for spruce needles 3+-1.5 days, for pine needles 6+-3 days. The dispersion of T released into the air and into the water is dealt with briefly. (orig.) [de

  15. Tritium behaviour in higher plants

    International Nuclear Information System (INIS)

    Guenot, J.

    1984-05-01

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

  16. 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

  17. Improvement of tritium accountancy technology for the ITER fuel cycle safety enhancement

    International Nuclear Information System (INIS)

    O'hira, Shigeru; Hayashi, T.; Nakamura, H.

    1999-01-01

    In order to improve the safe handling and control of tritium for ITER fuel cycle, effective 'in-situ' tritium accounting methods have been developed at Tritium Process Laboratory in Japan Atomic Energy Research Institute under one of the ITER-EDA R and D Tasks. A remote and multi-location analysis of process gases by an application of laser Raman spectroscopy developed and tested could provide a measurement of hydrogen isotope gases with a detection limit of 0.3 kPa for 120 seconds analytical periods. An 'in-situ' tritium inventory measurement by application of a 'self assaying' storage bed with 25 g tritium capacity could provide a measurement with a required detection limit less than 1% and a design proof of a bed with 100 g tritium capacity. (author)

  18. Improvement of tritium accountancy technology for the ITER fuel cycle safety enhancement

    International Nuclear Information System (INIS)

    O'Hira, S.; Hayashi, T.; Nakamura, H.

    2001-01-01

    In order to improve the safe handling and control of tritium for ITER fuel cycle, effective ''in-situ'' tritium accounting methods have been developed at Tritium Process Laboratory in Japan Atomic Energy Research Institute under one of the ITER-EDA R and D Tasks. A remote and multi-location analysis of process gases by an application of laser Raman spectroscopy developed and tested could provide a measurement of hydrogen isotope gases with a detection limit of 0.3 kPa for 120 seconds analytical periods. An ''in-situ'' tritium inventory measurement by application of a ''self assaying'' storage bed with 25 g tritium capacity could provide a measurement with a required detection limit less than 1 % and a design proof of a bed with 100 g tritium capacity. (author)

  19. 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

  20. Lithium test module on ITER: engineering design of the tritium recovery system

    International Nuclear Information System (INIS)

    Finn, P.A.

    1988-01-01

    The design presented is an overview of the tritium recovery system for a lithium module on an ITER type reactor. The design of a tritium recovery system for larger blanket units, sectors, etc. could use the information developed in this report. A goal of this design was to ensure that a reliable, integrated performance of the tritium recovery system could be demonstrated. An equally important goal was to measure and account for the tritium in the liquid lithium blanket module and its recovery system in order to validate the operation of the blanket module

  1. Tritium Experience in Large Tokamaks: Application to ITER

    International Nuclear Information System (INIS)

    Skinner, C.H.; Gentile, C.; Hosea, J.; Mueller, D; Gentile, C.; Federici, G.; Haanges, R.

    1998-05-01

    Recent experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports on an IEA workshop that brought together scientists and engineers to share experience and expertise on all fusion-related tritium issues. Extensive discussion periods were devoted to exploring outstanding issues and identifying potential R ampersand D avenues to address them. This paper summarizes the presentations, discussions, and recommendations

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  3. Tritium absorption and desorption in ITER relevant materials: comparative study of tungsten dust and massive samples

    Energy Technology Data Exchange (ETDEWEB)

    Grisolia, C., E-mail: christian.grisolia@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Hodille, E. [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Chene, J.; Garcia-Argote, S.; Pieters, G.; El-Kharbachi, A. [CEA Saclay, SCBM, iBiTec-S, PC n° 108, 91191 Gifsur-Yvette (France); Marchetti, L.; Martin, F.; Miserque, F. [CEA Saclay, DEN/DPC/SCCME/LECA, F-91191 Gif-sur-Yvette (France); Vrel, D.; Redolfi, M. [LSPM, Université Paris 13, Sorbonne Paris Cité, UPR 3407 CNRS, 93430 Villetaneuse (France); Malard, V. [CEA, DSV, IBEB, Lab Biochim System Perturb, Bagnols-sur-Cèze F-30207 (France); Dinescu, G.; Acsente, T. [NILPRP, 409 Atomistilor Street, 77125 Magurele, Bucharest (Romania); Gensdarmes, F.; Peillon, S. [IRSN, PSN-RES/SCA/LPMA, Saclay, Gif-sur-Yvette, 91192 (France); Pegourié, B. [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Rousseau, B. [CEA Saclay, SCBM, iBiTec-S, PC n° 108, 91191 Gifsur-Yvette (France)

    2015-08-15

    Tritium adsorption and desorption from well characterized tungsten dust are presented. The dust used are of different types prepared by planetary milling and by aggregation technique in plasma. For the milled powder, the surface specific area (SSA) is 15.5 m{sup 2}/g. The particles are poly-disperse with a maximum size of 200 nm for the milled powder and 100 nm for the aggregation one. Prior to tritiation the particles are carefully de-oxidized. Both samples are experiencing a high tritium inventory from 5 GBq/g to 35 GBq/g. From comparison with massive samples and considering that tritium inventory increases with SSA, it is shown that surface effects are predominant in the tritium trapping process. Extrapolation to the ITER environment is undertaken with the help of a Macroscopic Rate Equation model. It is shown that, during the life time of ITER, these particles can exceed rapidly 1 GBq/g.

  4. ITER plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.A.

    2000-01-01

    The ITER site has not yet been determined. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electrical power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, demineralized water, steam and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radiation exposure of workers and the public. To prevent interference between services of different types and for efficient arrangement of buildings, structures and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographical centre of the site. The locations of the buildings on the generic site were selected to meet all design requirements at minimum total project cost. A similar approach was used to determine the locations of services above, at and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography, resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  5. Development of tritium plant system for fusion reactors. Achievements in the 14-year US-Japan collaboration

    International Nuclear Information System (INIS)

    Nishi, Masataka; Yamanishi, Toshihiko; Shu, Wataru

    2003-01-01

    Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safe-operation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore, distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work. (author)

  6. Development of Tritium Plant System for Fusion Reactors - Achievements in the 14-year US-Japan Collaboration -

    Science.gov (United States)

    Nishi, Masataka; Yamanishi, Toshihiko; Shu, Wataru

    Fuel processing technology and tritium safe-handling technology have been developed through US/DOE-JAERI collaboration from 1987 till 2001, and the technologies to construct the tritium plant system of ITER have been made currently available. This paper overviews the major achievements of this collaborative researches over fourteen years, which were performed mainly at the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory (LANL). The tritium plant system consists mainly of a fuel processing system, which includes a fuel cycle system and a blanket tritium recovery system, and a tritium confinement/removal system. The fuel cycle system recovers fuel from plasma exhaust gas and recycles it. In the collaboration, major key components and subsystems were developed, and the performance of the integrated system was successfully demonstrated over its one-month operation in which plasma exhaust model gas was processed at a processing rate of up to 1/6 level of the ITER. The technological basis of the fuel cycle system was thus established. Blanket tritium recovery technology was also successfully demonstrated using the TSTA system. Through the successful safeoperation of the TSTA, reliability of tritium confinement/removal system was verified basically. In addition, much data to confirm or enhance safety were accumulated by experiments such as intentional tritium release in a large room. Furthermore,distribution of tritium contamination in the vacuum vessel of the TFTR, a large tokamak of the Princeton Plasma Physics Laboratory (PPPL), was investigated in this work.

  7. 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

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

    Science.gov (United States)

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

    2017-12-01

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

  9. 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

  10. In-vessel tritium retention and removal in ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Brooks, J.N. [Argonne National Lab., IL (United States); Iseli, M. [ITER Naka Joint Work Site, Naka-gun (Japan); Wu, C.H. [EFDA Close Support Unit, Garching (Germany)

    2001-07-01

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., {proportional_to}350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

  11. In-Vessel Tritium Retention and Removal in ITER-FEAT

    Science.gov (United States)

    Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

  12. In-vessel tritium retention and removal in ITER-FEAT

    International Nuclear Information System (INIS)

    Federici, G.; Brooks, J.N.; Iseli, M.; Wu, C.H.

    2001-01-01

    Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ∝350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

  13. Overview of R and D at TLK for process and analytical issues on tritium management in breeder blankets of ITER and DEMO

    International Nuclear Information System (INIS)

    Demange, D.; Alecu, C.G.; Bekris, N.; Borisevich, O.; Bornschein, B.; Fischer, S.; Gramlich, N.; Köllö, Z.; Le, T.L.; Michling, R.; Priester, F.; Röllig, M.; Schlösser, M.; Stämmler, S.; Sturm, M.; Wagner, R.; Welte, S.

    2012-01-01

    Highlights: ► We present advanced processes and analytics to improve tritium management. ► Membranes and membrane reactors can minimise tritium residence time and inventory. ► Spectroscopic methods can ensure on-line and near to real time tritium measurement. - Abstract: Safe, reliable, and efficient tritium management in the breeder blanket will have to face unprecedented technological challenges. Beside the efficiency for tritium recovery from the breeder blanket (Tritium Extraction (TES) and Coolant Purification Systems (CPS)), the accuracy for tritium tracking between the inner and the outer fuel cycle must also be demonstrated. This paper focuses on the recent R and D carried out at the Tritium Laboratory Karlsruhe to tackle these issues. For ITER, the recently consolidated TES and CPS designs comprise adsorption columns and getter beds operated in semi-continuous mode. Different approaches for the tritium accountancy stage (TAS) have been evaluated. Balancing static (batch-wise gas collection at the TBM outlets and the tritium plant) or dynamic (in/on-line) approaches with respect to the expected analytical performances and integration issues, the first conceptual design of the TAS for EU TBMs is presented. For DEMO, the overall strategy for tritium recovery and tracking has been revisited. The necessity for on-line real-time tritium accountancy and improved process efficiency suggest the use of continuous processes such as permeator and catalytic membrane reactor. The main benefits combining the PERMCAT process with advanced membranes is discussed with respect to process improvements and facilitated accountancy using spectroscopic methods.

  14. Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

    Directory of Open Access Journals (Sweden)

    G. De Temmerman

    2017-08-01

    Full Text Available As a licensed nuclear facility, ITER must limit the in-vessel tritium (T retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

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

    Czech Academy of Sciences Publication Activity Database

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

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

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

    NARCIS (Netherlands)

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

    2010-01-01

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

  17. 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.)

  18. Experimental evaluation of tritium permeation through stainless steel tubes of heat exchanger from primary to secondary water in ITER

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Nishi, Masataka

    2004-01-01

    Tritium permeation through heat exchanger from primary cooling water to secondary cooling water has been investigated experimentally with SS316L heat exchanger under simulated ITER (international thermonuclear experimental reactor) operation condition in order to establish the tritium permeation evaluation method through the heat exchanger. As the result, the permeation rate of aqueous tritium was found to be about three orders smaller than that of the gaseous tritium. Tritium permeation through the heat exchanger in ITER was then evaluated, and it was revealed that total tritium permeation amount based on obtained aqueous permeability was about one order less than that with the former method with the gaseous permeability and putting the permeation reduction factor as 1000. Evaluated tritium permeation amount into secondary water during 20 years was quite small, which could be considered as negligible from the safety viewpoint

  19. Development of tritium fuel processing system using electrolytic reactor for ITER

    International Nuclear Information System (INIS)

    Yamanishi, T.; Kawamura, Y.; Iwai, Y.

    2001-01-01

    The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

  20. Development of tritium fuel processing system using electrolytic reactor for ITER

    International Nuclear Information System (INIS)

    Yamanishi, Toshihiko; Kawamura, Y.; Iwai, Y.

    1999-01-01

    The system composed of a palladium diffuser and an electrolytic reactor was proposed, and was developed for a Fuel Cleanup system of ITER. The performance of the system was studied in a stand-alone test in detail. A fuel simulation loop of ITER was constructed by connecting the developed Fuel Cleanup and Hydrogen Isotope Separation systems; and the function of each system in the loop was demonstrated. For the tritium recovery from the exhaust gas at He glow discharge cleaning of vacuum chamber of ITER, a cryogenic molecular sieve bed system was proposed and demonstrated. (author)

  1. Behaviour of tritium in the plant domain

    International Nuclear Information System (INIS)

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

    1979-01-01

    An attempt is made to review in three different ways the present status regarding the uptake, retention and release patterns of plants of different trophic levels exposed to a tritiated atmosphere. The mean residence time for the first major component is found to be within 1-12 days for all the terrestrial vegetations. The second and the third components form only a small fraction of the total tritium uptake and show significantly different mean residence times. The aquatic plant, Hydrilla verticellata, also shows very small residence time for the first component (0.4 h). An isotopic discrimination of about 30-40% against the incorporation of tissue-bound tritium (TBT) is found in some algal and aquatic species, and this is enhanced to as high as 90-99% for the terrestrial vegetations. When conditions of constant specific activity are maintained, even for the terrestrial vegetations high specific activity ratios are obtained in the few cases cited. The nucleic acid fraction has shown relative specific activity ratios of 0.37 and 1.0 for pea seedlings and algae respectively. In maize, though carbohydrate is a major fraction fat has shown maximum fixation of TBT. Mathematical resolution procedures to determine the mean residence times for TFWT/TBT fractions either from separated or composite sample fractions have been reviewed. TFWT time-activity profiles are built up for succulent plants from diffusion, xylem flow and the utilization of TFWT in the synthesis of TBT. A good agreement between the calculated and experimental values is obtained using molecular diffusion coefficients for succulent plants. The applicability of this concept for larger trees is discussed

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  5. Distillation plant for tritium enrichment in metallic lithium

    International Nuclear Information System (INIS)

    Barnert, E.; Butzek, D.; Cordewiner, J.; Heinrichs, E.

    1984-06-01

    To close the external fuel cycle of fusion reactors, the tritium obtained from lithium must be separated off. One way of doing this is by high-temperature distillation and subsequent permeation. The construction of high-temperature distillation plant is described. For the time being, deuterium is processed instead of tritium. (orig.) [de

  6. Review of the ITER diagnostics suite for erosion, deposition, dust and tritium measurements

    Energy Technology Data Exchange (ETDEWEB)

    Reichle, R., E-mail: roger.reichle@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Andrew, P. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Bates, P. [F4E, Torres Diagonal Litoral B3, Barcelona (Spain); Bede, O.; Casal, N.; Choi, C.H.; Barnsley, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Damiani, C. [F4E, Torres Diagonal Litoral B3, Barcelona (Spain); Bertalot, L. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Dubus, G. [F4E, Torres Diagonal Litoral B3, Barcelona (Spain); Ferreol, J.; Jagannathan, G.; Kocan, M.; Leipold, F.; Lisgo, S.W.; Martin, V.; Palmer, J.; Pearce, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Philipps, V. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich GmbH, Association EURATOM – Forschungszentrum Jülich, D-52425 Jülich (Germany); Pitts, R.A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); and others

    2015-08-15

    Dust and tritium inventories in the vacuum vessel have upper limits in ITER that are set by nuclear safety requirements. Erosion, migration and re-deposition of wall material together with fuel co-deposition will be largely responsible for these inventories. The diagnostic suite required to monitor these processes, along with the set of the corresponding measurement requirements is currently under review given the recent decision by the ITER Organization to eliminate the first carbon/tungsten (C/W) divertor and begin operations with a full-W variant Pitts et al. [1]. This paper presents the result of this review as well as the status of the chosen diagnostics.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

  8. Tritium and heat management in ITER Test Blanket Systems port cell for maintenance operations

    Energy Technology Data Exchange (ETDEWEB)

    Giancarli, L.M., E-mail: luciano.giancarli@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Cortes, P.; Iseli, M.; Lepetit, L.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Livingston, D. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom); Nevière, J.C. [Comex-Nucleaire, 13115 Saint Paul Lez Durance (France); Pascal, R. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ricapito, I. [Fusion for Energy, Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Shu, W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wyse, S. [Frazer-Nash Consultancy Ltd., Stonebridge House, Dorking Business Park, Dorking, Surrey RH4 1HJ (United Kingdom)

    2014-10-15

    Highlights: •The ITER TBM Program is one of the ITER missions. •We model a TBM port cell with CFD to optimize the design choices. •The heat and tritium releases management in TBM port cells has been optimized. •It is possible to reduce the T-concentration below one DAC in TBM port cells. •The TBM port cells can have human access within 12 h after shutdown. -- Abstract: Three ITER equatorial port cells are dedicated to the assessment of six different designs of breeding blankets, known as Test Blanket Modules (TBMs). Several high temperature components and pipework will be present in each TBM port cell and will release a significant quantity of heat that has to be extracted in order to avoid the ambient air and concrete wall temperatures to exceed allowable limits. Moreover, from these components and pipes, a fraction of the contained tritium permeates and/or leaks into the port cell. This paper describes the optimization of the heat extraction management during operation, and the tritium concentration control required for entry into the port cell to proceed with the required maintenance operations after the plasma shutdown.

  9. Tritium concentration in the heavy water upgrading plants

    International Nuclear Information System (INIS)

    Croitoru, C.; Pop, F.; Titescu, Gh.; Dumitrescu, M.; Ciortea, C.; Stefanescu, I.; Peculea, M.; Pitigoi, Gh.; Trancota, D. . E-mail of corresponding author: croitoru@icsi.ro; Croitoru, C.)

    2005-01-01

    In the course of time heavy water used in CANDU nuclear power plants, as moderator or coolant, degrades, as a result of its impurification with light water and tritium. Concentration diminution below 99.8% mol for moderator and 99.75% mol for coolant causes an inefficient functioning of CANDU reactor. By isotopic distillation, light water is removed. Simultaneously tritium concentration takes place. The heavy water upgrading plant from Cernavoda is an isotopic separation cascade with two stages. The paper presents, for this plant, a theoretical study of the tritium concentration. (author)

  10. 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.)

  11. 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

  12. Evolution of ITER tritium confinement strategy and adaptation to Cadrache site conditions and French regulatory requirements

    International Nuclear Information System (INIS)

    Murdoch, D.

    2007-01-01

    The ITER Nuclear Buildings include the Tokamak, Tritium and Diagnostic Buildings (Tokamak Complex) and the Hot Cell and Low Level Radioactive Waste Buildings. The Tritium Confinement Strategy of the Nuclear Buildings comprises key features of the Atmosphere and Vent Detritiation Systems (ADS/VDS) and the Heating, Ventilation and Air Conditioning (HVAC) Systems. The designs developed during the ITER EDA (Engineering Design Activities) for these systems need to be adapted to the specific conditions of the Cadarache site and modified to conform with the regulatory requirements applicable to Installations Nucleaires de Base (INB) - Basic Nuclear Installations - in France. The highest priority for such adaptation has been identified as the Tritium Confinement of the Tokamak Complex and the progress in development of a robust, coherent design concept compliant with French practice is described in the paper. The Tokamak Complex HVAC concept for generic conditions was developed for operational cost minimisation under more extreme climatic conditions (primarily temperature) than those valid for Cadarache, and incorporated recirculation of a large fraction of the air flow through the HVAC systems to achieve this objective. Due to the impracticality of precluding the spread of contamination from areas of higher activity to less contaminated areas, this concept has been abandoned in favour of a once-through configuration, which requires a complete redesign, with revised air change rates, module sizes, layout, redundancy provisions and other features. The ADS/VDS concept developed for the generic design of the ITER Tokamak Complex is undergoing a radical revision in which the system architecture, module sizing and basic process are being optimised for the Cadarache conditions. Investigation is being launched into the implementation of a wet stripper concept to replace the molecular sieve (MS) beds incorporated in the generic design, where concerns have been raised over low

  13. 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

  14. 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

  15. Internal dose from tritium at Wolsung nuclear power plant

    International Nuclear Information System (INIS)

    Hee Geun Kim; Jeong Yull Dho; Myung Jae Song

    1995-01-01

    Tritium is produced in large quantities at heavy water nuclear power reactors via the neutron activation reaction 2 H(n,γ) 3 H. At Wolsung nuclear power plant which has a CANDU reactor, the tritium concentrations in coolant and in moderator systems are 1.5 Ci/Kg-D 2 O and 35 Ci/kg-D 2 O, respectively, after 12 years of operation. The airborne tritium concentration in main access area is normally less than 5 MPCa except short-term peaks. The average tritium concentrations in main access controlled areas are normally less than 100 MPCa. Tritium is mainly present in the air of workplace of CANDU reactors as a tritiated water vapour. Airborne tritiated water vapour enters the workers body via inhalation and absorption through skin and can result in a significant dose. The occupational doses from tritium at Wolsung NPP have been maintained below 1 man-Sv per year so far. The tritium contribution to the total plant man-Sv changes between 30 percent and 50 percent. For the mitigation of tritium inhalation, various protective equipment are being used at Wolsung NPP. The respirator system was devised at Wolsung NPP in order to remove tritiated water vapours from the inhaled air. A respirator is connected to a small plastic bottle filled with ice cubes. The system devised shows a good tritium removal efficiency. The air pressure drop through the ice cubes is minimal. The operation cost of the system is also very cheap. Further mitigation of tritium inhalation is heavily dependant on the source term reduction. One of the ultimate solutions is to introduce a tritium removal facility. (author). 7 figs., 3 tabs

  16. Improved iodine and tritium control in reprocessing plants

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  17. Studies on tritium incorporation into wheat plants after short-term exposure to atmospheric tritium

    International Nuclear Information System (INIS)

    Diabate, S.; Strack, S.; Raskob, W.

    1996-01-01

    The paper summarizes the results of a series of laboratory experiments to study the uptake, loss, conversion and translocation of tritium in wheat plants following a short-term exposure to atmospheric tritiated water vapour (HTO) under laboratory conditions. The experiments were accompanied by the development of a Plant-OBT-Model to calculate the tritium behaviour in wheat. Exposures of potted plants were carried out between anthesis and maturity, under day conditions at two different light intensities (900 μmol m -2 s -1 and 120 μmol m -2 s -1 photosynthetic active radiation) and under night conditions. In leaves, the tritium uptake into tissue water tritium (TWT) was about four times lower under night conditions than day conditions. Organically bound tritium (OBT) was generated in leaves, stems and ears under day as well as under night conditions. The initial relative OBT concentrations in leaves observed under night conditions were about 50% of those under day conditions. OBT was translocated into the grain in dependence on the growth rate of the grain. Due to incorporation of new organic matter with lower OBT concentration into the grain, the specific OBT concentrations decreased slightly until harvest but the total OBT was rather constant. Once translocation to grain has taken place, OBT is lost only slowly. The growth of the plants has been calibrated with the measured growth data of winter wheat and spring wheat. Subsequently, the tritium incorporation was calibrated using the results of the exposure experiments in the same year. The final OBT concentration in the grain can be predicted with sufficient precision. However, the modelling of the OBT formation and turnover processes right after exposure to tritium needs improvement. A comprehensive validation of the model with independent data sets is still necessary. (J.P.N.)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  19. Tritium technology. A Canadian overview

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-10-01

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

  20. Tritium technology. A Canadian overview

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    2002-01-01

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

  1. ITER SAFETY TASK NID-5D: Operational tritium loss and accident investigation for heat transport and water detritiation systems

    International Nuclear Information System (INIS)

    Kalyanam, K.M.; Fong, C.; Moledina, M.; Natalizio, A.

    1995-02-01

    The task objectives are to: a) determine major pathways for tritium loss during normal operation of the cooling systems and water detritiation system, b) estimate operational losses and environmental tritium releases from the heat transport and water detritiation systems of ITER, and c) prepare a preliminary Failure Modes and Effects Analysis (FMEA) for the ITER Water Detritiation System. The analysis will be used to estimate chronic environmental tritium releases (airborne and waterborne) for the ITER Cooling Systems and Water Detritiation System. The assessment will form the basis for demonstrating the acceptability of ITER for siting in the Early Safety and Environmental Characterization Study (ESECS), to be issued in early 1995. (author). 7 refs., 10 tabs., 11 figs

  2. Radiation risk analysis of tritium in PWR plants

    International Nuclear Information System (INIS)

    Yang Maochun; Wang Shimin

    1999-03-01

    Tritium is a common radionuclide in PWR nuclear power plant. In the normal operation conditions, its radiation risk to plant workers is the internal radiation exposure when tritium existing in air as HTO (hydrogen tritium oxide) is breathed in. As the HTO has the same physical and chemical characteristics as water, the main way that HTO entering the air is by evaporation. There are few opening systems in Nuclear Power Plant, the radiation risk of tritium mainly exists near the area of spent fuel pit and reactor pit. The highest possible radiation risk it may cause--the maximum concentration in air is the level when equilibrium is established between water and air phases for tritium. The author analyzed the relationship among the concentration of HTO in water, in air and the water temperature when equilibrium is established, the equilibrated HTO concentration in air increases with HTO concentration in water and water temperature. The analysis revealed that at 30 degree C, the equilibrated HTO concentration in air might reach 1 DAC (derived air concentration) when the HTO concentration in water is 28 GBq/m 3 . Owing to the operation of plant ventilation systems and the existence of moisture in the input air of the ventilation, the practical tritium concentration in air is much lower than its equilibrated levels, the radiation risk of tritium in PWR plant is quite limited. In 1997, Daya Bay Nuclear Power Plant's practical monitoring result of the HTO concentration in the air of the nuclear island and the urine of workers supported this conclusion. Based on this analysis, some suggestions to the reduction of tritium radiation risk were made

  3. Catalytic membrane reactors for tritium recovery from tritiated water in the ITER fuel cycle

    International Nuclear Information System (INIS)

    Tosti, S.; Violante, V.; Basile, A.; Chiappetta, G.; Castelli, S.; De Francesco, M.; Scaglione, S.; Sarto, F.

    2000-01-01

    Palladium and palladium-silver permeators have been obtained by coating porous ceramic tubes with a thin metal layer. Three coating techniques have been studied and characterized: chemical electroless deposition (PdAg film thickness of 10 μm), ion sputtering (about 1 μm) and rolling of thin metal sheets (50 μm). The Pd-ceramic membranes have been used for manufacturing catalytic membrane reactors (CMR) for hydrogen and its isotopes recovering and purifying. These composite membranes and the CMR have been studied and developed for a closed-loop process with reference to the design requirements of the international thermonuclear experimental reactor (ITER) blanket tritium recovery system in the enhanced performance phase of operation. The membranes and CMR have been tested in a pilot plant equipped with temperature, pressure and flow-rate on-line measuring and controlling devices. The conversion value for the water gas shift reaction in the CMR has been measured close to 100% (always above the equilibrium one, 80% at 350 deg. C): the effect of the membrane is very clear since the reaction is moved towards the products because of the continuous hydrogen separation. The rolled thin film membranes have separated the hydrogen from other gases with a complete selectivity and exhibited a slightly larger mass transfer resistance with respect to the electroless membranes. Preliminary tests on the sputtered membranes have also been carried out with a promising performance. Considerations on the use of different palladium alloy in order to improve the performances of the membranes in terms of permeation flux and mechanical strength, such as palladium/yttrium, are also reported

  4. 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.)

  5. Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket

    International Nuclear Information System (INIS)

    Spannagel, G.; Gierszewski, P.

    1991-01-01

    At the Karlsruhe Nuclear Research Center (KfK) a flexible tool is being developed to simulate the dynamics of tritium inventories. This tool can be applied to any tritium handling system, especially to the fuel cycle components of future nuclear fusion devices. This instrument of simulation will be validated in equipment to be operated at the Karlsruhe Tritium Laboratory. In this study tritium inventories in a NET/ITER type fuel cycle involving a lithium salt solution blanket are investigated. The salt solution blanket serves as an example because it offers technological properties which are attractive in modeling the process; the example does not impair the general validity of the tool. Usually, the operation strategy of complex structures will deteriorate due to failures of the subsystems involved. These failures together with the reduced availability ensuing from them will be simulated. The example of this study is restricted to reduced availabilities of two subsystems, namely the reactor and the tritium recovery system. For these subsystems the influence of statistically varying intervals of operation is considered. Strategies we selected which are representative of expected modes of operation. In the design of a fuel cycle, care will be taken that prescribed availabilities of the subsystems can be achieved; however, the description of reactor operation is a complex task since operation breaks down into several campaigns for which rules have been specified which enable determination of whether a campaign has been successful and can be stopped. Thus, it is difficult to predict the overall behavior prior to a simulation which includes stochastic elements. Using the example mentioned above the capabilities of the tool will be illustrated; besides the presentation of results of inventory simulation, the applicability of these data will be discussed. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  7. Measurement of Tritium Activity in Plants by Ice Extraction Method

    International Nuclear Information System (INIS)

    Pelled, O.; Ovad, S.; Tubul, Y.; Tsroya, S.; Gonen, R.; Abraham, A.; Weinstein, M.; German, U.

    2014-01-01

    Tritium is produced primarily by interactions of cosmic rays with the atmosphere. However, nuclear installations may add significantly tritium to the surroundings, increasing its concentration. The main sources of tritium released by man are linked to the nuclear power cycle: nuclear power stations, nuclear fuel reprocessing plants or tritium production plants. Tritium is found in the environment mainly as tritiated water, in gaseous or liquid form (HTO, T2O), in the surrounding air and in soil. It accumulates in plants, which may use as a measure to the level of tritium concentration in the environment. The most common routes of tritium uptake from the environment in plants are from atmospheric humidity and by precipitation water which entered the soil. The fraction of tritium bound to the plant tissue is small compared to that present as tritiated water in the plant (from 0.06% to 0.3% for growing crops). The tritiated water uptake is through the roots, as tritiated water from the soil follows a pathway similar to that of ordinary water. As most tritium in plants consists of tritiated water, the measurements of only the tissue free water tritium concentration (as HTO or T2O) gives an accurate estimate of the tritium content in the plant. Analyzing free tritium in biological matrices usually requires using the freeze-drying method to extract the water from the sample, and then measure the water collected in a cold trap with a Liquid Scintillation Counter (LSC). The 'freeze-drying' occurs because of the sublimation of the frozen water inside the plant, that takes place when the temperature is beneath the triple point and the vapour pressure is low. In the temperature range of -5° to -10° C the mechanism that plants use to avoid freezing is drawing of water from the cell protoplasm into the intercellular spaces. Changes in cell membrane permeability allow water to leave the cell and enter the spaces between the cells where it freezes instead of freezing within the

  8. Monitoring and management of tritium from the nuclear power plant effluent

    Science.gov (United States)

    Zhang, Qiaoe; Liu, Ting; Yang, Lili; Meng, De; Song, Dahu

    2018-01-01

    It is important to regulate tritium nuclides from the nuclear power plant effluent, the paper briefly analyzes the main source of tritium, and the regulatory requirements associated with tritium in our country and the United States. The monitoring methods of tritium from the nuclear power plant effluent are described, and the purpose to give some advice to our national nuclear power plant about the effluent of tritium monitoring and management.

  9. Conceptual design of tritium accountancy system for LLCB TBM

    International Nuclear Information System (INIS)

    Patel, Rudreksh; Sircar, Amit

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-09-01

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

  11. A Study on Thermal Desorption of Deuterium in D-loaded SS316LN for ITER Tritium Removal System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Myungchul; Kim, Heemoon; Ahn, Sangbok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Jaeyong; Lee, Sanghwa; LanAhn, Nguyen Thi [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    Because Type B radwaste includes tritium on its inside, especially at vicinity of surface, tritium removal from the radwaste is a matter of concern in terms of the radwaste processes. Tritium behavior in materials is related with temperature. Considering a diffusion process, it is expected that tritium removal efficiency is enhanced with increasing baking temperature. However, there is a limitation about temperature due to facility capacity and economic aspect. Therefore, it is necessary to investigate the effect of temperature on the desorption behavior of Tritium in ITER materials. TDS analysis was performed in SS316LN loaded at 120, 240 and 350 °C. D2 concentration and the desorption peak temperature increased with increasing loading temperature. Using peak shift method with three ramp rates of 0.166, 0.332, and 0.5 °C/sec, trap activation energy of D in SS316LN loaded at 350 °C was 56 kJ/mol.

  12. 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.)

  13. Experimental investigation of buried tritium in plant and animal tissues

    International Nuclear Information System (INIS)

    Kim, S. B.; Workman, W. J. G.; Davis, P. A.

    2008-01-01

    Buried exchangeable tritium appears as part of organically bound tritium (OBT) in the traditional experimental determination of OBT. Since buried tritium quickly exchanges with hydrogen atoms in the body following ingestion, assuming that it is part of OBT rather than part of tritiated water (HTO) could result in a significant overestimate of the ingestion dose. This paper documents an experimental investigation into the existence, amount and significance of buried tritium in plant and fish samples. OBT concentrations in the samples were determined in the traditional way and also following denaturing with five chemical solutions that break down large molecules and expose buried tritium to exchange with free hydrogen atoms. A comparison of the OBT concentrations before and after denaturing, together with the concentration of HTO in the supernatant obtained after denaturing, suggests that buried OBT may exist but makes up less than 5% of the OBT concentration in plants and at most 20% of the OBT concentration in fish. The effects of rinse time and rinse water volumes were investigated to optimize the removal of exchangeable OBT from the samples. (authors)

  14. Technological exploitation of Deuterium–Tritium operations at JET in support of ITER design, operation and safety

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P., E-mail: paola.batistoni@enea.it [ENEA, Dipartimento Fusione e Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Campling, D. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Conroy, S. [Department of Physics and Astronomy, Uppsala University, SE-75120 Uppsala (Sweden); Croft, D. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Giegerich, T. [Karlsruhe Institute of Technology, P.O.Box 3640, D-76021 Karlsruhe (Germany); Huddleston, T.; Lefebvre, X. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lengar, I. [Jozef Stefan Institute, Reactor Physics Department, Jamova 39, SI-1000 Ljubljana (Slovenia); Lilley, S. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Peacock, A. [JET Exploitation Unit, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pillon, M. [ENEA, Dipartimento Fusione e Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Popovichev, S.; Reynolds, S. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Vila, R. [Laboratorio Nacional de Fusión, CIEMAT, Madrid (Spain); Villari, R. [ENEA, Dipartimento Fusione e Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Bekris, N. [ITER Physics Department, EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2016-11-01

    Highlights: • Within the framework of the EUROfusion programme, a work-package of technology projects (WPJET3) is being carried out in conjunction with the planned Deuterium–Tritium experiment on JET (DTE2). • The objective is to maximise the scientific and technological return of DT operations at JET in support of ITER. • Preparatory experiments, analyses and studies are carried out in several fusion nuclear technology areas. • These are: neutronics, neutron induced activation and damage in ITER materials, nuclear safety, tritium retention, permeation and outgassing, and waste production. • This paper presents the progress since the start of the project in 2014. - Abstract: Within the framework of the EUROfusion programme, a work-package of technology projects (WPJET3) is being carried out in conjunction with the planned Deuterium–Tritium experiment on JET (DTE2) with the objective of maximising the scientific and technological return of DT operations at JET in support of ITER. This paper presents the progress since the start of the project in 2014 in the preparatory experiments, analyses and studies in the areas of neutronics, neutron induced activation and damage in ITER materials, nuclear safety, tritium retention, permeation and outgassing, and waste production in preparation of DTE2.

