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Sample records for tritium operations experience

  1. Tritium Room Air Monitor Operating Experience Review

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader; B. J. Denny

    2008-09-01

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

  2. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  3. The tritium operations experience on TFTR

    International Nuclear Information System (INIS)

    von Halle, A.; Gentile, C.

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  5. Design and operational experience with a portable tritium cleanup system

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Wilson, S.W.; Garcia, F.

    1991-06-01

    We built a portable tritium cleanup system to scavenge tritium from contaminated gases in any tritium-containing system in the LLNL Tritium Facility. The cleanup system uses standard catalytic oxidation of tritium to water followed by water removal with a molecular sieve dryer. The cleanup unit, complete with instrumentation, is contained in a portable cart that is rolled into place and connected to the apparatus to be cleaned. The cleanup systems is effective, low-tech, simple, and reliable. The nominal flow rate of the system is 30 liters/minute, and the decontamination factor is > 1000. In this paper we will show design information on our portable cleanup system, and will discuss our operational experience with it over the past several years

  6. Operational experience with two tritium-effluent-monitoring systems

    International Nuclear Information System (INIS)

    Haynie, J.S.; Gutierrez, J.A.

    1982-01-01

    Two new tritium stack monitoring systems were designed and built. The operational experience of a wide-range detector with a useful range of a few μCi/m 3 to 10 8 μCi/m 3 , and a second monitoring system using an improved Kanne chamber and a new electrometer, called a Model 39 Electrometer-Chargemeter are discussed. Both tritium chambers have been designed to have a reduced sensitivity to tritium contamination, a fast response, and an integrating chargemeter with digital readout for easy conversion to microcuries. The calibration of these monitors and advantages of using these chambers over conventional systems are discussed

  7. Operating Experience Review of Tritium-in-Water Monitors

    Energy Technology Data Exchange (ETDEWEB)

    S. A. Bruyere; L. C. Cadwallader

    2011-09-01

    Monitoring tritium facility and fusion experiment effluent streams is an environmental safety requirement. This paper presents data on the operating experience of a solid scintillant monitor for tritium in effluent water. Operating experiences were used to calculate an average monitor failure rate of 4E-05/hour for failure to function. Maintenance experiences were examined to find the active repair time for this type of monitor, which varied from 22 minutes for filter replacement to 11 days of downtime while waiting for spare parts to arrive on site. These data support planning for monitor use; the number of monitors needed, allocating technician time for maintenance, inventories of spare parts, and other issues.

  8. Operation of the tokamak fusion test reactor tritium systems during initial tritium experiments

    International Nuclear Information System (INIS)

    Anderson, J.L.; Gentile, C.; Kalish, M.; Kamperschroer, J.; Kozub, T.; LaMarche, P.; Murray, H.; Nagy, A.; Raftopoulos, S.; Rossmassler, R.; Sissingh, R.; Swanson, J.; Tulipano, F.; Viola, M.; Voorhees, D.; Walters, R.T.

    1995-01-01

    The high power D-T experiments on the tokamak fusion test reactor (TFTR) at the Princeton Plasma Physics Laboratory commenced in November 1993. During initial operation of the tritium systems a number of start-up problems surfaced and had to be corrected. These were corrected through a series of system modifications and upgrades and by repair of failed or inadequate components. Even as these operational concerns were being addressed, the tritium systems continued to support D-T operations on the tokamak. During the first six months of D-T operations more than 107kCi of tritium were processed successfully by the tritium systems. D-T experiments conducted at TFTR during this period provided significant new data. Fusion power in excess of 9MW was achieved in May 1994. This paper describes some of the early start-up issues, and reports on the operation of the tritium system and the tritium tracking and accounting system during the early phase of TFTR D-T experiments. (orig.)

  9. Operating experience and procedures at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Carlson, R.V.; Binning, K.E.; Cole, S.P.; Jenkins, E.M.; Wilhelm, R.C.; Cole, S.P.

    1988-01-01

    Operating procedures are important for the safe and efficient operation of the Tritium Systems Test Assembly (TSTA). TSTA has been operating for four years with tritium in a safe and efficient manner. The inventory of tritium in the process loop is 100 grams and several milestone runs have been completed. This paper describes the methods used to operate TSTA. 3 refs., 1 fig

  10. 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

  11. Operating experience with TFTR's Tritium Storage and Delivery System

    International Nuclear Information System (INIS)

    Voorhees, D.R.

    1995-01-01

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

  12. JET experiments with tritium and deuterium–tritium mixtures

    NARCIS (Netherlands)

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

    2016-01-01

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

  13. An assembly of tritium production experiment

    International Nuclear Information System (INIS)

    Abe, Toshihiko

    1981-01-01

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

  14. 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

  15. TFTR tritium operations lessons learned

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  16. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    International Nuclear Information System (INIS)

    Gentile, C.A.; LaMarche, P.H.

    1995-01-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During '' time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems

  17. Thought experiment with tritium

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  18. Tritium contamination experience in an operational D-T fusion reactor

    International Nuclear Information System (INIS)

    Gentile, C.A.; Ascione, G.

    1994-01-01

    During December 1993, the Tokamak Fusion Test Reactor (TFTR) injected a mixture of deuterium and tritium in the TFTR vacuum vessel for the purpose of creating D-T plasmas. The tritium used in these D-T plasmas was stored, delivered and processed in the TFTR tritium facility that includes the tritium vault, waste handling area, clean-up area, and gas holding tank room. During this time period, several components in the tritium process system were found to have tritium leaks which led to tritium deposition on process skids, components and floor area. Radiological surveys of surfaces contaminated with tritium oxide indicate a decrease in surface contamination in time (on the order of 12 to 36 hours) as the result of room ventilation. In instances where the facility HVAC system was maintained in the purge mode, a dramatic decrease in surface contamination was observed. Areas contaminated with tritium oxide (> 16.6 Bq/100 cm 2 ) were found to be clean ( 2 ) after several hours of continuous purging by the facility HVAC system. In instances where relative humidity was not decreased, the tritium surface contamination was found to be attenuated. During the months of December 1993, January and February 1994 tritium leaking components were either replaced, redesigned or repaired. During this time period, data were collected in the form of contamination surveys, real time tritium monitor output, and HVAC configuration indicating the correlation of purge ventilation leading to a decrease in tritium oxide surface contamination

  19. Tritium contamination experience in an operational D-T fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-09-01

    During December 1993, the Tokamak Fusion Test Reactor (TFTR) injected a mixture of deuterium and tritium in the TFTR vacuum vessel for the purpose of creating D-T plasmas. The tritium used in these D-T plasmas was stored, delivered and processed in the TFTR tritium facility that includes the tritium vault, waste handling area, clean-up area, and gas holding tank room. During this time period, several components in the tritium process system were found to have tritium leaks which led to tritium deposition on process skids, components and floor area. Radiological surveys of surfaces contaminated with tritium oxide indicate a decrease in surface contamination in time (on the order of 12 to 36 hours) as the result of room ventilation. In instances where the facility HVAC system was maintained in the purge mode, a dramatic decrease in surface contamination was observed. Areas contaminated with tritium oxide (> 16.6 Bq/100 cm{sup 2}) were found to be clean (< 16.6 Bq/100 cm{sub 2}) after several hours of continuous purging by the facility HVAC system. In instances where relative humidity was not decreased, the tritium surface contamination was found to be attenuated. During the months of December 1993, January and February 1994 tritium leaking components were either replaced, redesigned or repaired. During this time period, data were collected in the form of contamination surveys, real time tritium monitor output, and HVAC configuration indicating the correlation of purge ventilation leading to a decrease in tritium oxide surface contamination.

  20. Recommended radiological controls for tritium operations

    International Nuclear Information System (INIS)

    Mansfield, G.

    1992-01-01

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

  1. Tips for the fabrication of temporary tritium experiments

    International Nuclear Information System (INIS)

    Binning, K.E.; Jenkins, E.M.

    1988-01-01

    The Tritium System Test Assembly (TSTA) is a facility built for the demonstration of tritium handling systems necessary for tritium-burning fusion reactors. The facility has been in operation handling tritium for four years. The current inventory of tritium is approximately one hundred grams, with DOE approval for a maximum inventory of two hundred grams. Not all experiments performed at TSTA require the operation of the main process loop. During the last four years, many small scale experiments have been performed to test the compatibility and operation of tritium processing components in small self-contained experimental packages. These packages are fabricated inside secondary containment gloveboxes and can be operated for hours or months with little monitoring. Construction of these packages need to be tritium compatible, inexpensive, easy to build, and versatile. This paper discusses some of the problems and remedies encountered during the building of temporary experiments

  2. Tips for the fabrication of temporary tritium experiments

    International Nuclear Information System (INIS)

    Binning, K.E.; Jenkins, E.M.

    1988-01-01

    The Tritium System Test Assembly (TSTA) is a facility built for the demonstration of tritium handling systems necessary for tritium-burning fusion reactors. The facility has been in operation handling tritium for four years. The current inventory of tritium is approximately one hundred grams with DOE approval exists for a maximum inventory of two hundred grams. Not all experiments performed at TSTA require the operation of the main process loop. During the last four years, many small scale experiments have been performed to test the compatibility and operation of tritium processing components in small self-contained experimental packages. These packages are fabricated inside secondary containment gloveboxes and can be operated for hours or months with little monitoring. Construction of these packages need to be tritium compatible, inexpensive, easy to build, and versatile. This paper discusses some of the problems and remedies encountered during the building of temporary experiments

  3. Implanted-tritium permeation experiments

    International Nuclear Information System (INIS)

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

    1982-04-01

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

  4. Tritium experience in the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

    1998-01-01

    Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

  5. TSTA loop operation with 100 grams-level of tritium

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Fukui, Hiroshi; Hirata, Shingo

    1988-12-01

    A fully integrated loop operation test of Tritium systems Test Assembly (TSTA) with 107 grams of tritium was completed at Los Alamos National Laboratory (LANL) in June, 1988. In this test, a compound cryopump with a charcoal panel was incorporated into the main process loop for the first time. The objectives were (i) to demonstrate the compound cryopump system with different flow rates and impurities, (ii) to demonstrate the regeneration of the compound cryopump system, (iii) to accumulate operating experience with other process systems such as the fuel cleanup system, the isotope separation system, the tritium supply and recovery system, etc. and (iv) to improve the data-base on TSTA safety systems such as the secondary containment system, tritium waste treatment system and tritium monitoring system. This report briefly describes characteristics of the main subsystems observed during the milestone run. (author)

  6. Tritium handling and processing experience at TSTA

    International Nuclear Information System (INIS)

    Anderson, J.L.; Okuno, K.

    1994-01-01

    In 1987, the Japan Atomic Energy Research Institute (JAERI) and the US Department of Energy (DOE) signed a collaborative agreement (Annex IV) for the joint funding and operation of the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory (LANL) for a five year period ending June, 1992. After this initial five year collaboration, the Annex IV agreement was extended for another two year period ending June, 1994. During the first five years, a number of the integrated process loop tests of TSTA were conducted, as well as off-line testing of TSTA subsystems. During integrated loop testing the vacuum system, fuel cleanup systems, isotope separation system, transfer pumping system and gas analysis system, are interconnected and tested using 100 g-inventories of tritium to demonstrate steady-state operation of a tritium fuel processing cycle for a fusion reactor. These tests have resulted in a number of significant accomplishments and an experience data base on research, development and operation of the fuel processing system. One of the most significant accomplishments during the initial five year period was the continuous operation of the fuel processing loop for 25 days. During this 25-day extended operation, both the JAERI fuel cleanup system (J-FCU) and the original TSTA fuel cleanup system (FCU) were operated under similar conditions of flow, pressure, and impurity content of the DT gas. Both fuel cleanup systems were demonstrated to provide adequate impurity removal for plasma exhaust gas processing. The isotope separation system was operated continuously, producing pure tritium while rejecting protium as an impurity

  7. Initial experience of tritium exposure control at JET

    International Nuclear Information System (INIS)

    Patel, B.; Campling, D.C.; Schofield, P.A.; Macheta, P.; Sandland, K.

    1998-01-01

    Some of the safety procedures and controls in place for work with tritium are described, and initial operational experience of handling tritium is discussed. A description is given of work to rectify a water leak in a JET neutral beam heating component, which involved man-access to a confined volume to perform repairs, at tritium levels about 100 DAC (80 MBq/m 3 . HTO). Control measures involving use of purge and extract ventilation, and of personal protection using air-fed pressurized suits are described. Results are given of the internal doses to project staff and of atmospheric discharges of tritium during the repair outage. (P.A.)

  8. Tritium monitoring equipments for animal experiment facilities

    International Nuclear Information System (INIS)

    Sato, Hiroo

    1980-01-01

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

  9. Tritium Systems Test Assembly operator training program

    International Nuclear Information System (INIS)

    Kerstiens, F.L.

    1985-01-01

    Proper operator training is needed to help ensure the safe operation of fusion facilities by personnel who are qualified to carry out their assigned responsibilities. Operators control and monitor the Tritium Systems Test Assembly (TSTA) during normal, emergency, and maintenance phases. Their performance is critical both to operational safety, assuring no release of tritium to the atmosphere, and to the successful simulation of the fusion reaction progress. Through proper training we are helping assure that TSTA facility operators perform their assignments in a safe and efficient manner and that the operators maintain high levels of operational proficiency through continuing training, retraining, requalification, and recertification

  10. Tritium calorimeter setup and operation

    CERN Document Server

    Rodgers, D E

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-02-01

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

  12. Tritium calorimeter setup and operation

    International Nuclear Information System (INIS)

    Rodgers, David E.

    2002-01-01

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

  13. Five years of tritium handling experience at the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    Carlson, R.V.

    1989-01-01

    The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory is a facility designed to develop and demonstrate, in full scale, technologies necessary for safe and efficient operation of tritium systems required for tokamak fusion reactors. TSTA currently consists of systems for evacuating reactor exhaust gas with compound cryopumps; for removing impurities from plasma exhaust gas and recovering the chemically-combined tritium; for separating the isotopes of hydrogen; for transfer pumping; or storage of hydrogen isotopes; for gas analysis; and for assuring safety by the necessary control, monitoring, and tritium removal from effluent streams. TSTA also has several small scale experiments to develop and test new equipment and processes necessary for fusion reactors. In this paper, data on component reliability, failure types and rates, and waste quantities are presented. TSTA has developed a Quality Assurance program for preparing and controlling the documentation of the procedures required for the design, purchase, and operation of the tritium systems. Operational experience under normal, abnormal, and emergency conditions is presented. One unique aspect of operations at TSTA is that the design personnel for the TSTA systems are also part of the operating personnel. This has allowed for the relatively smooth transition from design to operations. TSTA has been operated initially as a research facility. As the system is better defined, operations are proceeding toward production modes. The DOE requirements for the operation of a tritium facility like TSTA include personnel training, emergency preparedness, radiation protection, safety analysis, and preoperational appraisals. The integration of these requirements into TSTA operations is discussed. 4 refs., 3 figs., 3 tabs

  14. Experiences with decontaminating tritium-handling apparatus

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  15. TSTA loop operation with 100 grams-level of tritium

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Hirata, Shingo; Naito, Taisei

    1988-10-01

    The first loop operation tests of Tritium Systems Test Assembly (TSTA) with 100 grams-level of tritium were carried out at Los Alamos National Laboratory(LANL) on June and July, 1987. The tests were one of the milestones for TSTA goal scheduled in June, 1987 through June, 1988. The objectives were (i) to operate TSTA process loop composed of tritium supply system, fuel gas purification system, hydrogen isotope separation system, etc, (ii) to demonstrate TSTA safety subsystems such as secondary containment system, tritium waste treatment system and tritium monitoring system, and (iii) to accumulate handling experience of a large amount of tritium. This report describes the plan and procedures of the milestone run done in June and the summary results especially on the safety aspects. Analysis of the emergency shutdown of the process loop, which happened in the June run, is also reported. A brief description of the process and safety subsystems as well as the summary of the TSTA safety analysis report is included. (author)

  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. A tritium vessel cleanup experiment in TFTR

    International Nuclear Information System (INIS)

    Caorlin, M.; Kamperschroer, J.; Owens, D.K.; Voorhees, D.; Mueller, D.; Ramsey, A.T.; La Marche, P.H.; Loughlin, M.J.

    1995-03-01

    A simple tritium cleanup experiment was carried out in TFTR following the initial high power deuterium-tritium discharges in December 1993. A series of 34 ohmic and deuterium neutral beam fueled shots was used to study the removal of tritium implanted into the wall and limiters. A very large plasma was created in each discharge to ''scrub'' an area as large as possible. Beam-fueled shots at 2.5 to 7.5 MW of injected power were used to monitor tritium concentration levels in the plasma by detection of DT-neutrons. The neutron signal decreased by a factor of 4 during the experiment, remaining well above the expected T-burnup level. The amount of tritium recovered at the end of the cleanup was about 8% of the amount previously injected with high power DT discharges. The experience gained suggests that measurements of tritium inventory in the torus are very difficult to execute and require dedicated systems with overall accuracy of 1%

  18. Tritium proof-of-principle injector experiment

    International Nuclear Information System (INIS)

    Fisher, P.W.; Milora, S.L.; Combs, S.K.; Carlson, R.V.; Coffin, D.O.

    1988-01-01

    The Tritium Proof-of-Principle (TPOP) pellet injector was designed and built by Oak Ridge National Laboratory (ORNL) to evaluate the production and acceleration of tritium pellets for fueling future fision reactors. The injector uses the pipe-gun concept to form pellets directly in a short liquid-helium-cooled section of the barrel. Pellets are accelerated by using high-pressure hydrogen supplied from a fast solenoid valve. A versatile, tritium-compatible gas-handling system provides all of the functions needed to operate the gun, including feed gas pressure control and flow control, plus helium separation and preparation of mixtures. These systems are contained in a glovebox for secondary containment of tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). 18 refs., 3 figs

  19. Transfer operations with tritium: a review

    International Nuclear Information System (INIS)

    Folkers, C.L.; Gede, V.P.

    1975-01-01

    Controlled thermonuclear reactors will involve pumping operations with tritium that may involve pressures ranging from submillipascals to megapascals. A variety of pumps are available that can cover portions of this range, and these can be staged to cover the entire pressure range. Some of these pumps can be adapted to virtually any size requirement currently anticipated. Special attention must be paid to operating features and construction materials. (U.S.)

  20. Design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

    International Nuclear Information System (INIS)

    Finn, P.A.; Reedy, G.T.; Homa, M.I.; Clemmer, R.G.; Pappas, G.; Slawecki, M.A.; Graczyk, D.G.; Bowers, D.L.; Clemmer, E.D.

    1983-01-01

    The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

  1. Operational Readiness Review: Savannah River Replacement Tritium Facility

    International Nuclear Information System (INIS)

    1993-02-01

    The Operational Readiness Review (ORR) is one of several activities to be completed prior to introducing tritium into the Replacement Tritium Facility (RTF) at the Savannah River Site (SRS). The Secretary of Energy will rely in part on the results of this ORR in deciding whether the startup criteria for RTF have been met. The RTF is a new underground facility built to safely service the remaining nuclear weapons stockpile. At RTF, tritium will be unloaded from old components, purified and enriched, and loaded into new or reclaimed reservoirs. The RTF will replace an aging facility at SRS that has processed tritium for more than 35 years. RTF has completed construction and is undergoing facility startup testing. The final stages of this testing will require the introduction of limited amounts of tritium. The US Department of Energy (DOE) ORR was conducted January 19 to February 4, 1993, in accordance with an ORR review plan which was developed considering previous readiness reviews. The plan also considered the Defense Nuclear Facilities Safety Board (DNFSB) Recommendations 90-4 and 92-6, and the judgements of experienced senior experts. The review covered three major areas: (1) Plant and Equipment Readiness, (2) Personnel Readiness, and (3) Management Systems. The ORR Team was comprised of approximately 30 members consisting of a Team Leader, Senior Safety Experts, and Technical Experts. The ORR objectives and criteria were based on DOE Orders, industry standards, Institute of Nuclear Power Operations guidelines, recommendations of external oversight groups, and experience of the team members

  2. Experiments on tritium behavior in beryllium, (1)

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  3. Design and operations at the National Tritium Labelling Facility

    International Nuclear Information System (INIS)

    Morimoto, H.; Williams, P.G.