  15. Conceptual design on interface between ITER and tritium extraction system of Chinese helium-cooled solid breeder test blanket module

    International Nuclear Information System (INIS)

    Zhang Long; Luo Tianyong; Feng Kaiming

    2010-01-01

    Tritium extraction system is essential for CN HCSB TBM for safety and technical reasons. Based on the assessments of system functions, integration issues and safety considerations, two main modifications of the system from previous design (Feng et al., 2007 ; Chen et al., 2008 ) are adopted: a)the TES has been split to 2 parts with one in port cell and another in tritium building. Q 2 O in the purge gas is reduced to Q 2 in a hot metal bed located in port cell; Q 2 is separated from the stream by a pair of cryogenic molecular sieve beds and a Pd/Ag diffuser located in tritium building. b)isotope separation process has been excluded. TES components sizes are estimated and space allocations are estimated. Required services and where and when they are needed are preliminary defined. Fluids delivered towards ITER tritium system are analyzed.

  16. Tritium conference days

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  17. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    2001-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  18. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    1999-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  19. 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

  20. Accounting strategy of tritium inventory in the heavy water detritiation pilot plant from ICIT Rm. Valcea

    International Nuclear Information System (INIS)

    Bidica, N.; Stefanescu, I.; Cristescu, I.; Bornea, A.; Zamfirache, M.; Lazar, A.; Vasut, F.; Pearsica, C.; Stefan, I.; Prisecaru, I.; Sindilar, G.

    2008-01-01

    In this paper we present a methodology for determination of tritium inventory in a tritium removal facility. The method proposed is based on the developing of computing models for accountancy of the mobile tritium inventory in the separation processes, of the stored tritium and of the trapped tritium inventory in the structure of the process system components. The configuration of the detritiation process is a combination of isotope catalytic exchange between water and hydrogen (LPCE) and the cryogenic distillation of hydrogen isotopes (CD). The computing model for tritium inventory in the LPCE process and the CD process will be developed basing on mass transfer coefficients in catalytic isotope exchange reactions and in dual-phase system (liquid-vapour) of hydrogen isotopes distillation process. Accounting of tritium inventory stored in metallic hydride will be based on in-bed calorimetry. Estimation of the trapped tritium inventory can be made by subtraction of the mobile and stored tritium inventories from the global tritium inventory of the plant area. Determinations of the global tritium inventory of the plant area will be made on a regular basis by measuring any tritium quantity entering or leaving the plant area. This methodology is intended to be applied to the Heavy Water Detritiation Pilot Plant from ICIT Rm. Valcea (Romania) and to the Cernavoda Tritium Removal Facility (which will be built in the next 5-7 years). (authors)

  1. Tritium in reprocessing plants: a study of the inventory, behavior, and the possibilities of separation of the tritium isotope

    International Nuclear Information System (INIS)

    Schnez, H.; Laser, M.; Merz, E.

    The path followed by tritium in reprocessing plants is described in quantitative terms based on the Purex and Thorex processes. Flowsheets are given for the Purex process which are on the one hand based on the present state of technology, but make provision at the same time for a recycling of the aqueous phase and a tritium separation. As an alternative approach the technical and economic aspects have been examined of a prior separation of the tritium after reduction of the fuel elements, followed by separation from the aqueous phase. The ultimate storage and transport of the separated tritium were included in the cost determination. The conclusion is reached as a result of the study that tritium separation is possible on scientific and technical grounds. The estimates made show the financial outlay to be less than 10 DM/GWh, but may on occasion be substantially higher, since no practical or industrial experience of the process is yet available

  2. Compartmental model for tritium persistence in the soil-plant system

    International Nuclear Information System (INIS)

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

    1977-01-01

    A three-component computer model for tritium persistence in the soil-plant system, on the basis of an exponential polynomial is attempted. A series of field experiments with four species of trees, viz. Cardia sebastina, Terminalia catappa, Aracaria bidwilli and Mangifera indica, were carried out to generate data for testing the model. It is observed that there are two short-term components and one long-term component for tritium mean residence time, corresponding to the three phases of tritium in the system, viz. Tissue-Free-Water-Tritium, labile component of Tissue-Bound-Tritium and non-labile component of Tissue-Bound-Tritium. (author)

  3. Tritium instrumentation for a fusion reactor power plant

    International Nuclear Information System (INIS)

    Shank, K.E.; Easterly, C.E.

    1976-09-01

    A review of tritium instrumentation is presented. This includes a discussion of currently available in-plant instrumentation and methods required for sampling stacks, monitoring process streams and reactor coolants, analyzing occupational work areas for air and surface contamination, and personnel monitoring. Off-site instrumentation and collection techniques are also presented. Conclusions are made concerning the adequacy of existing instrumentation in relation to the monitoring needs for fusion reactors

  4. ITER safety task NID-5a: on the effect of tritium sorption on building surfaces as a passive mechanism for reducing airborne concentrations

    International Nuclear Information System (INIS)

    Natalizio, A.

    1994-09-01

    As part of the source term analysis for ITER, it is important to investigate the potential for building surface interaction with tritium as a potential attenuation mechanism, which in the absence of engineered systems may be effective in reducing tritium environmental releases. (author). 2 refs., 8 tabs., 3 figs

  5. Investigation of tritium transfer to plants via the OBT/HTO and OBT/TFWT ratios

    International Nuclear Information System (INIS)

    Boyer, C.; Guetat, P.; Vichot, L.; Losset, Y.; Boyer, C.; Fromm, M.; Mavon, C.; Boyer, C.; Tatin-Froux, F.; Badot, P.M.

    2009-01-01

    After having recalled some concepts used to distinguish the various forms of water present in plants, and the reactions in presence of tritium, the authors discuss the biochemical behaviour of tritium. Then, they briefly report a literature survey and, for different plants and crops, indicate the values of the OBT/HTO (organically bound tritium/tritiated water) and OBT/TFWT (organically bound tritium/tissue free water tritium) ratios. They also report experimental studies performed by exposures of lettuces at the vicinity of a nuclear installation

  6. Progress in the integration of the ITER plant systems in auxiliary buildings

    International Nuclear Information System (INIS)

    Kotamäki, M.; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-01-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  7. Progress in the integration of the ITER plant systems in auxiliary buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kotamäki, M., E-mail: miikka.kotamaki@iter.org; Cordier, J.-J.; Kuehn, I.; Perrin, J.-L.; Sweeney, S.; Villedary, B.

    2016-11-01

    Highlights: • Usage of 3D CAD model in ITER configuration management presented. • 3D CAD models efficient in configuration and interface management. • Costly and schedule delaying changes avoided with proper interface management. • ITER buildings construction progressing. - Abstract: The ITER Tokamak machine is located in the center of Tokamak complex buildings consisting of Tokamak, Diagnostic, and Tritium buildings. Around the Tokamak complex there are over 30 auxiliary buildings housing various plant systems serving the Tokamak machine either directly or indirectly. The layout and space allocation of each auxiliary building and plant systems housed by the building are represented in the so-called Configuration Management Models (CMM). These are light 3D CAD models that define the required space envelope and the physical interfaces between the systems and the buildings and in-between the systems. The paper describes the CMM and interface management processes of the ITER auxiliary buildings and plant systems, and discusses the preparations for the plant installation phase. In addition, the current baseline configuration of the ITER plant systems in auxiliary buildings is described together with the recent developments in the configuration of different systems, as well as the current status of the construction of the buildings.

  8. Study of column construction and tritium inventory of cryogenic distillation columns for tritium plant of a fusion reactor

    International Nuclear Information System (INIS)

    Iwai, Yasunori; Yamanishi, Toshihiko; Okuno, Kenji

    1996-11-01

    Cryogenic distillation column system is believed to be best for large throughput of hydrogen isotope separation. The major disadvantage of the system is a large tritium inventory in liquid phase. From a viewpoint of safety of a fusion reactor, it is important to establish the design method of minimized the tritium inventory. Anumerical study to investigate the possible design improvement to reduce inventory while maintaining separation performance was conducted. The design conditions are based on ITER DDD report, and details are as follows, 1) Exhaust stream with less than 50 Ci/y loss of tritium, 2) 99.9% purity D2, and 3) 90% purity T2. 4) total inventory with less than 100g. In the design of ITER to process 32 mol/hr, 4 columns (3 of 4 columns make closed loop) are best and total inventory is 94g. Particularly recent design of ITER to process 320 mol/hr requires additional efforts to minimize the inventory. The simulation also suggests it is effective to reduce inventory to draw two different purity product streams. (author)

  9. ITER EDA Newsletter. V.3, no.3

    International Nuclear Information System (INIS)

    1994-03-01

    This ITER EDA Newsletter issue contains reports on (i) the completion of the ITER EDA Protocol 1, (ii) the signing of ITER EDA Protocol 2, (iii) a technical meeting on pumping and fuelling and (iv) a technical meeting on the ITER Tritium Plant

  10. Mean residence times for tritium in some terrestrial plants

    International Nuclear Information System (INIS)

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

    1975-01-01

    Uptake of tritiated water from a single or multiple exposure, its fixation and elimination from 3 terrestrial plants, Raphanus sativus L., Amaranthus viridis L. and Phyllanthus fraternus Webster under experimental field conditions are described. Tissue free water tritium (TFWT) attains peak concentration within 4 hr after a single exposure in R. sativus and then decreases with a single component from both leaves and root, with a biological half-time of 36.4 +- 4.2 and 48.5 +- 7.2 hr respectively. TFWT in leaves of A. viridis reaches its peak value within 0.5 hr of exposure and its decay exhibits 2 component fall, one having a very short half-life of 2.4 +- 0.7 hr and the other a long half-time of 86.1 +- 2.0 hr. The two component decay of tritium in TFWT is well represented in P. fraternus. The long-lived components are nearly 8 times of the short-lived ones, both in leaves and stems respectively. Generally, the long -lived component accounts for 15% of the peak TFWT. Tissue bound tritium (TBT) reaches to 4% of TFWT in Phyllanthus sp. TBT elimination time is many times longer than the experimental periods employed in the present study. (author)

  11. 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)

  12. ITER Construction--Plant System Integration

    International Nuclear Information System (INIS)

    Tada, E.; Matsuda, S.

    2009-01-01

    This brief paper introduces how the ITER will be built in the international collaboration. The ITER Organization plays a central role in constructing ITER and leading it into operation. Since most of the ITER components are to be provided in-kind from the member countries, integral project management should be scoped in advance of real work. Those include design, procurement, system assembly, testing, licensing and commissioning of ITER.

  13. 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

  14. The research status and prospect of tritium migration in soil-plant system

    International Nuclear Information System (INIS)

    Qiu Guohua

    2011-01-01

    In this paper, the research significance and progress of tritium migration in soil-plant system are briefly introduced, which includes spatial and temporal distribution, migration pattern and influence factors, chemical forms, mathematical models of tritium migration in different soil and plant. The research results are summarized, and the existing problems in research process are analyzed and discussed. (authors)

  15. ITER Safety Task NID-5A, Subtask 1-1: Source terms and energies - initial tritium source terms. Final report

    International Nuclear Information System (INIS)

    Fong, C.; Kalyanam, K.M.; Tanaka, M.R.; Sood, S.; Natalizio, A.; Delisle, M.

    1995-02-01

    The overall objective of the Early Safety and Environmental Characterization Study (ESECS) is to assess the environmental impact of tritium using appropriate assumptions on a hypothetical site for ITER, having the r eference s ite characteristics as proposed by the JCT. The objective of this work under the above subtask 1-1, NID-5a, is to determine environmental source terms (i.e., process source term x containment release fraction) for the fuel cycle and cooling systems. The work is based on inventories and process source terms (i.e., inventory x mobilization fraction), provided by others (under Task NID 3b). The results of this work form the basis for the determination, by others, of the off-site dose (i.e., environmental source term x dose/release ratio). For the determination of the environmental source terms, the TMAP4 code has been utilized (ref 1). This code is approved by ITER for safety assessment. 6 refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  17. 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

  18. 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)

  19. Tritium concentration in ambient air around Kaiga Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Srinivas S Kamath

    2018-01-01

    Full Text Available Tritium (3H is one of the important long-lived radioisotopes in the gaseous effluent from nuclear power plants. In this article, we present the results of 3H monitoring in ambient air samples around the Kaiga Nuclear Power Plant, on the West Coast of India. Air samples were collected by moisture condensation method and the 3H concentration was determined by liquid scintillation spectrometry. The 3H concentration in the 2.3–15 km zone of the power plant varied in the range of <0.04–6.64 Bq m−3 with a median of 0.67 Bq m−3. The samples collected from the 2.3–5 km zone of the power plant exhibit marginally higher concentration when compared to the 5–10 km and 10–15 km zones, which is as expected. The values observed in the present study for Kaiga region are similar to those reported from other nuclear power plants, both within India and other parts of the world.

  20. 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

  1. 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.

  2. 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.

  3. 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)

  4. 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

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

    International Nuclear Information System (INIS)

    O'Brien, C.J.

    1976-01-01

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

  6. 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.

  7. Tritium behavior pattern in some soil-plant systems in a tropical environment

    International Nuclear Information System (INIS)

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

    1975-01-01

    A study of the distribution pattern of tritium in the soil/plant environment gives valuable ecological information on the natural water balance. The results of such a study for the conditions obtaining in India are given in this paper. Field studies are carried out by injection of tritium into some soil/plant systems and following the transfer pathways. The method of extraction for tissue-free-water-tritium (TFWT) is based on the vacuum freeze-drying technique while the tissue-bound-tritium (TBT) is estimated by a modified version of the Shoniger method. The determination of residence time of tritium in aqueous and organic phase in a number of tropical trees has been carried out both for stem-injection as well as intake from the soil. From the results of this study the tree biomass and transpiration rates have been determined. The tritium profile over time, for an acute exposure in certain trees such as Morinda Tinetoria, Achras Sapota etc. shows significantly different patterns compared to the normal pattern shown by Mangifera Indica, Terminalia Catappa, Ficus Glomerata etc. The period of investigation in each case varied from 400 to 1000 h. In most of the cases, the TBT fractions were very low compared to TFWT fractions in the initial stages. The tritium behavior in the tree reflects significant characteristics of the tritium behavior in the soil system. The authors have found that the leaf sampling can be used as an indicator of total environmental tritium behavior. (author)

  8. A basic study on the ITER tritium storage vessel design and components

    International Nuclear Information System (INIS)

    Chung, H. S.; Ahn, D. H.; Kim, K. R.; Yim, S. P.; Paek, S. W.; Lee, M. S.; Lee, S. H.; Shim, M. H.

    2006-01-01

    The ZrCo getter beds are built of a primary vessel which contains the ZrCo powder mixed with Cu spheres of less than one mm diameter and of a secondary outer vessel. The purpose of the secondary outer vessel is to capture permeated or leaked tritium and to present a good thermal insulation when properly evacuated. A third volume, a helium filled loop, is installed in the primary volume to remove the decay heat and is used to perform tritium accountancy measurements

  9. 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

  10. Tritium behaviour in aquatic plants and animals in a freshwater marsh ecosystem

    International Nuclear Information System (INIS)

    Adams, L.W.; Peterle, T.J.; White, G.C.

    1979-01-01

    Ten curies of tritium as tritiated water (HTO) were experimentally added to an enclosed 2-ha Lake Erie marsh on 20 October 1973. Tritium kinetics in selected plants and animals were determined over a one-year period. Tritium levels in the marsh bottom sediment averaged 1.8 times the marsh water levels, with little evidence of tritium concentration above the marsh water tritium levels in the flora and fauna. The unbound tritium: marsh water tritium ratios in smartweed (Polygonum lapathifolium) and pickerelweed (Pontederia cordata) (both emergents) were lower than the same ratio for pondweed (Potamogeton crispus) (a submergent). There was some evidence of bound tritium buildup in midsummer, particularly in the pondweed. Tritium uptake into the unbound compartments of crayfish (Procambarus blandingi), carp (Cyprinus carpio) and bluegills (Lepomis macrochirus) was rapid. For crayfish, maximum HTO levels were observed on days 2 and 3 following treatment for muscle and viscera respectively. Unbound HTO in carp muscle peaked in 4 hours and the level in carp viscera reached a maximum in 2 days, in bluegill muscle and viscera on day 1. Unbound HTO in all species decreased following peak levels, paralleling marsh water HTO activity. Tritium uptake into the bound compartments was not as rapid nor were the levels as high as for unbound HTO in the fauna. The peak bound level in crayfish muscle was observed on day 10 (bound : unbound ratio of 0.34) and the maximum level in viscera was noted on day 20 (bound : unbound ratio of 0.23). Bound tritium in carp muscle and viscera reached maximum levels on day 20 (bound : unbound ratios of 0.25 and 0.39 respectively). In bluegills, peaks were reached on days 5 and 7 (bound : unbound ratios of 0.35 and 0.38 for muscle and viscera respectively). Bound tritium in all species decreased following maximum levels

  11. A methodology for determination of tritium inventory to the heavy water detritiation pilot plant from ICIT Rm. Valcea

    International Nuclear Information System (INIS)

    Bidica, N.; Stefanescu, I.; Bornea, A.; Zamfirache, M.; Lazar, A.; Vasut, F.; Pearsica, C.; Stefan, I.; Cristescu, I.; Prisecaru, I.; Sindilar, G.

    2007-01-01

    Full text: In this paper we present a methodology for determination of tritium inventory in a tritium removal facility. The method proposed is based on the developing of computing models for accountancy of the mobile tritium inventory in the separation processes of the stored tritium and of the trapped tritium inventory in the structure of the process system components. The configuration of the detritiation process is a combination of isotope catalytic exchange between water and hydrogen (LPCE) and the cryogenic distillation of hydrogen isotopes (CD). The computing model for tritium inventory in the LPCE process and the CD process will be developed based on mass transfer coefficients in catalytic isotope exchange reactions and in dual-phase system (liquid-vapour) of hydrogen isotopes distillation process. Accounting of tritium inventory stored in metallic hydride will be based on in-bed calorimetry. Estimation of the trapped tritium inventory can be made by subtraction of the mobile and stored tritium inventories from the global tritium inventory of the plant area. Determinations of the global tritium inventory of the plant area will be made on a regular basis by measuring any tritium amount entering or leaving the plant area. This methodology is intended to be applied to the Heavy Water Detritiation Pilot Plant from ICIT Rm. Valcea (Romania) and at the Cernavoda Tritium Removal Facility (which will be built in the next 5-7 years). (authors)

  12. Application of remote handling compatibility on ITER plant

    International Nuclear Information System (INIS)

    Sanders, S.; Rolfe, A.; Mills, S.F.; Tesini, A.

    2011-01-01

    The ITER plant will require fully remote maintenance during its operational life. For this to be effective, safe and efficient the plant will have to be developed in accordance with remote handling (RH) compatibility requirements. A system for ensuring RH compatibility on plant designed for Tokamaks was successfully developed and applied, inter alia, by the authors when working at the JET project. The experience gained in assuring RH compatibility of plant at JET is now being applied to RH relevant ITER plant. The methodologies required to ensure RH compatibility of plant include the standardization of common plant items, standardization of RH features, availability of common guidance on RH best practice and a protocol for design and interface review and approval. The protocol in use at ITER is covered by the ITER Remote Maintenance Management System (IRMMS) defines the processes and utilization of management controls including Plant Definition Forms (PDF), Task Definition Forms (TDFs) and RH Compatibility Assessment Forms (RHCA) and the ITER RH Code of Practice. This paper will describe specific examples where the authors have applied the methodology proven at JET to ensure remote handling compatibility on ITER plant. Examples studied are: ·ELM coils (to be installed in-vessel behind the Blanket Modules) - handling both in-vessel, in Casks and at the Hot Cell as well as fully remote installation and connection (mechanical and electrical) in-vessel. ·Neutral beam systems (in-vessel and in the NB Cell) - beam sources, cesium oven, beam line components (accessed in the NB Cell) and Duct Liner (remotely replaced from in-vessel). ·Divertor (in-vessel) - cooling pipe work and remotely operated electrical connector. The RH compatibility process can significantly affect plant design. This paper should therefore be of interest to all parties who develop ITER plant designs.

  13. A versatile model for tritium transfer from atmosphere to plant and soil

    International Nuclear Information System (INIS)

    Melintescu, A.; Galeriu, D.

    2004-01-01

    The need to increase the predictive power of risk assessment for large tritium releases implies a process level approach for model development. Tritium transfer for atmosphere to plant and the conversion in organically bound tritium depend strongly on plant characteristics, season, and meteorological conditions.In order to cope with this large variability and to avoid also, expensive calibration experiments, we developed a model using knowledge of plant physiology, agro-meteorology, soil sciences, hydrology, and climatology. The transfer of tritiated water to plant is modelled with resistance approach including sparse canopy. The canopy resistance is modelled using Jarvis-Calvet approach modified in order to directly use the canopy photosynthesis rate.The crop growth model WOFOST is used for photosynthesis rate both for canopy resistance and formation of organically bound tritium, also. Using this formalism, the tritium transfer parameters are directly linked to known processes and parameters from agricultural sciences. The model predictions for tritium in wheat are closed to a factor two to experimental data without any calibration. The model also is tested for rice and soya bean and can be applied for various plants and environmental conditions. For sparse canopy the model uses coupled equations between soil and plants. (author)

  14. Darlington tritium removal facility and station upgrading plant dynamic process simulation

    International Nuclear Information System (INIS)

    Busigin, A.; Williams, G. I. D.; Wong, T. C. W.; Kulczynski, D.; Reid, A.

    2008-01-01

    Ontario Power Generation Nuclear (OPGN) has a 4 x 880 MWe CANDU nuclear station at its Darlington Nuclear Div. located in Bowmanville. The station has been operating a Tritium Removal Facility (TRF) and a D 2 O station Upgrading Plant (SUP) since 1989. Both facilities were designed with a Distributed Control System (DCS) and programmable logic controllers (PLC) for process control. This control system was replaced with a DCS only, in 1998. A dynamic plant simulator was developed for the Darlington TRF (DTRF) and the SUP, as part of the computer control system replacement. The simulator was used to test the new software, required to eliminate the PLCs. The simulator is now used for operator training and testing of process control software changes prior to field installation. Dynamic simulation will be essential for the ITER isotope separation system, where the process is more dynamic than the relatively steady-state DTRF process. This paper describes the development and application of the DTRF and SUP dynamic simulator, its benefits, architecture, and the operational experience with the simulator. (authors)

  15. Energetic-economic analysis of inertial fusion plants with tritium commercial production

    International Nuclear Information System (INIS)

    Vezzani, M.; Cerullo, N.; Lanza, S.

    2000-01-01

    The realization of nuclear power plants based on fusion principles is expected to be, at the moment, very expensive. As a result the expected cost of electricity (COE) of fusion power plants is much higher than the COE of fission and fossil power plants. Thus it is necessary to study new solutions for fusion power plant designs to reduce the COE. An interesting solution for the first generation of fusion plants is to produce a surplus of tritium for commercial purposes. The present paper is concerned with the study of whether such a tritium surplus production can improve the plant economic balance, so that the COE is reduced, and to what extent. The result was that such a production allows a considerable reduction of COE and seems to be a good direction for development for the first generation of fusion power plants. To give an example, for a reference inertial confinement fusion (ICF) power plant the rise of the plant net tritium breeding ratio (TBR n ) from 1 to 1.2 would allow, in the conservative estimate of a tritium market price (C T ) of 5 M$/kg, a COE reduction of about 20%. In the estimate of a TBR n rise from 1 to 1.3 and of a C T value of 10 M$/kg, COE reduction could be more than 50%! In conclusion, the present paper points out the influence of TBR increase on COE reduction. Such a conclusion, which holds true for every fusion plant, is much more valid for ICF plants in which it is possible to reach higher TBR values and to use tritium extraction systems easily. Thus, considering the relevant economic advantages, a commercial tritium surplus production should not be disregarded for first generation fusion power plant designs, in particular for ICF plant designs

  16. A dynamic compartment mode for evaluating the contamination level of tritium in agricultural plants

    Energy Technology Data Exchange (ETDEWEB)

    Keum, Dong Kwon; Lee, Han Soo; Kang, Hee Seok; Jun, In; Choi, Yong Ho; Lee, Chang Woo

    2006-03-15

    This report describes a dynamic compartment model for evaluating the tritium level in agricultural plants after a short-term exposure to HTO vapor and its comparison with experimental results to test the predictive accuracy of the model. The model uses a time-dependent growth equation of a plant so that it can predict the contamination level of tritium depending on the stage of the growth of the plant, which is a major difference from some other compartment models using a constant crop yield. The model is able to calculate the time variable concentrations of the compartments representing the atmosphere, soil, and plants of four categories including leafy vegetables, root vegetables, grains, and tuber plants. Experimental results include the tissue free water tritium (TFWT) and the organically bound tritium (OBT) concentration of rice, soybean, cabbage, and radish exposed to HTO vapor for 1 h in the daytime at different growth stages. The model predictions showed that the model could simulate well not only the time-dependent tritium concentration of the plants but also the effect of the growth stage of the plant at the exposure time. Comparison of the model predictions with the experimental results suggested that the model could predict reasonably well the observed TFWT and OBT concentrations of the plants considered.

  17. ITER Safety Task NID-5A, Subtask 1-1: Source terms and energies - initial tritium source terms. Final report

    International Nuclear Information System (INIS)

    Fong, C.; Kalyanam, K.M.; Tanaka, M.R.; Sood, S.; Natalizio, A.; Delisle, M.

    1995-02-01

    The overall objective of the Early Safety and Environmental Characterization Study (ESECS) is to assess the environmental impact of tritium using appropriate assumptions on a hypothetical site for ITER, having the r eference s ite characteristics as proposed by the JCT. The objective of this work under the above subtask 1-1, NID-5a, is to determine environmental source terms (i.e., process source term x containment release fraction) for the fuel cycle and cooling systems. The work is based on inventories and process source terms (i.e., inventory x mobilization fraction), provided by others (under Task NID 3b). The results of this work form the basis for the determination, by others, of the off-site dose (i.e., environmental source term x dose/release ratio). For the determination of the environmental source terms, the TMAP4 code has been utilized (ref 1). This code is approved by ITER for safety assessment. Volume 3 is a compilation of appendices giving detailed results of the study

  18. ITER Safety Task NID-5A, Subtask 1-1: Source terms and energies - initial tritium source terms. Final report

    International Nuclear Information System (INIS)

    Fong, C.; Kalyanam, K.M.; Tanaka, M.R.; Sood, S.; Natalizio, A.; Delisle, M.

    1995-02-01

    The overall objective of the Early Safety and Environmental Characterization Study (ESECS) is to assess the environmental impact of tritium using appropriate assumptions on a hypothetical site for ITER, having the r eference s ite characteristics as proposed by the JCT. The objective of this work under the above subtask 1-1, NID-5a, is to determine environmental source terms (i.e., process source term x containment release fraction) for the fuel cycle and cooling systems. The work is based on inventories and process source terms (i.e., inventory x mobilization fraction), provided by others (under Task NID 3b). The results of this work form the basis for the determination, by others, of the off-site dose (i.e., environmental source term x dose/release ratio). For the determination of the environmental source terms, the TMAP4 code has been utilized (ref 1). This code is approved by ITER for safety assessment. Volume 2 is a compilation of appendices giving detailed results of the study. 5 figs

  19. Investigation of internal contamination by tritium in A-1 nuclear power plant personnel in 1974

    International Nuclear Information System (INIS)

    Ondris, D.; Herchl, M.; Homolova, E.

    1977-01-01

    The results are presented of the 1974 personnel monitoring of the Bohunice A-1 nuclear power plant staff for internal contamination with tritium. Totally, 650 urine samples taken from 103 workers were analyzed using the recommended ICRP procedure. In routine examinations, the highest dose equivalent value of tritium incorporated within two weeks did not exceed 10 mrem, i.e., the maximum annual dose equivalent did not exceed 260 mrem. 8.5 μCi tritium per 1 litre urine was considered to be an alarm value. In a selected group of 21 high-risk persons analyses were conducted before and after each operation associated with tritium hazards. The limit dose was set to 5.8 μCi.l -1 , i.e., the tritium concentration equivalent to 10% of the maximum permissible annual intake. In 18 workers where tritium risk was of a more serious nature the biological half-life was followed up, with the average biological half-life being 8.5 days, with 5 days for the minimum and 12 days for the maximum values. The results show that in 1974 the tritium burden did not exceed 1/10 of the maximum permissible dose for any of the A-1 nuclear power plant workers. (L.O.)

  20. Analysis of the tritium gas sample by gas chromatography. one of the basic steps in the fuel cycle of the ITER

    International Nuclear Information System (INIS)

    Szuecs, Z.

    2006-01-01

    Complete text of publication follows. The international consortium (EU, USA, Russia, Japan, South Korea, China and India) signed a formal agreement to build the International Thermonuclear Experimental Reactor (ITER) on November 21, 2006. The official objective of ITER is to 'demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes'. The deuterium-tritium fusion process releases roughly three times much energy as uranium 235 fission and millions of times more energy than a chemical reaction such as the burning of coal or hydrocarbons. It is the goal of a fusion power plant to harness this energy to produce electricity. The Atomki recognized this perspective trend at time and it was involved into the fusion research by the Hungarian Fusion Association from September 2005. This short communication is referring about the cooperation, which was realized between the Atomki and the Tritium Laboratory Karlsruhe (TLK) in field of the analysis of the tritium gas samples from spring of 2006. The tritium is one of the main important fuel materials for the fusion reactor. As a radioactive gas it needs special attention for its handling and analysis. In the TLK has been developed gas-chromatographic technology for checking of the tritium gas sampling, origin from the different parts of the Tritium Plant. However this method is already 14 years old and the old staff was changed 2 years ago, loosed some important knowhow at the same time. Therefore was necessary to check the system and to carry out some important maintenance. All of them are summarized in the following 4 points: 1. The continuously increasing of the retention time of the Porapack Q column created the question: Nor the resin phase of the column already damaged or the resin works properly and any other effect (leak on the pipe system or in the valves or in pressure regulators, changing of the inlet and outlet pressure and the changing of the ratio of the flow rates) occurs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  2. Development of ITER relevant laser techniques for deposited layer characterisation and tritium inventory

    NARCIS (Netherlands)

    Malaquias, A.; Philipps, V.; Huber, A.; Hakola, A.; Likonen, J.; Kolehmainen, J.; Tervakangas, S.; Aints, M.; Paris, P.; Laan, M.; Lissovski, A.; Almaviva, S.; Caneve, L.; Colao, F.; Maddaluno, G.; Kubkowska, M.; Gasior, P.; van der Meiden, H. J.; Lof, A. R.; van Emmichoven, P. A. Zeijlma; Petersson, P.; Rubel, M.; Fortuna, E.; Xiao, Q.

    2013-01-01

    Laser Induced Breakdown Spectroscopy (LIBS) is a potential candidate to monitor the layer composition and fuel retention during and after plasma shots on specific locations of the main chamber and divertor of ITER. This method is being investigated in a cooperative research programme on plasma

  3. 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)

  4. 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).

  5. Tritium control by water recycle in a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Hall, N.E.; Ward, G.N.

    1975-06-01

    A preliminary study was made of the use of water recycle within a reprocessing plant to control the escape of tritium and to consolidate it for disposal. Tritium distribution was evaluated in the leacher, high-level, and low-level systems for seven different flowsheet conditions. Tritium retention efficiency was also evaluated for these flowsheet conditions. Impact of tritiated water recycle on the plant design and operation is assessed. It is concluded that tritium control by water recycle is feasible. Achievement of satisfactory retention efficiencies and economic volumes of solidified tritium waste will require extension of existing technology and development of new technology. Evaluation of potential abnormal conditions indicate that releases from upsets need not be excessive. Some increase in occupational exposure will occur because of the pervasiveness, persistence, and ease of uptake of tritiated water vapor. Incentives for tritium control by water recycle may prove marginal if this increased exposure to plant personnel is significant compared to the small reduction in exposure to the general public. Recommendations are presented for further studies

  6. 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

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

    International Nuclear Information System (INIS)

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

    2009-02-01

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

  8. Comparison of the Tritium permeated from ITER Blanket in normal operation and its short range impact of HT over France, Swiss or Spain

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

    In this paper we assumes the hydrogen isotopes permeation from a liquid metal ITER breeder blanket (assuming normal operation and a LM as DCLL or HCLL blanket) as one of the possible sources of a leak and tritium release,mainly but not only. The paper presents a short range low impact of HT gas activity over France, Swiss or Spain from same cases in 2014 and 2015 releases from ITER. The permeation of hydrogen isotopes is an important experimental issue to take into account into the development of a Tritium Breeder Module for ITER [1]. Tritium cannot be confined -without an uncertainty of 5% in the flux permeation- and therefore HT can be detected (e.g. by ionization chamber) as permeates though the structure of RAFM steel towards the coolant [1]. HT from Pb15.7Li and permeated in Eurofer97 can contaminate the other parts of the system and may be delivered though the normal-vent detritiation system (NVDS). Real time forecast of transport of tritium in air from the fusion reactor towards off-site far downwind though extended tritium clouds into the low levels of the atmosphere is calculated for the short range (up to 24 hours) by the coupling of 2 models the European Centre for Medium Range Weather Forecast (ECMWF) [2] model and the FLEXPART lagrangian dispersion model [3] verified with NORMTRI simulation [4] and implemented in many different cases and scenarios [5, 6, 7]. As a function of daily weather conditions the release will affect just France or already can be delivered towards Swiss when cyclonic circulation, or towards the Iberian Peninsula or Balearic Islands (Spain) when high produce anticyclonic circulation of the air over the Mediterranean Sea. (Author)

  9. Integration of the ITER diagnostic plant systems with CODAC

    International Nuclear Information System (INIS)

    Simrock, S.; Barnsley, R.; Bertalot, L.; Hansalia, C.; Klotz, W.D.; Makijarvi, P.; Reichle, R.; Vayakis, G.; Yonekawa, I.; Walker, C.; Wallander, A.; Walsh, M.; Winter, A.