    1991-09-01

    The National Tritium Labelling Facility (NTLF) is a multipurpose facility engaged in tritium labeling research. It offers to the biomedical research community a fully equipped laboratory for the synthesis and analysis of tritium labeled compounds. The design of the tritiation system, its operations and some labeling techniques are presented

  4. Tritium operating safety seminar, Los Alamos, New Mexico, July 30, 1975

    International Nuclear Information System (INIS)

    1976-03-01

    A seminar for the exchange of information on tritium operating and safety problems was held at the Los Alamos Scientific Laboratory. The topics discussed are: (1) material use (tubing, lubricants, valves, seals, etc.); (2) hardware selection (valves, fittings, pumps, etc.); (3) biological effects; (4) high pressure; (5) operating procedures (high pressure tritium experiment at LLL); (6) incidents; and (7) emergency planning

  5. Experiences with decontaminating tritium-handling apparatus

    International Nuclear Information System (INIS)

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

    1991-07-01

    Tritium-handling apparatus has been decontaminated as part of the shutdown of the LLNL Tritium Facility. Two stainless-steel gloveboxes that had been used to process lithium deuteride-tritide (LiDT) salt were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. Further surface decontamination was performed by scrubbing the interior with paper towels and ethyl alcohol or Swish trademark. The surface contamination, as shown by swipe surveys, was reduced from 4x10 4 --10 6 disintegrations per minute (dpm)/cm 2 to 2x10 2 --4x10 4 dpm/cm 2 . Details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given

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

    International Nuclear Information System (INIS)

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

    1986-10-01

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

  7. Analysis of in-pile tritium release experiments

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. In-pile tritium permeation experiment

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  9. In-pile tritium permeation experiment

    Energy Technology Data Exchange (ETDEWEB)

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

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

  10. 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)

  11. The design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

    International Nuclear Information System (INIS)

    Finn, P.A.; Bowers, D.L.; Clemmer, E.D.; Clemmer, R.G.; Graczyk, D.G.; Homa, M.I.; Pappas, G.; Reedy, G.T.; Slawecki, M.A.

    1983-01-01

    The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time; unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. Results from deuterium-tritium tokamak confinement experiments

    International Nuclear Information System (INIS)

    Hawryluk, R.J.

    1997-02-01

    Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which enabled the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation which have both good confinement and are MHD stable have enabled a broad study of burning plasma physics issues. A review of the technical and scientific results from the deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) is given with particular emphasis on alpha-particle physics issues

  17. Preparations for deuterium tritium experiments on the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.; Barnes, G.

    1994-04-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR). These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinet trademark system, modification of the vacuum system to handle tritium, preparation and testing of the neutral beam system for tritium operation and a final deuterium-deuterium (D-D) run to simulate expected deuterium-tritium (D-T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D-T experiments using D-D have been performed. The physics objectives of D-T operation are production of ∼ 10 megawatts (MW) of fusion power, evaluation of confinement and heating in deuterium-tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α-particles. Experimental results and theoretical modeling in support of the D-T experiments are reviewed

  18. Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

    International Nuclear Information System (INIS)

    Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Jaeger, E.F.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Laughlin, M.J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D.E.; Marchlik, M.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M.P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Milora, S.L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Paul, S.F.; Pavlov, Y.I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Pitcher, S.; Popovichev, S.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.

    1994-01-01

    The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert TM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of ∼10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed

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

    International Nuclear Information System (INIS)

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

    1983-08-01

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

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

    International Nuclear Information System (INIS)

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

    1983-10-01

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

  1. Operation of the TSTA (Tritium Systems Test Assembly) with 100 gram tritium

    International Nuclear Information System (INIS)

    Sherman, R.H.; Bartlit, J.R.

    1988-01-01

    In March of 1988 full operation of the 4-column isotope separation system (ISS) was realized in runs that approximated the design load of tritium. Previous operations had been fraught with operating difficulties principally due to external systems. This report will examine the recent highly successful 6-day period of operation. During this time the system was cooled from room temperature, loaded with hydrogen isotopes including 109 grams of tritium, integrated with the transfer pumping, impurity injection, and impurity removal systems, as well as the remote computer control system. At the end of the operation 12 grams of tritium having a measured purity of 99.987% (remainder deuterium) were offloaded from the system. Observed profiles in the columns in general agree with computer models. A Height Equivalent to a Theoretical Plate (HETP) of 5.0 cm is confirmed. 3 refs., 5 figs

  2. Tritium depth profiling in carbon samples from fusion experiments

    International Nuclear Information System (INIS)

    Friedrich, M.; Pilz, W.; Sun, G.; Behrisch, R.; Garcia-Rosales, C.; Bekris, N.; Penzhorn, R.-D.

    2000-01-01

    Tritium depth profiling by accelerator mass spectrometry has been performed at the Rossendorf 3 MV Tandetron. Tritium particles are counted after the accelerator using a semiconductor detector, while deuterium and other light elements are simultaneously measured with the Faraday cup between the injection magnet and the accelerator. Depth profiles have been measured in carbon samples cut from the first wall tiles of the Garching fusion experiment ASDEX-Upgrade and of the European fusion experiment JET, Culham/UK. Tritium contents in the JET samples were up to six orders higher than in samples from ASDEX-Upgrade. Tritium beam currents from samples with high tritium content were measured partly in the Faraday cup before the accelerator. A dedicated tritium AMS facility with an air-insulated 100 kV tandem accelerator is under construction

  3. Accounting and Control of Tritium at the Tritium Process Laboratory (TPL) of JAERI - Results of 15-year Operation and Research Activity -

    Science.gov (United States)

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

    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 throughits15-yearsafe-operation. For future fusion fuel facility where kilograms 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.

  4. Conception and operation of a 10 kCi liquid tritium target for the study of tritium nucleus by electron diffusion: 3H(e,e)

    International Nuclear Information System (INIS)

    Juster, F.P.

    1986-06-01

    We describe the conception and operation of an experimental setup, specially suited for the study of the nuclear structure of tritium by elastic electron scattering at intermediate energy. The experiment has been conducted at the ALS 700 MeV electron linac (Saclay, France). The radioactive nature of tritium has led to the design of a new target, suited for handling reliably ten kilocuries of tritium (one gram). The tritium is contained in three sealed envelopes. Initially a high pressure gas (23 bars) at room temperature, the tritium is cooled down to liquefaction by thermal conduction through a solid, without breaking any seal. The beam path and the scattered trajectories cross thin metallic windows. Additional protection, during the presence of personnel, is provided by a heavy container, remotely operated. Any leak in the containement vessels is detected by changes in pressure and/or temperature gauges, monitored by two independent processors. These processors handle the operation of the outer container, the beam switching and the spare venting system. No tritium leak has been detected during a total six-week run. The tritium liquefies in a cylindrical target, 5 cm long and 1 cm in diameter. A beam of 10 microamperes, in the 200-700 MeV, has been measured. The charge and magnetic form factors of tritium have been measured up to a momentum transfer of 31.3 fm -2 [fr

  5. 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.)

  6. Observation of tritium in cold fusion experiments

    International Nuclear Information System (INIS)

    Krishnan, M.S.; Malhotra, S.K.; Gaonkar, D.G.; Sadhukhan, H.K.

    1990-01-01

    This paper describes the results of tritium measurements carried out during the electrolysis of heavy water in different electrolysers employing palladium and titanium as cathodes. The tritium level of electrolytes have been found to be many orders of magnitudes higher than what can be explained on the basis of isotope enrichment and evaporation during electrolysis. The neutron measurement results have also been included and these observations have been attributed to the phenomenon of cold fusion. (author). 6 refs., 1 tab

  7. 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

  8. 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

  9. Modeling and experiments on tritium permeation in fusion reactor blankets

    Science.gov (United States)

    Holland, D. F.; Longhurst, G. R.

    The determination of tritium loss from helium-cooled fusion breeding blankets are discussed. The issues are: (1) applicability of present models to permeation at low tritium pressures; (2) effectiveness of oxide layers in reducing permeation; (3) effectiveness of hydrogen addition as a means to lower tritium permeation; and (4) effectiveness of conversion to tritiated water and subsequent trapping to reduce permeation. Theoretical models applicable to these issues are discussed, and results of experiments in two areas are presented; permeation of mixtures of hydrogen isotopes and conversion to tritiated water.

  10. Modeling and experiments on tritium permeation in fusion reactor blankets

    International Nuclear Information System (INIS)

    Holland, D.F.; Longhurst, G.R.

    1985-01-01

    Issues are discussed that are critical in determining tritium loss from helium-cooled fusion breeding blankets. These issues are: (a) applicability of present models to permeation at low tritium pressures, (b) effectiveness of oxide layers in reducing permeation, (c) effectiveness of hydrogen addition as a means to lower tritium permeation, and (d) effectiveness of conversion to tritiated water and subsequent trapping as a means to reduce permeation. The paper discusses theoretical models applicable to these issues, and presents results of experiments in two areas: permeation of mixtures of hydrogen isotopes and conversion to tritiated water

  11. 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)

  12. 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)

  13. 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

  14. 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)

  15. 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.

  16. 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.

  17. Exposure Control for Operations and Maintenance at the Accelerator Production of Tritium

    International Nuclear Information System (INIS)

    McGuire, D.H.

    1998-09-01

    The APT will be designed and operated to support continuous tritium production. Tritium is an essential ingredient in U.S. nuclear weapons. The APT will be designed and staffed to support continuous production of tritium by trained, qualified, and certified personnel

  18. Status of the Los Alamos tritium beta decay experiment

    International Nuclear Information System (INIS)

    Robertson, R.G.H.; Bowles, T.J.; Wark, D.L.; Wilkerson, J.F.; Knapp, D.A.

    1989-01-01

    The Los Alamos tritium experiment employs a gaseous tritium source and a magnetic spectrometer to determine the mass of the electron antineutrino from the shape of the tritium beta spectrum. Since publication of the first result from this apparatus (m/sub nu/ < 27 eV at 95% confidence), work has concentrated on improving the data rates. A 96-element Si microstrip array detector has been installed to replace the single proportional counter at the spectrometer focus, resulting in greatly increased efficiency. Measurements of the 1s photoionization spectrum of Kr now obviate the need for reliance on the theoretical shakeup and shakeoff spectrum of Kr in determining the spectrometer resolution. 19 refs., 3 figs

  19. Tritium depth profiling by AMS in carbon samples from fusion experiments

    International Nuclear Information System (INIS)

    Friedrich, M.; Pilz, W.; Sun, G.; Behrisch, R.; Garcia-Rosales, C.

    2001-01-01

    Tritium depth profiling measurements by accelerator mass spectrometry have been performed at a facility installed at the Rossendorf 3 MV Tandetron. In order to achieve an uniform erosion at the target surface inside of a commercial Cs ion sputtering source and to avoid edge effects, the samples were mechanically scanned inside of a commercial Cs sputter ion source. The sputtered negative ions were mass analysed by the injection magnet of the Tandetron. The tritium ions are counted after the acceleration with semiconductor detectors. Depth profiles have been measured for carbon samples which had been exposed to the plasma at the first wall of the Garching fusion experiment ASDEX-Upgrade and from the European fusion experiment JET, Culham/UK. A dedicated AMS facility with an air-insulated 100 kV tandem accelerator for depth profiling measurements at samples with high tritium concentration is under construction. First results of test operation are presented. (orig.)

  20. 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.)

  1. Hydrogen, deuterium, and tritium isotope exchange experiments in JET

    Energy Technology Data Exchange (ETDEWEB)

    Horton, L.D.; Andrew, P.; Bracco, G.; Conroy, S.; Corti, S.; Ehrenberg, J.; Goodall, D.H.J.; Jarvis, O.N.; Lomas, P.; Loughlin, M.; Peacock, A.T.; Saibene, G.; Sadler, G.; Sartori, R.; Stamp, M.F.; Thomas, P.R.; Belle, P. van (JET Joint Untertaking, Abingdon, Oxfordshire (United Kingdom))

    1992-12-01

    Isotope exchange experiments have been performed in JET using hydrogen, deuterium, and, in the recent preliminary tritium experiment (PTE), tritium. The rate of change-over from one isotope to another involves two quite different time constants. We have modelled this behaviour using a multireservoir model which splits the accessible hydrogenic particles into two groups, each having a different rate of exchange of particles with the plasma. By applying this model to the sequence of discharges during and after the PTE, we can determine the parameters in the model. The resulting fit also gives a good representation of hydrogen/deuterium change-over experiments, indicating that the tritium behaves in the same manner as other hydrogen isotopes, at least as far as recycling is concerned. Discrepancies between the model and the actual measurements of tritium recovery after the PTE lead us to conclude that isotope exchange processes resulting from collisions of molecules with the vessel walls play a significant role in spreading tritrium around the machine. (orig.).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  3. Cleanup of a Department of Energy Nonreactor Nuclear Facility: Experience at the Los Alamos National Laboratory High Pressure Tritium Laboratory

    International Nuclear Information System (INIS)

    Horak, H.L.

    1995-01-01

    On October 25, 1990, Los Alamos National Laboratory (LANL) ceased programmatic operations at the High Pressure Tritium Laboratory (HPTL). Since that time, LANL has been preparing the facility for transfer into the Department of Energy's (DOE's) Decontamination and Decommissioning Program. LANL staff now has considerable operational experience with the cleanup of a 40-year-old facility used exclusively to conduct experiments in the use of tritium, the radioactive isotope of hydrogen. Tritium and its compounds have permeated the HPTL structure and equipment, have affected operations and procedures, and now dominate efforts at cleanup and disposal. At the time of shutdown, the HPTL still had a tritium inventory of over 100 grams in a variety of forms and containers

  4. Results of preliminary experiments on tritium decontamination by UV irradiation

    International Nuclear Information System (INIS)

    Oya, Yasuhisa; Shu, Wataru; O'hira, Shigeru; Hayashi, Takumi; Nishi, Masataka

    2000-03-01

    In the point of view of protection of workers from the radiation exposure and the limitation of the contamination with radioactive materials, it is important to decontaminate mobile tritium from plasma facing components of a nuclear fusion reactor at the beginning of their maintenance work. It is considered that the heating is the most effective method for decontamination. However, it is important to develop new decontamination method of adsorbed hydro-carbon based substances from the materials that cannot be heated or the inner pipe of double pipes. This report presents results of preliminary experiments performed for the development of the effective tritium decontamination technique pursuing under US/Japan collaborative program on technology for fusion-fuel processing (Annex IV). In the experiments, the effects of Ultra Violet (UV) irradiation on tritium removal from some kinds of materials, such as poly vinyl chloride -(CH 2 CHCl) n - film, polyethylene film and graphite samples coated by C 2 H 2 plasma were examined. As the result of UV irradiation, it was confined that hydrogen and carbon based compounds could be released from the specimen during UV irradiation. It is concluded that UV irradiation is one of the hopeful candidates for effective tritium decontamination. (author)

  5. 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

  6. TPE upgrade for enhancing operational safety and improving in-vessel tritium inventory assessment in fusion nuclear environment

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M., E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Taylor, C.N.; Moore-McAteer, L.; Pawelko, R.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Kolasinski, R.D.; Buchenauer, D.A. [Sandia National Laboratories, Hydrogen and Materials Science Department, Livermore, CA 94550 (United States); Cadwallader, L.C.; Merrill, B.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

    2016-11-01

    The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γ{sub i}{sup max} of >1.0 × 10{sup 23} m{sup −2} s{sup −1} and q{sub heat} of >1 MW m{sup −2} with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

  7. Characterization of erosion dust and tritiated products inside the jet vessel after the first tritium experiment

    International Nuclear Information System (INIS)

    Charuau, J.; Belot, Y.; Cetier, P.; Drezet, L.; Grivaud, L.

    1992-01-01

    An experiment was carried out to characterize the erosion products found in the JET vessel after the first tritium experiment. These products were analyzed for carbon, beryllium, Inconel metals and tritium. All these elements were present in airborne particles or deposited dust. The tritium was found as tritiated water vapour, and also strongly associated to the suspended or deposited particles. It was more abundant in fine than in coarse particles. The particulate tritium seems to be almost entirely 'insoluble' in a water solution

  8. Tritium handling experience in vacuum systems at TSTA [Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    Anderson, J.L.; Jenkins, E.M.; Walthers, C.R.; Yoshida, H.; Fukui, H.; Naruse, Y.

    1989-01-01

    Compound cryopumps have been added to the Tritium Systems Test Assembly (TSTA) integrated fusion fuel loop. Operations have been performed which closely simulate an actual fusion reactor pumping scenario. In addition, performance data have been taken that support the concept of using coconut charcoal as a sorbent at 4K for pumping helium. Later tests show that coconut charcoal may be used to co-pump D,T and He mixtures on a single 4K panel. Rotary spiral pumps have been used successfully in several applications at TSTA and have acquired more than 9000 hours of maintenance-free operation. Metal bellows pumps have been used to back the spiral pumps and have been relatively trouble free in loop operations. Bellows pumps also have more than 9000 hours of maintenance-free operation. 5 refs., 6 figs

  9. Operating experience

    International Nuclear Information System (INIS)

    McRae, L.P.; Six, D.E.

    1991-01-01

    In 1987, Westinghouse Hanford Company began operating a first-generation integrated safeguards system in the Plutonium Finishing Plant storage vaults. This Vault Safety and Inventory System is designed to integrate data into a computer-based nuclear material inventory monitoring system. The system gathers, in real time, measured physical parameters that generate nuclear material inventory status data for thousands of stored items and sends tailored report to the appropriate users. These data include canister temperature an bulge data reported to Plant Operations and Material Control and Accountability personnel, item presence and identification data reported to Material Control and Accountability personnel, and unauthorized item movement data reported to Security response forces and Material Control and Accountability personnel. The Westinghouse Hanford Company's experience and operational benefits in using this system for reduce radiation exposure, increase protection against insider threat, and real-time inventory control are discussed in this paper

  10. 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

  11. The MOZART in-pile tritium extraction experiment

    International Nuclear Information System (INIS)

    Briec, M.

    1990-01-01

    In-pile tritium extraction behavior of various ceramics was compared in the MOZART experiment. The influence of temperature and purge gas composition was studied. The experimental results are analyzed by taking into account the processes of diffusion in the grain and desorption at grain surface. This analysis confirms that a better knowledge of the desorption process is necessary for a satisfactory explanation of the experimental data

  12. Progress report on the Los Alamos tritium beta decay experiment

    International Nuclear Information System (INIS)

    Wilkerson, J.F.; Bowles, T.J.; Knapp, D.A.; Robertson, R.G.H.; Wark, D.L.

    1988-01-01

    Measurements near the endpoint of the tritium beta-decay spectrum using a gaseous molecular tritium source yield an essentially model-independent upper limit of 27 eV on the /ovr ν//sub e/ mass at the 95% confidence level. Since demonstrating from this initial measurement the successful operation of a gaseous source based system, most of our effort has been concentrated towards the upgrade and optimization of the experimental apparatus. The emphasis of this work has been to eliminate or further reduce effects that generate systematic errors. Based on realistic projections from our initial measurement, an ultimate sensitivity to neutrino mass of 10 eV is expected. 12 refs., 1 fig

  13. In-pile tritium permeation experiment

    International Nuclear Information System (INIS)

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

    1982-01-01

    To examine radiation and implantation effects simultaneously, an experiment has been designed which makes use of the Coupled Fast Reactivity Measurement Facility (CFRMF), a small pool reactor at the INEL. The neutron flux is low in this reactor, but the high cross section (5300 b) for the 3 He(n,p) 3 H reaction with thermal neutrons gives a sufficiently intense flux of protons and tritons to a simulated fusion first wall for meaningful results

  14. Tritium experiments on components for fusion fuel processing at the Tritium Systems Test Assembly

    International Nuclear Information System (INIS)

    Konishi, S.; Yoshida, H.; Naruse, Y.; Carlson, R.V.; Binning, K.E.; Bartlit, J.R.; Anderson, J.L.

    1990-01-01

    Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, ''JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing

  15. 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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. Tritium handling experience at Atomic Energy of Canada Limited

    Energy Technology Data Exchange (ETDEWEB)

    Suppiah, S.; McCrimmon, K.; Lalonde, S.; Ryland, D.; Boniface, H.; Muirhead, C.; Castillo, I. [Atomic Energy of Canad Limited - AECL, Chalk River Laboratories, Chalk River, ON (Canada)

    2015-03-15

    Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritium powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.

  19. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  20. TSTA Piping and Flame Arrestor Operating Experience Data

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee C.; Willms, R. Scott

    2014-10-01

    The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility operated from 1984 to 2001, running a prototype fusion fuel processing loop with ~100 grams of tritium as well as small experiments. There have been several operating experience reports written on this facility’s operation and maintenance experience. This paper describes analysis of two additional components from TSTA, small diameter gas piping that handled small amounts of tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The operating experiences and the component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.

  1. Tritium source-related systematic uncertainties of the KATRIN experiment

    Energy Technology Data Exchange (ETDEWEB)

    Seitz-Moskaliuk, Hendrik [Karlsruher Institut fuer Technologie, Institut fuer experimentelle Kernphysik, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Collaboration: KATRIN-Collaboration

    2016-07-01

    KATRIN will perform a direct, kinematics-based measurement of the neutrino mass with a sensitivity of 200 meV (90 % C. L.) reached after 3 years of measurement time. The neutrino mass is obtained by determining the shape of the spectrum of tritium β decay electrons close to the endpoint of 18.6 keV with a spectrometer of MAC-E filter type. To achieve the planned sensitivity, the systematic measurement uncertainties have to be carefully controlled and evaluated. Main sources of systematics are the MAC-E filter on the one hand and the source and transport section (STS) on the other hand. Most of the operational parameters of KATRIN have to be stable at or even below the per mille level and have to meet further strict requirements. This talk reviews the KATRIN systematics with a special focus on the STS. Early commissioning measurements to determine the main systematics are introduced.

  2. Analysis of simulated data for the KArlsruhe TRItium Neutrino experiment using Bayesian inference

    DEFF Research Database (Denmark)

    Riis, Anna Sejersen; Hannestad, Steen; Weinheimer, C.

    2011-01-01

    The KATRIN (Karlsruhe Tritium Neutrino) experiment will analyze the tritium β spectrum to determine the mass of the neutrino with a sensitivity of 0.2 eV (90% C.L.). This approach to a measurement of the absolute value of the neutrino mass relies only on the principle of energy conservation and can...

  3. TSTA piping and flame arrestor operating experience data

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee C., E-mail: Lee.Cadwallader@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Willms, R. Scott [ITER International Organization, Cadarache (France)

    2015-10-15

    Highlights: • Experiences from the Tritium Systems Test Assembly were examined. • Failure rates of copper piping and a flame arrestor were calculated. • The calculated failure rates compared well to similar data from the literature. • Tritium component failure rate data support fusion safety assessment. - Abstract: The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility was operated with tritium for its research and development program from 1984 to 2001, running a prototype fusion fuel processing loop with ∼100 g of tritium as well as small experiments. There have been several operating experience reports written on this facility's operation and maintenance experience. This paper describes reliability analysis of two additional components from TSTA, small diameter copper gas piping that handled tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.

  4. Tritium processing and management during D-T experiments on TFTR

    International Nuclear Information System (INIS)

    La Marche, P.H.; Anderson, J.L.; Gentile, C.A.; Hawryluk, R.J.; Hosea, J.; Kalish, M.; Kozub, T.; Murray, H.; Nagy, A.; Raftopoulos, S.

    1994-11-01

    TFTR performance has surpassed many of the previous tokamak records. This has been made possible by the use of tritium as fuel for DT plasma discharges. Stable operations of tritium systems provide for safe, routine DT operation of TFTR. In the preparation for DT operation, in the commissioning of the tritium systems and in the operation of the Nuclear Facility several key lessons have been learned. They include: the facility must take the lead in interpreting the applicable regulations and orders and then seek regulator approval; the use of ultra high vacuum technology in tritium system design and construction simplifies and enhances operations and maintenance; and central facility control under a single supervisory position is crucial to safely orchestrate operational and maintenance activities

  5. Tritium breeding experiments with lithium titanate in thermal-type mockups

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  6. Mesocosm experiments on tritium dynamics in carp fish

    International Nuclear Information System (INIS)

    Reji, T.K.; Vishnu, M.S.; Joshi, R.M.; Dileep, B.N.; Baburajan, A.; Ravi, P.M.

    2013-01-01

    Tritium dynamics in carp fish (Cyprinus carpio) was studied in a locally designed mesocosm simulating a lake condition. The fishes were reared in an experimental tank containing tritiated water. Tissue Free water tritium (TFWT) concentration and Organically Bound Tritium (OBT) was measured for 3 months period. TFWT reached equilibrium with exposed water within one day. Detectable amount of OBT was observed after two months of exposure. OBT to TFWT ratio was 0.1. Estimated OBT was in agreement with that calculated using IAEA specific activity model. (author)

  7. Systems for the safe operation of the JET tokamak with tritium

    International Nuclear Information System (INIS)

    Stork, D.; Ageladarakis, P.; Bell, A.C.

    1999-01-01

    In 1997, the JET device was operated for an extensive campaign with deuterium-tritium (D-T) plasmas (the DTE1 campaign). A comprehensive network of machine protection systems was necessary so that this experimental campaign could be executed safely without damage to the machine or release of activated material. This network had been developed over many years of JET deuterium plasma operation and therefore the modifications for D-T operation was not a significant problem. The DTE1 campaign was executed successfully and safely and the machine protection systems proved reliable and robust and, in the limited cases where they were required to act, functioned correctly. The machine protection systems at JET are described and their categorisation and development over time are summarised. The management, commissioning and operational experience during DTE1 are discussed and some examples of fault scenarios are described. The experience with protection systems at JET highlights the importance of correct design and philosophy decisions being taken at an early stage. It is shown that this experience will be invaluable data input to the safe operation of future large fusion machines. (orig.)

  8. Experiment designed to measure the RBE of tritium for the induction of myeloid leukaemia in animals

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J R; Myers, D K; Gragtmans, N J

    1986-01-01

    The range in RBE vales measured for tritium can be attributed to differences in the biological endpoints measured, the reference radiation to which the effects of tritium were compared, and the tritium dosimetry of the particular study. Since the principal risk of low-level irradiation is the induction of cancers, it would be desirable to utilise this endpoint in tritium RBE experiments if these experiments are to be used to evaluate the quality factor for tritium. Furthermore, it would be desirable to use 200 k Vp X rays as the reference radiation since this radiation was suggested by ICRP as the standard reference to be used in the calculation of dose equivalents. Acute myeloid leukaemia is one of the earliest recognised examples of radiogenic cancer in humans and this endpoint has also been the subject of animal studies. A brief review is given of these animal studies to see if this endpoint is suitable for an experiment to measure the tritium RBE relative to 200 k Vp X rays. It was concluded that the male CBA/H mouse would be a suitable species and an experiment involving 5000 animals in four to five year study would be required to provide a useful estimate of the RBE for tritium.

  9. Experiences made with tritium-containing water used as tracer in laboratory experiments with fluvioglacial gravels

    International Nuclear Information System (INIS)

    Klotz, D.; Rauert, W.

    1982-01-01

    Batch tests performed on 11 different Bavarian fluvioglacial gravels led to tritium distribution coefficients, which deviated not or only insignificantly from zero within the range of experimental accuracy applied to routine testings. The result of nine flow experiments in a gravelfilled column was a mean retardation factor of 1.01 +- 0.01. These experiments thus showed - as it had been expected - that 3 HHO is not significantly delayed with regard to the flow or movement of the water. (orig.) [de

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

    International Nuclear Information System (INIS)

    Raskob, W.; Hasemann, I.

    1992-08-01

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

  11. A new Disruption Mitigation System for deuterium–tritium operation at JET

    Energy Technology Data Exchange (ETDEWEB)

    Kruezi, Uron, E-mail: uron.kruezi@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Jachmich, Stefan [Laboratory for Plasma Physic, ERM/KMS, B-1000 Brussels (Belgium); Koslowski, Hans Rudolf [Forschungszentrum Jülich GmbH, IEK-4, 52425 Jülich (Germany); Lehnen, Michael [ITER Organization, Route de Vinon-sur-Verdon, CS90046, 13067 St. Paul Lez Durance Cedex (France); Brezinsek, Sebastijan [Forschungszentrum Jülich GmbH, IEK-4, 52425 Jülich (Germany); Matthews, Guy [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2015-10-15

    Highlights: • A Disruption Mitigation System based on massive gas injections has been designed. • The DMS has been installed at the JET-tokamak for routine machine protection. • The DMS is capable of a throughput of up to 4.6 kPa m{sup 3}. • The new DMS is compatible with the deuterium–tritium operation at JET. - Abstract: Disruptions, the fast accidental losses of plasma current and stored energy in tokamaks, represent a significant risk to the mechanical structure as well as the plasma facing components of reactor-scale fusion facilities like ITER. At JET, the tokamak experiment closest to ITER in terms of operating parameters and size, massive gas injection has been established as a disruption mitigation method. As a “last resort” measure it reduces thermal and electromagnetic loads during disruptions which can potentially have a serious impact on the beryllium and tungsten plasma-facing materials of the main chamber and divertor. For the planned deuterium–tritium experiments, a new Disruption Mitigation System (DMS) has been designed and installed and is presented in this article. The new DMS at JET consists of an all metal gate valve compatible with gas injections, a fast high pressure eddy current driven valve, a high voltage power supply and a gas handling system providing six supply lines for pure and mixed noble and flammable gases (Ar, Ne, Kr, D{sub 2}, etc.). The valve throughput varies with the injection pressure and gas type (efficiency – injected/charged gas 50–97%); the maximum injected amount of gas is approximately 4.6 kPa m{sup 3} (at maximum system pressure of 5.0 MPa).

  12. A new Disruption Mitigation System for deuterium–tritium operation at JET

    International Nuclear Information System (INIS)

    Kruezi, Uron; Jachmich, Stefan; Koslowski, Hans Rudolf; Lehnen, Michael; Brezinsek, Sebastijan; Matthews, Guy

    2015-01-01

    Highlights: • A Disruption Mitigation System based on massive gas injections has been designed. • The DMS has been installed at the JET-tokamak for routine machine protection. • The DMS is capable of a throughput of up to 4.6 kPa m"3. • The new DMS is compatible with the deuterium–tritium operation at JET. - Abstract: Disruptions, the fast accidental losses of plasma current and stored energy in tokamaks, represent a significant risk to the mechanical structure as well as the plasma facing components of reactor-scale fusion facilities like ITER. At JET, the tokamak experiment closest to ITER in terms of operating parameters and size, massive gas injection has been established as a disruption mitigation method. As a “last resort” measure it reduces thermal and electromagnetic loads during disruptions which can potentially have a serious impact on the beryllium and tungsten plasma-facing materials of the main chamber and divertor. For the planned deuterium–tritium experiments, a new Disruption Mitigation System (DMS) has been designed and installed and is presented in this article. The new DMS at JET consists of an all metal gate valve compatible with gas injections, a fast high pressure eddy current driven valve, a high voltage power supply and a gas handling system providing six supply lines for pure and mixed noble and flammable gases (Ar, Ne, Kr, D_2, etc.). The valve throughput varies with the injection pressure and gas type (efficiency – injected/charged gas 50–97%); the maximum injected amount of gas is approximately 4.6 kPa m"3 (at maximum system pressure of 5.0 MPa).

  13. 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.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  16. The tokamak fusion test reactor tritium systems test contractor operational readiness review

    International Nuclear Information System (INIS)

    Gentile, C.A.; Levine, J.; Norris, M.; Rehill, F.; Such, C.

    1993-01-01

    In preparation for D-T operations at TFTR, the TFTR project has successfully completed the C-ORR process which has led to the introduction of 200 curies of tritium to the site. Preparations for the C-ORR began approximately 2 years ago. During July 1992 a one-week Site Assistance Review was conducted by the C-ORR , and C-ORR Team consisting of 12 persons, all of whom were outside experts, many of whom were from other facilities within the DOE complex. During the July 1992 Site Assistance Review 201 findings were documented which fell into one of three categories. All of the 109 category one findings which were generated were required to be resolved prior to the introduction of tritium to the TFTR site. On April 5, 1993, the TFTR Tritium System Test C-ORR commenced. The results of the C-ORR as documented in the final report by the C-ORR was that category 1 findings were resolved, and it was the recommendation of the C-ORR Team to the PPPL ES ampersand H Board that TFTR initiate the Tritium Systems Test. DOE (Chicago Operations, Princeton Area Office) concurred with the C-ORR final report and on April 29, 1993, at 12:15 pm tritium was introduced to the TFTR site

  17. Development of a Novel Contamination Resistant Ion Chamber for Process Tritium Measurement and Use in the JET First Trace Tritium Experiment

    International Nuclear Information System (INIS)

    Worth, L.B.C.; Pearce, R.J.H.; Bruce, J.; Banks, J.; Scales, S.

    2005-01-01

    The accuracy of process measurements of tritium with conventional ion chambers is often affected by surface tritium contamination. The measurement of tritium in the exhaust of the JET torus is particularly difficult due to surface contamination with highly tritiated hydrocarbons. JET's first unsuccessful attempt to overcome the contamination problem was to use an ion chamber, with a heating element as the chamber wall so that it could be periodically decontaminated by baking. The newly developed ion chamber works on the principle of minimising the surface area within the boundary of the anode and cathode.This paper details the design of the ion chamber, which utilises a grid of 50-micron tungsten wire to define the ion chamber wall and the collector electrode. The effective surface area which, by contamination, is able to effect the measurement of tritium within the process gas has been reduced by a factor of ∼200 over a conventional ion chamber. It is concluded that the new process ion chamber enables sensitive accurate tritium measurements free from contamination issues. It will be a powerful new tool for future tritium experiments both to improve tritium tracking and to help in the understanding of tritium retention issues

  18. Tritium calibration of the LUX dark matter experiment

    Science.gov (United States)

    Akerib, D. S.; Araújo, H. M.; Bai, X.; Bailey, A. J.; Balajthy, J.; Beltrame, P.; Bernard, E. P.; Bernstein, A.; Biesiadzinski, T. P.; Boulton, E. M.; Bradley, A.; Bramante, R.; Cahn, S. B.; Carmona-Benitez, M. C.; Chan, C.; Chapman, J. J.; Chiller, A. A.; Chiller, C.; Currie, A.; Cutter, J. E.; Davison, T. J. R.; de Viveiros, L.; Dobi, A.; Dobson, J. E. Y.; Druszkiewicz, E.; Edwards, B. N.; Faham, C. H.; Fiorucci, S.; Gaitskell, R. J.; Gehman, V. M.; Ghag, C.; Gibson, K. R.; Gilchriese, M. G. D.; Hall, C. R.; Hanhardt, M.; Haselschwardt, S. J.; Hertel, S. A.; Hogan, D. P.; Horn, M.; Huang, D. Q.; Ignarra, C. M.; Ihm, M.; Jacobsen, R. G.; Ji, W.; Kazkaz, K.; Khaitan, D.; Knoche, R.; Larsen, N. A.; Lee, C.; Lenardo, B. G.; Lesko, K. T.; Lindote, A.; Lopes, M. I.; Malling, D. C.; Manalaysay, A. G.; Mannino, R. L.; Marzioni, M. F.; McKinsey, D. N.; Mei, D.-M.; Mock, J.; Moongweluwan, M.; Morad, J. A.; Murphy, A. St. J.; Nehrkorn, C.; Nelson, H. N.; Neves, F.; O'Sullivan, K.; Oliver-Mallory, K. C.; Ott, R. A.; Palladino, K. J.; Pangilinan, M.; Pease, E. K.; Phelps, P.; Reichhart, L.; Rhyne, C.; Shaw, S.; Shutt, T. A.; Silva, C.; Solovov, V. N.; Sorensen, P.; Stephenson, S.; Sumner, T. J.; Szydagis, M.; Taylor, D. J.; Taylor, W.; Tennyson, B. P.; Terman, P. A.; Tiedt, D. R.; To, W. H.; Tripathi, M.; Tvrznikova, L.; Uvarov, S.; Verbus, J. R.; Webb, R. C.; White, J. T.; Whitis, T. J.; Witherell, M. S.; Wolfs, F. L. H.; Young, S. K.; Zhang, C.; LUX Collaboration

    2016-04-01

    We present measurements of the electron-recoil (ER) response of the LUX dark matter detector based upon 170 000 highly pure and spatially uniform tritium decays. We reconstruct the tritium energy spectrum using the combined energy model and find good agreement with expectations. We report the average charge and light yields of ER events in liquid xenon at 180 and 105 V /cm and compare the results to the NEST model. We also measure the mean charge recombination fraction and its fluctuations, and we investigate the location and width of the LUX ER band. These results provide input to a reanalysis of the LUX run 3 weakly interacting massive particle search.

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

    International Nuclear Information System (INIS)

    Carlson, R.V.; Binning, K.E.; Konishi, S.; Yoshida, H.; Naruse, Y.

    1987-01-01

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

  20. PERMCAT experiments with tritium at high helium flow rates relevant for the tritium extraction systems using the CAPER facility at TLK

    Energy Technology Data Exchange (ETDEWEB)

    Bükki-Deme, András, E-mail: andras.buekki-deme@kit.edu; Demange, David; Le, Thanh-Long; Fanghänel, Eleonore; Simon, Karl-Heinz

    2016-11-01

    Highlights: • We examined PERMCAT reactor efficiency processing tritiated water at high Helium carrier flow rates. • We have found that – as expected from previous studies – that the swamping ratio (ratio between the impurity and purge side flow rates) has a key effect on the decontamination factors. • On the other hand, some rather unexpected effects tend to show that the limiting phenomena of such specific operation of PERMCAT reactors (at high impurity flow rates, thus short residence time) lies on the kinetics of the isotope exchange reactions. - Abstract: Experiments are still necessary to consolidate the processes retained for the Tritium Extraction Systems of the European ITER Test Blanket Modules (TBM). A PERMCAT reactor combines a catalyst promoting isotope exchange reactions and a Pd/Ag membrane allowing tritium recovery from complex gaseous mixtures containing tritium in different chemical forms. Originally developed for the Tokamak Exhaust Processing, the PERMCAT process is also candidate to detritiate the water arising from an adsorption column installed in the TBM ancillary systems. We discuss the results of an extensive experimental campaign using a PERMCAT reactor to process Q{sub 2}O containing impurity gas mixtures at high flow rates. Two different experimental configurations were studied, namely PERMCAT stand-alone, and PERMCAT in combination with a zeolite molecular sieve bed (MSB, previously loaded with Q{sub 2}O) under regeneration. On the one hand, many expected behaviors were observed, such as the key influence of the swamping ratio (ratio between the impurity and purge side flow rates) on the decontamination factors. On the other hand, some rather unexpected effects tend to show that the limiting phenomena of such specific operation of PERMCAT reactors (at high flow rates, thus short residence time) lies on the kinetics of the isotope exchange reactions.

  1. Dependence of CuO particle size and diameter of reaction tubing on tritium recovery for tritium safety operation

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Cui, E-mail: cdxohc10000@163.com [Shizuoka University, 836 Ohya, Suruga-ku Shizuoka 422-8529 (Japan); Uemura, Yuki; Yuyama, Kenta; Fujita, Hiroe; Sakurada, Shodai; Azuma, Keisuke [Shizuoka University, 836 Ohya, Suruga-ku Shizuoka 422-8529 (Japan); Taguchi, Akira; Hara, Masanori; Hatano, Yuji [University of Toyama, 3190 Gofuku, Toyama 939-8555 (Japan); Chikada, Takumi; Oya, Yasuhisa [Shizuoka University, 836 Ohya, Suruga-ku Shizuoka 422-8529 (Japan)

    2016-12-15

    Highlights: • Influence of CuO particle size and diameter of reaction tubing on the tritium recovery was evaluated. • Reaction rate constant of tritium with CuO particle has been calculated by the combination of experimental results and a simulation code. • Dependence of reaction tubing length on tritium conversion ratio has been explored. - Abstract: Usage of CuO and water bubbler is one of the conventional and convenient methods for tritium recovery. In present work, influence of CuO particle size and diameter of reaction tubing on the tritium recovery was evaluated. Reaction rate constant of tritium with CuO particle has been calculated by the combination of experimental results and a simulation code. Then, these results were applied for exploring the dependence of reaction tubing length on tritium conversion ratio. The results showed that the surface area of CuO has a great influence on the oxidation rate constant. The frequency factor of the reaction would be approximately doubled by reducing the CuO particle size from 1.0 mm to 0.2 mm. Cross section of reaction tubing mainly affected on the duration of tritium at the temperature below 600 K. Reaction tubing with length of 1 m at temperature of 600 K would be suitable for keeping the tritium conversion ratio above 99.9%. The length of reaction tubing can be reduced by using the smaller CuO particle or increasing the CuO temperature.