    2011-01-01

    ITER requires extensive diagnostic systems in order to meet the requirements for machine operation, protection, plasma control and physics studies. The realization of these systems is a major challenge not only because of the harsh environment and the nuclear requirements but also with respect to Instrumentation and Control (I and C) of all the 59 diagnostics plants. The Plant Systems I and C are mostly 'in-kind', i.e. procured by the seven ITER Domestic Agencies (DAs), while the Central I and C Systems are 'in-fund', i.e. procured by ITER Organization (IO). Standardization of Plant Systems I and C is of primary importance and has been one of the highest priority tasks of CODAC. The standards are published in the Plant Control Design Handbook (PCDH) which will be followed to ensure a homogeneous design, guarantee high availability and simplify maintenance and support future upgrades. Most important for a successful commissioning and operation of the ITER facility are the concepts of interfacing the diagnostics plant systems with CODAC and the standards for instrumentation and control which must be followed all contributing parties. In this paper, we will elaborate on the concepts of interfacing the diagnostics plant systems with CODAC and the standards that must be followed for the design.

  10. Accelerator production of tritium plant design and supporting engineering development and demonstration work

    International Nuclear Information System (INIS)

    Lisowski, P.W.

    1997-11-01

    Tritium is an isotope of hydrogen with a half life of 12.3 years. Because it is essential for US thermonuclear weapons to function, tritium must be periodically replenished. Since K reactor at Savannah River Site stopped operating in 1988, tritium has been recycled from dismantled nuclear weapons. This process is possible only as long as many weapons are being retired. Maintaining the stockpile at the level called for in the present Strategic Arms Reduction Treaty (START-I) will require the Department of Energy to have an operational tritium production capability in the 2005--2007 time frame. To make the required amount of tritium using an accelerator based system (APT), neutrons will be produced through high energy proton reactions with tungsten and lead. Those neutrons will be moderated and captured in 3 He to make tritium. The APT plant design will use a 1,700 MeV linear accelerator operated at 100 mA. In preparation for engineering design, starting in October 1997 and subsequent construction, a program of engineering development and demonstration is underway. That work includes assembly and testing of the first 20 MeV of the low energy plant linac at 100 mA, high-energy linac accelerating structure prototyping, radiofrequency power system improvements, neutronic efficiency measurements, and materials qualifications

  11. Hypothetical operation model for the multi-bed system of the Tritium plant based on the scheduling approach

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae-Uk, E-mail: eslee@dongguk.edu [Department of Chemical Engineering, Pohang University of Science and Technology, San 31, Hyoja-Dong, Pohang 790-784 (Korea, Republic of); Chang, Min Ho; Yun, Sei-Hun [National Fusion Research Institute, 169-148-gil Kwahak-ro, Yusong-gu, Daejon 34133 (Korea, Republic of); Lee, Euy Soo [Department of Chemical & Biochemical Engineering, Dongguk University, Seoul 100-715 (Korea, Republic of); Lee, In-Beum [Department of Chemical Engineering and Graduate School of Engineering Mastership, Pohang University of Science and Technology, San 31, Hyoja-Dong, Pohang 790-784 (Korea, Republic of); Lee, Kun-Hong [Department of Chemical Engineering, Pohang University of Science and Technology, San 31, Hyoja-Dong, Pohang 790-784 (Korea, Republic of)

    2016-11-01

    Highlights: • We introduce a mathematical model for the multi-bed storage system in the tritium plant. • We obtain details of operation by solving the model. • The model assesses diverse operation scenarios with respect to risk. - Abstract: In this paper, we describe our hypothetical operation model (HOM) for the multi-bed system of the storage and delivery system (SDS) of the ITER tritium plant. The multi-bed system consists of multiple getter beds (i.e., for batch operation) and buffer vessels (i.e., for continuous operation). Our newly developed HOM is formulated as a mixed-integer linear programming (MILP) model and has been extensively investigated to optimize chemical and petrochemical production planning and scheduling. Our model determines the timing, duration, and size of tasks corresponding to each set of equipment. Further, inventory levels for each set of equipment are calculated. Our proposed model considers the operation of one cycle of one set of getter beds and is implemented and assessed as a case study problem.

  12. Design and calculation of tritium extraction from liquid LiPb by bubble columns for ITER

    International Nuclear Information System (INIS)

    Xie, Bo

    2009-04-01

    A mathematical model describing the complex fluid-dynamics of a bubble extractor from liquid LiPb loop for ITER is presented. A parametric analysis of the extraction efficiency of a bubble column as a function of the process parameters is carried out and the design of a bubble extractor system is proposed. On this base, a mathematical model is built by taking into consideration the kinetics of deuterium desorption from liquid LiPb alloy. The calculation data of deuterium release-behavior from liquid LiPb under different operating conditions of temperature and deuterium partial pressures and helium gas flow-rates in the liquid LiPb alloy are obtained. These results have shown that the overall re- lease process is governed by the diffusion of deuterium atoms in the LiPb and by the heterogeneous reaction at the gas-eutectic interface of the deuterium atoms recombination under the probable working temperature range. (authors)

  13. Study of Oxidizing Agents for Tritium Removal in ITER -Compatible Conditions: Alternatives to Oxygen and Ozone

    International Nuclear Information System (INIS)

    Tabares, F. L.; Tafalla, D.; Ferreira, J. A.; Gomez-Aleixandre, C.; Maria Albella, J.; Soria, J.; Rodriguez-Ramos, I.

    2007-01-01

    In the present report, the studies of tritiated carbon-film removal by oxidizing agents other than Oxygen and Ozone in ITER are described. Exposure of laboratory produced a-C:H/D films and tokamak flakes (Asdex Upgrade and Textor) to nitric oxide, water and hydrogen peroxide has been carried out. Temperatures of exposure up to 350 degree centigree were used, and thermal desorption of the samples at temperatures up to 750 degree centigree was performed for sample characterization prior to and after the treatment. Elastic Recoil Detection Analysis (ERDA), Infrared Spectroscopy, XPS and Nano indentation hardness analysis were applied to the characterization of the physical and chemical changes of the samples. This work was done under the EFDA Task 04-1175. (Author) 8 refs

  14. Study of Oxidizing Agents for Tritium Removal in ITER -Compatible Conditions: Alternatives to Oxygen and Ozone

    Energy Technology Data Exchange (ETDEWEB)

    Tabares, F. L.; Tafalla, D.; Ferreira, J. A.; Gomez-Aleixandre, C.; Maria Albella, J.; Soria, J.; Rodriguez-Ramos, I.

    2007-07-20

    In the present report, the studies of tritiated carbon-film removal by oxidizing agents other than Oxygen and Ozone in ITER are described. Exposure of laboratory produced a-C:H/D films and tokamak flakes (Asdex Upgrade and Textor) to nitric oxide, water and hydrogen peroxide has been carried out. Temperatures of exposure up to 350 degree centigree were used, and thermal desorption of the samples at temperatures up to 750 degree centigree was performed for sample characterization prior to and after the treatment. Elastic Recoil Detection Analysis (ERDA), Infrared Spectroscopy, XPS and Nano indentation hardness analysis were applied to the characterization of the physical and chemical changes of the samples. This work was done under the EFDA Task 04-1175. (Author) 8 refs.

  15. Experimental investigation and modelling of tritium washout by precipitation in the area of the nuclear power plant of Paks, Hungary

    International Nuclear Information System (INIS)

    Koello, Z.; Palcsu, L.; Major, Z.; Papp, L.; Molnar, M.; Ranga, T.; Dombovari, P.; Manga, L.

    2011-01-01

    Tritium occurs in nature in trace amounts, but its concentration is changing due to natural and artificial sources. Studies focusing on natural tritium have to take into account the effect of artificial sources. Also, the impact of tritium is an important issue in environmental protection, e.g. in connection with the emissions from nuclear power plants. The present work focuses on the rain washout of tritium emitted from the Paks nuclear power plant in Hungary. Rainwater collectors were placed around the plant and after a period of precipitation, rainwater was collected and analysed for tritium content. Samples were analysed using low-level liquid scintillation counting, with some also subject to the more accurate 3 He ingrowth method. The results clearly show the trace of the tritium plume emitted from the plant; however, values are only about one order of magnitude higher than environmental background levels. A washout model was devised to estimate the distribution of tritium around the plant. The model gives slightly higher concentrations than those measured in the field, but in general the agreement is satisfactory. The modelled values demonstrate that the effect of the plant on rainwater tritium levels is negligible over a distance of some kilometres. - Research highlights: →The rainwater around a nuclear power plant was collected with a special rainwater collector →The rainwater after a rain event was analysed for tritium with LSC and with the helium ingrowth method. →The trace of the tritium plume is clearly detectable in the rainwater. →The agreement between a reversible washout model and experimental data is satisfactory. →According to the model the tritium plume is hardly detectable over some kilometers from the plant

  16. Experimental investigation and modelling of tritium washout by precipitation in the area of the nuclear power plant of Paks, Hungary

    Energy Technology Data Exchange (ETDEWEB)

    Koello, Z., E-mail: kolloz42@gmail.co [Hertelendi Laboratory of Environmental Studies, Institute of Nuclear Research of the Hungarian Academy of Sciences, Bem ter 18/c, Debrecen 4026 (Hungary); Palcsu, L.; Major, Z.; Papp, L.; Molnar, M. [Hertelendi Laboratory of Environmental Studies, Institute of Nuclear Research of the Hungarian Academy of Sciences, Bem ter 18/c, Debrecen 4026 (Hungary); Ranga, T.; Dombovari, P.; Manga, L. [Department of Radiation Protection, Nuclear Power Plant of Paks, Paks (Hungary)

    2011-01-15

    Tritium occurs in nature in trace amounts, but its concentration is changing due to natural and artificial sources. Studies focusing on natural tritium have to take into account the effect of artificial sources. Also, the impact of tritium is an important issue in environmental protection, e.g. in connection with the emissions from nuclear power plants. The present work focuses on the rain washout of tritium emitted from the Paks nuclear power plant in Hungary. Rainwater collectors were placed around the plant and after a period of precipitation, rainwater was collected and analysed for tritium content. Samples were analysed using low-level liquid scintillation counting, with some also subject to the more accurate {sup 3}He ingrowth method. The results clearly show the trace of the tritium plume emitted from the plant; however, values are only about one order of magnitude higher than environmental background levels. A washout model was devised to estimate the distribution of tritium around the plant. The model gives slightly higher concentrations than those measured in the field, but in general the agreement is satisfactory. The modelled values demonstrate that the effect of the plant on rainwater tritium levels is negligible over a distance of some kilometres. - Research highlights: {yields}The rainwater around a nuclear power plant was collected with a special rainwater collector {yields}The rainwater after a rain event was analysed for tritium with LSC and with the helium ingrowth method. {yields}The trace of the tritium plume is clearly detectable in the rainwater. {yields}The agreement between a reversible washout model and experimental data is satisfactory. {yields}According to the model the tritium plume is hardly detectable over some kilometers from the plant

  17. Experimental investigation and modelling of tritium washout by precipitation in the area of the nuclear power plant of Paks, Hungary.

    Science.gov (United States)

    Köllo, Z; Palcsu, L; Major, Z; Papp, L; Molnár, M; Ranga, T; Dombóvári, P; Manga, L

    2011-01-01

    Tritium occurs in nature in trace amounts, but its concentration is changing due to natural and artificial sources. Studies focusing on natural tritium have to take into account the effect of artificial sources. Also, the impact of tritium is an important issue in environmental protection, e.g. in connection with the emissions from nuclear power plants. The present work focuses on the rain washout of tritium emitted from the Paks nuclear power plant in Hungary. Rainwater collectors were placed around the plant and after a period of precipitation, rainwater was collected and analysed for tritium content. Samples were analysed using low-level liquid scintillation counting, with some also subject to the more accurate (3)He ingrowth method. The results clearly show the trace of the tritium plume emitted from the plant; however, values are only about one order of magnitude higher than environmental background levels. A washout model was devised to estimate the distribution of tritium around the plant. The model gives slightly higher concentrations than those measured in the field, but in general the agreement is satisfactory. The modelled values demonstrate that the effect of the plant on rainwater tritium levels is negligible over a distance of some kilometres. Copyright © 2010 Elsevier Ltd. All rights reserved.

  18. 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.

  19. Tritium breeding potential of the Princeton reference fusion power plant

    International Nuclear Information System (INIS)

    Greenspan, E.; Price, W.G. Jr.

    1974-04-01

    A variational method is used to investigate the tritium breeding potential of the blanket of a fusion reactor. Effectiveness functions indicating the changes in the breeding ratio (BR) due to material density perturbations are calculated with the code SWAN. Results are presented analyzing the sensitivity of the BR both to cross section variations and to material density perturbations. For example, SWAN indicates a 0.176 increase in BR for the replacement of 10% of the flibe by beryllium. Implications of the sensitivity figures for design modification and optimization are discussed. 15 refs., 7 figs

  20. Investigation of tritium transfer to plants via the OBT/HTO and OBT/TFWT ratios; Etude du transfert du tritium aux vegetaux via les ratios OBT/HTO ET OBT/TFWT

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, C.; Guetat, P.; Vichot, L.; Losset, Y. [CEA Valduc, UMR, 21 - Is-sur-Tille (France); Boyer, C.; Fromm, M.; Mavon, C. [UMR CEA E4, Universite de Franche-Comte, 16 route de Gray, F-25030 Besancon cedex (France); Boyer, C.; Tatin-Froux, F.; Badot, P.M. [CNRS-Universite de Franche Comte / UMR 6249 Chrono-environnement usc INRA, Place Leclerc, F-25030 Besancon cedex (France)

    2009-07-01

    After having recalled some concepts used to distinguish the various forms of water present in plants, and the reactions in presence of tritium, the authors discuss the biochemical behaviour of tritium. Then, they briefly report a literature survey and, for different plants and crops, indicate the values of the OBT/HTO (organically bound tritium/tritiated water) and OBT/TFWT (organically bound tritium/tissue free water tritium) ratios. They also report experimental studies performed by exposures of lettuces at the vicinity of a nuclear installation

  1. Tritium in Japanese precipitation following the March 2011 Fukushima Daiichi Nuclear Plant accident.

    Science.gov (United States)

    Matsumoto, Takuya; Maruoka, Teruyuki; Shimoda, Gen; Obata, Hajime; Kagi, Hiroyuki; Suzuki, Katsuhiko; Yamamoto, Koshi; Mitsuguchi, Takehiro; Hagino, Kyoko; Tomioka, Naotaka; Sambandam, Chinmaya; Brummer, Daniela; Klaus, Philipp Martin; Aggarwal, Pradeep

    2013-02-15

    Tritium concentrations in Japanese precipitation samples collected after the March 2011 accident at the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) were measured. Values exceeding the pre-accident background were detected at three out of seven localities (Tsukuba, Kashiwa and Hongo) southwest of the FNPP1 at distances varying between 170 and 220 km from the source. The highest tritium content was found in the first rainfall in Tsukuba after the accident; however concentrations were 500 times less than the regulatory limit for tritium in drinking water. Tritium concentrations decreased steadily and rapidly with time, becoming indistinguishable from the pre-accident values within five weeks. The atmospheric tritium activities in the vicinity of the FNPP1 during the earliest stage of the accident was estimated to be 1.5×10(3) Bq/m(3), which is potentially capable of producing rainwater exceeding the regulatory limit, but only in the immediate vicinity of the source. Copyright © 2012 Elsevier B.V. All rights reserved.

  2. Overview of the ITER Tokamak complex building and integration of plant systems toward construction

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2015-10-15

    The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed

  3. Engineering Design of ITER Prototype Fast Plant System Controller

    Science.gov (United States)

    Goncalves, B.; Sousa, J.; Carvalho, B.; Rodrigues, A. P.; Correia, M.; Batista, A.; Vega, J.; Ruiz, M.; Lopez, J. M.; Rojo, R. Castro; Wallander, A.; Utzel, N.; Neto, A.; Alves, D.; Valcarcel, D.

    2011-08-01

    The ITER control, data access and communication (CODAC) design team identified the need for two types of plant systems. A slow control plant system is based on industrial automation technology with maximum sampling rates below 100 Hz, and a fast control plant system is based on embedded technology with higher sampling rates and more stringent real-time requirements than that required for slow controllers. The latter is applicable to diagnostics and plant systems in closed-control loops whose cycle times are below 1 ms. Fast controllers will be dedicated industrial controllers with the ability to supervise other fast and/or slow controllers, interface to actuators and sensors and, if necessary, high performance networks. Two prototypes of a fast plant system controller specialized for data acquisition and constrained by ITER technological choices are being built using two different form factors. This prototyping activity contributes to the Plant Control Design Handbook effort of standardization, specifically regarding fast controller characteristics. Envisaging a general purpose fast controller design, diagnostic use cases with specific requirements were analyzed and will be presented along with the interface with CODAC and sensors. The requirements and constraints that real-time plasma control imposes on the design were also taken into consideration. Functional specifications and technology neutral architecture, together with its implications on the engineering design, were considered. The detailed engineering design compliant with ITER standards was performed and will be discussed in detail. Emphasis will be given to the integration of the controller in the standard CODAC environment. Requirements for the EPICS IOC providing the interface to the outside world, the prototype decisions on form factor, real-time operating system, and high-performance networks will also be discussed, as well as the requirements for data streaming to CODAC for visualization and

  4. Development of the ELEX process for tritium separation at reprocessing plants

    International Nuclear Information System (INIS)

    Bruggeman, A.; Meynendonckx, L.; Parmentier, C.; Goossens, W.R.A.; Baetsle, L.H.

    1985-01-01

    The ELEX process for isotopic enrichment and separation of tritium from aqueous reprocessing effluents is described. After the development of an appropriate hydrophobic catalyst and the study of the separate constituent steps of the ELEX process, an integrated bench-scale installation with a detritiation capacity of 10 mol water per hour was constructed. It comprises essentially a 1.5 kW electrolyser and two 2 cm diameter by 3 m high exchange columns. In this mini-pilot the ELEX process was successfully demonstrated by detritiating more than 1000 dm 3 water containing up to 100 mCi tritium per dm 3 , which is the feed concentration expected for application of the process in a reprocessing plant. The process decontamination factor was always larger than 100 and the overall tritium balance could be kept within the experimental errors of the various measurements. Depending on the duration of the runs, the volume reduction factor was between 10 and 15, but this factor will become much higher when the present electrolyser will be replaced by the low-volume one now under construction at S.C.K./C.E.N. The technical availability of the mini-pilot amounted to 99% or more for the last experiments. During the nearly 6000 hours of operation there were no tritium contamination problems. At present the construction of a 280 mol.h -1 or 0.12 m 3 H 2 O (HTO) per day pilot detritiation installation nears completion. In this unit, which will have a total tritium inventory of maximum 1000 Ci, the ELEX process will be demonstrated with a volume reduction factor of 100 and a process decontamination factor of 100. Again the feed concentration will be 100 Ci tritium per m 3 . In the future this installation will be supplemented with a pre-treatment unit for the removal of fission products, organics and nitric acid

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

    Directory of Open Access Journals (Sweden)

    E. Pajuste

    2017-08-01

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

  6. Organic tritium in freshwater ecosystems: long-term trends in the environment of French nuclear power plants

    International Nuclear Information System (INIS)

    Gontier, G.; Siclet, F.

    2011-01-01

    From 1977 to 2009, more than 600 measurements of organic tritium were performed on fish, aquatic plants and sediments upstream and downstream of the 15 French NPP located along rivers. Examination of the results shows that organic tritium activities have exponentially decreased over the last thirty years, in all components of aquatic ecosystems. Upstream of all NPP, OBT levels in sediments are higher than in plants and fish, themselves larger than HTO in surface water. The magnitude of these differences and the long-term trends depend on the river basin and can be explained by the varying nature of tritium sources. In river catchment, where atmospheric test fallout is the main source of tritium, the observed levels result from the exposure of aquatic organisms to two distinct tritium pools of different ages: atmospheric tritiated water (representing present fallout), and organic tritium from soils (formed over several decades) which supplies particulate matter to surface waters. In the Rhone and Rhine river basins, an additional source of organic tritium of very low bio-availability, probably originating from the luminescent paint industry, is responsible for the spiking of sediment organic matter up to 100 to 100 000 Bq.L -1 combustion water. The comparison of upstream and downstream NPP tritium levels shows that the influence of tritium discharges is detectable only in rivers, with low background OBT activities, i.e in basins other than the Rhone and Rhine. The observed increase in plant and fish OBT is lower than the added HTO activity in water due to discharge, which supports the absence of bioaccumulation for tritium originating from HTO and the absence of highly bio-available tritiated organic molecules in NPP discharges. (authors)

  7. ITER Fast Plant System Controller prototype based on PXIe platform

    International Nuclear Information System (INIS)

    Ruiz, M.; Vega, J.; Castro, R.; Sanz, D.; López, J.M.; Arcas, G. de; Barrera, E.; Nieto, J.; Gonçalves, B.; Sousa, J.; Carvalho, B.; Utzel, N.; Makijarvi, P.

    2012-01-01

    Highlights: ► Implementation of Fast Plant System Controller (FPSC) for ITER CODAC. ► Efficient data acquisition and data movement using EPICS. ► Performance of PCIe technologies in the implementation of FPSC. - Abstract: The ITER Fast Plant System Controller (FPSC) is based on embedded technologies. The FPSC will be devoted to both data acquisition tasks (sampling rates higher than 1 kHz) and control purposes (feedback loop actuators). Some of the essential requirements of these systems are: (a) data acquisition and data preprocessing; (b) interfacing with different networks and high speed links (Plant Operation Network, timing network based on IEEE1588, synchronous data transference and streaming/archiving networks); and (c) system setup and operation using EPICS (Experimental Physics and Industrial Control System) process variables. CIEMAT and UPM have implemented a prototype of FPSC using a PXIe (PCI eXtension for Instrumentation) form factor in a R and D project developed in two phases. The paper presents the main features of the two prototypes developed that have been named alpha and beta. The former was implemented using LabVIEW development tools as it was focused on modeling the FPSC software modules, using the graphical features of LabVIEW applications, and measuring the basic performance in the system. The alpha version prototype implements data acquisition with time-stamping, EPICS monitoring using waveform process variables (PVs), and archiving. The beta version prototype is a complete IOC implemented using EPICS with different software functional blocks. These functional blocks are integrated and managed using an ASYN driver solution and provide the basic functionalities required by ITER FPSC such as data acquisition, data archiving, data pre-processing (using both CPU and GPU) and streaming.

  8. Control system for technological processes in tritium processing plants with process analysis

    International Nuclear Information System (INIS)

    Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu; Bucur, Ciprian

    2005-01-01

    Integration of a large variety of installations and equipment into a unitary system for controlling the technological process in tritium processing nuclear facilities appears to be a rather complex approach particularly when experimental or new technologies are developed. Ensuring a high degree of versatility allowing easy modifications in configurations and process parameters is a major requirement imposed on experimental installations. The large amount of data which must be processed, stored and easily accessed for subsequent analyses imposes development of a large information network based on a highly integrated system containing the acquisition, control and technological process analysis data as well as data base system. On such a basis integrated systems of computation and control able to conduct the technological process could be developed as well protection systems for cases of failures or break down. The integrated system responds to the control and security requirements in case of emergency and of the technological processes specific to the industry that processes radioactive or toxic substances with severe consequences in case of technological failure as in the case of tritium processing nuclear plant. In order to lower the risk technological failure of these processes an integrated software, data base and process analysis system are developed, which, based on identification algorithm of the important parameters for protection and security systems, will display the process evolution trend. The system was checked on a existing plant that includes a removal tritium unit, finally used in a nuclear power plant, by simulating the failure events as well as the process. The system will also include a complete data base monitoring all the parameters and a process analysis software for the main modules of the tritium processing plant, namely, isotope separation, catalytic purification and cryogenic distillation

  9. Tritium activities in Canada

    International Nuclear Information System (INIS)

    Gierszewski, P.

    1995-01-01

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

  10. Tritium in Japanese precipitation following the March 2011 Fukushima Daiichi Nuclear Plant accident

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Takuya, E-mail: t.matsumoto@iaea.org [Isotope Hydrology Section, Division of Physical and Chemical Sciences, International Atomic Energy Agency, Vienna International Centre, 1400 Vienna (Austria); Maruoka, Teruyuki [Division of Integrative Environmental Sciences Graduate School of Life and Environmental Sciences, University of Tsukuba, 1-1-1 Tennodai, Tsukuba City, Ibaraki 305-8572 (Japan); Shimoda, Gen [Geological Survey of Japan, National Institute of Advanced Industrial Science and Technology, 1-1-1 Higashi, Tsukuba City, Ibaraki 305-8561 (Japan); Obata, Hajime [Atmosphere and Ocean Research Institute, The University of Tokyo, 5-1-5, Kashiwanoha, Kashiwa-shi, Chiba 277-8564 (Japan); Kagi, Hiroyuki [Geochemical Research Center, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-0033 (Japan); Suzuki, Katsuhiko [Japan Agency for Marin-Earth Science and Technology, 2-15, Natsushima, Yokosuka, Kanagawa 237-0061 (Japan); Yamamoto, Koshi [Graduate School of Environmental Studies, Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8601 (Japan); Mitsuguchi, Takehiro [215 Ooma Akadoji-cho Konan, 483-8226 (Japan); Usa Marine Biological Institute, Kochi University, 194 Inoshiri, Usa, Tosa, Kochi 781-1164 (Japan); Hagino, Kyoko; Tomioka, Naotaka [Institute for Study of the Earth' s Interior, Okayama University at Misasa, 827 Yamada, Misasa, Tottori 682-0193 (Japan); Sambandam, Chinmaya; Brummer, Daniela; Klaus, Philipp Martin; Aggarwal, Pradeep [Isotope Hydrology Section, Division of Physical and Chemical Sciences, International Atomic Energy Agency, Vienna International Centre, 1400 Vienna (Austria)

    2013-02-15

    Tritium concentrations in Japanese precipitation samples collected after the March 2011 accident at the Fukushima Dai-ichi Nuclear Power Plant (FNPP1) were measured. Values exceeding the pre-accident background were detected at three out of seven localities (Tsukuba, Kashiwa and Hongo) southwest of the FNPP1 at distances varying between 170 and 220 km from the source. The highest tritium content was found in the first rainfall in Tsukuba after the accident; however concentrations were 500 times less than the regulatory limit for tritium in drinking water. Tritium concentrations decreased steadily and rapidly with time, becoming indistinguishable from the pre-accident values within five weeks. The atmospheric tritium activities in the vicinity of the FNPP1 during the earliest stage of the accident was estimated to be 1.5 × 10{sup 3} Bq/m{sup 3}, which is potentially capable of producing rainwater exceeding the regulatory limit, but only in the immediate vicinity of the source. - Highlights: ► We measured the {sup 3}H contents of Japanese rain collected after the Fukushima accident. ► {sup 3}H level became 30 times higher than pre-accident level in the first rain at Tsukuba. ► Some locality within 220 km from the source showed elevated {sup 3}H levels. ► These high {sup 3}H signals disappear in a few weeks. ► Atmospheric {sup 3}H level at the source during the earliest stage was estimated to be 1500 Bq/m{sup 3}.

  11. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  12. 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

  13. 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

  14. Possibilities of tritium removal from waste waters of pressurized water reactors and fuel reprocessing plants

    International Nuclear Information System (INIS)

    Ribnikar, S.V.; Pupezin, J.D.

    1975-01-01

    Starting from parameters known for heavy water production processes, a parallel was made with separation of tritium from water. The quantity in common is the total cascade flow. The most efficient processes appear to be hydrogen sulfide, water exchange, hydrogen- and water distillation. Prospects of application of new processes are discussed briefly. Problems concerning detritiation of pressurized water reactors and large fuel reprocessing plants are analyzed. Detritiation of the former should not present problems. With the latter, economical detritiation can be achieved only after some plant flow patterns are changed. (U.S.)

  15. 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)

  16. Simulation and optimisation of the data acquisition system for tritium removal pilot plant

    International Nuclear Information System (INIS)

    Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu

    2004-01-01

    Optimization and simulation of systems especially in science and engineering can help to reduce risk and cost of design and testing processes. A huge number of codes has been developed to support modeling and simulation efforts. All of these software tools support the use of one or more mathematical model classes. Despite all of these efforts, it is hard to find simulation software, which is capable of combining several model classes in a real industry standard environment. The paper presents a simulation software product for controlling and data acquisition system of cryogenic installation process in the tritium removal pilot plant, using an industry standard programming environment widely applied to data acquisition, process control and data visualization, namely LabView. One of the problems in a tritium separation installation is controlling the temperature. To solve this problem it is necessary to develop a simulation system which includes the mathematical model for cryogenic distillation. Also with this simulation system we can approach the safety system which ensures the monitoring of radiations and toxic gases from installation. All elements used in controlling, modeling and simulation of the process, as well as, in the datalogging and supervisory control module from tritium removal installation are new. (authors)

  17. Atmospheric tritium concentrations under influence of AREVA NC La Hague reprocessing plant (France) and background levels.

    Science.gov (United States)

    Connan, O; Hébert, D; Solier, L; Maro, D; Pellerin, G; Voiseux, C; Lamotte, M; Laguionie, P

    2017-10-01

    In-air tritium measurements were conducted around the AREVA NC La Hague reprocessing plant, as well as on other sites that are not impacted by the nuclear industry in northwest of France. The results indicate that the dominant tritium form around the AREVA site is HT (86%). HT and HTO levels are lower than 5 and 1 Bq. m -3 for hourly samples taken in the plume. No tritiated organic molecules (TOM) were detected. 26 measurement campaigns were performed and links were established between near-field 85 Kr, HT and HTO activities. Environmental measurements are in line with those taken at the discharge stack, and tend to demonstrate that there are no rapid changes in the tritium forms released. Out of the influence of any nuclear activities, the levels measured were below 13 mBq.m -3 for HT and 5 mBq.m -3 for HTO (<0.5 Bq. L -1 ). HTO level in air seems to be influenced by HTO activities in surrounding seawater. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Method for calculating the steady-state distribution of tritium in a molten-salt breeder reactor plant

    International Nuclear Information System (INIS)

    Briggs, R.B.; Nestor, C.W.

    1975-04-01

    Tritium is produced in molten salt reactors primarily by fissioning of uranium and absorption of neutrons by the constituents of the fuel carrier salt. At the operating temperature of a large power reactor, tritium is expected to diffuse from the primary system through pipe and vessel walls to the surroundings and through heat exchanger tubes into the secondary system which contains a coolant salt. Some tritium will pass from the secondary system into the steam power system. This report describes a method for calculating the steady state distribution of tritium in a molten salt reactor plant and a computer program for making the calculations. The method takes into account the effects of various processes for removing tritium, the addition of hydrogen or hydrogenous compounds to the primary and secondary systems, and the chemistry of uranium in the fuel salt. Sample calculations indicate that 30 percent or more of the tritium might reach the steam system in a large power reactor unless special measures are taken to confine the tritium. (U.S.)

  19. Tritium analysis of divertor tiles used in JET ITER-like wall campaigns by means of β-ray induced x-ray spectrometry

    Science.gov (United States)

    Hatano, Y.; Yumizuru, K.; Koivuranta, S.; Likonen, J.; Hara, M.; Matsuyama, M.; Masuzaki, S.; Tokitani, M.; Asakura, N.; Isobe, K.; Hayashi, T.; Baron-Wiechec, A.; Widdowson, A.; contributors, JET

    2017-12-01

    Energy spectra of β-ray induced x-rays from divertor tiles used in ITER-like wall campaigns of the Joint European Torus were measured to examine tritium (T) penetration into tungsten (W) layers. The penetration depth of T evaluated from the intensity ratio of W(Lα) x-rays to W(Mα) x-rays showed clear correlation with poloidal position; the penetration depth at the upper divertor region reached several micrometers, while that at the lower divertor region was less than 500 nm. The deep penetration at the upper part was ascribed to the implantation of high energy T produced by DD fusion reactions. The poloidal distribution of total x-ray intensity indicated higher T retention in the inboard side than the outboard side of the divertor region.