  2. Experiments on a ceramic electrolysis cell and a palladium diffuser at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Konishi, Satoshi; Yoshida, Hiroshi; Ohno, Hideo; Naruse, Yuji; Coffin, D.O.; Walthers, C.R.; Binning, K.E.

    1985-01-01

    A ceramic electrolysis cell and a palladium diffuser are developed in Japan and is tested with tritium in Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory, in order to confirm the feasibility as possible upgrades for the fuel cleanup system (PCU). The ceramic electrolysis cell made of stabilized zirconia was operated at 630 0 C for an extended period with a mixture of 3% T 2 O in He carrier gas in the circulation system with oxidizing catalyst bed. The palladium diffuser was tested with circulated pure tritium gas at 280 0 C to verify the compatibility of the alloy with tritium, since the 3 He produced in the metal could cause a degradation. The isotopic effects were also measured for both devices

  3. Tritium processing in JT-60U

    International Nuclear Information System (INIS)

    Miya, Naoyuki; Masaki, Kei

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Briec, M.; Roth, E.

    1987-04-01

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

  5. Storage and Assay of Tritium in STAR

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. 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

  7. Computer aided design of operational units for tritium recovery from Li17Pb83 blanket of a DEMO fusion reactor

    International Nuclear Information System (INIS)

    Malara, C.; Viola, A.

    1995-01-01

    The problem of tritium recovery from Li 17 Pb 83 blanket of a DEMO fusion reactor is analyzed with the objective of limiting tritium permeation into the cooling water to acceptable levels. To this aim, a mathematical model describing the tritium behavior in blanket/recovery unit circuit has been formulated. By solving the model equations, tritium permeation rate into the cooling water and tritium inventory in the blanket are evaluated as a function of dimensionless parameters describing the combined effects of overall resistance for tritium transfer from Li 17 Pb 83 alloy to cooling water, circulating rate of the molten alloy in blanket/recovery unit circuit and extraction efficiency of tritium recovery unit. The extraction efficiency is, in turn, evaluated as a function of the operating conditions of recovery unit. The design of tritium recovery unit is then optimized on the basis of the above parametric analysis and the results are herein reported and discussed for a tritium permeation limit of 10 g/day into the cooling water. 14 refs., 9 figs., 2 tabs

  8. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  9. Construction and operation of a tritium extraction facility at the Savannah River Site. Final environmental impact statement

    International Nuclear Information System (INIS)

    1999-03-01

    DOE proposes to construct and operate a Tritium Extraction Facility (TEF) at H Area on the Savannah River Site (SRS) to provide the capability to extract tritium from commercial light water reactor (CLWR) targets and from targets of similar design. The proposed action is also DOE's preferred alternative. An action alternative is to construct and operate TEF at the Allied General Nuclear Services facility, which is adjacent to the eastern side of the SRS. Under the no-action alternative DOE could incorporate tritium extraction capabilities in the accelerator for production of tritium. This EIS is linked to the Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling, from which DOE determined that it would produce tritium either in an accelerator or in a commercial light water reactor. The purpose of the proposed action and alternatives evaluated in this EIS is to provide tritium extraction capability to support either tritium production technology. The EIS assesses the environmental impacts from the proposed action and the alternatives, including the no action alternative

  10. An experiment to measure the electron neutrino mass using a cryogenic tritium source

    International Nuclear Information System (INIS)

    Fackler, O.; Jeziorski, B.; Kolos, W.; Monkhorst, H.; Mugge, M.; Sticker, H.; Szalewicz, K.; White, R.M.; Woerner, R.

    1985-01-01

    An experiment has been performed to determine the electron neutrino mass with the precision of a few eV by measuring the tritium beta decay energy distribution near the endpoint. Key features of the experiment are a 2 eV resolution electrostatic spectrometer and a high-activity frozen tritium source. It is important that the source have electronic wavefunctions which can be accurately calculated. These calculations have been made for tritium and the HeT + daughter ion and allow determination of branching fractions to 0.1% and energy of the excited states to 0.1 eV. The excited final molecular state calculations and the experimental apparatus are discussed. 4 refs., 5 figs

  11. Determination of effective cross sections and production rates for tritium in the irradiation experiment TRIDEX

    International Nuclear Information System (INIS)

    Weise, L.

    1986-04-01

    In the framework of the development of a fusion reactor blanket the irradiation experiment TRIDEX (Tritium Recovery Irradiation DIDO Experiment) takes place at the Juelich Research Reactor FRJ-2 (DIDO). For this equipment the required neutronic calculations have been performed. The aim was the determination of the neutron spectrum and several therefrom derived integral parameters for the irradiation positions in interest. From the calculated effective cross sections for the formation of Tritium resulting from irradiated Lithium samples on one hand and from measured neutron flux densities on the other hand, all needed quantities of the Tritium production could be determined. The calculation of the neutron spectrum has been performed in a two-dimensional x-y-geometry. The neutron flux densities have been gained by gamma-spectrometric measurement of the activities in irradiated activation samples. (orig.) [de

  12. Tritium pellet injection sequences for TFTR

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. TRitium Activity Measurements with a PhotomultipliEr in Liquids–The TRAMPEL experiment

    International Nuclear Information System (INIS)

    Priester, Florian; Klein, Manuel

    2016-01-01

    Highlights: • We have set up a new test device for measuring of tritiated water samples. • The device is very compact and easy and reliable in operation. • Easy integration in flow-through systems is possible. • The device has been operated at Tritium Laboratory Karlsruhe for several months. • The lower detection limit has been improved with regard to predecessors experiments. - Abstract: A common technique for the determination of the activity of tritiated water (HTO) is liquid scintillation counting (LSC). This implies sample taking, sample preparation and radioactive waste processing afterwards. When handling highly tritiated water special care has to be taken because of possible harmfulness of the sample. Furthermore, LSC devices are mostly large, heavy and expensive. The TRAMPEL experiment aims at measuring the activity of tritiated water in-line without sample taking. The device is intended to be easy to use and operate, quite inexpensive and compact. The measurement principle is based on electrons from β-decay which induce light in commercially available scintillation fibres. The light is detected by a small photomultiplier tube (PMT). A proof-of-principle was set up for static measurements using standard stainless steel parts. The complete device has a volume of less than 0.5 l.

  14. TRitium Activity Measurements with a PhotomultipliEr in Liquids–The TRAMPEL experiment

    Energy Technology Data Exchange (ETDEWEB)

    Priester, Florian, E-mail: florian.priester@kit.edu; Klein, Manuel

    2016-11-01

    Highlights: • We have set up a new test device for measuring of tritiated water samples. • The device is very compact and easy and reliable in operation. • Easy integration in flow-through systems is possible. • The device has been operated at Tritium Laboratory Karlsruhe for several months. • The lower detection limit has been improved with regard to predecessors experiments. - Abstract: A common technique for the determination of the activity of tritiated water (HTO) is liquid scintillation counting (LSC). This implies sample taking, sample preparation and radioactive waste processing afterwards. When handling highly tritiated water special care has to be taken because of possible harmfulness of the sample. Furthermore, LSC devices are mostly large, heavy and expensive. The TRAMPEL experiment aims at measuring the activity of tritiated water in-line without sample taking. The device is intended to be easy to use and operate, quite inexpensive and compact. The measurement principle is based on electrons from β-decay which induce light in commercially available scintillation fibres. The light is detected by a small photomultiplier tube (PMT). A proof-of-principle was set up for static measurements using standard stainless steel parts. The complete device has a volume of less than 0.5 l.

  15. Second international comparison on measuring techniques of tritium production rate for fusion neutronics experiments (ICMT-2)

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Maekawa, Hiroshi

    1993-02-01

    An second international comparison on measuring techniques of tritium production rates for fusion neutronics experiments (ICMT-2) has been performed. The purpose is to evaluate the measurement accuracy of tritium production rates in the current measurement techniques. Two 14 MeV neutron source facilities, FNS at JAERI-Japan and LOTUS at EPFL-Switzerland, were used for this purpose. Nine groups out of seven countries participated in this program. A fusion simulated blanket assembly of simple-geometry was served as the test bed at each facility, in which Li-containing samples from the participants were irradiated in an uniform neutron field. The tritium production rates were determined by the participants using their own ways by using the liquid scintillation counting method. Tritiated water sample with unknown but the same concentration was also distributed and its concentration was measured to make a common reference. The standard deviation of measured tritium production rates among participants was about 10 % for both FNS and LOTUS irradiation levels: 4x10 -13 T-atoms/Li-atom and 1.6x10 -12 T-atoms/Li-atom at a sample, respectively. This standard deviation exceeds the expected deviation of 5 % in this program. It is presumed that the deviation of 10 % is caused mainly by the systematic and unknown errors in a process of tritium extraction from the irradiated samples depending on each organization. (author)

  16. Tritium breeding experiments in a fusion blanket assembly using a low-intensity neutron generator

    International Nuclear Information System (INIS)

    Dalton, A.W.; Woodley, H.J.; McGregor, B.J.

    1987-01-01

    Experiments have been carried out to determine the accuracy with which tritium production rates (TPRs) can be measured in a fusion blanket assembly of non-spherical geometry by a non-central low intensity D-T neutron source (2x10 10 neutrons per second). The tritium production was determined for samples of lithium carbonate containing high enrichments of 6 Li(96%) and 7 Li(99.9%). The measured data were used to check the accuracy with which the TPRs could be numerically predicted using current nuclear data and calculational methods. The numerical predictions from tritium production from the 7 Li samples agreed within the experimental errors of the measurements, but 6 Li measurements which differ by more than 20 per cent from the predicted values were observed in the lower half of the assembly

  17. 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.

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

    International Nuclear Information System (INIS)

    Iseli, M.; Esser, B.

    1989-01-01

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

  19. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

    Ono, F.; Yamawaki, M.

    1995-01-01

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

  20. Estimating cancer risk in relation to tritium exposure from routine operation of a nuclear-generating station in Pickering, Ontario.

    Science.gov (United States)

    Wanigaratne, S; Holowaty, E; Jiang, H; Norwood, T A; Pietrusiak, M A; Brown, P

    2013-09-01

    Evidence suggests that current levels of tritium emissions from CANDU reactors in Canada are not related to adverse health effects. However, these studies lack tritium-specific dose data and have small numbers of cases. The purpose of our study was to determine whether tritium emitted from a nuclear-generating station during routine operation is associated with risk of cancer in Pickering, Ontario. A retrospective cohort was formed through linkage of Pickering and north Oshawa residents (1985) to incident cancer cases (1985-2005). We examined all sites combined, leukemia, lung, thyroid and childhood cancers (6-19 years) for males and females as well as female breast cancer. Tritium estimates were based on an atmospheric dispersion model, incorporating characteristics of annual tritium emissions and meteorology. Tritium concentration estimates were assigned to each cohort member based on exact location of residence. Person-years analysis was used to determine whether observed cancer cases were higher than expected. Cox proportional hazards regression was used to determine whether tritium was associated with radiation-sensitive cancers in Pickering. Person-years analysis showed female childhood cancer cases to be significantly higher than expected (standardized incidence ratio [SIR] = 1.99, 95% confidence interval [CI]: 1.08-3.38). The issue of multiple comparisons is the most likely explanation for this finding. Cox models revealed that female lung cancer was significantly higher in Pickering versus north Oshawa (HR = 2.34, 95% CI: 1.23-4.46) and that tritium was not associated with increased risk. The improved methodology used in this study adds to our understanding of cancer risks associated with low-dose tritium exposure. Tritium estimates were not associated with increased risk of radiationsensitive cancers in Pickering.

  1. Tritium proof-of-principle pellet injector

    International Nuclear Information System (INIS)

    Fisher, P.W.

    1991-07-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  3. High pressure deuterium-tritium gas target vessels for muon-catalyzed fusion experiments

    International Nuclear Information System (INIS)

    Caffrey, A.J.; Spaletta, H.W.; Ware, A.G.; Zabriskie, J.M.; Hardwick, D.A.; Maltrud, H.R.; Paciotti, M.A.

    1989-01-01

    In experimental studies of muon-catalyzed fusion, the density of the hydrogen gas mixture is an important parameter. Catalysis of up to 150 fusions per muon has been observed in deuterium-tritium gas mixtures at liquid hydrogen density; at room temperature, such densities require a target gas pressure of the order of 1000 atmospheres (100 MPa, 15,000 psi). We report here the design considerations for hydrogen gas target vessels for muon-catalyzed fusion experiments that operate at 1000 and 10,000 atmospheres. The 1000 atmosphere high pressure target vessels are fabricated of Type A-286 stainless steel and lined with oxygen-free, high-conductivity (OFHC) copper to provide a barrier to hydrogen permeation of the stainless steel. The 10,000 atmosphere ultrahigh pressure target vessels are made from 18Ni (200 grade) maraging steel and are lined with OFHC copper, again to prevent hydrogen permeation of the steel. In addition to target design features, operating requirements, fabrication procedures, and secondary containment are discussed. 13 refs., 3 figs., 1 tab

  4. Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

    1993-01-01

    The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

  5. CANDU operating experience

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.; Thurygill, E.W.

    1980-05-01

    The CANDU-PHW program is based upon 38 years of heavy water reactor experience with 35 years of operating experience. Canada has had 72 reactor years of nuclear-electric operations experience with 10 nuclear units in 4 generating stations during a period of 18 years. All objectives have been met with outstanding performance: worker safety, public safety, environmental emissions, reliable electricity production, and low electricity cost. The achievement has been realized through total teamwork involving all scientific disciplines and all project functions (research, design, manufacturing, construction, and operation). (auth)

  6. Overview of tritium processing development at the tritium systems test assembly

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1986-01-01

    The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been operating with tritium since June 1984. Presently there are some 50 g of tritium in the main processing loop. This 50 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation system and to integrate the fuel cleanup system into the main fuel processing loop. In January 1986 two major experiments were conducted. During these experiments the fuel cleanup system was integrated, through the transfer pumping system, with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems leaves only the main vacuum system to be integrated into the TSTA fuel processing loop. In September 1986 another major tritium experiment was performed in which the integrated loop was operated, the tritium inventory increased to 50 g and additional measurements on the performance of the distillation system were taken. In the period June 1984 through September 1986 the TSTA system has processed well over 10 8 Ci of tritium. Total tritium emissions to the environment over this period have been less than 15 Ci. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flows have been uncovered

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  8. Temporal sealing material of tritium-contaminated stainless steel

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  9. CANDU operating experience

    International Nuclear Information System (INIS)

    McConnell, L.G.; Woodhead, L.W.; Fanjoy, G.R.

    1982-03-01

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

  10. Feasibility study of an experiment to measure the RBE of tritium for the induction of myeloid leukemia in animals

    International Nuclear Information System (INIS)

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

    1986-02-01

    A variety of RBE values ranging from about 1 to 3 for tritium have been measured by different investigators. The reason for this range in RBE can be attributed to differences in the biological endpoints measured, the reference radiation to which the effects of tritium were compared, and the tritium dosimetry of the particular study. Since the principal risk of low-level irradiation is the induction of cancers, it would be desirable to utilize this endpoint in tritium RBE experiments if these experiments are to be used to evaluate the quality factor for tritium. Furthermore, it would be desirable to use 200 kVp X-rays as the reference radiation since this radiation was suggested by ICRP as the standard reference to be used in the calculation of dose equivalents for purposes of radiation protection. Acute myeloid leukemia is one of the earliest recognized examples of radiogenic cancer in humans and this endpoint has also been the subject of animal studies. This report gives the results of a review of these animal studies to see if this endpoint is suitable for an experiment to measure the tritium RBE relative to 200 kVp X-rays. It was concluded that the male CBA/H mouse, would be a suitable species and an experiment involving 5000 animals in a four to five year study would be required to provide a useful estimate of the RBE for tritium. 72 refs

  11. 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

  12. Direct measurement of tritium production rate in LiPb with removed parasitic activities: Preliminary experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kuc, Tadeusz, E-mail: kuc@agh.edu.pl; Pohorecki, Władysław; Ostachowicz, Beata

    2014-10-15

    Liquid scintillation (LS) technique applied to direct measurement of tritium activity produced in LiPb eutectic in Frascati HCLL TBM mock-up neutronic experiment has been tested so far in the case of LS measurement after long period since irradiation. LiPb samples irradiated in neutron filed show, except of tritium, meaningful activity of other radioisotopes (parasitic). Parasitic activity, mainly from isotopes of lead ({sup 209}Pb, {sup 204m}Pb, {sup 203}Pb) calculated with the use of FISPACT, exceeds ca 5 times tritium activity 1.4 h after irradiation. We propose to remove disturbing radioisotopes in a chemical way to avoid long “cooling” of the irradiated samples before tritium measurement. Samples (1 g of LiPb) irradiated in reactor fast neutron flux were diluted and metallic cations removed by chemical precipitation. For this purpose we used: potassium iodide (KJ), strontium chloride (SrCl{sub 2}), APDC (C{sub 5}H{sub 8}NS{sub 2}·NH{sub 4}), NaDDTC (C{sub 5}H{sub 10}NNaS{sub 2}·3H{sub 2}O), and PAN (C{sub 15}H{sub 11}N{sub 3}O). Precipitation procedure in each case lasted ca 5–25 min, and the following filtration next 10–20 min. In each filtrate (ca 120 ml) we measured Pb concentration in total reflection X-ray fluorescence (TXRF) analyzer and parasitic activity (left after 21-day “cooling”) applying HPGe gamma spectrometer. Pb cations precipitated by SrCl{sub 2} and than by PAN lowered activity of Pb isotopes to less than 1% of the initial tritium activity. Another combination of reagents: NaDDTC followed by SrCl{sub 2} in a single and double step filtration reduced Pb concentration 10{sup 2} and 10{sup 4} times, respectively. Reduction of this order allows tritium radiometric measurement ca 3 h after irradiation with acceptable accuracy. This time can be shortened by applying correction for decay of known parasitic activity. Input of {sup 76}As and other less abundant radioisotopes can be eliminated using high purity LiPb. Tritium activity of

  13. Operating experience feedback

    International Nuclear Information System (INIS)

    Cimesa, S.

    2007-01-01

    Slovenian Nuclear Safety Administration (SNSA) has developed its own system for tracking, screening and evaluating the operating experiences of the nuclear installations. The SNSA staff regularly tracks the operating experiences throughout the world and screens them on the bases of applicability for the Slovenian nuclear facilities. The operating experiences, which pass the screening, are thoroughly evaluated and also recent operational events in these facilities are taken into account. If needed, more information is gathered to evaluate the conditions of the Slovenian facilities and appropriate corrective actions are considered. The result might be the identification of the need for modification at the licensee, the need for modification of internal procedures in the SNSA or even the proposal for the modification of regulations. Information system helps everybody to track the process of evaluation and proper logging of activities. (author)

  14. 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

  15. 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

  16. Operating experience with snubbers

    International Nuclear Information System (INIS)

    Levin, H.; Cudlin, R.

    1978-06-01

    Recent operating experience with hydraulic and mechanical snubbers has indicated that there is a need to evaluate current practice in the industry associated with snubber qualification testing programs, design and analysis procedures, selection and specification criteria, and the preservice inspection and inservice surveillance programs. The report provides a summary of operational experiences that represent problems that are generic throughout the industry. Generic Task A-13 is part of the NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants described in NUREG-0410. The report is based upon a rather large amount of data that have become available in the past four years. These data have been evaluated by the Division of Operating Reactors to develop a data base for use in connection with several NRC activities including Category A, Technical Activity A-13 (Snubbers); the Standard Review Plan; future Regulatory Guides; ASME Code Provisions; and various technical specifications of operating nuclear power plants

  17. Overview of the assessment of the french in-field tritium experiment with computer codes

    International Nuclear Information System (INIS)

    Crabol, B.; Graziani, G.; Edlund, O.