  20. Development of ITER PRM and standard parts catalogues in CATIA V5 for tritium-containing systems and components

    International Nuclear Information System (INIS)

    Lazar, Alin; Brad, Sebastian; Zamfirache, Marius; Soare, Sorin; Sofalca, Nicolae; Vijulie, Mihai

    2006-01-01

    CATIA V5 is a software chosen to perform the design and integration within ITER of both systems: fluid and mechanical systems. The broad range of applications provides the ability to develop the design process from the functional 2D design (P and ID) to the 3D plant layouts and detailed design. The 2D symbols for the equipment and piping components were developed in accordance with EN ISO 10628 standard, ISO 3511 (part I, II and IV) standard and DIN 28401 standard and considering the classification and settings done by FZK-team in the PRM-FZK-TLK in order to ensure compliance with the established design standards. These symbols were inserted in 2D catalogues linked with the main catalogue installed in PRM-FZK-TLK on CATIA SERVER at TLK-FZK and intensively tested in Piping and Instrumentation Diagrams. The 3D part for the equipment, instruments and piping components was developed according to the specifications, industrial conventions, terminology and literature available in library, internet and practice. All parts were parametrically built in order to allow the designer to modify the part geometry according to the design. The 3D parts were typed in accordance with IKARUS-Project Instructions by FZK team, inserted in 3D test catalogues and tested in Equipment and Systems workbench. The elements of P and ID, pipes, piping parts, instruments and equipment carry attribute information such as fluid type, pressure rating, insulation or material. The list of attributes is defined in the PRM and consists of the standard industrial attributes implemented by default in CATIA V5 and project specific attributes defined by Project Administration. 2D logical design and 3D Piping models are logically related through the common functions definition in the PRM. The task, TW5-TTFD-TPI-51, was performed collaboratively by MEdC/ICIT and FZK/TLK. (authors)

  1. 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.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. RAMI modeling of selected balance of plant systems for the proposed Accelerator Production of Tritium (APT) project

    International Nuclear Information System (INIS)

    Radder, J.A.; Cramer, D.S.

    1997-01-01

    In order to meet Department of Energy (DOE) Defense Program requirements for tritium in the 2005-2007 time frame, new production capability must be made available. The Accelerator Production of Tritium (APT) Plant is being considered as an alternative to nuclear reactor production of tritium, which has been the preferred method in the past. The proposed APT plant will use a high-power proton accelerator to generate thermal neutrons that will be captured in 3 He to produce tritium (3H). It is expected that the APT Plant will be built and operated at the DOE's Savannah River Site (SRS) in Aiken, South Carolina. Discussion is focused on Reliability, Availability, Maintainability, and Inspectability (RAMI) modeling of recent conceptual designs for balance of plant (BOP) systems in the proposed APT Plant. In the conceptual designs for balance of plant (BOP) systems in the proposed APT Plant. In the conceptual design phase, system RAMI estimates are necessary to identify the best possible system alternative and to provide a valid picture of the cost effectiveness of the proposed system for comparison with other system alternatives. RAMI estimates in the phase must necessarily be based on generic data. The objective of the RAMI analyses at the conceptual design stage is to assist the designers in achieving an optimum design which balances the reliability and maintainability requirements among the subsystems and components

  4. ITER EDA Newsletter. V.4, no.2

    International Nuclear Information System (INIS)

    1995-02-01

    This ITER EDA (Engineering Design Activities) Newsletter issue reports on (i) the exploits of the Special Working Group (SWG-2) designated in Protocol 1 to address task allocations and drafting of Protocol 2; and (ii) a report on the Tritium Plant Group Technical Meeting held at the Naka Joint Work Site on February 1-6, 1995

  5. Studies on transpiration rates and tritium concentration in transpired water in some plant species at Kaiga site

    International Nuclear Information System (INIS)

    Selvi, S.B.; Ravi, P.M.; Hegde, A.G.

    2005-01-01

    Transpiration is the driving force for uptake of water and hence that of tritiated water from environment. Transpiration rates and tritium concentration in transpired water in some plants at Kaiga site were estimated. Good correlation was observed between transpiration rates with humidity, temperature and leaf surface area. Transpiration rates varied seasonally and diurnally due to the influence of interdependent parameters such as temperature, humidity, water availability, etc. The ratio between the tritium concentrations in transpired plant water to that in air moisture ranged from 0.1 to 0.2. (author)

  6. Determination of total tritium in urine from residents living in the vicinity of nuclear power plants in Qinshan, China.

    Science.gov (United States)

    Shen, Bao-Ming; Ji, Yan-Qin; Tian, Qing; Shao, Xiang-Zhang; Yin, Liang-Liang; Su, Xu

    2015-01-16

    To estimate the tritium doses of the residents living in the vicinity of a nuclear power plant, urine samples of 34 adults were collected from residents living near the Qinshan nuclear power plant. The tritium-in-urine (HTO plus OBT) was measured by liquid scintillation counting. The doses of tritium-in-urine from participants living at 2, 10 and 22 km were in a range of 1.26-6.73 Bq/L, 1.31-3.09 Bq/L and 2.21-3.81 Bq/L, respectively, while the average activity concentrations of participants from the three groups were 3.53 ± 1.62, 2.09 ± 0.62 and 2.97 ± 0.78 Bq/L, respectively. The personal committed effective doses for males were 2.5 ± 1.7 nSv and for females they were 2.9 ± 1.3 nSv. These results indicate that tritium concentrations in urine samples from residents living at 2 km from a nuclear power plant are significantly higher than those at 10 km. It may be the downwind direction that caused a higher dose in participants living at 22 km. All the measured doses of tritium-in-urine are in a background level range.

  7. Tritium-related fusion technology programmes under EFDA-JET

    International Nuclear Information System (INIS)

    Coad, J. P.; Ciattaglia, S.; Piazza, G.; Rosanvallon, S.; Grisolia, Ch.; Laesser, R.

    2003-01-01

    The Fusion Technology Task Force (TFFT) has a wide-ranging series of programmes in the areas of waste management and safety, tritium recovery, tritium analysis and accounting, and testing components under development for ITER at JET. Examples have been presented here in the fields of waste management and safety. In waste management, the largest effort is currently on water de-tritiation, which is considered to be the most urgent and important topic affecting JET operations, and plant design is also required for ITER. It is also the most technically challenging of the waste detritiation issues. A complete design for a water de-tritiation plant for JET (a prototype for ITER), including optimised and tested catalysts, is expected within the next 2 years. TFFT safety programmes support the on-going work on safety in preparation for ITER, including tritium spreading and dust inhalation effects for worse-case accident scenarios. Effort is also going into documenting the operational experience of the JET machine with respect to reliability of mechanical components within the tritium boundary and radiation exposure, and inferring what lessons should be learnt for ITER

  8. Safe handling of tritium

    International Nuclear Information System (INIS)

    1991-01-01

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

  9. 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)

  10. 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)

  11. Tritium dispersion around the Angra Nuclear Power Plant: boundary simplification by Diffeomorph Conformal Transformations

    Energy Technology Data Exchange (ETDEWEB)

    Meneghetti, Andre; Bodmann, Bardo E.J.; Vilhena, Marco T. de, E-mail: andre.imef@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil)

    2017-07-01

    We present progress on research concerning dispersion of tritium around the Angra Nuclear Power Plant (Angra dos Reis, Rio de Janeiro state, Brazil). In particular, we are interested in studying how dispersion behaves in scenarios with complex orography. Our proposal is to transform a problem with curvilinear boundaries into an equivalent problem with plane parallel boundaries. We modify the coordinate system through a diffeomorph conformal transformation. Consequently, the operators of the dynamical equations change according to the additional terms from the affine connection. To de ne the transformation it is necessary to satisfy strong constraints, i.e., boundaries shall be 'smooth'. Our main purpose is to solve problems using a semi-analytical resolution. Currently, semi-analytic resolutions are applied only in problems that have domain with parallel planes. As a rst step into this direction in this work we present a numerical resolution. Even with restrictions, our model can be implemented in several situations. A at region is a particular case of a curvilinear domain and can be studied, where the height of the boundary layer above rivers, lakes, basins is typically smaller and thus implies a varying boundary layer height, for instance. Thus, even in at regions variations in the boundary layer occur, which characterizes a case of a curvilinear domain. Our specific interest is the region around the Angra Nuclear Power Plant that need a large source of water for their operation. There are several nuclear power plants worldwide, that are located in mountainous regions, as for example in Japan and Brazil. As one step into a new direction we focus in this work on complex relieves. We present a simulation of tritium dispersion specifically in the area where the Angra 2 Nuclear Power Plant of is located and where the relief is characterized by a considerable complexity. (author)

  12. Tritium dispersion around the Angra Nuclear Power Plant: boundary simplification by Diffeomorph Conformal Transformations

    International Nuclear Information System (INIS)

    Meneghetti, Andre; Bodmann, Bardo E.J.; Vilhena, Marco T. de

    2017-01-01

    We present progress on research concerning dispersion of tritium around the Angra Nuclear Power Plant (Angra dos Reis, Rio de Janeiro state, Brazil). In particular, we are interested in studying how dispersion behaves in scenarios with complex orography. Our proposal is to transform a problem with curvilinear boundaries into an equivalent problem with plane parallel boundaries. We modify the coordinate system through a diffeomorph conformal transformation. Consequently, the operators of the dynamical equations change according to the additional terms from the affine connection. To de ne the transformation it is necessary to satisfy strong constraints, i.e., boundaries shall be 'smooth'. Our main purpose is to solve problems using a semi-analytical resolution. Currently, semi-analytic resolutions are applied only in problems that have domain with parallel planes. As a rst step into this direction in this work we present a numerical resolution. Even with restrictions, our model can be implemented in several situations. A at region is a particular case of a curvilinear domain and can be studied, where the height of the boundary layer above rivers, lakes, basins is typically smaller and thus implies a varying boundary layer height, for instance. Thus, even in at regions variations in the boundary layer occur, which characterizes a case of a curvilinear domain. Our specific interest is the region around the Angra Nuclear Power Plant that need a large source of water for their operation. There are several nuclear power plants worldwide, that are located in mountainous regions, as for example in Japan and Brazil. As one step into a new direction we focus in this work on complex relieves. We present a simulation of tritium dispersion specifically in the area where the Angra 2 Nuclear Power Plant of is located and where the relief is characterized by a considerable complexity. (author)

  13. ITER safety challenges and opportunities

    International Nuclear Information System (INIS)

    Piet, S.J.

    1992-01-01

    This paper reports on results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ITER is capable of meeting anticipated regulatory dose limits, but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. Much research and development (R ampersand D) and design analysis is needed to establish that ITER meets regulatory requirements. There is a further oportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, three programmatic challenges and three technical challenges must be overcome. The first step is to fund a comprehensive safety and environmental ITER R ampersand D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of three key technical challenges is plasma engineering - burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost

  14. The effect of the accident of Fukushima Daiichi Nuclear Power Plants on Niigata city based on tritium concentration in precipitation

    Energy Technology Data Exchange (ETDEWEB)

    Kataoka, N. [Department of Chemistry and Chemical Engineering, Faculty of Engineering, Niigata University 8050 Ikarashi 2-III ocho, Niigata-shi, Nishiku, Niigata Pref. 950-2181 (Japan); Environmental Analytical Center of Niigata Prefecture 53-1 Ojigouya, Kounan-ku, Niigata Pref. 950-1144 (Japan); Imaizumi, H.; Kano, N.; Ying, W. [Department of Chemistry and Chemical Engineering, Faculty of Engineering, Niigata University 8050 Ikarashi 2-III ocho, Niigata-shi, Nishiku, Niigata Pref. 950-2181 (Japan)

    2014-07-01

    The maximum value of tritium concentration in precipitation was 200 Bq/kg in Niigata city when atmospheric nuclear-bomb tests were performed in 1960s. After that, tritium concentration continuously decreased and reached to environmental revel (0.5~1.0 Bq/kg). However, after the accident of Fukushima daiichi nuclear power plants, the tritium concentration in precipitation increased in Niigata city. Therefore the observation of tritium concentration had to be carried out. In our laboratory, we have investigated the tritium concentration in precipitation and also investigated the relation between the tritium concentration and other ion (Na⁺, Mg⁺, K⁺, Ca⁺, Cl⁻, NO3⁺ or SO₄²⁻) concentration in precipitation in Niigata city. In this study, precipitation in Niigata city was gathered monthly and the evaluation of tritium concentration in precipitation was performed. In addition, we also collected the precipitation hourly (short precipitation). Each water sample thus obtained was distilled with sodium peroxide and potassium permanganate. Then the water sample thus distilled was enriched in SPE electronic enrichment apparatus, and the tritium concentration in the sample thus treated was measured in a liquid scintillation counter. On the other hand, each ion concentration in the sample was measured by ion chromatography or Atomic Absorption Spectrometry. From the above the mentioned, the following five matters can be found. (1) The tritium concentration in the samples in March and April 2011 were twice or three times higher than that in March and April in annual years. In other words, it is considered that the thus high level concentration of tritium leads to the evaluation of the effect of the accident of the Fukushima nuclear power plants on Niigata city. (2) As to the sample, the concentration of the non-sea salt Ca²⁺ (nssCa²⁺) is similar to that in March and April in annual years. (3) For each short precipitation sample collected on March 15, 2011

  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. Modification of a solid polymer electrolyte (SPE) electrolyser to ensure tritium compatibility

    International Nuclear Information System (INIS)

    Eichelhardt, F.; Cristescu, I.; Michling, R.; Welte, S.

    2010-01-01

    A Water Detritiation System (WDS) is required for the ITER Tritium Plant in order to process tritiated water which is accumulated in various subsystems (e.g. the hall ventilation systems). For the ITER-WDS, the Combined Electrolysis Catalytic Exchange (CECE) process with an electrolyser unit as one of the major components is envisaged. An experimental WDS was built and commissioned at the Tritium Laboratory Karlsruhe (TLK) for the investigation of various subsystems of the CECE process in tritium environment. The TLK-WDS consists of an 8 m Liquid Phase Catalytic Exchange column and two Solid Polymer Electrolyte electrolysers, each with a maximum hydrogen output of 1 m 3 /h. The commercially available Hogen40 electrolyser units from Proton Energy Systems are not tritium compatible concerning materials, joints and quality documentation (e.g. necessary certificates). In order to process tritiated water with tritium concentrations up to 370 GBq/kg, tritium compatibility had to be ensured by appropriate modifications. Up to now, the modified system has been operated with tritiated water for 3500 h, the maximum tritium concentration in the electrolysers being 190 GBq/kg. This contribution reports on the necessary modifications of the electrolyser units and the experiences gained thereby. The results are equally important for the ITER-WDS, where the maximum tritium concentration in the feed water of the electrolyser units will be even higher with 11 TBq/kg.

  17. Experience in handling concentrated tritium

    International Nuclear Information System (INIS)

    Holtslander, W.J.

    1985-12-01

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

  18. 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

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

    International Nuclear Information System (INIS)

    Ciampichetti, A.; Nitti, F.S.; Aiello, A.; Ricapito, I.; Liger, K.; Demange, D.; Sedano, L.; Moreno, C.; Succi, M.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    El-Guebaly, Laila A.; Malang, Siegfried

    2009-01-01

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

  1. 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)

  2. Computational model for the tritium inventory management in a nuclear plant

    International Nuclear Information System (INIS)

    Zamfirache, M.; Bornea, A.; Stefanescu, I.; Stefan, L.; Bidica, N.; Vasut, F.; David, C.

    2010-01-01

    Full text: ICIT Rm. Valcea has built an experimental pilot plant having as the main objective the development of a technology for detritiation of heavy water processed in CANDU-type reactors at the nuclear power plant in Cernavoda, Romania. The aspects related to safeguards and safety for such a detritiation installation being of great importance, a complex computational model has been developed. The model allows real-time calculation of tritium inventory in a working installation. The applied detritiation technology is catalyzed isotopic exchange coupled with cryogenic distillation. The computational model was developed as based on the experience obtained in the pilot installation operation. The model uses a set of coefficients specific for isotopic exchange processes. The coefficients were experimentally determined in the pilot installation. The model is included in the monitoring system and uses as input data the parameters acquired in real-time from automation system of the pilot installation. A friendly interface has been created to visualize the final results as data or graphs. (authors)

  3. The ITER neutral beam test facility: Designs of the general infrastructure, cryosystem and cooling plant

    International Nuclear Information System (INIS)

    Cordier, J.J.; Hemsworth, R.; Chantant, M.; Gravil, B.; Henry, D.; Sabathier, F.; Doceul, L.; Thomas, E.; Houtte, D. van; Zaccaria, P.; Antoni, V.; Bello, S. Dal; Marcuzzi, D.; Antipenkov, A.; Day, C.; Dremel, M.; Mondino, P.L.

    2005-01-01

    The CEA Association is involved, in close collaboration with ENEA, FZK, IPP and UKAEA European Associations, in the first ITER neutral beam (NB) injector and the ITER neutral beam test facility design (EFDA task ref. TW3-THHN-IITF1). A total power of about 50 MW will have to be removed in steady state on the neutral beam test facility (NBTF). The main purpose of this task is to make progress with the detailed design of the first ITER NB injector and to start the conceptual design of the ITER NBTF. The general infrastructure layout of a generic site for the NBTF includes the test facility itself equipped with a dedicated beamline vessel [P.L. Zaccaria, et al., Maintenance schemes for the ITER neutral beam test facility, this conference] and integration studies of associated auxiliaries such as cooling plant, cryoplant and forepumping system

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

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

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

  5. 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.)

  6. Computer based plant display and digital control system of Wolsong NPP Tritium Removal Facility

    International Nuclear Information System (INIS)

    Jung, C.; Smith, B.; Tosello, G.; Grosbois, J. de; Ahn, J.

    2007-01-01

    The Wolsong Tritium Removal Facility (WTRF) is an AECL-designed, first-of-a-kind facility that removes tritium from the heavy water that is used in systems of the CANDUM reactors in operation at the Wolsong Nuclear Power Plant in South Korea. The Plant Display and Control System (PDCS) provides digital plant monitoring and control for the WTRF and offers the advantages of state-of-the-art digital control system technologies for operations and maintenance. The overall features of the PDCS will be described and some of the specific approaches taken on the project to save construction time and costs, to reduce in-service life-cycle costs and to improve quality will be presented. The PDCS consists of two separate computer sub-systems: the Digital Control System (DCS) and the Plant Display System (PDS). The PDS provides the computer-based Human Machine Interface (HMI) for operators, and permits efficient supervisory or device level monitoring and control. A System Maintenance Console (SMC) is included in the PDS for the purpose of software and hardware configuration and on-line maintenance. A Historical Data System (HDS) is also included in the PDS as a data-server that continuously captures and logs process data and events for long-term storage and on-demand selective retrieval. The PDCS of WTRF has been designed and implemented based on an off-the-self PDS/DCS product combination, the Delta-V System from Emerson. The design includes fully redundant Ethernet network communications, controllers, power supplies and redundancy on selected I/O modules. The DCS provides field bus communications to interface with 3rd party controllers supplied on specialized skids, and supports HART communication with field transmitters. The DCS control logic was configured using a modular and graphical approach. The control strategies are primarily device control modules implemented as autonomous control loops, and implemented using IEC 61131-3 Function Block Diagram (FBD) and Structured

  7. R and D of tritium technology as SHI (Sumitomo Heavy Industries)

    International Nuclear Information System (INIS)

    Yokogawa, N.

    1997-01-01

    Sumitomo Heavy Industries (SHI) participated in an R and D programme on tritium processing for the first time in 1967 by joining the advanced thermal reactor project. (The thermal reactor is cooled by light water and moderated by heavy water.) From that time SHI has developed various kind of tritium handling technologies. On the basis of cooperation with Sulzer (Sulzer Chemtech Ltd. Switzerland), SHI developed a system for removing waste water for fuel reprocessing plants by water distillation technology. In the field of fusion technology, SHI has developed a hydrogen isotope separation system by cryogenic distillation and thermal diffusion methods, and a tritium storage bed. Fundamental data required for the system design were obtained through the production and operation of the above prototype systems. Recently, SHI has also been taking part in the design and planning of ITER. In the future, along with ITER design, SHI will aim at developing tritium measuring technology. (author)

  8. 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

  9. ITER safety

    International Nuclear Information System (INIS)

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  10. 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.)

  11. Tritium concentrations in the atmospheric environment at Rokkasho, Japan before the final testing of the spent nuclear fuel reprocessing plant.

    Science.gov (United States)

    Akata, Naofumi; Kakiuchi, Hideki; Shima, Nagayoshi; Iyogi, Takashi; Momoshima, Noriyuki; Hisamatsu, Shun'ichi

    2011-09-01

    This study aimed at obtaining background tritium concentrations in precipitation and air at Rokkasho where the first commercial spent nuclear fuel reprocessing plant in Japan has been under construction. Tritium concentration in monthly precipitation during fiscal years 2001-2005 had a seasonal variation pattern which was high in spring and low in summer. The tritium concentration was higher than that observed at Chiba City as a whole. The seasonal peak concentration at Rokkasho was generally higher than that at Chiba City, while the baseline concentrations of both were similar. The reason for the difference may be the effect of air mass from the Asian continent which is considered to have high tritium concentration. Atmospheric tritium was operationally separated into HTO, HT and hydrocarbon (CH(3)T) fractions, and the samples collected every 3 d-14 d during fiscal year 2005 were analyzed for these fractions. The HTO concentration as radioactivity in water correlated well with that in the precipitation samples. The HT concentration was the highest among the chemical forms analyzed, followed by the HTO and CH(3)T concentrations. The HT and CH(3)T concentrations did not have clear seasonal variation patterns. The HT concentration followed the decline previously reported by Mason and Östlund with an apparent half-life of 4.8 y. The apparent and environmental half-lives of CH(3)T were estimated as 9.2 y and 36.5 y, respectively, by combining the present data with literature data. The Intergovernmental Panel on Climate Change used the atmospheric lifetime of 12 y for CH(4) to estimate global warming in its 2007 report. The longer environmental half-life of CH(3)T suggested its supply from other sources than past nuclear weapon testing in the atmosphere. Copyright © 2011 Elsevier Ltd. All rights reserved.

  12. Tritium concentrations in the atmospheric environment at Rokkasho, Japan before the final testing of the spent nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Akata, Naofumi; Kakiuchi, Hideki; Shima, Nagayoshi; Iyogi, Takashi; Momoshima, Noriyuki; Hisamatsu, Shun'ichi

    2011-01-01

    This study aimed at obtaining background tritium concentrations in precipitation and air at Rokkasho where the first commercial spent nuclear fuel reprocessing plant in Japan has been under construction. Tritium concentration in monthly precipitation during fiscal years 2001-2005 had a seasonal variation pattern which was high in spring and low in summer. The tritium concentration was higher than that observed at Chiba City as a whole. The seasonal peak concentration at Rokkasho was generally higher than that at Chiba City, while the baseline concentrations of both were similar. The reason for the difference may be the effect of air mass from the Asian continent which is considered to have high tritium concentration. Atmospheric tritium was operationally separated into HTO, HT and hydrocarbon (CH 3 T) fractions, and the samples collected every 3 d-14 d during fiscal year 2005 were analyzed for these fractions. The HTO concentration as radioactivity in water correlated well with that in the precipitation samples. The HT concentration was the highest among the chemical forms analyzed, followed by the HTO and CH 3 T concentrations. The HT and CH 3 T concentrations did not have clear seasonal variation patterns. The HT concentration followed the decline previously reported by Mason and Ostlund with an apparent half-life of 4.8 y. The apparent and environmental half-lives of CH 3 T were estimated as 9.2 y and 36.5 y, respectively, by combining the present data with literature data. The Intergovernmental Panel on Climate Change used the atmospheric lifetime of 12 y for CH 4 to estimate global warming in its 2007 report. The longer environmental half-life of CH 3 T suggested its supply from other sources than past nuclear weapon testing in the atmosphere. - Highlights: → We observed background tritium concentrations in atmospheric environment at Rokkasho, Japan. → Tritium concentration in precipitation was high in spring and low in summer. → The atmospheric HT

  13. Tritium concentrations in the atmospheric environment at Rokkasho, Japan before the final testing of the spent nuclear fuel reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Akata, Naofumi, E-mail: nao@ies.or.jp [Department of Radioecology, Institute for Environmental Sciences, 1-7 Ienomae, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Kakiuchi, Hideki [Department of Radioecology, Institute for Environmental Sciences, 1-7 Ienomae, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Shima, Nagayoshi [Entex Inc., 1-2-8 Asahi, Kashiwa, Chiba 277-0852 (Japan); Iyogi, Takashi [Department of Radioecology, Institute for Environmental Sciences, 1-7 Ienomae, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Momoshima, Noriyuki [Radioisotope Center, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Hisamatsu, Shun' ichi [Department of Radioecology, Institute for Environmental Sciences, 1-7 Ienomae, Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2011-09-15

    This study aimed at obtaining background tritium concentrations in precipitation and air at Rokkasho where the first commercial spent nuclear fuel reprocessing plant in Japan has been under construction. Tritium concentration in monthly precipitation during fiscal years 2001-2005 had a seasonal variation pattern which was high in spring and low in summer. The tritium concentration was higher than that observed at Chiba City as a whole. The seasonal peak concentration at Rokkasho was generally higher than that at Chiba City, while the baseline concentrations of both were similar. The reason for the difference may be the effect of air mass from the Asian continent which is considered to have high tritium concentration. Atmospheric tritium was operationally separated into HTO, HT and hydrocarbon (CH{sub 3}T) fractions, and the samples collected every 3 d-14 d during fiscal year 2005 were analyzed for these fractions. The HTO concentration as radioactivity in water correlated well with that in the precipitation samples. The HT concentration was the highest among the chemical forms analyzed, followed by the HTO and CH{sub 3}T concentrations. The HT and CH{sub 3}T concentrations did not have clear seasonal variation patterns. The HT concentration followed the decline previously reported by Mason and Ostlund with an apparent half-life of 4.8 y. The apparent and environmental half-lives of CH{sub 3}T were estimated as 9.2 y and 36.5 y, respectively, by combining the present data with literature data. The Intergovernmental Panel on Climate Change used the atmospheric lifetime of 12 y for CH{sub 4} to estimate global warming in its 2007 report. The longer environmental half-life of CH{sub 3}T suggested its supply from other sources than past nuclear weapon testing in the atmosphere. - Highlights: > We observed background tritium concentrations in atmospheric environment at Rokkasho, Japan. > Tritium concentration in precipitation was high in spring and low in summer. > The

  14. RAMI modeling of plant systems for proposed tritium production and extraction facilities

    International Nuclear Information System (INIS)

    Blanchard, A.

    2000-01-01

    The control of life-cycle cost is a primary concern during the development, construction, operation, and decommissioning of DOE systems and facilities. An effective tool that can be used to control these costs, beginning with the design stage, is called a reliability, availability, maintainability, and inspectability analysis or, simply, RAMI for short. In 1997, RAMI technology was introduced to the Savannah River Site with applications at the conceptual design stage beginning with the Accelerator Production of Tritium (APT) Project and later extended to the Commercial Light Water Reactor (CLWR) Tritium Extraction Facility (TEF) Project. More recently it has been applied to the as-build Water Treatment Facilities designed for ground water environmental restoration. This new technology and database was applied to the assessment of balance-of-plant systems for the APT Conceptual Design Report. Initial results from the Heat Removal System Assessment revealed that the system conceptual design would cause the APT to fall short of its annual production goal. Using RAM technology to immediately assess this situation, it was demonstrated that the product loss could be gained back by upgrading the system's chiller unit capacity at a cost of less than $1.3 million. The reclaimed production is worth approximately $100 million. The RAM technology has now been extended to assess the conceptual design for the CLWR-TEF Project. More specifically, this technology and database is being used to translate high level availability goals into lower level system design requirements that will ensure the TEF meets its production goal. Results, from the limited number of system assessments performed to date, have already been used to modify the conceptual design for a remote handling system, improving its availability to the point that a redundant system, with its associated costs of installation and operation may no longer be required. RAMI results were also used to justify the elimination

  15. The distribution of tritium in the terrestrial and aquatic environments of the Creys-Malville nuclear power plant (2002-2005)

    International Nuclear Information System (INIS)

    Jean-Baptiste, P.; Baumier, D.; Fourre, E.; Dapoigny, A.; Clavel, B.

    2007-01-01

    The Creys-Malville nuclear plant, located on the left bank of the Rhone, was shut down in 1998. The facilities are currently in their initial stage of dismantling. In order to establish a baseline for tritium in the vicinity of the site prior to the main dismantling phase, we carried out a monitoring program between 2002 and 2005 in the main terrestrial and aquatic compartments of the local environment. Tritium levels in the groundwaters and in the Rhone waters correspond to the regional tritium concentration in precipitation. The data obtained for the terrestrial environment are also in good agreement with the regional background and do not show any specific signature linked to the nuclear plant. The various aquatic compartments of the Rhone (fish, plant, sediment) are significantly enriched in tritium both upstream and downstream of the power plant: although Tissue-Free Water Tritium concentrations are in equilibrium with the river water, the non-exchangeable fraction of organic bound tritium in plants and fishes shows values which outpace the river water background by one to two orders of magnitude, and up to four to five orders of magnitude in the sediments. This tritium anomaly is not related to the nuclear plant, as it is already present at the Swiss border 100 km upstream of the site. Although fine particles of tritiated polystyrene entering the composition of the luminous paints used by the clock industry have been suspected on several occasions, the exact nature and the origin of this tritium source remain unknown and require further investigations

  16. Tritium breeding blanket

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  17. 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

  18. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1990-01-01

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

  19. Fuel cycle design for ITER and its extrapolation to DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Kyoto 611-0011 (Japan)], E-mail: s-konishi@iae.kyoto-u.ac.jp; Glugla, Manfred [Forschungszentrum Karlsruhe, P.O. Box 3640, D 76021 Karlsruhe (Germany); Hayashi, Takumi [Apan Atomic Energy AgencyTokai, Ibaraki 319-0015 Japan (Japan)

    2008-12-15

    ITER is the first fusion device that continuously processes DT plasma exhaust and supplies recycled fuel in a closed loop. All the tritium and deuterium in the exhaust are recovered, purified and returned to the tokamak with minimal delay, so that extended burn can be sustained with limited inventory. To maintain the safety of the entire facility, plant scale detritiation systems will also continuously run to remove tritium from the effluents at the maximum efficiency. In this entire tritium plant system, extremely high decontamination factor, that is the ratio of the tritium loss to the processing flow rate, is required for fuel economy and minimized tritium emissions, and the system design based on the state-of-the-art technology is expected to satisfy all the requirements without significant technical challenges. Considerable part of the fusion tritium system will be verified with ITER and its decades of operation experiences. Toward the DEMO plant that will actually generate energy and operate its closed fuel cycle, breeding blanket and power train that caries high temperature and pressure media from the fusion device to the generation system will be the major addition. For the tritium confinement, safety and environmental emission, particularly blanket, its coolant, and generation systems such as heat exchanger, steam generator and turbine will be the critical systems, because the tritium permeation from the breeder and handling large amount of high temperature, high pressure coolant will be further more difficult than that required for ITER. Detritiation of solid waste such as used blanket and divertor will be another issue for both tritium economy and safety. Unlike in the case of ITER that is regarded as experimental facility, DEMO will be expected to demonstrate the safety, reliability and social acceptance issue, even if economical feature is excluded. Fuel and environmental issue to be tested in the DEMO will determine the viability of the fusion as a

  20. Fuel cycle design for ITER and its extrapolation to DEMO

    International Nuclear Information System (INIS)

    Konishi, Satoshi; Glugla, Manfred; Hayashi, Takumi

    2008-01-01

    ITER is the first fusion device that continuously processes DT plasma exhaust and supplies recycled fuel in a closed loop. All the tritium and deuterium in the exhaust are recovered, purified and returned to the tokamak with minimal delay, so that extended burn can be sustained with limited inventory. To maintain the safety of the entire facility, plant scale detritiation systems will also continuously run to remove tritium from the effluents at the maximum efficiency. In this entire tritium plant system, extremely high decontamination factor, that is the ratio of the tritium loss to the processing flow rate, is required for fuel economy and minimized tritium emissions, and the system design based on the state-of-the-art technology is expected to satisfy all the requirements without significant technical challenges. Considerable part of the fusion tritium system will be verified with ITER and its decades of operation experiences. Toward the DEMO plant that will actually generate energy and operate its closed fuel cycle, breeding blanket and power train that caries high temperature and pressure media from the fusion device to the generation system will be the major addition. For the tritium confinement, safety and environmental emission, particularly blanket, its coolant, and generation systems such as heat exchanger, steam generator and turbine will be the critical systems, because the tritium permeation from the breeder and handling large amount of high temperature, high pressure coolant will be further more difficult than that required for ITER. Detritiation of solid waste such as used blanket and divertor will be another issue for both tritium economy and safety. Unlike in the case of ITER that is regarded as experimental facility, DEMO will be expected to demonstrate the safety, reliability and social acceptance issue, even if economical feature is excluded. Fuel and environmental issue to be tested in the DEMO will determine the viability of the fusion as a

  1. 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)

  2. Monitoring of tritium, 60Co and 137Cs in the vicinity of the warm water outlet of the Paks Nuclear Power Plant, Hungary.

    Science.gov (United States)

    Janovics, R; Bihari, Á; Papp, L; Dezső, Z; Major, Z; Sárkány, K E; Bujtás, T; Veres, M; Palcsu, L

    2014-02-01

    Danube water, sediment and various aquatic organisms (snail, mussel, predatory and omnivorous fish) were collected upstream (at a background site) and downstream of the outlet of the warm water channel of Paks Nuclear Power Plant. Gamma emitters, tissue free-water tritium (TFWT) and total organically-bound tritium (T-OBT) measurements were performed. A slight contribution of the power plant to the natural tritium background concentration was measured in water samples from the Danube section downstream of the warm water channel. Sediment samples also contained elevated tritium concentrations, along with a detectable amount of (60)Co. In the case of biota samples, TFWT exhibited only a very slight difference compared to the tritium concentration of the Danube water, however, the OBT was higher than the tritium concentration in the Danube, independent of the origin of the samples. The elevated OBT concentration in the mollusc samples downstream of the warm water channel may be attributed to the excess emission from the nuclear power plant. The whole data set obtained was used for dose rate calculations and will be contributed to the development of the ERICA database. Copyright © 2013 Elsevier Ltd. All rights reserved.