    1989-01-01

    In the framework of the international cooperation settled for the realization of the French tritium experiment, an expert group for the assessment of computer codes, including the Joint Research Center of Ispra (European Communities), Studsvik (Sweden) and the Atomic Energy Commission (France), has been organized. The aim of the group was as follows: - to help the design of the experiment by evaluating beforehand the consequences of the release, - to interpret the results of the experiment. This paper describes the last task and gives the main conclusions drawn from the work

  18. 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

  19. In-situ Tritium Measurements of the Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

    International Nuclear Information System (INIS)

    C.A. Gentile; C.H. Skinner; K.M. Young; M. Nishi; S. Langish; et al

    1999-01-01

    The Princeton Plasma Physics Laboratory (PPPL) Engineering and Research Staff in collaboration with members of the Japan Atomic Energy Research Institute (JAERI), Tritium Engineering Laboratory have commenced in-situ tritium measurements of the TFTR bumper limiter. The Tokamak Fusion Test Reactor (TFTR) operated with tritium from 1993 to 1997. During this time ∼ 53,000 Ci of tritium was injected into the TFTR vacuum vessel. After the cessation of TFTR plasma operations in April 1997 an aggressive tritium cleanup campaign lasting ∼ 3 months was initiated. The TFTR vacuum vessel was subjected to a regimen of glow discharge cleaning (GDC) and dry nitrogen and ''moist air'' purges. Currently ∼ 7,500 Ci of tritium remains in the vacuum vessel largely contained in the limiter tiles. The TFTR limiter is composed of 1,920 carbon tiles with an average weight of ∼ 600 grams each. The location and distribution of tritium on the TFTR carbon tiles are of considerable interest. Future magnetically confined fusion devices employing carbon as a limiter material may be considerably constrained due to potentially large tritium inventories being tenaciously held on the surface of the tiles. In-situ tritium measurements were conducted in TFTR bay L during August and November 1998. During the bay L measurement campaign open wall ion chambers and ultra thin thermoluminscent dosimeters (TLD) affixed to a boom and end effector were deployed into the vacuum vessel. The detectors were designed to make contact with the surface of the bumper limiter tile and to provide either real time (ion chamber) or passive (TLD) indication of the surface tritium concentration. The open wall ion chambers were positioned onto the surface of the tile in a manner which employed the surface of the tile as one of the walls of the chamber. The ion chambers, which are (electrically) gamma insensitive, were landed at four positions per tile. The geometry for landing the TLD's provided measurement at 24

  20. Operating experience: safety perspective

    International Nuclear Information System (INIS)

    Piplani, Vivek; Krishnamurthy, P.R.; Kumar, Neeraj; Upadhyay, Devendra

    2015-01-01

    Operating Experience (OE) provides valuable information for improving NPP safety. This may include events, precursors, deviations, deficiencies, problems, new insights to safety, good practices, lessons and corrective actions. As per INSAG-10, an OE program caters as a fundamental means for enhancing the defence-in-depth at NPPs and hence should be viewed as ‘Continuous Safety Performance Improvement Tool’. The ‘Convention on Nuclear Safety’ also recognizes the OE as a tool of high importance for enhancing the NPP safety and its Article 19 mandates each contracting party to establish an effective OE program at operating NPPs. The lessons drawn from major accidents at Three Mile Island, Chernobyl and Fukushima Daiichi NPPs had prompted nuclear stalwarts to change their safety perspective towards NPPs and to frame sound policies on issues like safety culture, severe accident prevention and mitigation. An effective OE program, besides correcting current/potential problems, help in proactively improving the NPP design, operating and maintenance procedures, practices, training, etc., and thus plays vital role in ensuring safe and efficient operation of NPPs. Further it enhances knowledge with regard to equipment operating characteristics, system performance trends and provides data for quantitative and qualitative safety analysis. Besides all above, an OE program inculcates a learning culture in the organisation and thus helps in continuously enhancing the expertise, technical competency and knowledge base of its staff. Nuclear and Radiation Facilities in India are regulated by Atomic Energy Regulatory Board (AERB). Operating Plants Safety Division (OPSD) of AERB is involved in managing operating experience activities. This paper provides insights about the operating experience program of OPSD, AERB (including its on-line data base namely OPSD STAR) and its utilisation in improving the regulations and safety at Indian NPPs/projects. (author)

  1. Tritium-doping enhancement of polystyrene by ultraviolet laser and hydrogen plasma irradiation for laser fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Iwasa, Yuki, E-mail: iwasa-y@ile.osaka-u.ac.jp [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Yamanoi, Kohei; Iwano, Keisuke; Empizo, Melvin John F.; Arikawa, Yasunobu; Fujioka, Shinsuke; Sarukura, Nobuhiko; Shiraga, Hiroyuki; Takagi, Masaru; Norimatsu, Takayoshi; Azechi, Hiroshi [Institute of Laser Engineering, Osaka University, 2-6 Yamadaoka, Suita, Osaka 565-0871 (Japan); Noborio, Kazuyuki; Hara, Masanori; Matsuyama, Masao [Hydrogen Isotope Research Center, Organization for Promotion of Research, University of Toyama, 3190 Gofuku, Toyama 930-8555 (Japan)

    2016-11-15

    Highlights: • Tritium-doped polystyrene films are fabricated by the Wilzbach method with UV laser and hydrogen plasma irradiation. • The 266-nm laser-irradiated, 355-nm laser-irradiated, and hydrogen plasma-irradiated polystyrene films exhibit higher PSL intensities and specific radioactivities than the non-irradiated sample. • Tritium doping by UV laser irradiation can be largely affected by the laser wavelength because of polystyrene’s absorption. • Hydrogen plasma irradiation results to a more uniform doping concentration even at low partial pressure and short irradiation time. • UV laser and plasma irradiations can be utilized to fabricate tritium-doped polystyrene shell targets for future laser fusion experiments. - Abstract: We investigate the tritium-doping enhancement of polystyrene by ultraviolet (UV) laser and hydrogen plasma irradiation. Tritium-doped polystyrene films are fabricated by the Wilzbach method with UV laser and hydrogen plasma. The 266-nm laser-irradiated, 355-nm laser-irradiated, and hydrogen plasma-irradiated polystyrene films exhibit higher PSL intensities and specific radioactivities than the non-irradiated sample. Tritium doping by UV laser irradiation can be largely affected by the laser wavelength because of polystyrene’s absorption. In addition, UV laser irradiation is more localized and concentrated at the spot of laser irradiation, while hydrogen plasma irradiation results to a more uniform doping concentration even at low partial pressure and short irradiation time. Both UV laser and plasma irradiations can nevertheless be utilized to fabricate tritium-doped polystyrene targets for future laser fusion experiments. With a high doping rate and efficiency, a 1% tritium-doped polystyrene shell target having 7.6 × 10{sup 11} Bq g{sup −1} specific radioactivity can be obtained at a short period of time thereby decreasing tritium consumption and safety management costs.

  2. ATLAS IBL operational experience

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00237659; The ATLAS collaboration

    2017-01-01

    The Insertable B-Layer (IBL) is the inner most pixel layer in the ATLAS experiment, which was installed at 3.3 cm radius from the beam axis in 2014 to improve the tracking performance. To cope with the high radiation and hit occupancy due to proximity to the interaction point, a new read-out chip and two different silicon sensor technologies (planar and 3D) have been developed for the IBL. After the long shut-down period over 2013 and 2014, the ATLAS experiment started data-taking in May 2015 for Run-2 of the Large Hadron Collider (LHC). The IBL has been operated successfully since the beginning of Run-2 and shows excellent performance with the low dead module fraction, high data-taking efficiency and improved tracking capability. The experience and challenges in the operation of the IBL is described as well as its performance.

  3. 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

  4. FFTF operational experience

    International Nuclear Information System (INIS)

    Newland, D.J.; Krupar, J.J.

    1984-01-01

    In April 1982, the FFTF began its first nominally 100 day irradiation cycle. Since that time the plant has operated very well with steadily increasing plant capacity factors during its first four cycles. One hundred fifty fuel assemblies (eighty of which are experiments) and over 32,000 individual fuel pins have been irradiated, some in excess of 100 MWd/Kg burnup. Specialized equipment and systems unique to sodium cooled reactor plants have performed well

  5. Operating experience in reprocessing

    International Nuclear Information System (INIS)

    Schueller, W.

    1983-01-01

    Since 1953, reprocessing has accumulated 180 years of operating experience in ten plants, six of them with 41 years of operation in reprocessing oxide fuel from light water reactors. After abortive, premature attempts at what is called commercial reprocessing, which had been oriented towards the market value of recoverable uranium and plutonium, non-military reprocessing technologies have proved their technical feasibility, since 1966 on a pilot scale and since 1976 on an industrial scale. Reprocessing experience obtained on uranium metal fuel with low and medium burnups can now certainly be extrapolated to oxide fuel with high burnup and from pilot plants to industrial scale plants using the same technologies. The perspectives of waste management of the nuclear power plants operated in the Federal Republic of Germany should be viewed realistically. The technical problems still to be solved are in a balanced relationship to the benefit arising to the national economy out of nuclear power generation and can be solved in time, provided there are clearcut political boundary conditions. (orig.) [de

  6. GNF2 Operating Experience

    International Nuclear Information System (INIS)

    Schardt, John

    2007-01-01

    GNF's latest generation fuel product, GNF2, is designed to deliver improved nuclear efficiency, higher bundle and cycle energy capability, and more operational flexibility. But along with high performance, our customers face a growing need for absolute fuel reliability. This is driven by a general sense in the industry that LWR fuel reliability has plateaued. Too many plants are operating with fuel leakers, and the impact on plant operations and operator focus is unacceptable. The industry has responded by implementing an INPO-coordinated program aimed at achieving leaker-free reliability by 2010. One focus area of the program is the relationship between fuel performance (i.e., duty) and reliability. The industry recognizes that the right balance between performance and problem-free fuel reliability is critical. In the development of GNF2, GNF understood the requirement for a balanced solution and utilized a product development and introduction strategy that specifically addressed reliability: evolutionary design features supported by an extensive experience base; thoroughly tested components; and defense-in-depth mitigation of all identified failure mechanisms. The final proof test that the balance has been achieved is the application of the design, initially through lead use assemblies (LUAs), in a variety of plants that reflect the diversity of the BWR fleet. Regular detailed surveillance of these bundles provides the verification that the proper balance between performance and reliability has been achieved. GNF currently has GNF2 lead use assemblies operating in five plants. Included are plants that have implemented extended power up-rates, plants on one and two-year operating cycles, and plants with and without NobleChem TM and zinc injection. The leading plant has undergone three pool-side inspections outages to date. This paper reviews the actions taken to insure GNF2's reliability, and the lead use assembly surveillance data accumulated to date to validate

  7. 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.

  8. Tritium monitoring techniques

    International Nuclear Information System (INIS)

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

    1996-05-01

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

  9. Tritium 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.

  10. Prospects for cosmic neutrino detection in tritium experiments in the case of hierarchical neutrino masses

    International Nuclear Information System (INIS)

    Blennow, Mattias

    2008-01-01

    We discuss the effects of neutrino mixing and the neutrino mass hierarchy when considering the capture of the cosmic neutrino background (CNB) on radioactive nuclei. The implications of mixing and hierarchy at future generations of tritium decay experiments are considered. We find that the CNB should be detectable at these experiments provided that the resolution for the kinetic energy of the outgoing electron can be pushed to a few 0.01 eV for the scenario with inverted neutrino mass hierarchy, about an order of magnitude better than that of the upcoming KATRIN experiment. Another order of magnitude improvement is needed in the case of normal neutrino mass hierarchy. We also note that mixing effects generally make the prospects for CNB detection worse due to an increased maximum energy of the normal beta decay background

  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 retention and clean-up in JET

    International Nuclear Information System (INIS)

    Andrew, P.; Brennan, P.D.; Coad, J.P.

    1999-01-01

    During 1997 JET operation with D-T plasmas, 35 g of tritium were introduced into the torus, mainly by gas puffing. It was found that during this period, the torus tritium inventory would accumulate at a rate of about 40% of the input. After tritium operation ceased, the experimental program continued with deuterium- and hydrogen-fuelled experiments, during which time the tritium inventory decreased to about 17% of the total input. Techniques aimed at detritiation of the torus included methods using deuterium gas (such as deuterium pulsing) which were used in the middle of the experimental campaign, and methods which could adversely affect the torus vacuum conditions (such as air purges) which were reserved for the period after the experimental campaign. Whilst it was found that the plasma tritium fraction could be reduced to below the 1% level in a few days, the tritium inventory reached a virtually steady level of about 6 g by the end of the campaign. (orig.)

  13. Reviewing operational experience feedback

    International Nuclear Information System (INIS)

    1991-04-01

    The purpose of this document is to provide detailed supplementary guidance to OSART experts to aid in the evaluation of operational experience feedback (OEF) programmes at nuclear power plants. The document begins by describing the objectives of an OEF programme. It goes on to indicate preparatory work and investigatory guidance for the expert. Section 5 describes attributes of an excellent OEF programme. Appended to these guidelines are examples of OEF documents from various plants. These are intended to help the expert by demonstrating the actual implementation of OEF in practice. These guidelines are in no way intended to conflict with existing national regulations and rules. A comprehensive OEF programme, as described in Section 2, would be impossible to evaluated in detail in the amount of time typically allocated for assessing OEF in an OSART review. The expert must use his or her time wisely by concentrating on those areas that appear to be the weakest

  14. Fusion fuel purification during the Tritium Systems Test Assembly 3-week loop experiment

    International Nuclear Information System (INIS)

    Willms, R.S.

    1989-01-01

    During the time period from April 19, 1989--May 5, 1989, the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL) conducted its longest continuous integrated loop operation to date. This provided an opportunity to test some hitherto unproven capabilities of the TSTA Fuel Cleanup System (FCU). Previous FCU tests were reported. The purpose of the FCU is to remove impurities from a stream of hydrogen isotopes (Q 2 ) representative of torus exhaust gas. During this run impurities loadings ranging from 60 to 179 sccm of 90% N 2 and 10% CH 4 were fed to the FCU. Each of the two FCU main flow molecular sieve beds (MSB's) were filled to breakthrough three times. The MSB's were regenerated during loop operations. 2 refs., 6 figs., 2 tabs

  15. Operating experience at CEBAF

    International Nuclear Information System (INIS)

    Legg, R.

    1996-01-01

    CEBAF, the Continuous Electron Beam Accelerator Facility, is a 5-pass, recirculating, superconducting rf linac designed to provide exceptional beam quality at 4 GeV up to 200 μA CW. It is made up of an injector, two 400-MeV linacs, and 9 recirculation arcs having a total beamline length of more than 4.5 km. On Nov. 5, 1995, CEBAF delivered a 4 GeV, 25-μA CW electron beam to the first of 3 experimental halls and the experimental physics program was started 10 days later. Accelerator availability during the first month of the experimental run exceeded 75%. Beam properties measured in the experimental hall to date are a one sigma momentum spread of 5x10 -5 and an rms emittance of 0.2 nanometer-radians, better than design specification. CW beam has been provided from all 5 passes at 800 MeV intervals. Outstanding performance of the superconducting linacs suggests a machine energy upgrade to 6 GeV in the near term with eventual machine operation at 8-10 GeV. Results from commissioning and operations experience since the last conference are presented

  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. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  19. Desactivation of tritium waters by rectification methods

    International Nuclear Information System (INIS)

    Egorov, A.I.; Tyunis, V.M.

    2002-01-01

    Results of experiments into the basic rectification processes dedicated to tritium separation from reactor, technological and waste waters are presented. Coefficients of separation for rectification of water (1.028), ammonia (1.05), azeotrope H 2 O - HTO - HNO 3 (1.098) and D 2 O - DTO - DNO 3 (1.039) are performed. Operating schemes of tritium separating units are reviewed [ru

  20. Results of tritium tests performed on Sandia Laboratories decontamination system

    International Nuclear Information System (INIS)

    Gildea, P.D.; Wall, W.R.; Gede, V.P.

    1978-05-01

    The Tritium Research Laboratory (TRL), a facility for performing experiments using gram amounts of tritium, became operational on October 1, 1977. As secondary containment, the TRL employs sealed glove boxes connected on demand to two central decontamination systems, the Gas Purification System and the Vacuum Effluent Recovery System. Performance tests on these systems show the tritium removal systems can achieve concentration reduction factors (ratio of inlet to exhaust concentrations) much in excess of 1000 per pass at inlet concentrations of 1 part per million or less for both tritium and tritiated methane

  1. Development of a low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic

    International Nuclear Information System (INIS)

    Pawelko, R.; Shimada, M.; Katayama, K.; Fukada, S.; Terai, T.

    2014-01-01

    A new experimental system designed to investigate tritium mass transfer properties in materials important to fusion technology is operational at the Safety and Tritium Applied Research (STAR) facility located at the Idaho National Laboratory (INL). The tritium permeation measurement system was developed as part of the Japan/US TITAN collaboration to investigate tritium mass transfer properties in liquid lead lithium eutectic (LLE) alloy. The system is similar to a hydrogen/deuterium permeation measurement system developed at Kyushu University and also incorporates lessons learned from previous tritium permeation experiments conducted at the STAR facility. This paper describes the experimental system that is configured specifically to measure tritium mass transfer properties at low tritium partial pressures. We present preliminary tritium permeation results for α-Fe and α-Fe/LLE samples at 600degC and at tritium partial pressures between 1.0E-3 and 2.4 Pain helium. The preliminary results are compared with literature data. (author)

  2. Screening effect as an explanation of results of Simpson experiment on tritium β decay

    International Nuclear Information System (INIS)

    Drukarev, E.G.; Strikman, M.I.

    1989-01-01

    Results of the work carried out by Simpson who observed a wide maximum in electron spectrum as a result of tritium decay in the range 0.75< T<1.5 keV are discussed. The results are not confirmed by the subsequent experiments. It is shown in the work that after correct account of interaction in finite state the discrepancy between experimental data and theoretical predictions for T ≥ 1 keV disappears. It is also shown that results at T<1 keV depend greatly on the effects of interaction with medium. Gross evaluation of the effects eliminates the discrepancy in the range 0.75< T<1 keV

  3. 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

  4. 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

  5. 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%

  6. 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

  7. 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

  8. Tevatron operational experiences

    International Nuclear Information System (INIS)

    Norris, B.L.; Theilacker, J.C.

    1989-02-01

    Fermilabs superconducting accelerator, the Tevatron has been operational for nearly six years. The history of its operation is presented. Several long shutdowns for superconducting dipole repairs are discussed. The dominant factor influencing the repair was conductor motion which fatigued the cable in the magnet ends. Borescoping and x-raying techniques were used to determine which magnet ends required repair. Detailed downtime logs were kept for each of the running periods. A discussion of the sources of downtime and a comparison for different operating modes is presented

  9. 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

  10. Operating experience feedback in TVO

    Energy Technology Data Exchange (ETDEWEB)

    Piirto, A [Teollisuuden Voima Oy (Finland)

    1997-12-31

    TVO is a power company operating with two 710 MW BWR units at Olkiluoto. For operating experience feedback TVO has not established a separate organizational unit but rather relies on a group of persons representing various technical disciplines. The ``Operating Experience Group`` meets at about three-week intervals to handle the reports of events (in plant and external) which have been selected for handling by an engineer responsible for experience feedback. 7 charts.

  11. TRIGA reactor operating experience

    International Nuclear Information System (INIS)

    Anderson, T.V.

    1970-01-01

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  12. 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.

  13. Tritium sorption by cement and subsequent release

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-01

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

  15. Handling of tritium at TFTR

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  16. Direct tritium measurement in lithium titanate for breeding blanket mock-up experiments with D-T neutrons

    International Nuclear Information System (INIS)

    Klix, A.; Ochiai, K.; Nishitani, T.; Takahashi, A.

    2004-01-01

    At Fusion Neutronics Source (FNS) of JAERI, tritium breeding experiments with blanket mock-ups consisting of advanced fusion reactor materials are in progress. The breeding zones are thin layers of lithium titanate which is one of the candidate tritium breeder materials for the DEMO fusion power reactor. It is anticipated that the application of small pellet-shaped lithium titanate detectors manufactured from the same material as the breeding layer will reduce experimental uncertainties arising from necessary corrections due to different isotopic lithium volume concentrations in breeding material and detector. Therefore, a method was developed to measure the local tritium production by means of lithium titanate pellet detectors and a liquid scintillation counting technique. The lithium titanate pellets were dissolved in concentrated hydrochloric acid solution and the resulting acidic solution was neutralized. Two ways of further processing were followed: direct incorporation into a liquid scintillation cocktail and distillation of the solution followed by mixing with liquid scintillator. Two types of lithium titanate pellets were investigated with different 6 Li enrichment and manufacturing procedure. It was found that lithium titanate is suitable for tritium production measurements. However some discrepancies in the measurement accuracy remained with one of the investigated pellet detectors when compared with a well-established lithium carbonate measurement technique and this issue needs further investigation

  17. Fuel cleanup system for the tritium systems test assembly: design and experiments

    International Nuclear Information System (INIS)

    Kerr, E.C.; Bartlit, J.R.; Sherman, R.H.

    1980-01-01

    A major subsystem of the Tritium Systems Test Assembly is the Fuel Cleanup System (FCU) whose functons are to: (1) remove impurities in the form of argon and tritiated methane, water, and ammonia from the reactor exhaust stream and (2) recover tritium for reuse from the tritiated impurities. To do this, a hybrid cleanup system has been designed which utilizes and will test concurrently two differing technologies - one based on disposable, hot metal (U and Ti) getter beds and a second based on regenerable cryogenic asdorption beds followed by catalytic oxidation of impurities to DTO and stackable gases and freezout of the resultant DTO to recover essentially all tritium for reuse

  18. Operating Experience at NPP Krsko

    International Nuclear Information System (INIS)

    Kavsek, D.; Bach, B.

    1998-01-01

    Systematic analysis of operational experience by assessment of internal and industry events and the feedback of lessons learned is one of the essential activities in the improvement of the operational safety and reliability of the nuclear power plant. At NPP Krsko we have developed a document called ''Operating Experience Assessment Program''. Its purpose is to establish administrative guidance for the processing of operating events including on-site and industry events. Assessment of internal events is based on the following methods: Event and Causal Factor Charting, Change Analysis, Barrier Analysis, MORT (Management Oversight and Risk Tree Analysis) and Human Performance Evaluation. The operating experience group has developed a sophisticated program entitled ''Operating experience tracking system'' (OETS) in response to the need for a more efficient way of processing internal and industry operating experience information. The Operating Experience Tracking System is used to initiate and track operational events including recommended actions follow up. Six screens of the system contain diverse essential information which allows tracking of operational events and enables different kinds of browsing. OETS is a part of the NPP Krsko nuclear network system and can be easily accessed by all plant personnel. (author)

  19. Tritium assay of Li2O pellets in the LBM/LOTUS experiments

    International Nuclear Information System (INIS)

    Quanci, J.; Azam, S.; Bertone, P.

    1986-01-01

    One of the objectives of the Lithium Blanket Module (LBM) program is to test the ability of advanced neutronics codes to model the tritium breeding characteristics of a fusion blanket exposed to a toroidal fusion neutron source. The LBM consists of over 20,000 cylindrical lithium oxide pellets and numerous diagnostic pellets and wafers. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a Haefely sealed neutron generator that gives a point deuterium-tritium neutron source up to 5 x 10 12 14-MeV n/s. Both Princeton Plasma Physics Lab. (PPPL) and EPFL assayed the tritium bred at various positions in the LBM. EPFL employed a dissolution technique while PPPL recovered the tritium by a thermal extraction method

  20. 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

  1. BEATRIX-II: In-situ tritium recovery data correction

    International Nuclear Information System (INIS)

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

    1993-09-01

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

  2. NPD Operating Experience

    Energy Technology Data Exchange (ETDEWEB)

    Horton, E. P. [Hydro-Electric Power Commission of Ontario, Rolphton, ON (Canada)

    1968-04-15

    NPD has demonstrated high-capacity factor operation and for the past three years has achieved an average net capacity factor of 98% for the ''winter-peak'' period. The net capacity factor for the year 1966 was 88% and for the period from the end of commissioning (October 1962) to the end of 1966 was 71%. The output of the station has been stretched from 22 MW(e) gross to 25 MW(e) gross. This was aided by the installation of an internal steam separator in the turbine but no basic modifications to the reactor-boiler systems were required. The turbine has also been modified by the installation of chrome steel diaphragms as a solution to an erosion problem. The station also continues as a test facility to develop new components and techniques. This includes the recent successful replacement of two reactor pressure tubes and the conversion of the reactor vault ventilation system to a ''dry'' atmosphere using a molecular sieve to collect heavy-water leakage and control the concentration of acidic oxides of nitrogen. Fuel performance has been excellent and the average burn-up in the core is now 84 MWh/kg U which is slightly above the equilibrium design value. Only three fuel bundles have been found with sheath failures and none of these was due to a deficiency in the fuel but was as a result of handling problems with the refuelling equipment. In spite of undesirably high maintenance time, the fuelling machines have now inserted over 1000 fuel bundles into the reactor ''on power''. Heavy-water loss rates have been acceptable and are improving. The average loss rate from leaks during 1966 was 210 g/h. A proposal to modify the NPD heavy-water heat transport system to allow boiling is under consideration. (author)

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

    International Nuclear Information System (INIS)

    Raskob, W.

    1994-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  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 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

  7. FDMH - The tritium model in RODOS

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  8. Vacuum system operating experience review for fusion applications

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1994-03-01

    This report presents a review of vacuum system operating experiences from particle accelerator, fusion experiment, space simulation chamber, and other applications. Safety relevant operating experiences and accident information are discussed. Quantitative order-of-magnitude estimates of vacuum system component failure rates and accident initiating event frequencies are presented for use in risk assessment, reliability, and availability studies. Safety concerns with vacuum systems are discussed, including personnel safety, foreign material intrusion, and factors relevant to vacuum systems being the primary confinement boundary for tritium and activated dusts. This information should be useful to fusion system designers and safety analysts, such as the team working on the Engineering Design Activities for the International Thermonuclear Experimental Reactor

  9. 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)

  10. Predicted fate of tritium residuum from groundwater tracer experiments in the Amargosa Desert, southern Nevada

    International Nuclear Information System (INIS)

    Brikowski, T.

    1993-07-01

    Analytic solutions are used in this study to evaluate potential groundwater transport of tritium used in goundwater tracer tests southwest of the Nevada Test Site. Possible transport from this site is of interest because initial radionuclide concentrations were high and the site is close to goundwater discharge points (12 km). Anecdotal evidence indicates that 90 percent of these tracers were removed by pumping at the completion of the tests; this study examines the probable transport of the tracers with and without the removal. Classical dispersive transport analytic solutions are used, treating the tracer test as a point slug injection. Input parameters for the solutions were measured at the site, and consideration of parameter uncertainty is incorporated in the results. With removal of the tracer, the maximum expected region with above-Safe Drinking Water Act (40 CFR 121) concentrations of tritium extends 5 km from the injection point, and does not reach any sites of public access. Detectable tritium from the tests is likely to have reached the Ash Meadows fault zone, but flow along the fault probably diluted the tracer to below detection limits before arrival at springs along the fault. Arrival at the springs would have occurred 20 to 25 years after the tests. Without removal of the tracer, the solutions indicate that tritium concentrations just above Safe Drinking Water Act standards would have reached the Ash Meadows fault zone. In this case, detectable tritium might have been found in Devil's Hole or Longstreet Spring, the nearest points of possible public exposure

  11. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    Jackson, H.A.; Woodhead, L.W.; Fanjoy, G.R.

    1984-03-01

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations for the workers and the public

  12. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    Bartholomew, R.W.; Woodhead, L.W.; Horton, E.P.; Nichols, M.J.; Daly, I.N.

    1987-01-01

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on worker and public safety, operating performance and costs, and reliability of system components

  13. New results from the Mainz neutrino mass experiment and perspective of a new large tritium-β-spectrometer

    International Nuclear Information System (INIS)

    Bonn, J.; Bornschein, B.; Bornschein, L.; Fickinger, L.; Kraus, Ch.; Otten, E.W.; Ulrich, H.; Weinheimer, Ch.; Kazachenko, O.; Kovalik, A.

    2001-01-01

    The Mainz neutrino mass experiment investigates the endpoint region of the tritium β decay spectrum to determine the mass of the electron antineutrino. By the recent upgrade the former problem of de-wetting T 2 films has been solved and the signal-to-background-ratio was improved by a factor of 10. The latest measurement leads to m ν 2 = -1.1 ± 2.6 stat ± 1.8 sys eV 2 /c 4 (preliminary), which corresponds to an upper limit of m ν 2 (95 % C.L.) (preliminary). Some indication for the anomaly, reported by the Troitsk group, was found, but its postulated half year period is contradicted by our data. The perspectives of a new Large Tritium-β-Spectrometer to reach sub eV sensitivity will be presented. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-15

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

  15. Treatment of tritiated exhaust gases at the Tritium Laboratory Karlsruhe

    Energy Technology Data Exchange (ETDEWEB)

    Hutter, E.; Besserer, U. [Kernforschungszentrum Karlsruhe GmbH (Germany); Jacqmin, G. [NUKEM GmbH, Industreistr, Alzenau (Germany)

    1995-02-01

    The Tritium Laboratory Karlsruhe (TLK) accomplished commissioning; tritium involving activities will start this year. The laboratory is destined mainly to investigating processing of fusion reactor fuel and to developing analytic devices for determination of tritium and tritiated species in view of control and accountancy requirements. The area for experimental work in the laboratory is about 800 m{sup 2}. The tritium infrastructure including systems for tritium storage, transfer within the laboratory and processing by cleanup and isotope separation methods has been installed on an additional 400 m{sup 2} area. All tritium processing systems (=primary systems), either of the tritium infrastructure or of the experiments, are enclosed in secondary containments which consist of gloveboxes, each of them connected to the central depressurization system, a part integrated in the central detritiation system. The atmosphere of each glovebox is cleaned in a closed cycle by local detritiation units controlled by two tritium monitors. Additionally, the TLK is equipped with a central detritiation system in which all gases discharged from the primary systems and the secondary systems are processed. All detritiation units consist of a catalyst for oxidizing gaseous tritium or tritiated hydrocarbons to water, a heat exchanger for cooling the catalyst reactor exhaust gas to room temperature, and a molecular sieve bed for adsorbing the water. Experiments with tracer amounts of tritium have shown that decontamination factors >3000 can be achieved with the TLK detritiation units. The central detritiation system was carefully tested and adjusted under normal and abnormal operation conditions. Test results and the behavior of the tritium barrier preventing tritiated exhaust gases from escaping into the atmosphere will be reported.

  16. ARIES-I tritium system

    International Nuclear Information System (INIS)

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

    1990-09-01

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

  17. Installation and operation of a chain of detection of tritium for groundwater dating at Madagascar-INSTN

    International Nuclear Information System (INIS)

    ANDRIAMIHARITSOA, G.

    2008-01-01

    The present study aims at installing and operating the Madagascar -INSTN tritium line for groundwater dating. The laboratory was first installed in July 2007 and is operational since January 2008. The objective of the laboratory is to determine the tritium activity in water sample by electrolytic enrichment prior to Liquid Scintillation Counting. The spike analyses showed that the mean value of the enrichment factor is 23.78 and that of the enrichment parameter is 0.88 for the first analysis. Such factor value is much higher than the usual determined value of 18 for a first enrichment run and shows that the electrolytic cells are working properly. the standard deviations are 1.6 and 0.02 for the enrichment factor and parameter, respectively. This indicates that the cells performance is approximately the same. The dead water activity is very low, with a value of 1.07cpm, while the inside and outside contamination control water activities are 0.5 cpm and 0.4 cpm, respectively. These values indicate that the laboratory is free of any contamination. [fr

  18. Tritium in the Savannah River Site environment

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-05-01

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

  19. Tritium in the Savannah River Site environment

    International Nuclear Information System (INIS)

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

    1991-05-01

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

  20. 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)

  1. Experiment to study the β-decay of free atomic and molecular tritium

    International Nuclear Information System (INIS)

    Robertson, R.G.H.; Bowles, T.J.; Maley, M.; Browne, J.C.; Burritt, T.; Toevs, J.; Stelts, M.; Helfrick, J.; Knapp, D.; Ledebuhr, A.G.

    1982-01-01

    An apparatus is described which will allow the measurement of the β-decay of free tritium atoms and molecules for determining antineutrino mass. It consists of an RF dissociator, a long cylindrical decay region open at both ends, a guide field, and a magnetic spectrometer

  2. 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.

  3. Overview of the tritium system of Ignitor

    International Nuclear Information System (INIS)

    Rizzello, C.; Tosti, S.

    2008-01-01

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

  4. Operating practical experience at Argentina

    International Nuclear Information System (INIS)

    Quihillalt, Oscar

    1997-01-01

    Operating experiences of Atucha-1 and Embalse Nuclear Power Plants were discussed in this work. The technical and economic aspects, such as reliability, availability, personnel training, operating costs, prices and market, which exercise influence upon Argentina nuclear energy policy, mainly on the power electric generation by nuclear power plants were considered. Finally the current status of the nucleoelectric sector in Argentina and forecasting were analysed

  5. Design of a tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

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

    1991-01-01

    The TFTR tritium pellet injector (TPI) is designed to provide a tritium pellet fueling capability with pellet speeds in the 1- to 3 km/s-range for the TFTR D-T phase. The existing TFTR deuterium pellet injector is being modified at Oak Ridge National Laboratory to provide a fourshot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns a two -stage light gas gun driver. The pipe gun concept has been qualified for tritium operation by the tritium proof-of-principle injector experiments conducted on the Tritium Systems Test Assembly at Los Alamos National Laboratory. In these experiments, tritium and D-T pellets were accelerated to speeds near 1.5 km/s. The TPI is being designed for pellet sizes in the range from 3.43 to 4.0 mm in diameter in arbitrarily programmable firing sequences at speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation will be controlled by a programmable logic controller. 7 refs., 4 figs

  6. Fort Saint Vrain operational experience

    International Nuclear Information System (INIS)

    Fuller, C.H.

    1989-01-01

    Fort St. Vrain (FSV), on the system of the Public Service Company of Colorado, is the only high temperature gas-cooled (HTGR) power reactor in the United States. The plant features a helium-cooled reactor with a uranium-thorium fuel cycle. The paper describes the experience made during its operation. (author). 2 refs, 4 figs, 2 tabs

  7. Comparing PRAs with operating experience

    International Nuclear Information System (INIS)

    Picard, R.R.; Martz, H.F.

    1998-01-01

    Probabilistic Risk Assessment is widely used to estimate the frequencies of rare events, such as nuclear power plant accidents. An obvious question concerns the extent to which PRAs conform to operating experience--that is, do PRAs agree with reality? The authors discuss a formal methodology to address this issue and examine its performance using plant-specific data

  8. 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

  9. Investigations of the applicability of a new accountancy tool in a closed tritium loop

    Energy Technology Data Exchange (ETDEWEB)

    Ebenhöch, S., E-mail: sylvia.ebenhoech@kit.edu; Niemes, S.; Priester, F.; Röllig, M.

    2016-11-01

    Highlights: • We have set up a new test device for measuring of tritiated gas samples. • The device is very compact and easy and reliable in operation. • Easy integration in flow-through systems. • The device has been operated at Tritium Laboratory Karlsruhe for several months. • The lower detection limit has been improved with regard to predecessors experiments. - Abstract: A commonly used activity monitoring method for tritium accountancy and process monitoring in tritium technology is ionization counting. Despite the wide use of ionization chambers (IC), they have several drawbacks like a strong gas species and pressure dependency. Furthermore, if compact systems are needed, there is also the necessity for process gas pressures >10 kPa. To encounter these drawbacks, the TRitium Activity Chamber Experiment (TRACE) has been developed at the Tritium Laboratory Karlsruhe (TLK) as a compact tritium monitor based on the beta induced X-ray spectrometry (BIXS) principle. TRACE can be used as an accountancy tool in tritium-processing facilities like the KArlsruhe TRItium Neutrino (KATRIN) experiment. In contrast to ICs TRACE shows a linear response to pressure changes up to approx. 1 kPa. The results of performed flow-through measurements confirm that TRACE is a complement for ICs in the low-pressure regime. Furthermore the gas species dependency of TRACE is investigated both with tritium measurements and with Monte Carlo simulations.

  10. ERB-II operating experience

    International Nuclear Information System (INIS)

    Smith, R.N.; Cissel, D.W.; Smith, R.R.

    1977-01-01

    As originally designed and operated, EBR-II successfully demonstrated the concept of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle (mini-nuclear park). Subsequent operation has been as an irradiation facility, a role which will continue into the foreseeable future. Since the beginning of operation in 1961, operating experience of EBR-II has been very satisfactory. Most of the components and systems have performed well. In particular, the mechanical performance of heat-removal systems has been excellent. A review of the operating experience reveals that all the original design objectives have been successfully demonstrated. To date, no failures or incidents resulting in serious in-core or out-of-core consequences have occurred. No water-to-sodium leaks have been detected over the life of the plant. At the present time, the facility is operating very well and continuously except for short shutdowns required by maintenance, refueling, modification, and minor repair. A plant factor of 76.9% was achieved for the calendar year 1976

  11. Emergency planning and operating experience

    International Nuclear Information System (INIS)

    Halpern, O.; Breniere, J.

    1984-01-01

    The purpose of this paper is to derive lessons from operating experience for the planning of emergency measures. This operating experience has two facets: it is obtained not only from the various incidents and accidents which have occurred in countries with nuclear power programmes and from the resulting application of emergency plans but also from the different exercises and simulations carried out in France and in other countries. Experience generally confirms the main approaches selected for emergency plans. The lessons to be derived are of three types: first, it appears necessary to set forth precisely the responsibilities of each person involved in order to prevent a watering-down of decisions in the event of an accident; secondly, considerable improvements need to be made in the different communication networks to be used; and thirdly, small accidents with minor radiological consequences deserve as systematic and thorough an approach as large and more improbable accidents. (author)

  12. 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)

  13. 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

  14. Direct search for neutrino mass and anomaly in the tritium beta-spectrum: Status of 'Troitsk neutrino mass' experiment

    International Nuclear Information System (INIS)

    Lobashev, V.M.; Aseev, V.N.; Belesev, A.I.; Berlev, A.I.; Geraskin, E.V.; Golubev, A.A.; Kazachenko, O.V.; Kuznetsov, Yu.E.; Ostroumov, R.P.; Rivkis, L.A.; Stern, B.E.; Titov, N.A.; Zadoroghny, C.V.; Zakharov, Yu.I.

    2000-01-01

    Results of the 'Troitsk ν-mass' experiment on search for the neutrino rest mass in the tritium beta-decay are presented. New data on the time dependence of the anomalous, bump-like structure at the end of the beta spectrum reported earlier are discussed. Possible systematics is considered in view of contradiction of 'Troitsk nu-mass' observation with those of 'Mainz neutrino' set-up. An upper limit for electron antineutrino rest mass remains at m ν 2 at 95% C.L

  15. 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

  16. FDNH - the tritium module in RODOS

    International Nuclear Information System (INIS)

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

    2001-01-01

    Under the auspices of its RTD (Research and Technological Development) Framework Programmes, the European Commission has supported the development of the RODOS (Real-time On-line Decision Support) system for off-site emergency management. The project started in 1989 focusing on PWR/LWR type accidents and using experience from the Chernobyl accident. In 1997 it was realised that tritium should be included in the list of radionuclides, as large tritium sources exists in Europe and to allow a potential expansion of the RODOS system for application on future fusion reactor accidents. The National Institute for Physics and Nuclear Engineering (IFIN-HH) in Romania - in close co-operation with the Research Centre Karlsruhe (FZK) - was charged to develop the tritium module, based on previous experience in environmental tritium modelling and the operation of CANDU reactors in Romania (with potential tritium accidents). At present, the Food and Dose Module Hydrogen -(FDMH) - for tritium applications - is integrated and documented in the RODOS system. It calculates the time dependent tritium concentration (as tritiated water or organically bound tritium) in crops (as much as 22 different species) and up to 12 animal products, inhalation doses and ingestion dose from up to 34 diet items for various groups of the population and for up to 2520 locations around the source, following an accidental emission of tritiated water. FDMH incorporates many improved techniques in radiological assessment and makes intensively use of interdisciplinary research. It is developed in a modular structure with a variable time grid according to the physical processes. Differing from other models, using generic transfer parameters or parameters fitted on individual experiments, FDMH derives tritium transfer rates based on physical and physiological process analysis, using scientifically accepted results from interdisciplinary research on, among others, land-atmosphere interaction, water cycle in the

  17. Tritium assay of Li/sub 2/O in the LBM/LOTUS experiments

    International Nuclear Information System (INIS)

    Quanci, J.; Azam, S.; Bertone, P.