  3. Progress in XRCS-Survey plant instrumentation and control design for ITER

    International Nuclear Information System (INIS)

    Varshney, Sanjeev; Jha, Shivakant; Simrock, Stefan; Barnsley, Robin; Martin, Vincent; Mishra, Sapna; Patil, Prabhakant; Patel, Shreyas; Kumar, Vinay

    2016-01-01

    Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.

  4. Progress in XRCS-Survey plant instrumentation and control design for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Varshney, Sanjeev, E-mail: sanjeev.varshney@iter-india.org [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Jha, Shivakant [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Simrock, Stefan; Barnsley, Robin; Martin, Vincent [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Mishra, Sapna [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India); Patil, Prabhakant [ITER-Organization, Route de Vinon sur Verdon, CS 90 046, 13067 St. Paul-Lez-Durance, Cedex (France); Patel, Shreyas; Kumar, Vinay [ITER-India, Institute for Plasma Research, Bhat, Gandhinagar, 382 428 (India)

    2016-11-15

    Highlights: • An identification of the major process functions system compliant to Plant Control Design Handbook (PCDH) has been made for XRCS-Survey plant I&C. • I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using Enterprise architect (EA). • I&C architecture, interface with ITER networks and Plants, configuration of cubicles are discussed towards nine design review deliverables. - Abstract: A real time, plasma impurity survey system based on X-ray Crystal Spectroscopy (XRCS) has been designed for ITER and will be made available in the set of first plasma diagnostics for measuring impurity ion concentrations and their in-flux. For the purpose of developing a component level design of XRCS-Survey plant I&C system that is compliant to the rules and guidelines defined in the Plant Control Design Handbook (PCDH), firstly an identification of the major process functions has been made. The preliminary plant I&C Functional Breakdown Structure (FBS) and Operation Procedure (OP) have been drafted using a system engineering tool, Enterprise Architect (EA). Conceptual I&C architecture, interface with the ITER networks and other Plants have been discussed along with the basic configuration of I&C cubicles aiming towards nine I&C deliverables for the design review.

  5. Pre-operational HTO/HT surveys in the vicinity of the Chalk River Laboratories tritium extraction plant

    International Nuclear Information System (INIS)

    Workman, W.J.G.; Brown, R.M.

    1993-08-01

    Surveys of the concentrations of HT and HTO in the atmosphere downwind of the Chalk River Laboratories reactor facilities were carried out in 1986 November, and in 1989 March, April and September under different conditions of air temperature, wind direction, and snow or vegetative cover. HT usually amounted to 1-5% of total tritium, but values up to 20% were observed, probably resulting from preferential removal of HTO. In all of the surveys, the greater persistence in the atmosphere of HT than of HTO was evident. The existing levels of HT are such that they will not be augmented significantly by chronic releases from the Tritium Extraction Plant (TEP) when it comes into operation. Hence, operation of the TEP will not facilitate studies of the environmental behaviour of chronically released HT. However, longer term studies of the distribution of HT from the existing facilities would be worthwhile. Soil and vegetation HTO levels in the study area are reported. Further studies of the distribution of tritium between the air, soil and vegetation in areas subjected to chronic exposure would be valuable

  6. Tritium transport and control in the FED

    International Nuclear Information System (INIS)

    Rogers, M.L.

    1981-01-01

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

  7. Treatment and separation of radioactive fission products tritium, rare gases and iodine in nuclear fuel reprocessing plants

    Energy Technology Data Exchange (ETDEWEB)

    Schnez, H.

    1975-07-15

    Rare gases must be separated from the process off-gases of the head-end of the Purex and Thorex processes. To achieve high decontamination factors, the quantity of off-gas should be kept as low as possible. For rare gas separation, there are two possible methods of routing the off-gas: (a) the open flushing gas circuit, in which the purified off-gas (generally air) is passed off via the stack and (b) the closed circuit in which the off-gas (nitrogen or rare gases) is recycled to the dissolver after purification. Tritium must not be entrained into the second extraction cycle or be emitted with off-gases in the form of water vapor (HTO) or HT, but must remain completely in the aqueous phase. Most of the process water is recycled, as a result of which the tritium becomes concentrated in it. This tritiated water is then subjected to tritium rectification at a suitable point in the process. Iodine is very difficult to isolate to a small number of process stages. Present aim is to release the iodine in the dissolver stage into the off-gas, so as to prevent it being entrained into the extraction part. By the injection of hot nitrogen or water vapor into the dissolver or into iodine-containing condensates, all of the iodine is passed into the gaseous phase. Scrubbers can also be used together with iodine-containing condensates to adjust the scrubbing solution. Capital cost of separation plants account for 1 to 10 percent of the total cost of the reprocessing installation, and even more if a sophisticated tritium separation system is required. (DLC)

  8. ITER safety challenges and opportunities

    International Nuclear Information System (INIS)

    Piet, S.J.

    1991-01-01

    Results of the Conceptual Design Activity (CDA) for the International Thermonuclear Experimental Reactor (ITER) suggest challenges and opportunities. ''ITER is capable of meeting anticipated regulatory dose limits,'' but proof is difficult because of large radioactive inventories needing stringent radioactivity confinement. We need much research and development (R ampersand D) and design analysis to establish that ITER meets regulatory requirements. We have a further opportunity to do more to prove more of fusion's potential safety and environmental advantages and maximize the amount of ITER technology on the path toward fusion power plants. To fulfill these tasks, we need to overcome three programmatic challenges and three technical challenges. The first programmatic challenge is to fund a comprehensive safety and environmental ITER R ampersand D plan. Second is to strengthen safety and environment work and personnel in the international team. Third is to establish an external consultant group to advise the ITER Joint Team on designing ITER to meet safety requirements for siting by any of the Parties. The first of the three key technical challenges is plasma engineering -- burn control, plasma shutdown, disruptions, tritium burn fraction, and steady state operation. The second is the divertor, including tritium inventory, activation hazards, chemical reactions, and coolant disturbances. The third technical challenge is optimization of design requirements considering safety risk, technical risk, and cost. Some design requirements are now too strict; some are too lax. Fuel cycle design requirements are presently too strict, mandating inappropriate T separation from H and D. Heat sink requirements are presently too lax; they should be strengthened to ensure that maximum loss of coolant accident temperatures drop

  9. Determination of tritium concentrations in humans before the development of a nuclear power plant in Turkey.

    Science.gov (United States)

    Dizman, Serdar; Yilmaz, Adnan; Keser, Recep

    2015-01-01

    The most widely used method to determine the level of tritium in humans is testing urine. Tritium concentrations in urine samples of 100 persons aged 18-66 years selected randomly from a pilot region in Turkey were analysed. The average activity concentration of urine samples was 4.66 ± 1.94 Bq L(-1) and the maximum activity concentration was 27.91 Bq L(-1). The minimum detectable activity was 2.38 Bq L(-1). The annual effective dose from tritium was also evaluated on the basis of the measurement results and reference values recommended by the International Commission on Radiological Protection. The effective doses for males and females were 4.56 and 3.54 nSv, respectively. These results were lower than the permissible annual effective dose for members of the public.

  10. 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.)

  11. Fusion reactor start-up without an external tritium source

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

  12. Control, monitoring and data acquisition systems in pilot plant for tritium and deuterium separation

    International Nuclear Information System (INIS)

    Retevoi, Carmen; Balteanu, Ovidiu Ioan

    1999-01-01

    To achieve the control, monitoring and data acquisition for a pilot plant for tritium and deuterium separation we have developed a system based on computer processing which transfers and treats all the data from the physical system. It consists of six basic elements: 1. a process computer ; 2. a National Instruments Amplifier/Multiplexed - SCXI 1000 with a SCXI 1100 Module with 32 differential input channels; 3. a Honeywell Digital Process Recorder - DPR 250, with 32 universal input, 12 digital input and 12 internal relays; 4. a control system for 4 throttle valves; 5. a National Instruments Data Acquisition board - AT-MIO-16XE-10, with 8 differential channels; 6. a system consisting of up to 20 digital programming current units for carbon RTD's. All the parameters from transducers, sensors and transmitters are introduced into the multiplexer and beyond into the acquisition data board. With LabVIEW soft support (National Instrument product), we made a graphic interface which displays the plant and all the parameters and their points of measure and cumulates all these data into a file. On the other hand all the pressure flow and level values are monitored by the recorder DPR 250, which has a RS232/RS485 port for PC communication. The temperatures are measured with carbon RTD's and a system comprising 20 programming current units connected by RS485 serial bus and a RS485/RS232 converter directly to the serial port of process computer. A special program makes the voltage/temperature conversion. The control system for throttle valves comprises a central unit, which communicates by RS232 bus with 4 controllers commanding 4 stepping motors. Every stepping motor is linked by a reductor to the throttle valve. This system can operate in either manual or automatic mode. The central unit can communicate with process computer via RS232 link. In this way a process computer can receive all the parameters by means of RS232/RS245 link or directly through the multiplexer and

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  14. Tritium research activities in Korea

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  15. Target/Blanket Design for the Accelerator Production of Tritium Plant

    International Nuclear Information System (INIS)

    Cappiello, M. W.

    1997-01-01

    The Accelerator Production of Tritium Target/Blanket (T/B) system is comprised of the T/B assembly and the attendant heat removal systems. The T/B assembly produces tritium using a high energy proton beam, and a spallation neutron source. The supporting heat removal systems safely remove the heat deposited by the proton beam during both normal and off-normal conditions. All systems reside within the T/B building, which is located at the end of a linear accelerator. Protons are accelerated to an energy of 1700 MeV at a current of 100 mA and are directed onto the T/B assembly. The protons interact with tungsten and lead nuclei to produce neutrons through the process of nuclear spallation. Neutron capture in 3 He gas produces tritium which is removed on a continual basis in an adjacent Tritium Separation Facility (TSF). The T/B assembly is modular to allow for replacement of spent components and minimization of waste. Systems and components are designed with safety as a primary consideration to minimize risk to the workers and the public

  16. ITER driver blanket, European Community design

    International Nuclear Information System (INIS)

    Simbolotti, G.; Zampaglione, V.; Ferrari, M.; Gallina, M.; Mazzone, G.; Nardi, C.; Petrizzi, L.; Rado, V.; Violante, V.; Daenner, W.; Lorenzetto, P.; Gierszewski, P.; Grattarola, M.; Rosatelli, F.; Secolo, F.; Zacchia, F.; Caira, M.; Sorabella, L.

    1993-01-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  17. Enhanced configuration of a water detritiation system; impact on ITER Isotope Separation System based cryogenic distillation

    Energy Technology Data Exchange (ETDEWEB)

    Cristescu, Ion, E-mail: ion.cristescu@kit.edu

    2016-11-01

    Highlights: • An enhanced configuration of ITER WDS has been developed. • The proposed configuration allows minimization of hazards due to the reduction of tritium inventory. • The load on the tritium recovery system (ITER ISS) is minimized with benefits on mitigation of the explosion hazards. - Abstract: Tritiated water is generated in the ITER systems by various sources and may contain deuterium and tritium at various concentrations. The reference process for the ITER Water Detritiation System is based on Combined Electrolysis Catalytic Exchange (CECE) configuration. During long time operation of the CECE process, the accumulation of deuterium in the electrolysis unit and consequently along the Liquid Phase Catalytic Exchange (LPCE) column is unavoidable with consequences on the overall detritiation factor of the system. Beside the deuterium issue in the process, the large amount of the tritiated water with tritium activity up to 500 Ci/kg in the electrolysis cells is a concern from the safety aspect of the plant. The enhanced configuration of a system for processing tritiated water allows mitigation of the effects due to deuterium accumulation and also reduction of tritium inventory within the electrolysis system. In addition the benefits concerning to the interface between the water detritiation system and tritium recovery based cryogenic distillation are also presented.

  18. Liquid Scoping Study for Tritium-Lean, Fast Ignition Inertial Fusion Energy Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, R C; Latkowski, J F; Durbin, S G; Meier, W R; Reyes, S

    2001-08-14

    In a thick-liquid protected chamber design, such as HYLIFE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. The molten-salt absorbs some of the material and energy given off by the target explosion. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be Self-sufficient in the tritium breeding (i.e., the amount of tritium bred in target exceeds the amount burned). These ''tritium-lean'' targets contain approximately 0.5% tritium and 99.5% deuterium, but require a large pr of 10-20 g/cm{sup 2}. Although most of the yield is provided by D-T reactions, the majority of fusion reactions are D-D, which produces a net surplus of tritium. This aspect allows for greater freedom when selecting a liquid for the protective blanket (lithium-bearing compounds are not required). This study assesses characteristics of many single, binary, and ternary molten-salts. Using the NIST Properties of Molten Salts Database, approximately 4300 molten-salts were included in the study [1]. As an initial screening, salts were evaluated for their safety and environmental (S&E) characteristics, which included an assessment of waste disposal rating, contact dose, and radioactive afterheat. Salts that passed the S&E criteria were then evaluated for neutron shielding ability and pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lift. This assessment left us with 57 molten-salts to recommend for further analysis. Many of these molten-salts contain elements such as sodium, lithium, beryllium, boron, fluorine, and oxygen. Recommendations for further analysis are also made.

  19. Liquid Scoping Study for Tritium-Lean, Fast Ignition Inertial Fusion Energy Power Plants

    International Nuclear Information System (INIS)

    Schmitt, R C; Latkowski, J F; Durbin, S G; Meier, W R; Reyes, S

    2001-01-01

    In a thick-liquid protected chamber design, such as HYLIFE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. The molten-salt absorbs some of the material and energy given off by the target explosion. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be Self-sufficient in the tritium breeding (i.e., the amount of tritium bred in target exceeds the amount burned). These ''tritium-lean'' targets contain approximately 0.5% tritium and 99.5% deuterium, but require a large pr of 10-20 g/cm 2 . Although most of the yield is provided by D-T reactions, the majority of fusion reactions are D-D, which produces a net surplus of tritium. This aspect allows for greater freedom when selecting a liquid for the protective blanket (lithium-bearing compounds are not required). This study assesses characteristics of many single, binary, and ternary molten-salts. Using the NIST Properties of Molten Salts Database, approximately 4300 molten-salts were included in the study [1]. As an initial screening, salts were evaluated for their safety and environmental (SandE) characteristics, which included an assessment of waste disposal rating, contact dose, and radioactive afterheat. Salts that passed the SandE criteria were then evaluated for neutron shielding ability and pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lift. This assessment left us with 57 molten-salts to recommend for further analysis. Many of these molten-salts contain elements such as sodium, lithium, beryllium, boron, fluorine, and oxygen. Recommendations for further analysis are also made

  20. A survey of methods to immobilize tritium and carbon-14 arising from a nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Taylor, P.

    1991-02-01

    This report reviews the literature on methods to separate and immobilize tritium ( 3 H) and carbon-14 ( 14 C) released from U0 2 fuel in a nuclear fuel reprocessing plant. It was prepared as part of a broader review of fuel reprocessing waste management methods that might find future application in Canada. The calculated inventories of both 3 H and 14 C in used fuel are low; special measures to limit releases of these radionuclides from reprocessing plants are not currently in place, and may not be necessary in future. If required, however, several possible approaches to the concentration and immobilization of both radionuclides are available for development. Technology to control these radionuclides in reactor process streams is in general more highly developed than for reprocessing plant effluent, and some control methods may be adaptable to reprocessing applications

  1. JET experiments with tritium and deuterium–tritium mixtures

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. ITER fuel cycle systems layout

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-10-01

    The ITER fuel cycle building (FCB) will contain the following systems: fuel purification - permeator based; fuel purification - molecular sieves; impurity treatment; waste water storage and treatment; isotope separation; waste water tritium extraction; tritium extraction from solid breeder; tritium extraction from test modules; tritium storage, shipping and receiving; tritium laboratory; atmosphere detritiation systems; fuel cycle control centre; tritiated equipment maintenance space; control maintenance space; health physics laboratory; access, access control and facilities. The layout of the FCB and the requirements for these systems are described. (10 figs.)

  4. Tritium inventory of a target factor in an ICF power plant

    International Nuclear Information System (INIS)

    Sherohman, J.W.

    1981-01-01

    A preliminary parametric study has been performed to estimate the tritium inventory of a conjectured Target Factory. The inventory of a proposed tritiated fuel processing system was determined as a function of production efficiency, storage factor, and time interval for the slowest processing step. Results indicated that a study of this type will be beneficial in evaluating possible processing schemes for the production of tritiated laser fusion targets

  5. Tritium inventory tracking and management

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  6. Experimental investigation and numerical modelling of tritium wash-out by precipitation in the area of the nuclear power plant of Paks, Hungary

    International Nuclear Information System (INIS)

    Koelloe, Z.; Palcsu, L.; Major, Z.; Papp, L.; Molnar, M.

    2009-01-01

    Complete text of publication follows. Tritium is an important radioactive isotope because its natural occurrence in the air and precipitation due to natural and artificial sources. In order to investigate the natural changes in tritium concentration, the artificial component has to be known. Some field experiments were carried out before to investigate the washout of tritium by precipitation emitted from artificial sources, but none of them were carried out around a real power plant, measuring the deposition pattern. We collected rainwater around the nuclear power plant of Paks (Paks NPP), and analyzed them for tritium. The rainwater samplers were constructed at ATOMKI. Their structure consists of a funnel, support parts and a long tube, acting as the storage vessel for the rainwater. One end of the tube is connected to the funnel, the other end is open. This way when 'new' rainwater falls, it pushes out the 'old' water, causing that always the last rainwater is in the tube (if enough rain is falling). In total 54 samplers were placed out around the Paks NPP in two half circles, with radiuses 400 and 800 m, pointing east. We collected samples after a rain period on 7-8 June 2009. They were prepared and measured with liquid scintillation counting (LSC) for tritium. We measured some samples also with the 3 He-ingrowth method, to ensure better accuracy. The measurement results in Fig. 1 clearly show the trace of the tritium plume emitted from the plant. However, the highest values are not very high, compared to environmental levels, and considering the fact that all the samples were collected from the area of the plant. A numerical model was coded to calculate the washout of tritium theoretically from the meteorological and emission data, and to estimate the effect of the plant in larger distances. In Fig. 1 it is apparent that the model, however does not describe the data precisely, but gives reasonable results, especially for the outer circle. The calculations also

  7. Tritium and hydrogen behaviour at Phenix power plant. Application to development and validation of KUMAR type models

    International Nuclear Information System (INIS)

    Tibi, A.; Misraki, J.; Feron, D.

    1984-04-01

    Experimentations at Phenix reactor confirmed the fitness of the KUMAR model for predicting the behaviour of hydrogen and tritium, and thus, prevision of the tritium distribution at Super Phenix reactor: calculation of the tritium content of a regenerated secondary cold trap, behaviour of hydrogen during power operation, the primary cold trap being deliberately outage, and estimation of the tritium and hydrogen sources and permeation transfer ratios [fr

  8. Export Control Requirements for Tritium Processing Design and R&D

    Energy Technology Data Exchange (ETDEWEB)

    Hollis, William Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maynard, Sarah-Jane Wadsworth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-04-05

    This document will address requirements of export control associated with tritium plant design and processes. Los Alamos National Laboratory has been working in the area of tritium plant system design and research and development (R&D) since the early 1970’s at the Tritium Systems Test Assembly (TSTA). This work has continued to the current date with projects associated with the ITER project and other Office of Science Fusion Energy Science (OS-FES) funded programs. ITER is currently the highest funding area for the DOE OS-FES. Although export control issues have been integrated into these projects in the past a general guidance document has not been available for reference in this area. To address concerns with currently funded tritium plant programs and assist future projects for FES, this document will identify the key reference documents and specific sections within related to tritium research. Guidance as to the application of these sections will be discussed with specific detail to publications and work with foreign nationals.

  9. Export Control Requirements for Tritium Processing Design and R&D

    Energy Technology Data Exchange (ETDEWEB)

    Hollis, William Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maynard, Sarah-Jane Wadsworth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-30

    This document will address requirements of export control associated with tritium plant design and processes. Los Alamos National Laboratory has been working in the area of tritium plant system design and research and development (R&D) since the early 1970’s at the Tritium Systems Test Assembly (TSTA). This work has continued to the current date with projects associated with the ITER project and other Office of Science Fusion Energy Science (OS-FES) funded programs. ITER is currently the highest funding area for the DOE OS-FES. Although export control issues have been integrated into these projects in the past a general guidance document has not been available for reference in this area. To address concerns with currently funded tritium plant programs and assist future projects for FES, this document will identify the key reference documents and specific sections within related to tritium research. Guidance as to the application of these sections will be discussed with specific detail to publications and work with foreign nationals.

  10. Tritium in HTR systems

    International Nuclear Information System (INIS)

    Steinwarz, W.

    1987-07-01

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

  11. Comparison of balance of tritium activity in waste water from nuclear power plants and at selected monitoring sites in the Vltava River, Elbe River and Jihlava (Dyje) River catchments in the Czech Republic.

    Science.gov (United States)

    Hanslík, Eduard; Marešová, Diana; Juranová, Eva; Sedlářová, Barbora

    2017-12-01

    During the routine operation, nuclear power plants discharge waste water containing a certain amount of radioactivity, whose main component is the artificial radionuclide tritium. The amounts of tritium released into the environment are kept within the legal requirements, which minimize the noxious effects of radioactivity, but the activity concentration is well measurable in surface water of the recipient. This study compares amount of tritium activity in waste water from nuclear power plants and the tritium activity detected at selected relevant sites of surface water quality monitoring. The situation is assessed in the catchment of the Vltava and Elbe Rivers, affected by the Temelín Nuclear Power Plant as well as in the Jihlava River catchment (the Danube River catchment respectively), where the waste water of the Dukovany Nuclear Power Plant is discharged. The results show a good agreement of the amount of released tritium stated by the power plant operator and the tritium amount detected in the surface water and highlighted the importance of a robust independent monitoring of tritium discharged from a nuclear power plant which could be carried out by water management authorities. The outputs of independent monitoring allow validating the values reported by a polluter and expand opportunities of using tritium as e.g. tracer. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Development step toward fusion power plant and role of experimental reactor ITER

    International Nuclear Information System (INIS)

    Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

    2005-01-01

    The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

  13. Tritium sources

    International Nuclear Information System (INIS)

    Glodic, S.; Boreli, F.

    1993-01-01

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

  14. Tritium storage

    International Nuclear Information System (INIS)

    Hircq, B.

    1989-01-01

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

  15. The ITER fusion reactor and its role in the development of a fusion power plant

    International Nuclear Information System (INIS)

    McLean, A.

    2002-01-01

    Energy from nuclear fusion is the future source of sustained, full life-cycle environmentally benign, intrinsically safe, base-load power production. The nuclear fusion process powers our sun, innumerable other stars in the sky, and some day, it will power the Earth, its cities and our homes. The International Thermonuclear Experimental Reactor, ITER, represents the next step toward fulfilling that promise. ITER will be a test bed for key steppingstones toward engineering feasibility of a demonstration fusion power plant (DEMO) in a single experimental step. It will establish the physics basis for steady state Tokamak magnetic containment fusion reactors to follow it, exploring ion temperature, plasma density and containment time regimes beyond the breakeven power condition, and culminating in experimental fusion self-ignition. (author)

  16. Environmental monitoring for tritium in tritium separation facility

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  17. Environmental monitoring for tritium at tritium separation facility

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  18. Tritium concentration in the air at Rokkasho, Aomori before nuclear fuel reprocessing plant operation

    International Nuclear Information System (INIS)

    Kakiuchi, Hideki; Iyogi, Takashi; Hisamatsu, Shun'ichi; Ichinohe, Takaaki

    2007-01-01

    Three different chemical forms in the atmosphere, water vapor (HTO), molecular hydrogen (HT) and hydrocarbons (CH 3 T), were separately collected at Rokkasho, Aomori Prefecture, Japan, from April 2005 to December 2005, and their radioactivity was measured to clarify their regional features. Water vapor was collected by passing through a cold trap cooled at -15degC and then a column (50 mm φ) packed with 500 g molecular sieve 3A (MS-3A). Molecular hydrogen and hydrocarbons were separately oxidized to water by catalysts, and then collected with MS-3A columns as a form of water. Hydrogen and CH 4 gases were added as carriers prior to the oxidation by the conventional method. Since H 2 is highly flammable, the method was modified so as not to use it. Tritium-free water vapor was added to dried air as a carrier of water produced by oxidation of HT after collecting air moisture. Conversion of HT to HTO was carried out with a Pt honeycomb catalyst, which can oxidize HT completely at 100degC. Hydrocarbons were oxidized with a Pd catalyst at 350degC and the resulting water was trapped by a MS-3A column. Water in the MS-3A column was desorbed out by heating the column at 400degC with flowing N 2 gas, and was recovered on a cold trap. Tritium concentrations were determined by liquid scintillation counting with a background counter. (author)

  19. Liquid Wall Options for Tritium-Lean Fast Ignition Inertial Fusion Energy Power Plants

    International Nuclear Information System (INIS)

    Reyes, S.; Schmitt, R.C.; Latkowski, J.F.; Durbin, S.G.' Sanz, J.

    2002-01-01

    In an inertial fusion energy (FE) thick-liquid chamber design such as HYLEE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be self-sufficient in terms of tritium breeding (i.e., the amount of tritium bred in target exceeds the amount burned). This aspect allows for greater freedom when selecting a liquid for the protective blanket, given that lithium-bearing compounds are no longer required. The present work assesses the characteristics of many single, binary, and ternary molten-salts using the NIST Properties of Molten Salts Database. As an initial screening, salts were evaluated for their safety and environmental (S and E) characteristics, which included an assessment of waste disposal rating, contact dose, and radioactive afterheat. Salts that passed the S and E criteria were then evaluated for required pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lifting power. The results of the assessment are used to identify those molten-salts that are suitable for potential liquid-chamber fast-ignition IFE concepts, from both the S and E and pumping power perspective. Recommendations for further analysis are also made

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Tritium analysis in environmental samples around Nuclear Power Plants and nationwide surveillance of radionuclides in some environmental samples(meat and drinking water)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Yong Woo; Han, Man Jung; Cho, Seong Won; Cho, Hong Jun; Oh, Hyeon Kyun; Lee, Jeong Min; Chang, Jae Sook [KORTIC, Taejon (Korea, Republic of)

    2001-12-15

    12 kind of environmental samples such as soil, underground water, seawater, etc. around the Nuclear Power Plants(NPP) and surface seawater around the Korea peninsula were sampled, For the samples of rain, pine-needle, air, seawater, underground water, chinese cabbage, grain of rice and milk sampled around NPP, and surface seawater and rain sampled all around country, tritium concentration was measured, The tritium concentration in the tap water and the gamma activity in the domestic and imported beef that were sampled at ward in the large city in Korea(Seoul, Pusan, Taegu, Taejun, Inchun, Kwangju) were analyzed for the meat and drinking waters. As the results of analyzing, tritium concentration in rain and tap water were very low all around country, but a little higher around the NPP than general surrounding. At the Wolsung NPP, tritium concentration was descend according to distance from the stack. Tritium activity of surface seawater around the Korea peninsula was also, very low. The measured radioactive elements in the beef is the same as the radioactive elements on the earth surface.

  3. Effects of tritium in elastomers

    International Nuclear Information System (INIS)

    Zapp, P.E.

    1982-01-01

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

  4. Ontario Hydro diversifies into tritium

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

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

  5. 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

  6. Tritium trick

    Science.gov (United States)

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

    1971-01-01

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

  7. 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)

  8. Oxidation of atmospheric molecular tritium in plant leaves, lichens and mosses

    Energy Technology Data Exchange (ETDEWEB)

    Ichimasa, Michiko; Ichimasa, Yusuke; Yagi, Yoshimi; Ko, Rinkei [Ibaraki Univ., Mito (Japan). Faculty of Science; Suzuki, Masatomo; Akita, Yasukazu

    1989-12-01

    The oxidation of atmospheric molecular tritium (HT) in vegetation was determined by in vitro experiments for pine needles, pine bark, lichens attached to pine trees, taken from a coastal pine forest in Ibaraki prefecture and comparison of such measurements was made with those in soil. The oxidation of HT in pine needles was extremely low, being only about 1/40000 that in the surface soil of a pine forest, whereas its oxidation in pine bark with a lichen was almost 1000-7000 times higher than that in pine needles. HT oxidation in pine bark, a lichen and a moss was determined in each case under light and dark conditions and was found to be essentially the same. All mosses and lichens examined in the present study were found to have unusually high levels of HT oxidation whether their habitat was tree or ground surface. (author).

  9. Oxidation of atmospheric molecular tritium in plant leaves, lichens and mosses

    International Nuclear Information System (INIS)

    Ichimasa, Michiko; Ichimasa, Yusuke; Yagi, Yoshimi; Ko, Rinkei; Suzuki, Masatomo; Akita, Yasukazu.

    1989-01-01

    The oxidation of atmospheric molecular tritium (HT) in vegetation was determined by in vitro experiments for pine needles, pine bark, lichens attached to pine trees, taken from a coastal pine forest in Ibaraki prefecture and comparison of such measurements was made with those in soil. The oxidation of HT in pine needles was extremely low, being only about 1/40000 that in the surface soil of a pine forest, whereas its oxidation in pine bark with a lichen was almost 1000-7000 times higher than that in pine needles. HT oxidation in pine bark, a lichen and a moss was determined in each case under light and dark conditions and was found to be essentially the same. All mosses and lichens examined in the present study were found to have unusually high levels of HT oxidation whether their habitat was tree or ground surface. (author)

  10. The ITER Fast Plant System Controller ATCA prototype Real-Time Software Architecture

    International Nuclear Information System (INIS)

    Carvalho, B.B.; Santos, B.; Carvalho, P.F.; Neto, A.; Boncagni, L.; Batista, A.J.N.; Correia, M.; Sousa, J.; Gonçalves, B.

    2013-01-01

    Highlights: ► High performance ATCA systems for fast control and data acquisition. ► IEEE1588 timing system and synchronization. ► Plasma control algorithms. ► Real-time control software frameworks. ► Targeted for nuclear fusion experiments with long duration discharges. -- Abstract: IPFN is developing a prototype Fast Plant System Controller (FPSC) based in ATCA embedded technologies dedicated to ITER CODAC data acquisition and control tasks in the sub-millisecond range. The main goal is to demonstrate the usability of the ATCA standard and its enhanced specifications for the high speed, very high density parallel data acquisition needs of the most demanding ITER tokamak plasma Instrumentation and Control (I and C) systems. This effort included the in-house development of a new family of high performance ATCA I/O and timing boards. The standard ITER software system CODAC Core System (CCS) v3.1, with the control based in the EPICS system does not cover yet the real-time requirements fulfilled by this hardware, so a new set of software components was developed for this specific platform, attempting to integrate and leverage the new features in CSS, for example the Multithreaded Application Real Time executor (MARTe) software framework, the new Data Archiving Network (DAN) solution, an ATCA IEEE-1588-2008 timing interface, and the Intelligent Platform Management Interface (IPMI) for system monitoring and remote management. This paper presents the overall software architecture for the ATCA FPSC, as well a discussion on the ITER constrains and design choices and finally a detailed description of the software components already developed

  11. The ITER Fast Plant System Controller ATCA prototype Real-Time Software Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, B.B., E-mail: bernardo@ipfn.ist.utl.pt [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Santos, B.; Carvalho, P.F.; Neto, A. [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal); Boncagni, L. [Associazione Euratom-ENEA sulla Fusione, Frascati Research Centre, Division of Fusion Physics, Frascati, Rome (Italy); Batista, A.J.N.; Correia, M.; Sousa, J.; Gonçalves, B. [Associacao EURATOM/IST Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Universidade Tecnica de Lisboa, P-1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► High performance ATCA systems for fast control and data acquisition. ► IEEE1588 timing system and synchronization. ► Plasma control algorithms. ► Real-time control software frameworks. ► Targeted for nuclear fusion experiments with long duration discharges. -- Abstract: IPFN is developing a prototype Fast Plant System Controller (FPSC) based in ATCA embedded technologies dedicated to ITER CODAC data acquisition and control tasks in the sub-millisecond range. The main goal is to demonstrate the usability of the ATCA standard and its enhanced specifications for the high speed, very high density parallel data acquisition needs of the most demanding ITER tokamak plasma Instrumentation and Control (I and C) systems. This effort included the in-house development of a new family of high performance ATCA I/O and timing boards. The standard ITER software system CODAC Core System (CCS) v3.1, with the control based in the EPICS system does not cover yet the real-time requirements fulfilled by this hardware, so a new set of software components was developed for this specific platform, attempting to integrate and leverage the new features in CSS, for example the Multithreaded Application Real Time executor (MARTe) software framework, the new Data Archiving Network (DAN) solution, an ATCA IEEE-1588-2008 timing interface, and the Intelligent Platform Management Interface (IPMI) for system monitoring and remote management. This paper presents the overall software architecture for the ATCA FPSC, as well a discussion on the ITER constrains and design choices and finally a detailed description of the software components already developed.

  12. Developments in data acquisition systems with LabView datalogging and supervisory control module for tritium removal plant, with data base and process analysis

    International Nuclear Information System (INIS)

    Moraru, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu; Bucur, Ciprian; Hartescu, Florin

    2006-01-01

    Full text: The implementation of the new trends for tritium processing nuclear plants, and especially those with an experimental character or of new technology development, shows a very high complexity due to issues raised by the integration of a high diversity of instrumentation and equipment into a unitary control system of the technological process. Keeping the system's flexibility is a demand of the experimental plants for which the change of configuration, process and parameters is something usual. The big amount of data that needs to be monitored, stored and accessed for ulterior analyses demands the achievement of an information network where the data acquiring, control and analysis systems of the technological process can be integrated with a data base system. Thus, integrated computing and control systems needed for the control of the technological process will be executed, to be continued with the execution of failure protection system, by choosing methods corresponding to the technological processes within the tritium processing nuclear plants. (authors)

  13. Management of tritium at nuclear facilities

    International Nuclear Information System (INIS)

    1984-01-01

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

  14. Fuel cycle design evolution from FDR-ITER to RTO/RC-ITER

    International Nuclear Information System (INIS)

    Murdoch, D.K.; Glugla, M.; Kveton, O.

    2000-01-01

    Instantaneous fuelling and plasma exhaust flow rates for the reduced technical objective/reduced cost version of International Thermonuclear Experimental Reactor (RTO/RC-ITER) are similar to those described in the Final Design Report (FDR) of ITER, despite the reduction in fusion power by a factor of about two. However, the reduced pulse length and the lower fraction of campaign time spent in burn mode, together with the lower integrated operating lifetime proposed, will generate cost savings in several systems of the fuel cycle. As the quantity of tritium handled per pulse is now smaller, this could be buffered, allowing systems in the tritium plant still to operate in steady state mode as in the FDR design, thereby increasing the potential for downsizing of system capacities. The lower operating time fraction will increase performance margins for some systems, for example, the Torus Exhaust Gas Processing System (TEGPS) which was designed to meet a specified daily release rate for the FDR design conditions which were more onerous than RTO/RC-ITER. As no break through of tritium into cooling water is now expected, the duties of the Water and Atmosphere Detritiation Systems are considerably reduced, and design concepts which are simpler, cheaper and more amenable to modular implementation can be adopted

  15. Tritium-assisted fusion breeders

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1983-08-01

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

  16. Recent progress of China HCCB TBM tritium system

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  17. ITER Safety and Licensing

    International Nuclear Information System (INIS)

    Girard, J-.P; Taylor, N.; Garin, P.; Uzan-Elbez, J.; GULDEN, W.; Rodriguez-Rodrigo, L.

    2006-01-01

    The site for the construction of ITER has been chosen in June 2005. The facility will be implemented in Europe, south of France close to Marseille. The generic safety scheme is now under revision to adapt the design to the host country regulation. Even though ITER will be an international organization, it will have to comply with the French requirements in the fields of public and occupational health and safety, nuclear safety, radiation protection, licensing, nuclear substances and environmental protection. The organization of the central team together with its partners organized in domestic agencies for the in-kind procurement of components is a key issue for the success of the experimentation. ITER is the first facility that will achieve sustained nuclear fusion. It is both important for the experimental one-of-a-kind device, ITER itself, and for the future of fusion power plants to well understand the key safety issues of this potential new source of energy production. The main safety concern is confinement of the tritium, activated dust in the vacuum vessel and activated corrosion products in the coolant of the plasma-facing components. This is achieved in the design through multiple confinement barriers to implement the defence in depth approach. It will be demonstrated in documents submitted to the French regulator that these barriers maintain their function in all postulated incident and accident conditions. The licensing process started by examination of the safety options. This step has been performed by Europe during the candidature phase in 2002. In parallel to the final design, and taking into account the local regulations, the Preliminary Safety Report (RPrS) will be drafted with support of the European partner and others in the framework of ITER Task Agreements. Together with the license application, the RPrS will be forwarded to the regulatory bodies, which will launch public hearings and a safety review. Both processes must succeed in order to

  18. Considerations affecting deep-well disposal of tritium-bearing low-level aqueous waste from nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Trevorrow, L.E.; Warner, D.L.; Steindler, M.J.

    1977-03-01

    Present concepts of disposal of low-level aqueous wastes (LLAW) that contain much of the fission-product tritium from light water reactors involve dispersal to the atmosphere or to surface streams at fuel reprocessing plants. These concepts have been challenged in recent years. Deep-well injection of low-level aqueous wastes, an alternative to biospheric dispersal, is the subject of this presentation. Many factors must be considered in assessing its feasibility, including technology, costs, environmental impact, legal and regulatory constraints, and siting. Examination of these factors indicates that the technology of deep-well injection, extensively developed for other industrial wastes, would require little innovation before application to low-level aqueous wastes. Costs would be low, of the order of magnitude of 10 -4 mill/kWh. The environmental impact of normal deep-well disposal would be small, compared with dispersal to the atmosphere or to surface streams; abnormal operation would not be expected to produce catastrophic results. Geologically suitable sites are abundant in the U.S., but a well would best be co-located with the fuel-reprocessing plant where the LLAW is produced. Legal and regulatory constraints now being developed will be the most important determinants of the feasibility of applying the method

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

    OpenAIRE

    中村 博文; 西 正孝

    2003-01-01

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

  20. Tritium Removal from Carbon Plasma Facing Components

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  1. The background concentration of the tritium in surface water before operation of the nuclear power plant Temelin

    International Nuclear Information System (INIS)

    Tomasek, M.; Wilhelmova, L.

    1997-01-01

    In this announcement the results of the tritium determination in surface waters in the period 1991-1996 are summarized. The water samples from Vltava river under the estuary of waste canal of the NPP Temelin and on the entrance in Prague water-work were taken away. Moreover, the samples from local rivers in the area of NPP Temelin construction which can be contaminated by scouring after condensation of the gaseous exhalation of the NPP were taken away. The samples by standard method in month interval were taken away. From every sample for the measurement of the tritium activity two parallel samples were prepared. The tritium activity with the scintillation spectrometer at the optimal conditions was measured. From measured values seasonal character and gradual decreasing of tritium activity in observed period is evident. Maximal activities obviously in the summer months are observed. From the exponential regression half time of decreasing of the tritium activity 8.1 year was calculated [sk

  2. Tritium autoradiography

    International Nuclear Information System (INIS)

    Caskey, G.R. Jr.

    1981-01-01

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

  3. Monitoring system in Labview data logging and supervisory control module and Labview 8 with process analyses of the Cryogenic Pilot Plant for Tritium Removal

    International Nuclear Information System (INIS)

    Moraru, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Bucur, Ciprian; Stefan, Liviu; Bornea, Anisia; Stefanescu, Ioan

    2008-01-01

    The system responds to the monitoring requirements of the technological processes specific to the nuclear installation that processes radioactive substances, with severe consequences in case of technological failure, as is the case with a tritium processing nuclear plant. The big amount of data that needs to be processed, stored and accessed for real time simulation and optimization demands the achievement of the virtual technologic platform where the data acquiring, control and analysis systems of the technological process can be integrated with an advanced technological monitoring system. Thus, integrated computing and monitoring systems needed for the supervising of the technological process will be effected, to be then continued with optimization of the system, by implementing new and performing methods corresponding to the technological processes within the tritium removal processing nuclear plants. (authors)

  4. NetCDF based data archiving system applied to ITER Fast Plant System Control prototype

    International Nuclear Information System (INIS)

    Castro, R.; Vega, J.; Ruiz, M.; De Arcas, G.; Barrera, E.; López, J.M.; Sanz, D.; Gonçalves, B.; Santos, B.; Utzel, N.; Makijarvi, P.

    2012-01-01

    Highlights: ► Implementation of a data archiving solution for a Fast Plant System Controller (FPSC) for ITER CODAC. ► Data archiving solution based on scientific NetCDF-4 file format and Lustre storage clustering. ► EPICS control based solution. ► Tests results and detailed analysis of using NetCDF-4 and clustering technologies on fast acquisition data archiving. - Abstract: EURATOM/CIEMAT and Technical University of Madrid (UPM) have been involved in the development of a FPSC (Fast Plant System Control) prototype for ITER, based on PXIe (PCI eXtensions for Instrumentation). One of the main focuses of this project has been data acquisition and all the related issues, including scientific data archiving. Additionally, a new data archiving solution has been developed to demonstrate the obtainable performances and possible bottlenecks of scientific data archiving in Fast Plant System Control. The presented system implements a fault tolerant architecture over a GEthernet network where FPSC data are reliably archived on remote, while remaining accessible to be redistributed, within the duration of a pulse. The storing service is supported by a clustering solution to guaranty scalability, so that FPSC management and configuration may be simplified, and a unique view of all archived data provided. All the involved components have been integrated under EPICS (Experimental Physics and Industrial Control System), implementing in each case the necessary extensions, state machines and configuration process variables. The prototyped solution is based on the NetCDF-4 (Network Common Data Format) file format in order to incorporate important features, such as scientific data models support, huge size files management, platform independent codification, or single-writer/multiple-readers concurrency. In this contribution, a complete description of the above mentioned solution is presented, together with the most relevant results of the tests performed, while focusing in the

  5. JET Tokamak, preparation of a safety case for tritium operations

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, Helen, E-mail: helen.boyer@ccfe.ac.uk [CCFE, Culham Science Centre (United Kingdom); Plummer, David; Johnston, Jane [CCFE, Culham Science Centre (United Kingdom)

    2016-11-01

    Highlights: • A safety case incorporating technical and ITER related upgrades. • Hazard analysis reworked to include new modelling assessments. • Fitness for purpose assessment of safety controls. - Abstract: A new Safety Case is required to permit tritium operations on JET during the forthcoming DTE2 campaign. The outputs, benefits and lessons learned associated with the production of this Safety Case are presented. The changes that have occurred to the Safety Case methodology since the last JET tritium Safety Case are reviewed. Consideration is given to the effects of modifications, particularly ITER related changes, made to the JET and the impact these have on the hazard assessments as well as normal operations. Several specialized assessments, including recent MELCOR modelling, have been undertaken to support the production of this Safety Case and the impact of these assessments is outlined. Discussion of the preliminary actions being taken to progress implementation of this Safety Case is provided, highlighting new methods to improve the dissemination of the key Safety Case results to the plant operators. Finally, the work required to complete this Safety Case, before the next tritium campaign, is summarized.

  6. The path from ITER to a power plant - initial results from the ARIES ''Pathways'' program

    International Nuclear Information System (INIS)

    Najmabadi, F.

    2007-01-01

    The US national power plant studies program, ARIES, has initiated a 3-year integrated study, called the ''Pathways Program'' to investigate what the fusion program needs to do, in addition to successful operation of the ITER, in order to transform fusion into a commercial reality. The US power industry and regulatory agencies view the demonstration power plant, DEMO, as a device which is build and operated by industry, possibly with government participation, to demonstrate the commercial readiness of fusion power. As such, the ''Pathways'' programs will investigate what is needed, in addition to successful operation of ITER, to convince industry to move forward with a fusion DEMO. While many reports exists that provide a strategic view of the needs for fusion development; in the ITER era, a much more detailed view is needed to provide the necessary information for program planning. By comparing the anticipated results from ITER and existing facilities with the requirements for a power plant in the first phase of the Pathways study, we will develop a comprehensive list of remaining R and D items for developing fusion, will identify metrics for distributing resources among R and D issues, and will identify which of those items can/should be done in existing or simulation facilities. In the second phase of the study, we will develop potential embodiments for the fusion test facility (ies) and explore their cost/performance parametrically. An important by-product of this study is the identification of key R and D issues that can be performed and resolved in existing facilities to make the fusion facility cheaper and/or a higher performance device. This paper summarizes the results from the first phase of our study. We have adopted a ''holistic'' or integrated approach with the focus on the needs of the customer. In such an approach, the remaining R and D should generate all of the information needed by industry to move forward with the DEMO, i.e., data needed to

  7. ITER EDA status

    International Nuclear Information System (INIS)

    Aymar, R.

    2001-01-01

    The Project has focused on drafting the Plant Description Document (PDD), which will be published as the Technical Basis for the ITER Final Design Report (FDR), and its related documentation in time for the ITER review process. The preparations have involved continued intensive detailed design work, analyses and assessments by the Home Teams and the Joint Central Team, who have co-operated closely and efficiently. The main technical document has been completed in time for circulation, as planned, to TAC members for their review at TAC-17 (19-22 February 2001). Some of the supporting documents, such as the Plant Design Specification (PDS), Design Requirements and Guidelines (DRG1 and DRG2), and the Plant Safety Requirement (PSR) are also available for reference in draft form. A summary paper of the PDD for the Council's information is available as a separate document. A new documentation structure for the Project has been established. This hierarchical structure for documentation facilitates the entire organization in a way that allows better change control and avoids duplications. The initiative was intended to make this documentation system valid for the construction and operation phases of ITER. As requested, the Director and the JCT have been assisting the Explorations to plan for future joint technical activities during the Negotiations, and to consider technical issues important for ITER construction and operation for their introduction in the draft of a future joint implementation agreement. As charged by the Explorers, the Director has held discussions with the Home Team Leaders in order to prepare for the staffing of the International Team and Participants Teams during the Negotiations (Co-ordinated Technical Activities, CTA) and also in view of informing all ITER staff about their future directions in a timely fashion. One important element of the work was the completion by the Parties' industries of costing studies of about 83 ''procurement packages

  8. EPICS device support module as ATCA system manager for the ITER fast plant system controller

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Paulo F., E-mail: pricardofc@ipfn.ist.utl.pt [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico – Universidade Técnica de Lisboa, Lisboa (Portugal); Santos, Bruno; Gonçalves, Bruno; Carvalho, Bernardo B.; Sousa, Jorge; Rodrigues, A.P.; Batista, António J.N.; Correia, Miguel; Combo, Álvaro [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico – Universidade Técnica de Lisboa, Lisboa (Portugal); Correia, Carlos M.B.A. [Centro de Instrumentação, Departamento de Física, Universidade de Coimbra, Coimbra (Portugal); Varandas, Carlos A.F. [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico – Universidade Técnica de Lisboa, Lisboa (Portugal)

    2013-10-15

    Highlights: ► In Nuclear Fusion, demanding security and high-availability requirements call for redundancy to be available. ► ATCA based Nuclear Fusion Systems are composed by several electronic and mechanical component. ► Control and monitoring of ATCA electronic systems are recommended. ► ITER Fast Plant System Controller Project CODAC system prototype. ► EPICS device support module as External ATCA system manager solution. -- Abstract: This paper presents an Enhanced Physics and Industrial Control System (EPICS) device support module for the International Thermonuclear Experimental Reactor (ITER) Fast Plant System Controller (FPSC) project based in Advanced Telecommunications Computing Architecture (ATCA) specification. The developed EPICS device support module provides an External System Manager (ESM) solution for monitoring and control the ITER FPSC ATCA shelf system and data acquisition boards in order to take proper action and report problems to a control room operator or high level management unit in case of any system failure occurrence. EPICS device support module acts as a Channel Access (CA) server to report problems and publish ATCA system data information to the control room operator, high level management unit or other CA network clients such as Control System Studio Operator Interfaces (CSS OPIs), Best Ever Alarm System Toolkit (BEAST), Best Ever Archive Utility (BEAUTY) or other CA client applications. EPICS device support module communicates with the ATCA Shelf manager (ShM) using HTTP protocol to send and receive commands through POST method in order to get and set system and shelf components properties such as fan speeds measurements, temperatures readings, module status and ATCA boards acquisition and configuration parameters. All system properties, states, commands and parameters are available through the EPICS device support module CA server in EPICS Process Variables (PV) and signals format. ATCA ShM receives the HTTP protocol

  9. Separation of tritium from other hydrogen isotopes

    International Nuclear Information System (INIS)

    Roth, E.

    1988-01-01

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

  10. Safety and environmental aspects of deuterium--tritium fusion power plants: work shop summary

    International Nuclear Information System (INIS)

    1978-05-01

    In September of 1977 a workshop was held on the safety and environmental aspects of fusion power plants to consider potential safety and environmental problems of fusion power plants and to reveal solutions or methods of solving those problems. The objective was to promote incorporation of safety and environmental protection into reactor design, thereby reducing the expense and delay of backfitting safety systems after reactor designs are complete. A dialogue was established between fusion reactor designers and safety and environmental researchers. Four topics, each with several subdivisions, were selected for discussion: radiation exposure, accidents, environmental effects, and plant safety. For each topic, discussion focused on the significance of the problem, and adequacy of current technology to solve the problem, design solutions available and research needed to solve the problem

  11. 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

  12. Fusion Technology for ITER, the ITER Project. Further Development Towards a DEMO Fusion Power Plant (3/4)

    CERN Multimedia

    CERN. Geneva

    2011-01-01

    This is the second half of a lecture series on fusion and will concentrate on fusion technology. The early phase of fusion development was concentrated on physics. However, during the 1980s it was realized that if one wanted to enter the area of fusion reactor plasmas, even in an experimental machine, a significant advance in fusion technologies would be needed. After several conceptual studies of reactor class fusion devices in the 1980s the engineering design phase of ITER started in earnest during the 1990s. The design team was in the beginning confronted with many challenges in the fusion technology area as well as in physics for which no readily available solution existed and in a few cases it was thought that solutions may be impossible to find. However, after the initial 3 years of intensive design and R&D work in an international framework utilizing basic fusion technology R&D from the previous decade it became clear that for all problems a conceptual solution could be found and further devel...

  13. Tritium immobilisation

    International Nuclear Information System (INIS)

    Bridger, N.J.

    1982-01-01

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

  14. Using the WOFOST plant growth model to built the data base of the terrestrial tritium and food chain modules in RODOS

    International Nuclear Information System (INIS)

    Melintescu, A.; Galeriu, D.; Marica, A.

    2003-01-01

    The European Commission Project RODOS implements a coherent methodology for a Real-time On-line Decision Support System for Nuclear Emergency across Europe. Within this system there is a special module to model the transfer of tritiated water from releases to terrestrial foods. In order to model the transfer of tritiated water from air to various plants, the conversion to organically bound tritium, and the migration to edible parts of the plant, both the mean dynamics of leaf area index and a physiological description of canopy photosynthesis are required. The WOFOST crop growth model has been selected as a basis for deriving tritium transfer dynamics to plants. Its ability to reproduce site-specific biomass growth of various plants (not only from Europe) is demonstrated in this paper, as well as its compatibility to other photosynthesis models. We have tested that this model can simulate limited fertilization situations via the adaptation of two important parameters. After adaptation of model parameters to site-specific plant growth data, multi-annual mean dynamics can be obtained using meteorological data for subsequent years. (authors)

  15. Use of recovered cold nitrogen for tritium separation plant safety and environmental protection

    Energy Technology Data Exchange (ETDEWEB)

    Popescu, Gheorghe, E-mail: popescu.gheorghe@icsi.ro; Gherghinescu, Sorin, E-mail: sorin.gherghinescu@icsi.ro

    2013-10-15

    Highlights: • Cold exhausted gas from the liquid nitrogen storage tank used to speed up the regeneration stage, if molecular sieve adsorbers are further used. • Cold nitrogen use for safer plant through ITGR (atmosphere detritiation system) efficiency and availability upgrade of the system. • Easier decontamination in flushing the installation with water and significantly mitigate the risk of radiation for the operators. • Drying surfaces contaminated with liquid DTO, causing a decrease in the specific activity of the washing water used in decontamination. -- Abstract: This work started with the identification of some operating parameters deviations from the purification module of the Pilot Detritiation Plant, in confront with their design levels, namely: the tritiated heavy water vapor degree of retention in the deuterium flow was more than 5 ppm, less than the design value of max. 2 ppm; the cooling duration from 128 °C to the 20 °C was 20 h, more than the design value of max. 5 h. On meeting these challenges we came up with the proposal of cold nitrogen utilization to increase the adsorption specific capacity of the molecular sieve, on one hand, and to shorten its cooling time after regeneration, on the other hand. The theoretical model of this solution was made by using Langmuir theory and then its validation was done experimentally by cooling one of the two adsorbers with cold nitrogen vapors coming from the liquid nitrogen tank; in the second part of the experiment the adsorber was replaced with an experimental cryoadsorber. Further we present a number of other opportunities that this solution brings, such as: reducing the risk of fire/explosion in the nuclear detritiation plant, environmental protection by increasing the efficiency and availability of the atmospheric detritiation system, solving the problem of dismantling, transport and storing contaminated adsorbers in the event of radiological accident, easier decontamination in flushing the

  16. Tritium concentration monitor

    International Nuclear Information System (INIS)

    Shono, Kosuke.

    1991-01-01

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

  17. 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)

  18. Construction and commissioning of a hydrogen cryogenic distillation system for tritium recovery at ICIT Rm. Valcea

    Energy Technology Data Exchange (ETDEWEB)

    Ana, George, E-mail: george.ana@icsi.ro [Institute for Cryogenic and Isotopic Technologies, Rm. Valcea (Romania); Cristescu, Ion [Karlsruhe Istitute for Technologies, Tritium Laboratory, Eggenstein-Leopoldshaffen (Germany); Draghia, Mirela [ISTECH, Timisoara (Romania); Bucur, Ciprian; Balteanu, Ovidiu; Vijulie, Mihai; Popescu, Gheorghe; Costeanu, Claudiu; Sofilca, Nicolae; Stefan, Iulia; Daramus, Robert; Niculescu, Alina; Oubraham, Anisoara; Spiridon, Ionut; Vasut, Felicia; Moraru, Carmen; Brad, Sebastian [Institute for Cryogenic and Isotopic Technologies, Rm. Valcea (Romania); Pasca, Gheorghe [ISTECH, Timisoara (Romania)

    2016-05-15

    Highlights: • Cryogenic distillation (CD) process is being employed for tritium separation from tritiated hydrogen mixtures. • Process control and safety phylosophy with the detritiation plant from Rm. Vâlcea. • Tests undertaken prior to commissioning of the CD system from Rm. Vâlcea. • Preliminary experiments with the CD system (non-radiological). - Abstract: Cryogenic distillation (CD) of hydrogen in combination with Liquid Phase Catalytic Exchange (LPCE) or Combined Electrolytic Catalytic Exchange (CECE) process is used for tritium removal/recovery from tritiated water. Tritiated water is being obtained after long time operation of CANDU reactors, or in case of ITER mainly by the Detritiation System (DS). The cryogenic distillation system (CDS) used to remove/recover tritium from a hydrogen stream consists of a cascade of cryogenic distillation columns and a refrigeration unit which provides the cooling capacity for the condensers of CD columns. The columns, together with the condensers and the process heat-exchangers are accommodated in a vacuumed cold box. In the particularly case of the ICIT Plant, the cryogenic distillation cascade consists of four columns with diameters between 100–7 mm and it has been designed to process up to 10 mc/h of tritiated deuterium. This paper will present the steps undertaken for construction and commissioning of a pilot plant for tritium removal/recovery by cryogenic distillation of hydrogen. The paper will show besides preliminary data obtained during commissioning, also general characteristics of the plant and its equipments.

  19. Dynamic informational system for control and monitoring the tritium removal pilot plant with data transfer and process analyses

    International Nuclear Information System (INIS)

    Retevoi, Carmen Maria; Stefan, Iuliana; Balteanu, Ovidiu; Stefan, Liviu

    2005-01-01

    The dynamic informational system with datalogging and supervisory control module includes a motion control module and is a new conception used in tritium removal installation with isotopic exchange and cryogenic distillation. The control system includes an event-driven engine that maintains a real-time database, logs historical data, processes alarm information, and communicates with I/O devices. Also, it displays the operator interfaces and performs tasks that are defined for advanced control algorithms, supervisory control, analysis, and display with data transfer from data acquisition room to the control room. By using the parameters, we compute the deuterium and tritium concentration, respectively, of the liquid at the inlet of the isotopic exchange column and, consequently, we can compute at the outlet of the column, the tritium concentration in the water vapors. (authors)

  20. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  1. The movement of tritium in ecological systems

    International Nuclear Information System (INIS)

    Polevoy, Y; Laichter, Y.

    1988-11-01

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

  2. Tritium resources available for fusion reactors

    Science.gov (United States)

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

    2018-02-01

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

  3. Plasma wall interaction and tritium retention in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Amarescu, E.; Ascione, G.

    1996-01-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced an impressive number of record breaking results including core fusion power, ∼ 2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently, the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. The authors present recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues

  4. Plasma wall interaction and tritium retention in TFTR

    International Nuclear Information System (INIS)

    Skinner, C.H.; Amarescu, E.; Ascione, G.

    1997-01-01

    The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, ∝2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented. (orig.)

  5. Performance of a full-scale ITER metal hydride storage bed in comparison with requirements

    International Nuclear Information System (INIS)

    Beloglazov, S.; Glugla, M.; Fanghaenel, E.; Perevezentsev, A.; Wagner, R.

    2008-01-01

    The storage of hydrogen isotopes as metal hydride is the technique chosen for the ITER Tritium Plant Storage and Delivery System (SDS). A prototype storage bed of a full-scale has been designed, manufactured and intensively tested at the Tritium Laboratory, addressing main performance parameters specified for the ITER application. The main requirements for the hydrogen storage bed are a strict physical limitation of the tritium storage capacity (currently 70 g T 2 ), a high supply flow rate of hydrogen isotopes, in-situ calorimetry capabilities with an accuracy of 1 g and a fully tritium compatible design. The pressure composition isotherm of the ZrCo hydrogen system, as a reference material for ITER, is characterised by significant slope. As a result technical implementation of the ZrCo hydride bed in the SDS system requires further considerations. The paper presents the experience from the operation of ZrCo getter bed including loading/de-loading operation, calorimetric loop performance, and active gas cooling of the bed for fast absorption operation. The implications of hydride material characteristics on the SDS system configuration and design are discussed. (authors)

  6. Development of tritium handing technology(II)

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-02-01

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

  7. 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)

  8. Modeling tritium transport in the environment

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.

    1986-01-01

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

  9. Comparison of Tritium Component Failure Rate Data

    International Nuclear Information System (INIS)

    Lee C. Cadwallader

    2004-01-01

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

  10. Tritium retention in next step devices and the requirements for mitigation and removal techniques

    International Nuclear Information System (INIS)

    Counsell, G; Coad, P; Grisola, C; Hopf, C

    2006-01-01

    Mechanisms underlying the retention of fuel species in tokamaks with carbon plasma-facing components are presented, together with estimates for the corresponding retention of tritium in ITER. The consequential requirement for new and improved schemes to reduce the tritium inventory is highlighted and the results of ongoing studies into a range of techniques are presented, together with estimates of the tritium removal rate in ITER in each case. Finally, an approach involving the integration of many tritium removal techniques into the ITER operational schedule is proposed as a means to extend the period of operations before major intervention is required

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  13. Power converters for ITER

    CERN Document Server

    Benfatto, I

    2006-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium– tritium burning plasma operation. After a short description of ITER objectives, the main design parameters and the construction schedule, the paper describes the electrical characteristics of the French 400 kV grid at Cadarache: the European site proposed for ITER. Moreover, the paper describes the main requirements and features of the power converters designed for the ITER coil and additional heating power supplies, characterized by a total installed power of about 1.8 GVA, modular design with basic units up to 90 MVA continuous duty, dc currents up to 68 kA, and voltages from 1 kV to 1 MV dc.

  14. Tritium in surface water of the Yenisei river Basin

    International Nuclear Information System (INIS)

    Bondareva, L.G.; Bolsunovsky, A.Ya.

    2005-01-01

    The paper reports an investigation of the tritium content in the surface waters of the Yenisei River basin near the Mining-and-Chemical Combine (MCC). In 2001-2003 the maximum tritium concentration in the Yenisei River did not exceed 4±1 Bq/L. It has been found that there are surface waters containing enhanced tritium, up to 168 Bq/L, as compared with the background values for the Yenisei River. There are two possible sources of tritium input. First, the last operating reactor of the MCC, which still uses the Yenisei water as coolant. Second, tritium may come from the deep aquifers at the Severny testing site. For the first time tritium has been found in two aquatic plant species of the Yenisei River with maximal tritium concentration 304 Bq/Kg wet weight. Concentration factors of tritium for aquatic plants are much higher than 1

  15. Magmatic tritium

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  16. Universal tritium transmitter

    International Nuclear Information System (INIS)

    Cordaro, J. V.; Wood, M.

    2008-01-01

    sufficient time to thermally equilibrate. Amplifiers, transistors, resistors all need time to stabilize before the electrometer circuit will measure accurately in the 10 -15 and 10 -14 ampere range. Existing electrometers give the user no indication when the unit has stabilized and is acceptable for low level measurements. Savannah River National Laboratory (SRNL) funded through the NNSA Plant Directed Research and Development (PDRD) program, has developed a truly Universal Tritium Transmitter (UTT) capable of solving many known problems with existing commercial electrometers. This UTT pushes the state-of-the-art in electrometer design and incorporates solutions to deficiencies found in commercial electrometers. (authors)

  17. Parameter study on Japanese proposal of ITER hydrogen isotope separation system

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Enoeda, Mikio; Tanaka, Shigeru; Ohokawa, Yoshinao; Ohara, Atsushi; Nagakura, Masaaki; Naito, Taisei; Nagashima, Kazuhiro.

    1991-01-01

    As part of Japanese design contribution in the ITER activity, conceptual design of an entire ITER tritium system and their safety analysis have been carried out through the three-year period since 1988. The tritium system includes the following subsystems; - Fuelling (gas puffing and pellet injection) subsystem, - Torus vacuum pumping subsystem, - Plasma exhaust gas purification subsystem, - Hydrogen isotope separation subsystem, - NBI gas processing subsystem, - Blanket tritium recovery subsystem, - Tritiated water processing subsystem, - Tritium safety subsystem. Hydrogen isotope separation system is a key subsystem in the ITER tritium system because it is connected to all above subsystems. This report describes an analytical study on the Japanese concept of hydrogen isotope separation system. (author)

  18. Tritium processing using metal hydrides

    International Nuclear Information System (INIS)

    Mallett, M.W.

    1986-01-01

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

  19. 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)

  20. Biological effects of tritium

    International Nuclear Information System (INIS)

    Nieto, M.

    1985-01-01

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

  1. ITER EDA Newsletter. V. 4, no. 7

    International Nuclear Information System (INIS)

    1995-07-01

    This ITER EDA (Engineering Design Activities) Newsletter issue contains reports on (i) the 8th meeting of the ITER Technical Advisory Committee (TAC-8) held on June 29 - July 7, 1995 at the ITER San Diego Work Site, (ii) the 8th meeting of the ITER Management Advisory Committee (MAC-8) held at the ITER San Diego Work Site on July 9-10, 1995, (iii) the 33rd meeting of the International Fusion Research Council (FRC), held July 11, 1995 at the IAEA Headquarters in Vienna, Austria, and (iv) the ITER participation in the fifth topical meeting on Tritium Technology in Fission, Fusion and Isotopic Applications

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  3. 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

  4. Measurement of tritium concentration in urine

    International Nuclear Information System (INIS)

    Sekiyama, Shigenobu; Deshimaru, Takehide

    1979-01-01

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

  5. Tritium accountancy

    International Nuclear Information System (INIS)

    Avenhaus, R.; Spannagel, G.

    1995-01-01

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

  6. Tritium-labelled abscisic acid

    International Nuclear Information System (INIS)

    Pluciennik, H.; Michalski, L.

    1991-01-01

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

  7. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  8. 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

  9. ITER...ation

    International Nuclear Information System (INIS)

    Troyon, F.

    1997-01-01

    Recurrent attacks against ITER, the new generation of tokamak are a mix of political and scientific arguments. This short article draws a historical review of the European fusion program. This program has allowed to build and manage several installations in the aim of getting experimental results necessary to lead the program forwards. ITER will bring together a fusion reactor core with technologies such as materials, superconductive coils, heating devices and instrumentation in order to validate and delimit the operating range. ITER will be a logical and decisive step towards the use of controlled fusion. (A.C.)

  10. Turkey Point tritium. Progress report

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. Experimental validation of a method for performance monitoring of the impurity processing stage in the TEP system of ITER

    International Nuclear Information System (INIS)

    Bornschein, B.; Corneli, D.; Glugla, M.; Guenther, K.; Le, T.L.; Simon, K.H.

    2007-01-01

    The Tokamak Exhaust Processing (TEP) system within the Tritium Plant of ITER needs to be designed such that tritium is recovered from all exhaust gases produced during different modes and operational conditions of the vacuum vessel. The reference process for the TEP system of ITER is called CAPER and comprises three different, consecutive steps to recover hydrogen isotopes at highest purity for direct transfer to the cryogenic Isotope Separation system. The second step ('impurity processing', IP) is carried out in a closed loop involving heterogeneously catalyzed cracking or conversion reactions to liberate tritium from tritiated hydrocarbons or tritiated water combined with permeation of hydrogen isotopes through a Pd/Ag permeator. This combination shifts chemical equilibria towards dehydrogenation and, therefore, enables detritiation factors higher than 1000 in the IP stage. Such a high decontamination factor requires the optimal performance of the permeator, which on the other hand is operated under conditions which provoke coking of the permeator membrane by hydrocarbon cracking. For this reason the permeator in the impurity processing loop needs to be repeatedly regenerated in order to sustain decontamination factors higher/in the order of 1000. At the Tritium Laboratory Karlsruhe (TLK) a method to measure the actual performance of the second stage of the CAPER process has been developed. This method has been successfully tested with the CAPER facility and appears feasible for the TEP system of ITER

  12. Interaction among competitive producers in the electricity market: An iterative market model for the strategic management of thermal power plants

    International Nuclear Information System (INIS)

    Carraretto, Cristian; Zigante, Andrea

    2006-01-01

    The liberalization of the electricity sector requires utilities to develop sound operation strategies for their power plants. In this paper, attention is focused on the problem of optimizing the management of the thermal power plants belonging to a strategic producer that competes with other strategic companies and a set of smaller non-strategic ones in the day-ahead market. The market model suggested here determines an equilibrium condition over the selected period of analysis, in which no producer can increase profits by changing its supply offers given all rivals' bids. Power plants technical and operating constraints are considered. An iterative procedure, based on the dynamic programming, is used to find the optimum production plans of each producer. Some combinations of power plants and number of producers are analyzed, to simulate for instance the decommissioning of old expensive power plants, the installation of new more efficient capacity, the severance of large dominant producers into smaller utilities, the access of new producers to the market. Their effect on power plants management, market equilibrium, electricity quantities traded and prices is discussed. (author)

  13. 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.)

  14. Recent environmental tritium levels in Japan

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  15. STAR facility tritium accountancy

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  16. 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

  17. Tritium inventory measurements by 'in-bed' gas flowing calorimetry

    International Nuclear Information System (INIS)

    Hayashi, T.; Suzuki, T.; Yamada, M.; Okuno, K.