    1986-01-01

    The Lithium Blanket Module (LBM) is an assembly of over 20,000 cylindrical lithium oxide pellets in an array representative of a limited-coverage breeding zone for a toroidal fusion device. A principal objective of the LBM program is to test the ability of advanced neutronics coding to model the tritium breeding characteristics of a fusion device blanket. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a 14 MeV point-neutron source. Princeton Plasma Physics Laboratory (PPPL) and EPFL assayed the tritium bred in lithium oxide diagnostic samples placed at various positions in the LBM. PPPL employed a thermal extraction technique while EPFL used a dissolution method. The results for the assay are reported and compared to MCNP Monte Carlo neutronics calculations for the LBM/LOTUS system

  18. Cellular responses to tritium exposure in rainbow trout: HTO- and OBT-spiked feed exposure experiments

    Energy Technology Data Exchange (ETDEWEB)

    Festarini, A.; Shultz, C.; Stuart, M.; Kim, S.B., E-mail: amy.festarini@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Ferreri, C. [National Research Council of Italy, Dept. of Chemical Sciences and Materials Technologies, Bologna (Italy)

    2016-06-15

    Biological effects were evaluated in rainbow trout (Oncorhynchus mykiss) exposed to tritiated water (HTO) or food spiked with organically bound tritium (OBT). An HTO exposure study was conducted using a tritium activity concentration of 7000 Bq/L, and an OBT exposure study was conducted using a tritium activity concentration of 30 000 Bq/L. Following 140 days of in vivo HTO exposure, liver, heart, spleen, kidney, and brain cells did not show statistically significant differences in viability; kidney, liver, and spleen cells did not show significant differences in DNA double-strand break repair activity compared with control cells. Membrane fatty acid composition analysis was conducted on liver cells and no effects of HTO exposure could be detected. Following 140 days of in vivo OBT exposure, viability and DNA double-strand break repair activity were not statistically different from controls in liver, heart, spleen, kidney, and brain cells. Changes, however, were noted in the fatty acid composition of liver and muscle tissues. For both studies, all measurements were performed on each tissue and on a fraction of the same tissue that was exposed to a gamma 4 Gy dose in vitro to test for adaptive responses, and no effects were observed except for fatty acid composition. The findings demonstrated that membrane fatty acid composition is a sensitive marker and that microscopic evaluation of gamma-H2AX foci is more sensitive than the flow cytometric approach. These studies are the first to correlate uptake and depuration with biological health indicators in edible fish for tritium exposures within worldwide drinking water guidelines. (author)

  19. Cellular responses to tritium exposure in rainbow trout: HTO- and OBT-spiked feed exposure experiments

    International Nuclear Information System (INIS)

    Festarini, A.; Shultz, C.; Stuart, M.; Kim, S.B.; Ferreri, C.

    2016-01-01

    Biological effects were evaluated in rainbow trout (Oncorhynchus mykiss) exposed to tritiated water (HTO) or food spiked with organically bound tritium (OBT). An HTO exposure study was conducted using a tritium activity concentration of 7000 Bq/L, and an OBT exposure study was conducted using a tritium activity concentration of 30 000 Bq/L. Following 140 days of in vivo HTO exposure, liver, heart, spleen, kidney, and brain cells did not show statistically significant differences in viability; kidney, liver, and spleen cells did not show significant differences in DNA double-strand break repair activity compared with control cells. Membrane fatty acid composition analysis was conducted on liver cells and no effects of HTO exposure could be detected. Following 140 days of in vivo OBT exposure, viability and DNA double-strand break repair activity were not statistically different from controls in liver, heart, spleen, kidney, and brain cells. Changes, however, were noted in the fatty acid composition of liver and muscle tissues. For both studies, all measurements were performed on each tissue and on a fraction of the same tissue that was exposed to a gamma 4 Gy dose in vitro to test for adaptive responses, and no effects were observed except for fatty acid composition. The findings demonstrated that membrane fatty acid composition is a sensitive marker and that microscopic evaluation of gamma-H2AX foci is more sensitive than the flow cytometric approach. These studies are the first to correlate uptake and depuration with biological health indicators in edible fish for tritium exposures within worldwide drinking water guidelines. (author)

  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. In situ tritium from Li2O and Li2ZrO3 irradiated in a fast neutron flux: BEATRIX-II, Phase 1 and 2

    International Nuclear Information System (INIS)

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

    1994-06-01

    BEATRIX-II was an in situ tritium recovery experiment. This in situ tritium recovery experiment provided data on the performance of Li 2 O and Li 2 ZrO 3 under irradiation conditions covering a range of sweep gas compositions and temperatures. The experiment consisted of two separate irradiation cycles which in turn included two vented tritium recovery canisters each. Phase 1 operated for 300 Effective Full Power Days (EFPD) while Phase 2 operated for 203 EFPD of irradiation. The tritium recovery behavior of Li 2 O and Li 2 ZrO 3 was characterized using temperature transients, sweep gas composition changes and reactor power changes to effect changes in the tritium inventory. The high neutron flux level in FFTF resulted in high tritium generation rates which combined with a responsive tritium measurement system allowed detailed observations on the tritium recovery behavior of Li 2 O and Li 2 ZrO 3 . During the course of the experiment a number of observations were made which did not appear consistent with a simplified view of the tritium recovery behavior of these materials. These observations included small negative tritium recovery peaks preceding the typical primary positive peaks; and, for specific temperature ranges, changes in tritium inventory which where opposite to the characteristic expected changes

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  6. Tritium transport and release from lithium ceramic breeder materials

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. 13C-Tracer Experiments in DIII-D Preliminary to Thermal Oxidation Experiments to Understand Tritium Recovery in DIII-D, JET, C-Mod, and MAST

    International Nuclear Information System (INIS)

    Stangeby, P.; Allen, S.; Bekris, N.; Brooks, N.; Christie, K.; Chrobak, C.; Coad, J.; Counsell, G.; Davis, J.; Elder, J.; Fenstermacher, M.; Groth, M.; Haasz, A.; Likonen, J.; Lipschultz, B.; McLean, A.; Philipps, V.; Porter, G.; Rudakov, D.; Shea, J.; Wampler, W.; Watkins, J.; West, W.; Whyte, D.

    2006-01-01

    Retention of tritium in carbon co-deposits is a serious concern for ITER. Developing a reliable in-situ removal method of the co-deposited tritium would allow the use of carbon plasma-facing components which have proven reliable in high heat flux conditions and compatible with high performance plasmas. Thermal oxidation is a potential solution, capable of reaching even hidden locations. It is necessary to establish the least severe conditions to achieve adequate tritium recovery, minimizing damage and reconditioning time. The first step in this multi-machine project is 13 C-tracer experiments in DIII-D, JET, C-Mod and MAST. In DIII-D and JET, 13 CH 4 has been (and in C-Mod and MAST, will be) injected toroidally symmetrically, facilitating quantification and interpretation of the results. Tiles have been removed, analyzed for 13 C content and will next be evaluated in a thermal oxidation test facility in Toronto with regard to the ability of different severities of oxidation exposure to remove the different types of (known and measured) 13 C co-deposit. Removal of D/T from B on Mo tiles from C-Mod will also be tested. OEDGE interpretive code analysis of the 13 C deposition patterns is used to generate the understanding needed to apply findings to ITER. First results are reported here for the 13 C injection experiments IN DIII-D

  8. 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.

  9. 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

  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. Tritium safety issues for TFCX

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  12. 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)

  13. Tritium isotopic exchange in air detritiation dryers

    International Nuclear Information System (INIS)

    Everatt, A.E.; Johnson, R.E.; Senohrabek, J.A.; Shultz, C.M.

    1989-02-01

    Isotopic exchange between tritiated and non-tritiated water species in a molecular sieve bed has been demonstrated. At high humidities (+6 degrees Celsius dew point) the rate of tritium isotopic exchange in a 2.4 L molecular sieve bed has been demonstrated to be at least 50% of published exchange rates. In an industrial-sized air detritiation dryer, utilizing the pretreatment technique of H 2 O steam washing to elute the residual tritium, a DF of 12 600 has been demonstrated when operating at an inlet vapor tritium concentration of 14 Ci/kg and at inlet and outlet dew points of 4.8 and -54 degrees Celsius, respectively. In the NPD dryer bed studied, which was not optimally designed for full benefit from isotopic exchange, at least one order of magnitude in additional detritiation is attributed to isotopic exchange in the unsaturated zone. The technique of eluting the residual tritium from an industrial sized bed by H 2 O washing at high temperature, high humidity and low bed loading has been demonstrated to be a fast and effective way of removing tritium from a molecular sieve bed during regeneration. The isotopic exchange model accurately predicted the exchange between tritiated and non-tritiated water species in a molecular sieve bed where there is no net adsorption or desorption. The model's prediction of the tritium breakthrough trend observed in the NPD tests was poor; however, a forced fit can be achieved if the exchange rates in the MTZ and the unsaturated zone are manipulated. More experiments are needed to determine the relative rates of tritium exchange in the saturated, mass transfer, and unsaturated zones of a dryer bed

  14. 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

  15. Operating modes of electrochemical H-concentration probes for tritium sensors

    International Nuclear Information System (INIS)

    Juhera, E.; Colominas, S.; Abellà, J.

    2015-01-01

    Highlights: • Synthesis and chemical characterization of Sr(Ce_0_._9–Zr_0_._1)_0_._9_5Yb_0_._0_5O_3_−_α proton conductor ceramic. • Evaluation of the sensor performance at different hydrogen concentrations. • Two different operating modes of the sensors: amperometric and potentiometric. • In amperometric mode sensor sensitivity can be tuned by changing the applied voltage. - Abstract: Potentiometric hydrogen sensors using different solid-state electrolytes have been designed and tested at the Electrochemical Methods Lab at Institut Quimic de Sarria (IQS). The most promising element (Sr(Ce_0_._9–Zr_0_._1)_0_._9_5Yb_0_._0_5O_3_−_α) has been selected for this work in order to evaluate the sensor performance at different hydrogen concentrations in two different operating modes: amperometric and potentiometric. In addition, the sensor response has been evaluated at different working temperatures (500, 575 and 650 °C). The experiments performed proved that when the sensor was used in a potentiometric mode, there is a threshold hydrogen concentration that the sensor can detect depending on the working conditions; 15 mbar at 575 °C and 10 mbar 650 °C. At 500 °C the minimum working temperature of this ceramic has not been achieved, so large deviations between experimental data and theoretical calculations has been obtained. When the sensor was used in an amperometric mode the obtained currents increased as a function of the applied voltage. At a fixed potential, the higher the temperature the higher the current was. So the sensor sensitivity can be tuned by changing the applied voltage at a fixed temperature and hydrogen concentration.

  16. Tritium Research Laboratory safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    Wright, D.A.

    1979-03-01

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

  17. Tritium Research Laboratory safety analysis report

    International Nuclear Information System (INIS)

    Wright, D.A.

    1979-03-01

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

  18. Operation experience at the UWTF

    International Nuclear Information System (INIS)

    Ueno, Kazuhiro; Inada, Kameji; Ohmori, Kouji; Usui, Kazuya; Irinouchi, Sigenori; Asami, Makoto; Tohchi, Katsunori

    2003-01-01

    This report describes the operation experience on the volume reduction of metal wastes and used air filters contaminated with uranium at the Uranium contaminated Waste Treatment Facility (UWTF) in JNC Tokai Works. The UWTF consists of the metal waste treatment system and the filter-waste treatment system. The former treats metal wastes, the latter treats used air filters. Metal wastes are unpacked from drums, cut, and then compacted. Used air filters are separated into filter media and frames. Then the filter media are compacted and the frames are crushed. The operation of the UWTF was started in June 1998. The following volumes of wastes had been treated at the UWTF from the beginning of the operation to March 2003 (for about 5 years). (1) 1,524 drums of the metal wastes had been reduced to 410 drums. The volume reduction factor was 3.7. (2) 372 drums of the used air filters had been reduced to 39 drums. The volume reduction factor was 9.5. These systems have been operated without trouble for 5 years and have demonstrated to be able to reduce the volumes of the wastes to designed values. The volume reduction technologies for metal wastes and used air filters contaminated with uranium were successfully demonstrated at the UWTF. (author)

  19. ATLAS Pixel Detector Operational Experience

    CERN Document Server

    Di Girolamo, B; The ATLAS collaboration

    2011-01-01

    The ATLAS Pixel Detector is the innermost detector of the ATLAS experiment at the Large Hadron Collider at CERN, providing high-resolution measurements of charged particle tracks in the high radiation environment close to the collision region. This capability is vital for the identification and measurement of proper decay times of long-lived particles such as b-hadrons, and thus vital for the ATLAS physics program. The detector provides hermetic coverage with three cylindrical layers and three layers of forward and backward pixel detectors. It consists of approximately 80 million pixels that are individually read out via chips bump-bonded to 1744 n-in-n silicon substrates. In this talk, results from the successful operation of the Pixel Detector at the LHC will be presented, including monitoring, calibration procedures, timing optimization and detector performance. The detector performance is excellent: 96.9% of the pixels are operational, noise occupancy and hit efficiency exceed the design specification, an...

  20. Tritium waste control: April--September 1977

    International Nuclear Information System (INIS)

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  2. Comparison of inventory of tritium in various ceramic breeder blankets

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Regeneration and tritium recovery from the large JET neutral injection cryopump system after the FTE

    International Nuclear Information System (INIS)

    Obert, W.; Bell, A.; Davies, J.; Mayaux, C.; Perinic, G.; Saibene, G.; Sartori, R.; Thompson, E.; Anderson, J.; Jenkins, E.; Walthers, C.

    1992-01-01

    Neutral Beam Injection (NBI) was used to introduce tritium into the plasma for the First Tritium Experiment In addition to the decisive advantage of depositing the tritium into the centre of the plasma, the use of NBI also minimized the total quantity of tritium introduced into the Torus and the contamination of the vacuum vessel. However, because of the relatively low gas efficiency of the positive ion injection system approximately 95% of the total quantity of tritium introduced was pumped by the large condensation cryopumps which form an integral part of the injector. Several hardware and associated software changes were implemented in order to making provision for possible fault scenarios during operation with tritium and to ensure complete regeneration of the tritium from the cryopumps. The tritium released after all subsequent regeneration's has been monitored carefully in order to determine the amount of tritium retained by the black anodized liquid nitrogen panel surfaces of the cryopump and to compare it with experiments at TSTA on JET samples before the FTE

  4. 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)

  5. 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)

  6. Stack Monitor Operating Experience Review

    International Nuclear Information System (INIS)

    Cadwallader, L.C.; Bruyere, S.A.

    2009-01-01

    Stack monitors are used to sense radioactive particulates and gases in effluent air being vented from rooms of nuclear facilities. These monitors record the levels and types of effluents to the environment. This paper presents the results of a stack monitor operating experience review of the U.S. Department of Energy (DOE) Occurrence Reporting and Processing System (ORPS) database records from the past 18 years. Regulations regarding these monitors are briefly described. Operating experiences reported by the U.S. DOE and in engineering literature sources were reviewed to determine the strengths and weaknesses of these monitors. Electrical faults, radiation instrumentation faults, and human errors are the three leading causes of failures. A representative 'all modes' failure rate is 1E-04/hr. Repair time estimates vary from an average repair time of 17.5 hours (with spare parts on hand) to 160 hours (without spare parts on hand). These data should support the use of stack monitors in any nuclear facility, including the National Ignition Facility and the international ITER project.

  7. Operation experience with elevated ammonia

    International Nuclear Information System (INIS)

    Vankova, Katerina; Kysela, Jan; Malac, Miroslav; Petrecky, Igor; Svarc, Vladimir

    2011-01-01

    The 10 VVER units in the Czech and Slovak Republics are all in very good water chemistry and radiation condition, yet questions have arisen regarding the optimization of cycle chemistry and improved operation in these units. To address these issues, a comprehensive experimental program for different water chemistries of the primary circuit was carried out at the Rez Nuclear Research Institute, Czech Republic, with the goal of judging the influence of various water chemistries on radiation build-up. Four types of water chemistries were compared: standard VVER water chemistry (in common use), direct hydrogen dosing without ammonia, standard VVER water chemistry with elevated ammonia levels, and zinc dosing to standard VVER water chemistry. The test results showed that the types of water chemistry other than the common one have benefits for the operation of the nuclear power plant (NPP) primary circuit. Operation experience with elevated ammonia at NPP Dukovany Units 3 and 4 is presented which validates the experimental results, demonstrating improved corrosion product volume activity. (orig.)

  8. Tritium emissions from a detritiation facility

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  9. TRIO experiment

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Malecha, R.F.

    1984-09-01

    The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an anaytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion

  10. Tritium production in fusion reactors

    International Nuclear Information System (INIS)

    Roth, E.

    1981-08-01

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

  11. 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

  12. Tritium Systems Test Facility

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  13. 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)

  14. Catalyst study for the decontamination of glove-box atmospheres containing tritium at MPC levels

    International Nuclear Information System (INIS)

    Chobot, J.; Montel, J.; Sannier, J.

    1988-01-01

    The BEATRICE loop was designed for studying the conversion of tritium at very low activity levels using catalytic oxidation followed by water trapping. The purpose is to study kinetic parameters required for the design of the NET tritium cleanup system with the two main objectives to operate without isotopic swamping and to determine the ability of efficient conversion at room temperature. From experiments carried out between 20 and 250 0 C it is concluded that two palladium/alumina and platinum/alumina catalysts are very efficient in removing tritium from contaminated gas mixtures down to a few MPC levels without isotopic swamping and even at room temperature. However at room temperature, in relation to tritium species trapped on the catalyst surface a progressive deactivation with time occurs. This phenomenon may be a concern for process efficiency and tritium inventory and regeneration conditions have to be determined in order to demonstrate industrial feasibility of operating at room temperature

  15. Catalyst study for the decontamination of glove-boxe atmospheres containing tritium at MPC levels

    International Nuclear Information System (INIS)

    Chabot, J.; Montel, J.; Sannier, J.

    1988-01-01

    The BEATRICE loop was designed for studying the conversion of tritium at very low activity levels using catalytic oxidation followed by water trapping. The purpose is to study kinetic parameters required for the design of the NET tritium clean-up system with the two main objectives to operate without isotopic swamping and to determine the ability of efficient conversion at room temperature. From experiments carried out between 20 and 250 0 C it is concluded that two palladium/alumina and platinum/alumina catalysts are very efficient in removing tritium from contaminated gas mixtures down to a few MPC levels without isotopic swamping and even at room temperature. However at room temperature, in relation to tritium species trapped on the catalyst surface a progressive deactivation with time occurs. This phenomenon may be a concern for process efficiency and tritium inventory and best regeneration conditions have to be determined in order to demonstrate industrial feasibility of operating at room temperature

  16. Operating experience of Fugen Nuclear Power Station

    International Nuclear Information System (INIS)

    Ohteru, Shigeru; Kaneko, Jun; Kawahara, Toshio; Matsumoto, Mitsuo

    1987-01-01

    The prototype ATR 'Fugen' developed as one of the national project has verified the performance and reliability of the advanced thermal reactor system through the operation for about eight years since 1979, and the elucidation of the characteristics in plutonium utilization and the development and verification of the tuilizing techniques have been advanced. Besides, the operational results and the achievement of the technical development are successively reflected to the design of a demonstration reactor. In this paper, the outline of Fugan and the operational results are reported. The ATR Fugen Power Station is that of the prototype reactor of heavy water moderated, boiling light water cooled, pressure tube type, having the electric output of 165 MW. It started the full scale operation on March 20, 1979, and as of January, 1987, the total generated electric power reached about 7 billion kWh, the time of power generation was about 43,000 h, and the average capacity factor was 60.6 %. Plutonium utilization techniques, the flow characteristics and the dynamic plant characteristics of a pressure tube type reactor, the operational characteristics of a heavy water system and the techniques of handling heavy water containing tritium, and the operational reliability and maintainability of the machinery and equipment installed have been studied. (Kako, I.)

  17. 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)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  19. Regulatory challenges in using nuclear operating experience

    International Nuclear Information System (INIS)

    2006-01-01

    There can be no doubt that the systematic evaluation of operating experience by the operator and the regulator is essential for continued safe operation of nuclear power plants. Recent concerns have been voiced that the operating experience information and insights are not being used effectively to promote safety. If these concerns foreshadow a real trend in OECD countries toward complacency in reporting and analysing operating events and taking corrective actions, then past experience suggests that similar or even more serious events will recur. This report discusses how the regulator can take actions to assure that operators have effective programmes to collect and analyse operating experience and, just as important, for taking steps to follow up with actions to prevent the events and conditions from recurring. These regulatory actions include special inspections of an operator operating experience programme and discussion with senior plant managers to emphasize the importance of having an effective operating experience programme. In addition to overseeing the operator programmes, the regulator has the broader responsibility for assuring that industry-wide trends, both national and international are monitored. To meet these responsibilities, the regulatory body must have its own operating experience programme, and this report discusses the important attributes of such regulatory programmes. It is especially important for the regulator to have the capability for assessing the full scope of operating experience issues, including those that may not be included in an operator operating experience programme, such as new research results, international operating experience, and broad industry trend information. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  1. Tritium effluent removal system

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  2. Tritium accountancy and unmeasurable inventories in fusion reactors

    International Nuclear Information System (INIS)

    Avenhaus, R.; Spannagel, G.