    1996-01-01

    In order to establish the 'in-bed' tritium accounting technology for the ITER scale tritium storage system, a gas flowing calorimetry has been studied using a scaled ZrCo bed (25 g tritium capacity). The basic calorimetric characteristics, steady state temperature raise of He gas stream flowing through a secondary coil line fixed in the ZrCo tritide, was measured and correlated with the stored tritium inventory. The results shows that about 4 degrees raise of He stream temperature can be detected for each gram of tritium storage. The sensitivity of this calorimetry is about 0.05 g of tritium, calculated by 0.2 degrees of temperature sensor error. The accuracy is better than 0.25 g of tritium on 25 g storage, evaluated by 2 times of standard deviation from the repeat measurements. This accuracy of < 1% on full storage capacity is satisfied the target accountability to measure ± 1 gram of tritium on 100 g storage for ITER. 13 refs., 7 figs

  18. 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

  19. Oxidative Tritium Decontamination System

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  20. 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.

  1. 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.

  2. Study of the comparative dynamics of the incorporation of tissue free-water tritium (TFWT) in bulrushes (Typha latifolia) and carp (Cyprinus carpio) in the Almaraz nuclear power plant cooling reservoir

    Energy Technology Data Exchange (ETDEWEB)

    Baeza, A. [Department of Applied Physics, Faculty of Veterinary, University of Extremadura, Avda de la Universidad s/n, 10071 Caceres (Spain)], E-mail: ymiralle@unex.es; Garcia, E. [Department of Applied Physics, Technical Forest Engineering School, University of Extremadura, 10600 (Plasencia) Caceres (Spain); Paniagua, J.M. [Department of Applied Physics, Polytechnic School, University of Extremadura, Avda de la Universidad s/n, 10071 Caceres (Spain); Rodriguez, A. [Department of Applied Physics, Faculty of Veterinary, University of Extremadura, Avda de la Universidad s/n, 10071 Caceres (Spain)

    2009-03-15

    The Almaraz nuclear power plant (Spain) uses the water of Arrocampo reservoir for cooling, and consequently raises the radioactive levels of the aquatic ecosystem of this reservoir. From July 2002 to June 2005, monthly samples of surface water, bulrushes (Typha latifolia) and carp (Cyprinus carpio) were collected from this reservoir. They were analyzed to determine the temporal evolution of the levels of {sup 3}H in surface water and of its transfer from the surface water to free-water in the tissues (TFWT) of the aforementioned two organisms. The tritium levels in the surface water oscillate with a biannual period, with their values in the study period ranging between 53 and 433 Bq/L. The incorporation of tritium to bulrushes and carp was fairly similar, the respective mean concentration factors being 0.74 and 0.8 (unitless, as Bq/L tissue water per Bq/L reservoir water). The temporal evolution of the levels fairly closely followed that observed for the surface water tritium, although detailed analysis showed the dominant periodicity for the bulrushes to be annual. This difference reflects the influence on the incorporation of tritium to bulrushes of diverse environmental and metabolic factors, especially evapotranspiration and the seasonal growth of this plant.

  3. Study of the comparative dynamics of the incorporation of tissue free-water tritium (TFWT) in bulrushes (Typha latifolia) and carp (Cyprinus carpio) in the Almaraz nuclear power plant cooling reservoir

    International Nuclear Information System (INIS)

    Baeza, A.; Garcia, E.; Paniagua, J.M.; Rodriguez, A.

    2009-01-01

    The Almaraz nuclear power plant (Spain) uses the water of Arrocampo reservoir for cooling, and consequently raises the radioactive levels of the aquatic ecosystem of this reservoir. From July 2002 to June 2005, monthly samples of surface water, bulrushes (Typha latifolia) and carp (Cyprinus carpio) were collected from this reservoir. They were analyzed to determine the temporal evolution of the levels of 3 H in surface water and of its transfer from the surface water to free-water in the tissues (TFWT) of the aforementioned two organisms. The tritium levels in the surface water oscillate with a biannual period, with their values in the study period ranging between 53 and 433 Bq/L. The incorporation of tritium to bulrushes and carp was fairly similar, the respective mean concentration factors being 0.74 and 0.8 (unitless, as Bq/L tissue water per Bq/L reservoir water). The temporal evolution of the levels fairly closely followed that observed for the surface water tritium, although detailed analysis showed the dominant periodicity for the bulrushes to be annual. This difference reflects the influence on the incorporation of tritium to bulrushes of diverse environmental and metabolic factors, especially evapotranspiration and the seasonal growth of this plant

  4. Environmental monitoring for tritium separation facility

    International Nuclear Information System (INIS)

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

    2001-01-01

    The Cryogenic Pilot is an experimental project within the nuclear energy national research program, which has the aim of developing technologies for tritium and deuterium separation by cryogenic distillation. The experimental installation is located 15 km near the highest city of the area and 1 km near Olt River. An important chemical activity is developed in the area and almost all the neighbors of the Experimental Cryogenic Pilot are chemical plants. It is necessary to emphasize this aspect because the sewage system is connected with the other tree chemical plants from the neighborhood. In this work, a low background liquid scintillation is used to determine tritium activity concentration according to ISO 9698/1998. We measured drinking water, precipitation, river water, underground and waste water. The tritium level was between 10 TU and 27 TU what indicates that there is no sources of tritium contamination in the neighborhood of Cryogenic Pilot. In order to determine baseline levels we decided to monitor monthly each location. In this paper it is presented the standard method used for tritium determination in water samples, the precautions needed in order to achieve reliable results, and the evolution of tritium level in different location near the Tritium and Deuterium Cryogenic Separation Experimental Pilot. (authors)

  5. Tritium Decay Helium-3 Effects in Tungsten

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-01

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

  6. The Tritium White Paper

    International Nuclear Information System (INIS)

    2009-01-01

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

  7. Tritium Accounting Stability of a ZrCo Bed with 'In-Bed' Gas Flowing Calorimetry

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Suzuki, Takumi; Yamada, Masayuki; Nishi, Masataka

    2005-01-01

    Zirconium-Cobalt (ZrCo) tritium storage bed with 'in-bed' gas flowing calorimetry has been developed as a self-assaying system for the Tritium Storage and Delivery System of ITER. The basic tritium accounting characteristics have been investigated and practical data on the accounting stability has been accumulated under gram level tritium storage for more than 8 years. The initial sensitivity of tritium was about 0.05 g and the accuracy (standard deviation of repeat measurements: two sigma) was about 0.15 g at full tritium storage of 25 g. This initial accounting performance has been maintained after tritium storage for more than 8 years by keeping constant accounting conditions at each inventory measurement. Almost no aging effect of tritium was found except accumulation of 3 He in the primary ZrCo tritide vessel, which was easily evacuated to keep initial accounting condition before each inventory measurement

  8. 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.)

  9. 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

  10. Review of general tritium accountancy techniques

    International Nuclear Information System (INIS)

    Vassallo, G.; Engelmann, U.

    1995-01-01

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

  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. Photoproduction of tritium

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  13. Investigation of tritium in the aquatic environment

    International Nuclear Information System (INIS)

    Cohen, L.K.

    1977-01-01

    The behavior, cycling and distribution of tritium in an aquatic ecosystem was studied in the field and in the laboratory from 1969 through 1971. Field studies were conducted in the Hudson River Estuary, encompassing a 30 mile region centered about the Indian Point Nuclear Plant. Samples of water, bottom sediment, rooted emergent aquatic plants, fish, and precipitation were collected over a year and a half period from more than 15 locations. Specialized equipment and systems were built to combust and freeze-dry aquatic media to remove and recover the loose water and convert the bound tritium into an aqueous form. An electrolysis system was set up to enrich the tritium concentrations in the aqueous samples to improve the analytical sensitivity. Liquid scintillation techniques were refined to measure the tritium activity in the samples. Over 300 samples were analyzed during the course of the study

  14. FDMH - The tritium model in RODOS

    International Nuclear Information System (INIS)

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

    2000-01-01

    Under the auspices of its RTD (Research and Technological Development) Framework Programmes, the European Commission has supported the development of the RODOS (Real-time On-line DecisiOn Support) system for off-site emergency management. The project started in 1989 focusing on PWR/LWR type accidents and using experience from the Chernobyl accident. In 1996 it was realised that tritium should be included in the list of radionuclides, as large tritium sources exist in Europe and to allow a potential expansion of the RODOS system for application on future fusion reactor accidents. The National Institute for Physics and Nuclear Engineering (IFIN-HH) in Romania - in close co-operation with the Research Centre Karlsruhe (FZK) - was charged to develop the tritium module, based on previous experience in environmental tritium modelling and the operation of CANDU reactor-based NPP in Romania (with potential tritium accidents). Tritium, being an isotope of hydrogen, is incorporated immediately in the life cycle and its transport into the biosphere differs considerably from other radionuclides treated by the RODOS system. Concentrations in the individual compartments may change very rapidly (hours) under varying environmental conditions and conversion to organic forms by biochemical and metabolic processes takes place in plants and animals. Consequently, the tritium code in RODOS was developed as a separate module and harmonisation in data sets and interfaces with other food chain modules integrated in RODOS was ensured. Presently, the tritium module - FDMH- is integrated and documented in the RODOS system, delivering time dependent tritium concentration (as tritiated water or organically bound tritium) in plant and animal products, inhalation dose and ingestion dose for various groups of population, after an accident emitting tritiated water and for up to 2520 locations around the source. FDMH incorporates many improved techniques in radiological assessment and makes

  15. 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.)

  16. Development of compact tritium confinement system using gas separation membrane

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Okuno, Kenji

    1994-01-01

    In order to develop more compact and cost-effective tritium confinement system for fusion reactor, a new system using gas separation membranes has been studied at the Tritium Process Laboratory in the Japan Atomic Energy Research Institute. The preliminary result showed that the gas separation membrane system could reduce processing volume of tritium contaminated gas to more than one order of magnitude compared with the conventional system, and that most of tritiated water vapor (humidity) could be directly recovered by water condenser before passing through dryer such as molecular sieves. More detail investigations of gas separation characteristics of membrane were started to design ITER Atmospheric Detritiation System (ADS). Furthermore, a scaled polyimide membrane module (hollow-filament type) loop was just installed to investigate the actual tritium confinement performance under various ITER-ADS conditions. (author)

  17. 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.

  18. Tritiated hydrogen gas storage systems for a fusion plant

    International Nuclear Information System (INIS)

    Bramy, W.; Hircq, B.; Peyrat, M.; Leger, D.

    1992-01-01

    This paper reports that USSI INGENIERIE has carried out a study financed by European Communities Commission concerning the NET/ITER project, on tritium Fuel Management and Storage systems of the International Thermonuclear Experimental Reactor. A processing block diagram for hydrogen isotopes represents all interfaces and possible links between these systems and tritiated gas mixtures flowing through the Fusion plant. Large quantities of hydrogen isotopes (up to several thousand moles of protium, deuterium and tritium) in gaseous form associated with torus fuelling and exhaust pellet injection, and neutral beam injection, must be stored and managed in such a plant

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  20. The tritium systems test assembly: Overview and recent results

    International Nuclear Information System (INIS)

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

    1988-01-01

    The fusion technology development program for tritium in the US is centered around the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The TSTA is a full-scale system of reactor exhaust gas reprocessing for an ITER-sized machine. That is, TSTA has the capacity to process tritium in a closed loop mode at the rate of 1 kg per day, requiring a tritium inventory of about 100 g. The TSTA program also interacts with all other tritium-related fusion technology programs in the US and all major programs abroad. This report summarizes the current status, results and interactions of the TSTA. Special emphasis is given to operations in May/June using large compound cryopumps that completed the fuel loop integration of all TSTA subsystems for the first time. 6 refs., 2 figs

  1. Tritium pellet injector results

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  2. ITER: the Sun rises over nuclear fusion with West

    International Nuclear Information System (INIS)

    Sacco, Laurent

    2013-01-01

    The ITER project is considered as a critical step on the way to commercial production of electricity by a thermonuclear reactor based on controlled fusion. This project notably requires the development of a divertor which is the objective of the West project which will use the famous Cadarache superconductive magnet reactor, Tore Supra. After having outlined the future lack of fossil energies at the world scale, presented the operation principles of tokamaks and recalled some results obtained in their development, this article justifies the use of superconductive magnets. It presents the ITER project as a step in the production of thermonuclear electricity. ITER will be in fact a proof that such plants can be realised, and it should be followed by Demo, a demonstration power plant, by 2050. The article presents the West project, a test bench for ITER, which introduced modifications in the Tore Supra reactor to create conditions almost similar to that existing at the surface of the Sun. It notably comprises a divertor made of tungsten for the fusion with tritium. It finally outlines that the fusion will be a hot one, not a cold one

  3. Development and test of prototype components for ITER; Entwicklung und Test von Prototypkomponenten fuer ITER

    Energy Technology Data Exchange (ETDEWEB)

    Biel, Wolfgang; Behr, Wilfried; Castano-Bardawil, David; and others

    2015-08-15

    The scientific program of the project is divided into the following partial projects: (1.) ITER Diagnostic Port Plug for the charge-exchange spectroscopy (CXRS) with the subthemes: (a) Development of prototypes for critical mechanical components, (b) development of a roboter for the laser welding of vacuum seals and pipings at the Port Plug, (c) mirror studies, (d) CXRS prototype spectrometer, (2.) ITER tritium retention diagnostics (TR), (3.) ITER disruption mitigation ventile (DMV).

  4. A system dynamics model for tritium cycle of pulsed fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  5. A system dynamics model for tritium cycle of pulsed fusion reactor

    International Nuclear Information System (INIS)

    Zhu, Zuolong; Nie, Baojie; Chen, Dehong

    2017-01-01

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

  6. Tritium inventory and recovery in next-step fusion devices

    International Nuclear Information System (INIS)

    Causey, R.A.; Brooks, J.N.; Federici, G.

    2002-01-01

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

  7. CFTSIM-ITER dynamic fuel cycle model

    International Nuclear Information System (INIS)

    Busigin, A.; Gierszewski, P.

    1998-01-01

    Dynamic system models have been developed for specific tritium systems with considerable detail and for integrated fuel cycles with lesser detail (e.g. D. Holland, B. Merrill, Analysis of tritium migration and deposition in fusion reactor systems, Proceedings of the Ninth Symposium Eng. Problems of Fusion Research (1981); M.A. Abdou, E. Vold, C. Gung, M. Youssef, K. Shin, DT fuel self-sufficiency in fusion reactors, Fusion Technol. (1986); G. Spannagel, P. Gierszewski, Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket, Fusion Eng. Des. (1991); W. Kuan, M.A. Abdou, R.S. Willms, Dynamic simulation of a proposed ITER tritium processing system, Fusion Technol. (1995)). In order to provide a tool to understand and optimize the behavior of the ITER fuel cycle, a dynamic fuel cycle model called CFTSIM is under development. The CFTSIM code incorporates more detailed ITER models, specifically for the important isotope separation system, and also has an easier-to-use graphical interface. This paper provides an overview of CFTSIM Version 1.0. The models included are those with significant and varying tritium inventories over a test campaign: fueling, plasma and first wall, pumping, fuel cleanup, isotope separation and storage. An illustration of the results is shown. (orig.)

  8. A metabolic derivation of tritium transfer factors in animal products

    International Nuclear Information System (INIS)

    Galeriu, D.; Melintescu, A.; Crout, N. M. J.; Bersford, N. A.; Peterson, S. R.; Hess, M. van

    2001-01-01

    Tritium is a potentially important environmental contaminant arising from the nuclear industry. Because tritium is an isotope of hydrogen, its behaviour in the environment is controlled by the behaviour of hydrogen. Chronic releases of tritium to the atmosphere, in particular, will result in tritium-to-hydrogen (T/H) ratios in plants and animals that are more or less in equilibrium with T/H ratios in the air moisture. Tritium is thus a potentially important contaminant of plant and animal food products. The transfer of tritium from air moisture to plants is quite well understood. In contrast, although a number of regulatory agencies have published transfer coefficient values for diet tritium transfer for a limited number of animal products, a fresh evaluation of these transfers needs to be made In this paper we present an approach for the derivation of tritium transfer coefficients which is based on the metabolism of hydrogen in animals in conjunction with experimental data on tritium transfer. The derived transfer coefficients separately account for transfer to and from free (i.e. water) and organically bound tritium. The predicted transfer coefficients are compared to available data independent of model development. Agreement is good, with the exception of the transfer coefficient for transfer from tritiated water to organically bound tritium in ruminants, which may be attributable to the particular characteristics of ruminant digestion. We show that transfer coefficients will vary in response to the metabolic status of an animal (e.g. stage of lactation, digestibility of diet, etc.) and that the use of a single transfer coefficient from diet to animal product is not appropriate for tritium. It is possible to derive concentration ratio values which relate the concentration of tritiated water and organically bound tritium in an animal product to the corresponding concentrations in the animals diet. These concentration ratios are shown to be less subject to

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

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Nishi, Masataka

    2003-11-01

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

  10. PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Smith, P.; Sheetz, S.

    2013-09-30

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

  11. 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)

  12. 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

  13. Tritium sampling and measurement

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Morgan, L.W.G.; Packer, L.W.

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, L.W.G., E-mail: Lee.Morgan@CCFE.ac.uk; Packer, L.W.

    2014-10-15

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

  16. Development of tritium-handling technique

    International Nuclear Information System (INIS)

    Ohmura, Hiroshi; Hosaka, Akio; Okamoto, Takahumi

    1988-01-01

    The overview of developing activities for tritium-handling techniques in IHI are presented. To establish a fusion power plant, tritium handling is one of the key technologies. Recently in JAERI, conceptual design of FER (Fusion Experimental Reactor) has been carried out, and the FER system requires a processing system for a large amount of tritium. IHI concentrate on investigation of fuel gas purification, isotope separation and storage systems under contract with Toshiba Corporation. Design results of the systems and each components are reviewed. IHI has been developing fundamental handling techniques which are the ZrNi bed for hydrogen isotope storage and isotope separation by laser. The ZrNi bed with a tritium storage capacity of 1000 Ci has been constructed and recovery capability of the hydrogen isotope until 10 -4 Torr {0.013 Pa} was confirmed. In laser isotope separation, the optimum laser wave length has been determined. (author)

  17. Tritium Measurements in Slovenia - Chronology Till 2004

    International Nuclear Information System (INIS)

    Logar, Jasmina Kozar; Vaupotic, Janja; Kobal, Ivan

    2005-01-01

    Almost all the analyses of tritium in Slovenia have been performed by the tritium laboratory at the Jozef Stefan Institute. Nearly 90 % of its measurements have been covered by two national programs, both approved by the Slovenian Nuclear Safety Administration: the radioactive monitoring program in the environs of Krsko Nuclear Power Plant (KNPP) and the program of global radioactive contamination monitoring in the environment. These programs include samples of groundwaters, surface waters, precipitation and drinking waters, as well as liquid and gaseous effluents from KNPP. Tritium was determined in some research projects and in hydrological studies of thermal waters, groundwater and coalmine waters. Tritium in the Karst region was mapped as well as the springs of entire territory of Slovenia. Around 5500 samples have been analyzed up to 2004

  18. Technologies for immobilization and disposal of tritium

    International Nuclear Information System (INIS)

    Coppari, N.R.

    1996-01-01

    This study was done within a program one of whose objectives was to know the state of the technology development for tritium separation in the moderator circuit at HWR and to define the possible technologies to be applied to the Argentine nuclear power plants. Within this framework the strategies adopted by each country and the available technologies for a safe disposal of tritium, not only in its gaseous state tritium but also as tritiated water were analyzed. It is considered that if the selected separation method is such that the tritium is in its gaseous state, the hydride formation for long periods of immobilization should be studied. whereas if it were triated water immobilization should be studied to choose the technology between cementation and drying agents, in both cases the final disposal site will have to be selected. (author). 8 refs

  19. Separation of tritium from aqueous effluents

    International Nuclear Information System (INIS)

    Bruggeman, A.; Leysen, R.; Meynendonckx, L.; Parmentier, C.; Bellien, H.; Smets, D.; Stevens, J.

    1984-01-01

    This report describes the further development of the so-called ELEX process, carried out from 1 July 1980 until 31 December 1982. The ELEX process is the combination of electrolysis with the catalytic tritium exchange between hydrogen and water in order to accumulate the tritium in the liquid phase. The experimental study of the catalytic tritium exchange between hydrogen and liquid water was continued and the overall exchange rate could be substantially increased. An alternative process based on bithermal exchange of tritium has been evaluated. In the 10 mol h -1 mini-pilot bench scale detritiation unit the ELEX process was successfully demonstrated by detritiating up to now more than 1m 3 of water containing up to 100 mCi tritium per dm 3 , which is the feed concentration to be expected for application of the process in a reprocessing plant. A 280 mol h -1 pilot detritiation installation now being constructed is described. This installation will realize a volume reduction factor of 100 and a process decontamination factor of 100. The maximum total tritium inventory will be about 1000 Ci. The plant consists mainly of a 80 kW electrolyser and a 10 cm diameter exchange column and can be considered as the ultimate step before industrial application of the ELEX process

  20. Study of wall conditioning in tokamaks with application to ITER

    International Nuclear Information System (INIS)

    Kogut, Dmitri

    2014-01-01

    Thesis is devoted to studies of performance and efficiency of wall conditioning techniques in fusion reactors, such as ITER. Conditioning is necessary to control the state of the surface of plasma facing components to ensure plasma initiation and performance. Conditioning and operation of the JET tokamak with ITER-relevant material mix is extensively studied. A 2D model of glow conditioning discharges is developed and validated; it predicts reasonably uniform discharges in ITER. In the nuclear phase of ITER operation conditioning will be needed to control tritium inventory. It is shown here that isotopic exchange is an efficient mean to eliminate tritium from the walls by replacing it with deuterium. Extrapolations for tritium removal are comparable with expected retention per a nominal plasma pulse in ITER. A 1D model of hydrogen isotopic exchange in beryllium is developed and validated. It shows that fluence and temperature of the surface influence efficiency of the isotopic exchange. (author) [fr

  1. 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.

  2. 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

  3. Environmental tritium in trees

    International Nuclear Information System (INIS)

    Brown, R.M.

    1979-01-01

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

  4. Tritium monitoring techniques

    International Nuclear Information System (INIS)

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

    1996-05-01

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

  5. Preliminary RAMI analysis of DFLL TBS for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dagui [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, 230031 (China); Yuan, Run [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Jiaqun, E-mail: jiaqun.wang@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Fang; Wang, Jin [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2016-11-15

    Highlights: • We performed the functional analysis of the DFLL TBS. • We performed a failure mode analysis of the DFLL TBS. • We estimated the reliability and availability of the DFLL TBS. • The ITER RAMI approach was applied to the DFLL TBS for technical risk control in the design phase. - Abstract: ITER is the first fusion machine fully designed to prove the physics and technological basis for next fusion power plants. Among the main technical objectives of ITER is to test and validate design concepts of tritium breeding blankets relevant to the fusion power plants. To achieve this goal, China has proposed the dual functional lithium-lead test blanket module (DFLL TBM) concept design. The DFLL TBM and its associated ancillary system were called DFLL TBS. The DFLL TBS play a key role in next fusion reactor. In order to ensure reliable and available of DFLL TBS, the risk control project of DFLL TBS has been put on the schedule. As the stage of the ITER technical risk control policy, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was used to control the technical risk of ITER. In this paper, the RAMI approach was performed on the conceptual design of DFLL TBS. A functional breakdown was prepared on DFLL TBS, and the system was divided into 3 main functions and 72 basic functions. Based on the result of functional breakdown of DFLL TBS, the reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the DFLL TBS expected after implementation of mitigation actions was calculated to be 98.57% over 2 years based on the ITER reliability database. A Failure Modes Effects and Criticality Analysis (FMECA) was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.

  6. 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

  7. Tritium in metals

    International Nuclear Information System (INIS)

    Schober, T.

    1990-01-01

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

  8. Tritium sources; Izvori tricijuma

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

  9. High-pressure tritium

    International Nuclear Information System (INIS)

    Coffin, D.O.

    1976-01-01

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

  10. Tritium forms discrimination in ryegrass under constant tritium exposure: From seed germination to seedling autotrophy.

    Science.gov (United States)

    Renard, H; Maro, D; Le Dizès, S; Escobar-Gutiérrez, A; Voiseux, C; Solier, L; Hébert, D; Rozet, M; Cossonnet, C; Barillot, R

    2017-10-01

    Uncertainties remain regarding the fate of atmospheric tritium after it has been assimilated in grasslands (ryegrass) in the form of TFWT (Tissue Free Water Tritium) or OBT (Organically Bound Tritium). One such uncertainty relates to the tritium forms discrimination during transfer from TFWT to OBT resulting from photosynthesis (OBT photo ), corresponding to the OBT photo /TFWT ratio. In this study, the OBT/TFWT ratio is determined by experiments in the laboratory using a ryegrass model and hydroponic cultures, with constant activity of tritium in the form of tritiated water (denoted as HTO) in the "water" compartment (liquid HTO) and "air" compartment (HTO vapour in the air). The OBT photo /TFWT ratio and the exchangeable OBT fraction are measured for three parts of the plant: the leaf, seed and root. Plant growth is modelled using dehydrated biomass measurements taken over time in the laboratory and integrating physiological functions of the plant during the first ten days after germination. The results suggest that there is no measurable discrimination of tritium in the plant organic matter produced by photosynthesis. Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Tritium kinetics in a freshwater marsh ecosystem

    International Nuclear Information System (INIS)

    Adams, L.W.

    1976-01-01

    Ten curies of tritium (as tritiated water, HTO) were applied to a 2-ha enclosed Lake Erie marsh in northwestern Ohio on 29 October 1973. Tritium kinetics in the marsh water, bottom sediment, and selected aquatic plants and animals were determined. Following HTO application, peak tritium levels in the sediment were observed on day 13 in the top 1-cm layer, on day 27 at the 5-cm depth, and on day 64 at the 10-cm depth. Peak levels at 15 and 20 cm were not discernible, although there was some movement of HTO to the 20-cm depth. A model based on diffusion theory described tritium movement through the sediment. Unbound and bound tritium levels in curly-leaf pondweed (Potamogeton crispus), pickerelweed (Pontederia cordata), and smartweed (Polygonum lapathifolium) generally tended to follow tritium levels in marsh water. The unbound tritium:marsh water tritium ratio was significantly larger (P < 0.001) in curly-leaf pondweed than in either of the two emergents. Tritium uptake into the unbound compartments of crayfish (Procambarus blandingi), carp (Cyprinus carpio), and bluegills (Lepomis macrochirus) was rapid. For crayfish, maximum HTO levels were observed on days 3 and 2 for viscera and muscle, respectively. Unbound HTO in carp viscera peaked on day 2, and levels in carp muscle reached a maximum in 4 hours. Maximum levels of unbound HTO in bluegill viscera and muscle were observed on day 1. After peak levels were obtained, unbound HTO paralleled marsh water HTO activity in all species. Tritium uptake into the bound compartments was not as rapid nor were the levels as high as for unbound HTO in any of the species. Peak bound levels in crayfish viscera were observed on day 20 and maximum levels in muscle were noted on day 10. Bound tritium in carp viscera and muscle reached maximum levels on day 20. In bluegills, peaks were reached on days 7 and 5 for viscera and muscle, respectively. Bound tritium in all species decreased following maximum levels

  12. Development of effective tool for iterative design of human machine interfaces in nuclear power plant

    International Nuclear Information System (INIS)

    Nakagawa, Takashi; Matsuo, Satoko; Yoshikawa, Hidekazu; Wu, Wei; Kameda, Akiyuki; Fumizawa, Motoo

    2000-01-01

    The authors have developed SEAMAID, which is a Simulation-based Evaluation and Analysis support system for MAn-machine Interface Design (SEAMAID) in the domain of nuclear power plants. The SEAMAID simulated the interaction between an operator and human machine interfaces (HMI), and supports to evaluate the HMI by using the simulation results. In this paper, a case study of evaluation for conventional center control room design was conducted. The authors were confirmed that SEAMAID is a useful tool for improvements of HMI design (J.P.N.)

  13. Tritium management in fusion reactors

    International Nuclear Information System (INIS)

    Galloway, T.R.

    1978-05-01

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

  14. 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

  15. Tritium in the environment. The IRSN's opinion on key issues and on research and development perspectives

    International Nuclear Information System (INIS)

    2010-01-01

    This report states the opinion of the IRSN on issues related to the behaviour of tritium in the environment, and to the associated risks. This report is based on a set of studies and researches performed on this radionuclide. Thus, the authors address the status of knowledge on the evolution of tritium released by nuclear activities (measurement techniques), the risk of bioaccumulation of tritium by living organisms within ecosystems (behaviour of tritium in the atmosphere, in soils, in ground plants, in continental and sea aquatic media), and the knowledge of risks due to tritium absorbed by living organisms (dose assessment, knowledge of tritium harmful effects and relative biological effectiveness)

  16. Accounting control of tritium at the tritium process laboratory (TPL) of JAERI. Results of 15-year operation and research activity

    International Nuclear Information System (INIS)

    Nishi, Masataka; Yamanishi, Toshihiko; Hayashi, Takumi; Yamada, Masayuki; Suzuki, Takumi

    2003-01-01

    Research and development work of fuel processing technology and tritium safe-handling technology necessary for fusion reactors has been performed at the Tritium Process Laboratory (TPL) of JAERI. TPL is the first facility in Japan permitted to handle tritium of more than 1g (about 0.36PBq), and its operation itself is also important for the development of fusion reactor facility in the viewpoint of tritium control. Various experiments have been carried out at TPL safely since 1988 controlling 22PBq of tritium as the maximum observing regulations. In addition to the regulatory accounting and control, detailed independent control in TPL was planned and was established through its 15-year safe-operation. For future fusion fuel facility where kilo-grams of tritium will be handled, method of tritium accounting has been researched and some new technologies have been developed at TPL. Results of TPL operation and of the research activity in it contributed the completion of the engineering design of ITER. Further research activity on tritium accounting and control is in progress in TPL for the future fusion reactors. (author)

  17. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Science.gov (United States)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-08-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  18. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    International Nuclear Information System (INIS)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

    2014-01-01

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed

  19. Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

    2014-08-21

    In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

  20. FDNH - the tritium module in RODOS

    International Nuclear Information System (INIS)

    Galeriu, D.; Melintescu, A.; Turcanu, C. O.; Raskob, W.

    2001-01-01

    soil-plant-atmosphere system, plant physiology, photosynthesis, growth and hydrogen metabolism in mammals. A unique feature of FDMH is the coherent modelling of tritium uptake by plant canopies and its conversion to organic matter, using a physiological plant parameter data base which can reproduce plant growth under various pedo-climatic conditions. By this approach, the difficulties of scaling from leaf to canopy are avoided and the model parameters are tested by concomitant reproduction of plant growth, using an appropriate crop growth model- developed at process level. In order to predict the tritium transfer in animal products in the absence of a complete experimental database, results from basic research on hydrogen metabolism in mammals is applied. Both forms of tritium are considered and the transfer and the conversion from tritiated water (HTO) or organics (OBT) in feed to HTO and OBT in animal products are explicitly introduced. Incorporating the environmental tritium dynamics with time steps ranging from less than one hour up to days, FDMH illustrates seasonal and diurnal effects on public dose related to the time of the accident. Due to the novel modelling approach, FDMH can be easily customised for any European site and can predict the time evolution of tritiated water or organically bound tritium in such details that it can be easily used in establishing countermeasures. The present model as integrated in the RODOS platform contains a database for Central Europe but it is not directly coupled to real-time weather prognosis data, due to external constraints. In order to increase the model flexibility and reliability some upgrades are now on going and an international, stand-alone version is in preparation. (authors)

  1. JET experience on managing radioactive waste and implications for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Reynolds, Stephen, E-mail: Stephen.reynolds@ccfe.ac.uk [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE/Power and Active Operations Department, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Newman, Mark; Coombs, Dave; Witts, David [EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); CCFE/Power and Active Operations Department, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2016-11-01

    construction include a water de-tritiation facility and a materials de-tritiation facility, both of which are relevant for ITER. The procedures used to manage radioactive waste from generation to off-site disposal have been assessed for relevance to ITER and a number have been shown to be significant. The procedures and de-tritiation factors demonstrated by radioactive waste treatment plants currently under construction will be important to tritium recovery and waste minimisation in ITER and DEMO.

  2. Use of tritium in the production of selfluminous compounds

    International Nuclear Information System (INIS)

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

    1994-12-01

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

  3. Considerations for tritium protection at a fusion reactor

    International Nuclear Information System (INIS)

    Easterly, C.E.

    1981-01-01

    The purpose of this paper is to indicate the general direction of current fusion reactor concepts regarding tritium, and to indicate some options in tritium control strategies. Certain strategies, in addition to providing reduced potential health hazard, afford the potential for engineering alternatives for in-plant tritium control systems. The overall coupling of containment cleanup systems and health protection must continue to develop with increased knowledge of the health effects of different tritium species and the consequent systems options available subsequent to this understanding

  4. The APT program plan: Providing an assured tritium production capability

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Anderson, J.L.; Bishop, W.P.; Boggs, B.; Hall, K.

    1996-01-01

    Tritium is a radioactive hydrogen isotope used in all U.S. nuclear weapons. Because the half-life of tritium is short, 12.3 yr, it must be periodically replenished. To provide a new source, the U.S. Department of Energy (DOE) is sponsoring conceptual design and engineering development and demonstration activities for a plant that will use a high-power proton linear accelerator to produce tritium and will go on-line no later than 2007. The APT project is in the process of completing the conceptual design for a tritium production plant. In addition, there are several important areas under engineering development and demonstration that will ensure an efficient, cost-effective plant design and provide an adequate margin of tritium production. Information provided from this work will be used by the DOE in its 1998 choice of production technology implementation

  5. Design and technical status of the EU contribution to ITER

    International Nuclear Information System (INIS)

    Gasparotto, Maurizio; Federici, Gianfranco; Casci, Federico Riccardo

    2009-01-01

    Europe is involved in the procurement of most of the high-technology items for the ITER device (e.g. parts of the superconducting Toroidal (TF) and Poloidal Field (PF) coils, the vacuum vessel (VV), the in-vessel components, the remote handling, the additional heating systems, the tritium plant and cryoplant and finally parts of the diagnostics). In many cases the technologies required to manufacture these components are well established, in others there is still ongoing design and R and D work to select and optimise the final design solutions and to consolidate the underlying technologies as, for example, in the areas of heating and current drive, plasma diagnostics, shield blanket and first wall, remote handling, etc. A design review has recently been conducted by the ITER Organisation, with the support of the Domestic Agencies (DAs) established by the countries participating to ITER, to address the remaining outstanding technical issues and understand the associated implications for design, machine performance, schedule and cost. This paper provides an update of the design and technical status of EU contributions to ITER.

  6. Develop of omni-tritium sample preparation device

    International Nuclear Information System (INIS)

    Tian Junhua; Zheng Min; Zhang Dong

    2008-06-01

    The content of total tritium analysis is required in order to know the tritium contaminated degree of biological samples accurately. But the conversion and collection of organic tritium are difficult. A device to treat total tritium samples was developed. Plant samples were treated by combustion and catalysis. After expelling the free HTO in the samples when heated in abundant oxygen, the samples were ignited. Combustion gas passed the catalysts at 800 degree C and its oxidation was catalyzed, and then the combined tritium in tissues was converted into HTO. HTO was collected by water-cooling tube and condenser. For other samples, HTO was treated and collected by high temperature (The highest temperature is 1000 degree C)-catalysis-double condensation method. This device had solved the problem that organic tritium is difficult to gather. (authors)

  7. Study on conceptual design system of tritium production fusion reactor

    International Nuclear Information System (INIS)

    He Kaihui

    2004-11-01

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

  8. Is Carbon a Realistic Choice for ITER's Divertor?

    International Nuclear Information System (INIS)

    Skinner, C.H.; Federici, G.

    2005-01-01

    Tritium retention by co-deposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point, ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of co-deposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives

  9. 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)

  10. A Survey of Tritium in Irish Seawater

    International Nuclear Information System (INIS)

    Currivan, L.; Kelleher, K.; McGinnity, P.; Wong, J.; McMahon, C.

    2013-07-01

    This report provides a comprehensive record of the study and measurements of tritium in Irish seawater undertaken by the Radiological Protection Institute of Ireland RPII. The majority of the samples analysed were found to have tritium concentrations below the limit of detection and a conservative assessment of radiation dose arising showed a negligible impact to the public. Tritium is discharged in large quantities from various nuclear facilities, and mostly in liquid form. For this reason it is included in the list of radioactive substances of interest to the OSPAR (Oslo-Paris) Convention to protect the marine environment of the North-East Atlantic. To fulfil its role within OSPAR, to provide technical support to the Irish Government, RPII carried out a project to determine the levels of tritium in seawater from around the Irish coast to supplement its routine marine monitoring programme. A total of 85 seawater samples were collected over a three year period and analysed at the RPII's laboratory. Given that the operational discharges for tritium from the nuclear fuel reprocessing plant at Sellafield, UK, are expected to increase due to current and planned decommissioning activities RPII will continue to monitor tritium levels in seawater around the Irish coast, including the Irish Sea, as part of its routine marine monitoring programme

  11. ITER: Chronicle of a bankruptcy foretold

    International Nuclear Information System (INIS)

    Petit, Jean-Pierre

    2011-01-01

    By notably referring to Cedric Reux's research thesis and to Andrew Thornton's research thesis, the author shows that the ITER project is a very dangerous one as the use of a tokamak as an energy plant could lead to a catastrophe. He first recalls and describes the ITER's operation principle (use of nuclear fusion instead of nuclear fission) by recalling historical developments of this approach. He describes a tokamak operation, evokes some lessons learned from the experience, discusses tritium reactions, and the problem of plasma pollution. He then refers to Reux and Thornton, and also to an other author (Boozer from Columbia University) for their theoretical works on tokamaks, and quotes some important statements of their works. As quoted in this document, all these authors highlight the severe problem related to disruptions and to their consequences. The author quotes another article presented during a conference of the risk and consequences of disruptions in large tokamaks, and notably discusses the risk related to fatigue. He also illustrates the involved processes by commenting processes occurring at the Sun surface

  12. 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

  13. Radionuclide Basics: Tritium

    Science.gov (United States)

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

  14. Consequences of the technology survey and gap analysis on the EU DEMO R&D programme in tritium, matter injection and vacuum

    Energy Technology Data Exchange (ETDEWEB)

    Day, Chr., E-mail: Christian.Day@kit.edu [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Butler, B. [Culham Science Centre (CCFE), Abingdon (United Kingdom); Giegerich, T. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Lang, P.T. [Max-Planck-Institute of Plasma Physics (IPP), Garching (Germany); Lawless, R. [Culham Science Centre (CCFE), Abingdon (United Kingdom); Meszaros, B. [EUROfusion Consortium, Programme Management Unit, Garching (Germany)

    2016-11-01

    Highlights: • The inner fuel cycle architecture of DEMO is developed in a systems engineering approach as a functional break-down diagram, driven by the need for inventory minimisation. • Technologies to fulfil the required functions are discussed and ranked. • Prime technologies are identified and an associated R&D programme is developed. • The core challenges of a DEMO fuel cycle beyond those already addressed in ITER are discussed. - Abstract: In the framework of the EUROfusion Programme, EU is preparing the conceptual design of the inner fuel cycle of a pulsed tokamak DEMO. This paper illustrates a quantified process to shape a R&D programme that exploits as much as possible previous R&D. In an initial step, the high-level requirements are collected and a novel DEMO inner fuel cycle architecture with its three sub-systems vacuum pumping, matter injection (fuelling and injection of plasma enhancement gases) and tritium systems (tritium plant and breeder coolant purification) is delineated, driven by the DEMO key challenge to reduce tritium inventory. Then, a technology survey is carried out to review potential existing solutions for the required process functions and to assess their maturity and risks. Finally, a decision-making scheme is applied to select the most promising candidates. ITER technology is exploited where possible. As a primary result, a fuel cycle architecture is suggested with an advanced tritium plant that avoids full isotope separation in the main loop and with a Direct Internal Recycling path in the vacuum systems to shorten cycle times. For core fuelling, classical inboard pellet injection technology is selected, in principle similar to that proposed for ITER but aiming for higher launch speeds to achieve deep fuelling of the DEMO plasma. Based on these findings, a tailored R&D programme is shaped that tackles the key questions until 2020.

  15. In-vessel tritium

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. Protection against tritium radiations

    International Nuclear Information System (INIS)

    Bal, Georges

    1964-05-01

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

  17. Tritium in water ecosystems of Ural

    International Nuclear Information System (INIS)

    Chebotina, M.Ya.; Nikolin, O.A.

    2005-01-01

    The paper provides the data on tritium monitoring in water ecosystems of the Ural region. The study area comprises the Beloyarsk Atomic Power Plant (cooling reservoir and the Olkhovsk bog-river ecosystem), a territory around the 'Mayak' Enterprise, and control territory, for comparison, located in the North of Sverdlovsk region. It was found that a large area in the Ural region, particularly near the power plant and the 'Mayak,, was characterized by increased tritium content in water as compared with technogenic background is typical for control areas. It may be considered that nearly all the tritium within the study area including control ones are of anthropogenic origin taking into account the act that the global background level for the radionuclide is 1 Bq/l.(author)

  18. A Li-particulate blanket concept for ITER

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.

    1989-01-01

    The Li-particulate blanket design concept the authors proposed for the International Thermonuclear Experimental Reactor (ITER) uses a dilute suspension of fine solid breeder particles in a carrier gas as the combined coolant and lithium breeder stream. This blanket concept has a simple mechanical and hydraulic configuration, low inventory of bred tritium, and simple tritium extraction system. Existing technology can be used to implement the design for ITER. The concept has the potential to be a highly reliable shield and blanket design for ITER with relatively low development and capital costs

  19. Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

    International Nuclear Information System (INIS)

    Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

    2002-01-01

    Kazakhstan takes part in ITER project in partnership with Russian Federation since the year of 1994. At present the technical stage of the project is completed and ITER Council should take a decision on the site for international reactor. Four countries such as Canada, Japan, Spain and France have offered their territories for being used as site for launching ITER construction. ITER partners started preparing new international agreement that will cover activities on construction, operation and decommissioning of ITER. It will also include the list of research and experimental work that is conducted in support of ITER project. Kazakhstan has already made an important contribution into technical stage realization of ITER project due to scientific and technical researches conducted by National Nuclear Center, by Institute of Experimental and Theoretical Physics and by JSC 'Ulba Metallurgical plant' ('UMP'). Research activity carried out for the support of ITER project is performed in accordance with the following main trends: Tritium safety (permeability and retentin of hydrogen isotopes during in-pile irradiation in various structural materials, co-deposed layers and protective coatings); Verification of computer codes (LOCA type) loss of coolant accidents modeling in ITER reactor; Investigation of liquid metal blanket of thermonuclear reactor (tritium production in lithium containing eutectics Li17Pb83 and ceramics Li 2 TiO 3 , study of tritium permeability). At present the working group of ITER project participants started introducing proposals for cost distribution and for placing the orders on reactor construction. Further Kazakhstan participation in ITER project may be in manufacturing high-tech parts and assemblies from commercial grades of beryllium. They will be used for armouring the reactor first wall, for its thermal protection and for protection of superconductor's components for magnetic systems that are at JSC U MP'. Scientific and technical support of

  20. ITER, a major step toward nuclear fusion energy

    International Nuclear Information System (INIS)

    Ikeda, K.; Holtkamp, N.; Pick, M.; Gauche, F.; Garin, P.; Bigot, B.; Luciani, J.F.; Mougniot, J.C.; Watteau, J.P.; Saoutic, B.; Becoulet, A.; Libeyre, P.; Beaumont, B.; Simonin, A.; Giancarli, L.; Rosenvallon, S.; Gastaldi, O.; Marbach, G.; Boudot, C.; Ioki, K.; Mitchell, N.; Girard, J.Ph.; Giraud, B.; Lignini, F.; Giguet, E.; Bofusch, E.; Friconneau, J.P.; Di Pace, L.; Pampin, R.; Cook, I.; Maisonnier, D.; Campbell, D.; Hayward, J.; Li Puma, A.; Norajitra, P.; Sardain, P.; Tran, M.Q.; Ward, D.; Moslang, A.; Carre, F.; Serpantie, J.P.

    2007-01-01

    This document gathers together a series of articles dedicated to ITER. They are organized into 5 parts. The first part describes the potential of fusion as a source of energy that will be able to face the challenge of a continuously increasing demand. After a reminder of the main fusion reactions and the conditions to obtain fusion, the second part focuses on the magnetic fusion based concepts with a special emphasis on the tokamak configuration. In the third part the main components of ITER are described: first the plasma facing components, then the vacuum vessel, the superconducting magnets and the heating systems. In the fourth part short papers concerning ITER safety, the maintenance through remote handling systems, the tritium breeding blanket, are given, along with a full article on the waste management. It is interesting to notice that the nuclear wastes will represent: -) between 1600 and 3800 tons of housekeeping and process wastes produced during the 20 years of operation of ITER (20% very low level waste, 75% low or medium activity with short life and 5% medium activity with long life), -) about 750 tons from component replacement during ITER active operation, and -) about 30000 tons from the decommissioning of ITER. The last part presents the European concepts for a power plant based on a fusion reactor. A basic design is given along with a state of the art of the research on the materials that will be used for the structures. It is highlighted that synergies between fission and fusion technologies exist in at least 4 areas: nuclear design code system, high temperature materials, safety approach, and in-service inspection, maintenance and dismantling. (A.C.)

  1. 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

  2. Results of tritium measurement in environmental samples and drainage

    International Nuclear Information System (INIS)

    Koike, Ryoji; Hirai, Yasuo

    1983-01-01

    In Ibaraki prefecture, the tritium concentration in the drainage from the nuclear facilities has been measured since 1974. Then, with the start of operation of the fuel reprocessing plant in 1977, the tritium concentration in environmental samples was to be measured also in order to examine the effect of the drainage on the environment. The results of the tritium measurement in Ibaraki prefecture up to about 1980 are described: sampling points, sampling and measuring methods, the tritium concentration in the drainage, air, inland water and seawater, respectively. The drainages have been taken from Japan Atomic Power Company, Japan Atomic Energy Research Institute, and Power Reactor and Nuclear Fuel Development Corporation (with the fuel reprocessing plant). The samples of air, inland water and seawater have been taken in the areas concerned. The tritium concentration was measured by a low-background liquid scintillation counter. The measured values in the environment have been generally at low level, not different from other areas. (Mori, K.)

  3. Tritium breeding in fusion reactors

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1982-10-01

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

  4. ITER, a major step toward nuclear fusion energy; ITER, une etape majeure vers l'energie de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, K.; Holtkamp, N.; Pick, M.; Gauche, F.; Garin, P.; Bigot, B.; Luciani, J.F.; Mougniot, J.C.; Watteau, J.P.; Saoutic, B.; Becoulet, A.; Libeyre, P.; Beaumont, B.; Simonin, A.; Giancarli, L.; Rosenvallon, S.; Gastaldi, O.; Marbach, G.; Boudot, C.; Ioki, K.; Mitchell, N.; Girard, J.Ph.; Giraud, B.; Lignini, F.; Giguet, E.; Bofusch, E.; Friconneau, J.P.; Di Pace, L.; Pampin, R.; Cook, I.; Maisonnier, D.; Campbell, D.; Hayward, J.; Li Puma, A.; Norajitra, P.; Sardain, P.; Tran, M.Q.; Ward, D.; Moslang, A.; Carre, F.; Serpantie, J.P

    2007-01-15

    This document gathers together a series of articles dedicated to ITER. They are organized into 5 parts. The first part describes the potential of fusion as a source of energy that will be able to face the challenge of a continuously increasing demand. After a reminder of the main fusion reactions and the conditions to obtain fusion, the second part focuses on the magnetic fusion based concepts with a special emphasis on the tokamak configuration. In the third part the main components of ITER are described: first the plasma facing components, then the vacuum vessel, the superconducting magnets and the heating systems. In the fourth part short papers concerning ITER safety, the maintenance through remote handling systems, the tritium breeding blanket, are given, along with a full article on the waste management. It is interesting to notice that the nuclear wastes will represent: -) between 1600 and 3800 tons of housekeeping and process wastes produced during the 20 years of operation of ITER (20% very low level waste, 75% low or medium activity with short life and 5% medium activity with long life), -) about 750 tons from component replacement during ITER active operation, and -) about 30000 tons from the decommissioning of ITER. The last part presents the European concepts for a power plant based on a fusion reactor. A basic design is given along with a state of the art of the research on the materials that will be used for the structures. It is highlighted that synergies between fission and fusion technologies exist in at least 4 areas: nuclear design code system, high temperature materials, safety approach, and in-service inspection, maintenance and dismantling. (A.C.)

  5. Statistical Analysis of Environmental Tritium around Wolsong Site

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of)

    2010-04-15

    To find the relationship among airborne tritium, tritium in rainwater, TFWT (Tissue Free Water Tritium) and TBT (Tissue Bound Tritium), statistical analysis is conducted based on tritium data measured at KHNP employees' house around Wolsong nuclear power plants during 10 years from 1999 to 2008. The results show that tritium in such media exhibits a strong seasonal and annual periodicity. Tritium concentration in rainwater is observed to be highly correlated with TFWT and directly transmitted to TFWT without delay. The response of environmental radioactivity of tritium around Wolsong site is analyzed using time-series technique and non-parametric trend analysis. Tritium in the atmosphere and rainwater is strongly auto-correlated by seasonal and annual periodicity. TFWT concentration in pine needle is proven to be more sensitive to rainfall phenomenon than other weather variables. Non-parametric trend analysis of TFWT concentration within pine needle shows a increasing slope in terms of confidence level of 95%. This study demonstrates a usefulness of time-series and trend analysis for the interpretation of environmental radioactivity relationship with various environmental media.

  6. 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.)

  7. Development of organic tritium light technology at Ontario Hydro

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. Tritium permeation through iron

    International Nuclear Information System (INIS)

    Hagi, Hideki; Hayashi, Yasunori

    1989-01-01

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

  9. Metabolism and dosimetry of tritium

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

    2012-01-01

    a particular skeleton ad-hoc for repeated execution turns out to be considerably complicated, and raises general questions about introducing state into a stateless parallel computation. In addition, one would strongly prefer an approach which leaves the original skeleton intact, and only uses it as a building...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...

  11. Problems of anthropogenic tritium limitation

    Directory of Open Access Journals (Sweden)

    Kochetkov О.A.

    2013-12-01

    Full Text Available This article contains the current situation in respect to the environmental concentrations of anthropogenic and natural tritium. There are presented and analyzed domestic standards for НТО of all Radiation Safety Standards (NRB, as well as the regulations analyzed for tritium in drinking water taken in other countries today. This article deals with the experience of limitation of tritium and focuses on the main problem of rationing of tritium — rationing of organically bound tritium.

  12. Tritium removal by CO2 laser heating

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Mueller, D.

    1997-01-01

    Efficient techniques for rapid tritium removal will be necessary for ITER to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 ms is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

  13. Tritium removal by CO2 laser heating

    International Nuclear Information System (INIS)

    Skinner, C.H.; Kugel, H.; Mueller, D.

    1997-10-01

    Efficient techniques for rapid tritium removal will be necessary for ITER (International Thermonuclear Experimental Reactor) to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO 2 or Nd:YAG laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm 2 flux with an exposure time of order 10 msec is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally

  14. Development and test of prototype components for ITER

    International Nuclear Information System (INIS)

    Biel, Wolfgang; Behr, Wilfried; Castano-Bardawil, David

    2015-08-01

    The scientific program of the project is divided into the following partial projects: (1.) ITER Diagnostic Port Plug for the charge-exchange spectroscopy (CXRS) with the subthemes: (a) Development of prototypes for critical mechanical components, (b) development of a roboter for the laser welding of vacuum seals and pipings at the Port Plug, (c) mirror studies, (d) CXRS prototype spectrometer, (2.) ITER tritium retention diagnostics (TR), (3.) ITER disruption mitigation ventile (DMV).

  15. Development of pellet injection systems for ITER

    International Nuclear Information System (INIS)

    Combs, S.K.; Gouge, M.J.; Baylor, L.R.

    1995-01-01

    Oak Ridge National Laboratory (ORNL) has been developing innovative pellet injection systems for plasma fueling experiments on magnetic fusion confinement devices for about 20 years. Recently, the ORNL development has focused on meeting the complex fueling needs of the International Thermonuclear Experimental Reactor (ITER). In this paper, we describe the ongoing research and development activities that will lead to a ITER prototype pellet injector test stand. The present effort addresses three main areas: (1) an improved pellet feed and delivery system for centrifuge injectors, (2) a long-pulse (up to steady-state) hydrogen extruder system, and (3) tritium extruder technology. The final prototype system must be fully tritium compatible and will be used to demonstrate the operating parameters and the reliability required for the ITER fueling application

  16. Control of the tritium path in process heat HTR's

    International Nuclear Information System (INIS)

    Kirch, N.; Scheidler, G.

    1985-01-01

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

  17. ITER waste management

    International Nuclear Information System (INIS)

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

    2010-01-01

    ITER will produce solid radioactive waste during its operation (arising from the replacement of components and from process and housekeeping waste) and during decommissioning (de-activation phase and dismantling). The waste will be activated by neutrons of energies up to 14 MeV and potentially contaminated by activated corrosion products, activated dust and tritium. This paper describes the waste origin, the waste classification as a function of the French national agency for radioactive waste management (ANDRA), the optimization process put in place to reduce the waste radiotoxicity and volumes, the estimated waste amount based on the current design and maintenance procedure, and the overall strategy from component removal to final disposal anticipated at this stage of the project.

  18. ITER safety and operational scenario

    International Nuclear Information System (INIS)

    Shimomura, Y.; Saji, G.

    1998-01-01

    The safety and environmental characteristics of ITER and its operational scenario are described. Fusion has built-in safety characteristics without depending on layers of safety protection systems. Safety considerations are integrated in the design by making use of the intrinsic safety characteristics of fusion adequate to the moderate hazard inventories. In addition to this, a systematic nuclear safety approach has been applied to the design of ITER. The safety assessment of the design shows how ITER will safely accommodate uncertainties, flexibility of plasma operations, and experimental components, which is fundamental in ITER, the first experimental fusion reactor. The operation of ITER will progress step by step from hydrogen plasma operation with low plasma current, low magnetic field, short pulse and low duty factor without fusion power to deuterium-tritium plasma operation with full plasma current, full magnetic field, long pulse and high duty factor with full fusion power. In each step, characteristics of plasma and optimization of plasma operation will be studied which will significantly reduce uncertainties and frequency/severity of plasma transient events in the next step. This approach enhances reliability of ITER operation. (orig.)

  19. ITER primary cryopump test facility

    International Nuclear Information System (INIS)

    Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

    1998-01-01

    A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

  20. The ITER remote maintenance system

    International Nuclear Information System (INIS)

    Tesini, A.; Palmer, J.

    2007-01-01

    ITER is a joint international research and development project that aims to demonstrate the scientific and technological feasibility of fusion power. As soon as the plasma operation begins using tritium, the replacement of the vacuum vessel internal components will need to be done with remote handling techniques. To accomplish these operations ITER has equipped itself with a Remote Maintenance System; this includes the Remote Handling equipment set and the Hot Cell facility. Both need to work in a cooperative way, with the aim of minimizing the machine shutdown periods and to maximize the machine availability. The ITER Remote Handling equipment set is required to be available, robust, reliable and retrievable. The machine components, to be remotely handle-able, are required to be designed simply so as to ease their maintenance. The baseline ITER Remote Handling equipment is described. The ITER Hot Cell Facility is required to provide a controlled and shielded area for the execution of repair operations (carried out using dedicated remote handling equipment) on those activated components which need to be returned to service, inside the vacuum vessel. The Hot Cell provides also the equipment and space for the processing and temporary storage of the operational and decommissioning radwaste. A conceptual ITER Hot Cell Facility is described. (orig.)

  1. ITER plasma facing components

    International Nuclear Information System (INIS)

    Kuroda, T.; Vieider, G.; Akiba, M.

    1991-01-01

    This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

  2. Tritium permeation and recovery

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  3. Tritium protective clothing

    International Nuclear Information System (INIS)

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

    1979-06-01

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

  4. Tritium-v. 2

    International Nuclear Information System (INIS)

    1987-01-01

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

  5. Tritium waste package

    Science.gov (United States)

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

    1995-01-01

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

  6. Tritium protective clothing

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-06-01

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

  7. Follow-up Study of ITER Safety Analysis : Large In-vessel First Wall Pipe Break with Wet Confinement Bypass

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Sung Bo; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    Previous researches have been analyzed risk assessments of fusion reactors that are dangerous in the severe accidents where the radioactive material released from confinement building to the environment. To simulate the severe accidents in ITER, a number of thermal hydraulics simulation codes were used. Before construction of the fusion reactor, to obtain ITER license about safety issue, MELCOR is chosen as one of the several codes to be used to perform ITER safety analyses. Qualification of the simulation code is to simulate the cooling system in ITER, the transport of radionuclides during design basis accidents (DBAs) including beyond design basis accidents (BDBAs). MELCOR is fully integrated code that models the accidents in Light Water Reactor (LWR). To analyze the accidents in ITER, MELCOR 1.8.2 version is modified. In the nuclear fusion system, the amount of released radioactive material is criteria for safety permission. Tritium (or tritiated water: HTO) and radioactive dust aerosol are the source of radioactive leakage. In the Generic Site Safety Report (GSSR) for the ITER plant, Table I lists the release guidelines for tritium and activation products for normal operation, incidents and accidents. Several accident analyses have been studied to know how much radioactive material could be released from the severe accidents. In the present work, The MELCOR input deck of large First Wall (FW) coolant leak (pipe break) is used to study and radioactive material leakage thorough bypass accident are studied to follow up the ITER safety analysis. In this research, follow-up study of the in-vessel inboard/inboard-outboard FW pipe break was analyzed to investigate the amount of leakage of radioactive aerosol. All of the accident cases released the lower amount of radioactive aerosol compared to the IAEA guide lines. In addition, the OBB pipe break made lower HTO aerosol leakage because of condensation of HTO and adsorption between coolant and aerosol.

  8. 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).

  9. Thin film tritium dosimetry

    Science.gov (United States)

    Moran, Paul R.

    1976-01-01

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

  10. Management of tritium wastes

    International Nuclear Information System (INIS)

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

    1991-07-01

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

  11. PRODUCTION OF TRITIUM

    Science.gov (United States)

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

    1963-02-26

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

  12. Status of development of functional materials with perspective on beyond ITER

    International Nuclear Information System (INIS)

    Shikama, T.; Knitter, R.; Moeslang, A.; Konys, J.; Deli, L.; Muroga, T.; Kawamura, H.; Kohyama, A.

    2007-01-01

    Any engineering system is composed of functional materials as well as of structural materials, and more advanced systems tend to demand a more important and versatile role to functional materials. In nuclear fusion systems, examples of principle functional materials will be breeders and neutron multipliers for tritium production, coatings on structural materials for corrosion-resistance, MHD-loss-reduction and control of tritium permeation, thermal insertions for heat transport control, and optical and electrical materials for plasma and environmental diagnostics. For incarnation of a nuclear fusion power plant, namely DEMO, development of the functional materials with appropriate properties is essential. A role of functional materials depends strongly on a specific design of DEMO, namely designs of systems for tritium-breeding, system-cooling and heat-transfer. In the framework of ITER project, development of tritium blanket modules (TBM) is underway. Also, in parallel with the ITER project, a complemental program called the BA (Broader Approach) is launched for realization of a DEMO nuclear fusion reactor in an appropriate time schedule, where key issues of the nuclear fusion engineering needed for the DEMO will be studied under EU/Japan collaboration. In the meantime, technologies and materials needed for diagnostics and control of burning plasma are extensively discussed under the framework of International Tokamak Physics Activity (ITPA). The present paper reviews a present status of development of functional materials from views of internationally coordinated activities based on fundamental aspects of the DEMO demands as well as from views of activities based on specific but currently dominant DEMO designs. Examples of functional materials reviewed here are solid breeders, beryllium and beryllium alloys, coating layers on structural materials, thermal inserts, and some electrical and optical materials. (orig.)

  13. Fusion Plasma Physics and ITER - An Introduction (1/4)

    CERN Multimedia

    CERN. Geneva

    2011-01-01

    In November 2006, ministers representing the world’s major fusion research communities signed the agreement formally establishing the international project ITER. Sited at Cadarache in France, the project involves China, the European Union (including Switzerland), India, Japan, the Russian Federation, South Korea and the United States. ITER is a critical step in the development of fusion energy: its role is to confirm the feasibility of exploiting magnetic confinement fusion for the production of energy for peaceful purposes by providing an integrated demonstration of the physics and technology required for a fusion power plant. The ITER tokamak is designed to study the “burning plasma” regime in deuterium-tritium (D-T) plasmas by achieving a fusion amplification factor, Q (the ratio of fusion output power to plasma heating input power), of 10 for several hundreds of seconds with a nominal fusion power output of 500MW. It is also intended to allow the study of steady-state plasma operation at Q≥5 by me...

  14. Tritium isotope fractionation in biological systems and in analytical procedures

    International Nuclear Information System (INIS)

    Kim, M.A.; Baumgaertner, Franz

    1989-01-01

    The organically bound tritium (OBT) is evaluated in biological systems by determining the tritium distribution ratio (R-value), i.e. tritium concentrations in organic substance to cell water. The determination of the R-value always involves isotope fractionation is applied analytical procedures and hence the evaluation of the true OBT -value in a given biological system appears more complicated than hitherto known in the literature. The present work concentrates on the tritium isotope fractionation in the cell water separation and on the resulting effects on the R-value. The analytical procedures examined are vacuum freeze drying under equilibrium and non-equilibrium conditions and azeotropic distillation. The vaporization isotope effects are determined separately in the phase transition of solid or liquid to gas in pure tritium water systems as well as in real biological systems, e.g. corn plant. The results are systematically analyzed and the influence of isotope effects on the R-value is rigorously quantified

  15. 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

  16. Comparison of tritium production facilities

    International Nuclear Information System (INIS)

    He Kaihui; Huang Jinhua

    2002-01-01

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

  17. 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)

  18. 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.

  19. 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

  20. Final programmatic environmental impact statement for tritium supply and recycling

    International Nuclear Information System (INIS)

    1995-10-01

    Tritium, a radioactive gas used in all of the Nation's nuclear weapons, has a short half-life and must be replaced periodically in order for the weapon to operate as designed. Currently, there is no capability to produce the required amounts of tritium within the Nuclear Weapons Complex. The PEIS for Tritium Supply and Recycling evaluates the alternatives for the siting, construction, and operation of tritium supply and recycling facilities at each of five candidate sites: the Idaho National Engineering Laboratory, the Nevada Test Site, the Oak Ridge Reservation, the Pantex Plant, and the Savannah River Site. Alternatives for new tritium supply and recycling facilities consist of four different tritium supply technologies: Heavy Water Reactor, Modular High Temperature Gas-Cooled Reactor, Advanced Light Water Reactor, and Accelerator Production of Tritium. The PEIS also evaluates the impacts of the DOE purchase of an existing operating or partially completed commercial light water reactor or the DOE purchase of irradiation services contracted from commercial power reactors. Additionally, the PEIS includes an analysis of multipurpose reactors that would produce tritium, dispose of plutonium, and produce electricity. Evaluation of impacts on land resources, site infrastructure, air quality and acoustics, water resources, geology and soils, biotic resources, cultural and paleontological resources, socioeconomics, radiological and hazardous chemical impacts during normal operation and accidents to workers and the public, waste management, and intersite transport are included in the assessment

  1. Draft programmatic environmental impact statement for tritium supply and recycling

    International Nuclear Information System (INIS)

    1995-02-01

    Tritium, a radioactive gas used in all of the Nation's nuclear weapons, has a short half-life and must be replaced periodically in order for the weapon to operate as designed. Currently, the Nation has no tritium production capability. The Tritium Supply and Recycling PEIS evaluates the alternatives for the siting, construction, and operation of tritium supply and recycling facilities at each of five candidate sites: the Idaho Engineering Laboratory, the Nevada Test Site, the Oak Ridge Reservation, the Pantex Plant, and the Savannah River Site. Alternatives for new tritium supply and recycling facilities consist of four different tritium supply technologies; Heavy Water Reactor, Modular High Temperature Gas-cooled Reactor, Advanced Light Water Reactor, and Accelerator Production of Tritium. The PEIS also evaluates the impacts of using a commercial light water reactor, either as a contingency in the event of a national emergency or if purchased by the DOE and converted to defense purposes. Additionally, the PEIS includes an analysis of multi-purpose reactors which would produce tritium, dispose of plutonium and produce electricity. Volume I contains the findings of these analyses, Volume II contains the Appendices and supporting data

  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. Development of tritium permeation barriers on Al base in Europe

    Science.gov (United States)

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

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

  4. ITER diagnostic system: Vacuum interface

    International Nuclear Information System (INIS)

    Patel, K.M.; Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L.; Drevon, J.M.; Encheva, A.; Kashchuk, Y.; Maquet, Ph.; Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J.

    2013-01-01

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10 −7 Pa, irrespective of plasma operation, and a leak rate of less than 10 −10 Pa m 3 s −1 . In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described

  5. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  6. Radwaste management aspects of the test blanket systems in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Laan, J.G. van der, E-mail: JaapG.vanderLaan@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Canas, D. [CEA, DEN/DADN, centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Chaudhari, V. [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India); Iseli, M. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Kawamura, Y. [Japan Atomic Energy Agency, Naka-shi, Ibaraki-ken 311-0193 (Japan); Lee, D.W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Petit, P. [European Commission, DG ENER, Brussels (Belgium); Pitcher, C.S.; Torcy, D. [ITER Organization, Route de Vinon sur Verdon, F-13067 Saint Paul Lez Durance (France); Ugolini, D. [Fusion for Energy, Barcelona (Spain); Zhang, H. [China Nuclear Energy Industry Corporation, Beijing 100032 (China)

    2016-11-01

    Highlights: • Test Blanket Systems are operated in ITER to test tritium breeding technologies. • The in-vessel parts of TBS become radio-active during the ITER nuclear phase. • For each TBM campaign the TBM, its shield and the Pipe Forests are removed. • High tritium contents and novel materials are specific TBS radwaste features. • A preliminary assessment confirmed RW routing, provided its proper conditioning. - Abstract: Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.

  7. Incorporation of tritium due to foliar exposure in certain vegetation

    International Nuclear Information System (INIS)

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

    1981-01-01

    Tritium uptake, release and incorporation patterns through the foliage of seedlings of certain edible vegetation were investigated, for exposure periods ranging from an hour to about 20 hours. A large number of plants belonging to the family of lettuce (Lactuca sativa L.), cabbage (Brassica Oleracea L.) and capsicum (Capsicum fruitescens L.) were exposed to tritiated air under dynamic and static conditions. The half times for tissue-free-water-tritium (TFWT) were found to be about 46 and 32 minutes for capsicum and lettuce and 45 minutes for cabbage. Tissue-bound-tritium (TBT) in the seedlings and the grown plants showed different incorporation rates as a result of foliar exposure. The relative concentration factors were larger by a factor of ten for TFWT in the leaves of the grown plants than in the shoots of the seedlings. However, tissue-bound-tritium concentration values in the shoots/stems of the young and grown plants were of the same order, as related to the tissue-free-water-tritium concentrations. Thus the study indicates a larger translocation of tritium from aqueous to organic phase in the leaves of the grown plants than in the shoots of young seedlings. (auth.)

  8. Tritium emissions reduction facility (TERF)

    International Nuclear Information System (INIS)

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

    1993-01-01

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