    1997-12-01

    For the time being fusion technology development involves relatively small quantities of tritium. Consequently, it is sufficient to apply so-called ''conventional'' accountancy tools. However, it is foreseeable that tritium operations - and thus the amount of tritium - will increase substantially. An advanced accountancy methodology will satisfy the resulting new requirements. In this study such an advanced accountancy methodolody is developed and applied to the situation envisaged with idealized experiments of the Karlsruhe Tritium Laboratory (TLK) as well as an idealized ITER-type fuel cycle. Firstly, this task comprises modeling of fuel cycle operations, providing the ''true'' data of the in-process inventories. As both the fuel cycle subsystems and networking themselves are susceptible to changes, a measurement model takes care of the true data, handles data reduction, and applies mathematical methods to confirm the final inventories on a statistical basis. Then, in a third step, the test statistics might verify whether or not a tritium anomaly, e.g. a tritium loss has occurred. Since the statistical analysis generates problems the solutions of which are not part of the standard statistical literature in the Annex the results of the related original work is presented in mathematical-abstract form. (orig.)

  3. Tritium analysis at TFTR

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  4. Study on tritium recovery from breeder materials

    International Nuclear Information System (INIS)

    Moriyama, H.; Moritani, K.

    1997-01-01

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

  5. Tritium breeding materials

    International Nuclear Information System (INIS)

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

    1984-03-01

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

  6. Tritium breeding materials

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  7. Tritium monitoring at the Sandia Tritium Research Laboratory

    International Nuclear Information System (INIS)

    Devlin, T.K.

    1978-10-01

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

  8. Tritium behavior intentionally released in the room

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  9. Tritium research and technology facilities at the JRC-Ispra

    International Nuclear Information System (INIS)

    Dworschak, H.; Mannone, F.; Perujo, A.; Pierini, G.; Reiter, F.; Vassallo, G.; Viola, A.; Camposilvan, J.; Douglas, K.; Grassi, G.; Lolli Ceroni, P.; Simonetta, A.; Spelta, B.

    1990-01-01

    A set of experiments which are of prominent interest for the development of nuclear fusion technology in Europe are planned by the JRC-Ispra for the near future, in the frame of experimental activities to be performed in ETHEL, the European Tritium Handling Experimental Laboratory under construction at the Ispra site. These experiments already included for the most part as JRC-Task Action Sheets in the 1989-1991 European Technology Programme Actions will initiate in ETHEL on a fully active laboratory scale starting mid-1991. They will concern the following research areas: Recycling of tritium from first wall materials; Tritium recovery from water cooled Pb-17Li blankets; Detritiation of ventilation atmospheres; Plasma exhaust processing; Tritiazed waste management. In view of fully active tritium experiments in ETHEL and to obtain information of the basic processes involved, since 1985 preparatory experimental studies are being performed at the JRC-Ispra laboratories using hydrogen and deuterium. Furthermore, always with regard to ETHEL experiments, particular attention is given to possible technical and managerial problems which potentially may arise in this context. To identify at an early stage such problems a questionnaire has been developed and distributed to researchers in conjunction with an ETHEL information packet. The questionnaire demands information regarding the scope, design and operation of the intended experiment as well as planning and required support to be supplied by ETHEL. A brief description of experimental preparatory studies and future tritium handling experiments in ETHEL as well of the ETHEL facility is here presented. (orig.)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  11. Tritium - is it underestimated

    International Nuclear Information System (INIS)

    Whitlock, G.D.

    1980-01-01

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

  12. 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.

  13. Tritium accountancy in fusion systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  14. Tritium in the food chain

    International Nuclear Information System (INIS)

    Koenig, L.A.

    1988-01-01

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

  15. 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)

  16. Rate of generation of tritium during the operation of Tsuruga Power Station Unit No. 2

    International Nuclear Information System (INIS)

    Funamoto, Hisao; Yoshinari, Masaharu; Fukuda, Masayuki; Makino, Shinichi; Watari, Tuneo

    1994-01-01

    Total amount of 3 H activity in primary coolant due to the operation of Tsuruga Power Station Unit No. 2 was estimated. The 3 H inventory was measured for samples from the spent fuel pool, primary coolant and miscellaneous tanks. From the result of the measurement and the data of environmental release of 3 H, the rate of generation of 3 H in the reactor was found to be 25 TBq/GWa. Since Tsuruga Power Station Unit No. 2 is a PWR type reactor, we presume that most of the 3 H in primary coolant is formed by 10 B(n, 2α) 3 H reaction. It is necessary to release about 23 TBq/GWa of 3 H to maintain the station inventory at the present level. (author)

  17. 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

  18. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

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

    1985-08-01

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

  19. Tritium control and accountability instructions

    International Nuclear Information System (INIS)

    Wall, W.R.

    1981-03-01

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

  20. Tritium recovery from tritiated water with a two-stage palladium membrane reactor

    International Nuclear Information System (INIS)

    Birdsell, S.A.; Willms, R.S.

    1997-01-01

    A process to recover tritium from tritiated water has been successfully demonstrated at TSTA. The 2-stage palladium membrane reactor (PMR) is capable of recovering tritium from water without generating additional waste. This device can be used to recover tritium from the substantial amount of tritiated water that is expected to be generated in the International Thermonuclear Experimental Reactor both from torus exhaust and auxiliary operations. A large quantity of tritiated waste water exists world wide because the predominant method of cleaning up tritiated streams is to oxidize tritium to tritiated water. The latter can be collected with high efficiency for subsequent disposal. The PMR is a combined catalytic reactor/permeator. Cold (non-tritium) water processing experiments were run in preparation for the tritiated water processing tests. Tritium was recovered from a container of molecular sieve loaded with 2,050 g (2,550 std. L) of water and 4.5 g of tritium. During this experiment, 27% (694 std. L) of the water was processed resulting in recovery of 1.2 g of tritium. The maximum water processing rate for the PMR system used was determined to be 0.5 slpm. This correlates well with the maximum processing rate determined from the smaller PMR system on the cold test bench and has resulted in valuable scale-up and design information

  1. Tritium recovery from tritiated water with a two-stage palladium membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Birdsell, S.A.; Willms, R.S.

    1997-04-01

    A process to recover tritium from tritiated water has been successfully demonstrated at TSTA. The 2-stage palladium membrane reactor (PMR) is capable of recovering tritium from water without generating additional waste. This device can be used to recover tritium from the substantial amount of tritiated water that is expected to be generated in the International Thermonuclear Experimental Reactor both from torus exhaust and auxiliary operations. A large quantity of tritiated waste water exists world wide because the predominant method of cleaning up tritiated streams is to oxidize tritium to tritiated water. The latter can be collected with high efficiency for subsequent disposal. The PMR is a combined catalytic reactor/permeator. Cold (non-tritium) water processing experiments were run in preparation for the tritiated water processing tests. Tritium was recovered from a container of molecular sieve loaded with 2,050 g (2,550 std. L) of water and 4.5 g of tritium. During this experiment, 27% (694 std. L) of the water was processed resulting in recovery of 1.2 g of tritium. The maximum water processing rate for the PMR system used was determined to be 0.5 slpm. This correlates well with the maximum processing rate determined from the smaller PMR system on the cold test bench and has resulted in valuable scale-up and design information.

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

    International Nuclear Information System (INIS)

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

    1990-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-05-01

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

  4. Tritium oxidation and exchange: preliminary studies

    International Nuclear Information System (INIS)

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

    1978-05-01

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

  5. Operating experience feedback program at Olkiluoto NPP

    International Nuclear Information System (INIS)

    Kosonen, Mikko

    2002-01-01

    Recent review and development of the operating experience feedback program will be described. The development of the program has been based on several reviews by outside organizations. Main conclusions from these review reports and from the self assessment of safety performance, safety problems and safety culture on the basis of the operational events made by ASSET-method will be described. An approach to gather and analyze small events - so-called near misses - will be described. The operating experience program has been divided into internal and external operating experience. ASSET-methodology and a computer program assisting the analysis are used for the internal operating experience events. Noteworthy incidents occurred during outage are analyzed also by ASSET-method. Screening and pre analysis of the external operating experience relies on co-operation with ERFATOM, an organization of Nordic utilities for the exchange of nuclear industry experience. A short presentation on the performance of the Olkiluoto units will conclude the presentation. (author)

  6. Recent U.S. reactor operating experience

    International Nuclear Information System (INIS)

    Stello, V. Jr.

    1977-01-01

    A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

  7. BWR and ABWR operating experience

    International Nuclear Information System (INIS)

    Nagai, Kimio; Takayama, Yoshito; Shimizu, Shunichi

    1998-01-01

    The first commercial operation of a nuclear power plant in Japan was in 1969. At present, there are 52 nuclear power plants operating in the country, accounting for about 35% of the total electricity generated and about 20 % of the total capacity of electricity generation facilities in Japan. Moreover, Japan has had the highest facility utilization rate, which is an indicator of operational safety, among the top four nuclear-generation countries (U.S.A., France, Germany, and Japan) every year since 1993, and has maintained a utilization rate of more than 80% since 1995. Toshiba has supported plant operation and maintenance in 19 nuclear power plants. The overall facility utilization rate of these plants attained a record of 86.4% in fiscal 1996. Furthermore, the averaged rates over the past three fiscal years from 1995 to 1997 have been the highest in the world among plant constructors such as GE, WH, and Siemens. (author)

  8. Tritium monitor for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jalbert, R.A.

    1982-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

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

    International Nuclear Information System (INIS)

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

    1977-04-01

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

  11. Nuclear units operating improvement by using operating experience

    International Nuclear Information System (INIS)

    Rotaru, I.; Bilegan, I.C.

    1997-01-01

    The paper presents how the information experience can be used to improve the operation of nuclear units. This areas include the following items: conservative decision making; supervisory oversight; teamwork; control room distraction; communications; expectations and standards; operator training and fundamental knowledge, procedure quality and adherence; plant status awareness. For each of these topics, the information illustrate which are the principles, the lessons learned from operating experience and the most appropriate exemplifying documents. (authors)

  12. Operational experience with superconducting synchrotron magnets

    International Nuclear Information System (INIS)

    Martin, P.S.

    1987-03-01

    The operational experience with the Fermilab Tevatron is presented, with emphasis on reliability and failure modes. Comprisons are made between the operating efficiencies for the superconducting machine and for he conventional Main Ring

  13. Operational experience with superconducting synchrotron magnets

    International Nuclear Information System (INIS)

    Martin, P.S.

    1987-01-01

    The operational experience with the Fermilab Tevatron is presented, with emphasis on reliability and failure modes. Comparisons are made between the operating efficiencies for the superconducting machine and for the conventional Main Ring

  14. 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)

  15. Tritium system design studies of fusion experimental breeder

    International Nuclear Information System (INIS)

    Deng Baiquan; Huang Jinhua

    2003-01-01

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

  16. Magnet operating experience review for fusion applications

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1991-11-01

    This report presents a review of magnet operating experiences for normal-conducting and superconducting magnets from fusion, particle accelerator, medical technology, and magnetohydrodynamics research areas. Safety relevant magnet operating experiences are presented to provide feedback on field performance of existing designs and to point out the operational safety concerns. Quantitative estimates of magnet component failure rates and accident event frequencies are also presented, based on field experience and on performance of similar components in other industries

  17. Organically bound tritium (OBT) formation in rainbow trout (Oncorhynchus mykiss): HTO and OBT-spiked food exposure experiments

    International Nuclear Information System (INIS)

    Kim, S.B.; Shultz, C.; Stuart, M.; McNamara, E.; Festarini, A.; Bureau, D.P.

    2013-01-01

    In order to determine the rate of organically bound tritium (OBT) formation, rainbow trout (Oncorhynchus mykiss) were exposed to tritiated water (HTO) or OBT-spiked food. The HTO (in water) exposure study was conducted using a tritium activity concentration of approximately 7000 Bq/L and the OBT (in food) exposure study was conducted using a tritium activity concentration of approximately 30,000 Bq/L. Fish in both studies were expected to be exposed to similar tritium levels assuming 25% incorporation of the tritiated amino acids found in the food. Four different sampling campaigns of HTO exposure (Day 10, 30, 70, 140) and five different sampling campaigns of OBT-spiked food exposure (Day 9, 30, 70, 100, 140) were conducted to measure HTO and OBT activity concentrations in fish tissues. OBT depuration was also evaluated over a period of 30 days following the 140 d exposure studies. The results suggested that the OBT formation rate was slower when the fish were exposed to HTO compared to when the fish were ingesting OBT. In addition, the results indicated that OBT can bioaccumulate in fish tissues following OBT-spiked food exposure. - Highlights: ► The rate of organically bound tritium (OBT) formation was determined in rainbow trout. ► Rainbow trout were exposed to tritium in the form of tritiated water (HTO) and OBT-spiked food. ► OBT formation rate was slower when the fish were exposed to HTO compared to when the fish were ingesting OBT.

  18. 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.

  19. Imaging of tritium implanted into graphite

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  20. EBR-II operating experience

    International Nuclear Information System (INIS)

    Smith, C.R.F.

    1978-07-01

    Operation of the EBR-2 reactor is presented concerning the performance of the heat removal system; reactor materials; fuel handling system; sodium purification and sampling system; cover-gas purification; plant diagnostics and instrumentation; recent improvements in identifying fission product sources in EBR-2; and EBR-2 safety

  1. Catalytic oxidation efficiencies for tritium and tritiated methane in a mature, industrial-scale decontamination system

    International Nuclear Information System (INIS)

    Mintz, J.M.; Gildea, P.D.

    1981-01-01

    Almost all tritium decontamination systems proposed for fusion facilities employ catalytic oxidation to water, followed by drying, to remove tritium and tritiated hydrocarbons from gas streams. One such large-scale system, the gas purification system (GPS), has been operating in the Tritium Research Laboratory (TRL) at Sandia National Laboratories, Livermore, CA, since October 1977. A series of experiments have recently been conducted there to assesss the current operating characteristics of the GPS catalyst. The experiments used tritium and tritiated methane and covered a range of temperatures, flow rates, and concentration levels. When contrasted with 1977 data, the results indicate that no measurable degradation of catalyst function had occurred. However, some reduction in active metal surface area, as indicated by B.E.T. surface area measurements (approx. 100 → 90m 2 /g) and AES scans (approx. 1.4 → 0.9 at. % Pt), had occurred. Kinetic rate coefficients were also derived and a rough temperature dependence obtained

  2. Catalytic oxidation efficiencies for tritium and tritiated methane in a mature, industrial-scale decontamination system

    International Nuclear Information System (INIS)

    Mintz, J.M.; Gildea, P.D.

    1980-10-01

    Almost all tritium decontamination systems proposed for fusion facilities employ catalytic oxidation to water, followed by drying, to remove tritium and tritiated hydrocarbons from gas streams. One such large-scale system, the gas purification system (GPS), has been operating in the Tritium Research Laboratory (TRL) at Sandia National Laboratories, Livermore, CA, since October 1977. A series of experiments have recently been conducted there to assess the current operating characteristics of the GPS catalyst. The experiments used tritium and tritiated methane and covered a range of temperatures, flow rates, and concentration levels. When contrasted with 1977 data, the results indicate that no measurable degradation of catalyst function had occurred. However, some reduction in active metal surface area, as indicated by B.E.T. surface area measurements (approx. 100 → 90 m 2 /g) and AES scans (approx. 1.4 → 0.9 at% Pt), had occurred. Kinetic rate coefficients were also derived and a rough temperature dependence obtained

  3. The Safety and Tritium Applied Research (STAR) Facility: Status-2004

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Sharpe, J.P.; Schuetz, S.T.; Petti, D.A.

    2005-01-01

    The Safety and Tritium Applied Research (STAR) Facility, a US DOE National User Facility at the Idaho National Engineering and Environmental Laboratory (INEEL), comprises capabilities and infrastructure to support both tritium and non-tritium research activities important to the development of safe and environmentally friendly fusion energy. Research thrusts include (1) interactions of tritium and deuterium with plasma-facing-component (PFC) materials, (2) fusion safety issues [PFC material chemical reactivity and dust/debris generation, activation product mobilization, tritium behavior in fusion systems], and (3) molten salts and fusion liquids for tritium breeder and coolant applications. This paper updates the status of STAR and the capabilities for ongoing research activities, with an emphasis on the development, testing and integration of the infrastructure to support tritium research activities. Key elements of this infrastructure include a tritium storage and assay system, a tritium cleanup system to process glovebox and experiment tritiated effluent gases, and facility tritium monitoring systems

  4. Importance of root HTO uptake in controlling land-surface tritium dynamics after an-acute HT deposition: a numerical experiment

    International Nuclear Information System (INIS)

    Ota, Masakazu; Nagai, Haruyasu; Koarashi, Jun

    2012-01-01

    To investigate the role of belowground root uptake of tritiated water (HTO) in controlling land-surface tritium (T) dynamics, a sophisticated numerical model predicting tritium behavior in an atmosphere-vegetation-soil system was developed, and numerical experiments were conducted using the model. The developed model covered physical tritiated hydrogen (HT) transport in a multilayered atmosphere and soil, as well as microbial oxidation of HT to HTO in the soil, and it was incorporated into a well-established HTO-transfer organically bound tritium (OBT)-formation model. The model performance was tested through the simulation of an existing HT-release experiment. Numerical experiments involving a hypothetical acute HT exposure to a grassland field with a range of rooting depths showed that the HTO release from the leaves to the atmosphere, driven by the root uptake of the deposited HTO, can exceed the HTO evaporation from the ground surface to the atmosphere when root water absorption preferentially occurs beneath the ground surface. Such enhanced soil-leaf-atmosphere HTO transport, caused by the enhanced root HTO uptake, increased HTO concentrations in both the surface atmosphere and in the cellular water of the leaf. Consequently, leaf OBT assimilation calculated for shallow rooting depths increased by nearly an order of magnitude compared to that for large rooting depths. - Highlights: ► A model that calculates HT deposition from atmosphere to soil was developed. ► Tritium dynamics after an-acute HT deposition was studied by numerical experiments. ► OBT formation highly depends on magnitude of uptake of the deposited HTO by roots.

  5. Upgrade to the Tritium Remote Control and Monitoring System for TFTR D and D

    International Nuclear Information System (INIS)

    Sichta, P.; Oliaro, G.; Sengupta, S.

    2002-01-01

    Since 1988, the Tritium Remote Control and Monitoring System (TRECAMS) has performed crucial functions in support of D-T [deuterium-tritium] operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). Although plasma operations on TFTR were completed in 1997, the need for TRECAMS continued. During this period TRECAMS supported the TFTR tritium systems, the TFTR's Shutdown and Safing phase, and the TFTR Decontamination and Decommissioning (D and D) project. The most critical function of the TRECAMS in the post-TFTR era has been to provide a real-time indication of the airborne tritium levels in the tritium areas and the (HVAC) stacks. TRECAMS is a critical tool in conducting safe TFTR D and D tritium-line breaks and other tritium-related work activities. Beginning in 1998, the failure rate of the system's hardware sharply increased. Furthermore, the specialized knowledge required to maintain the original software and hardware was diminishing. It soon became apparent that a failure of the TRECAMS could significantly impact the TFTR D and D project's cost and schedule. To preclude this, the TRECAMS hardware and software was upgraded in the year 2000 to use modern components. This paper will describe that successful upgrade, including a review of the engineering processes and our operating experiences with the upgraded system

  6. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

  7. Experiences of an operating association

    International Nuclear Information System (INIS)

    Zefferer, H.

    1994-01-01

    The Austrian Draukraftwerke operate a coal power station with a capacity 110 MW el at St. Andraein Kaernten. In 1989, the first stage of a scheme for supplying the town St. Andraewith long-distance energy was started from this power station. To provide for the intermittent downtimes between the plant's operating times, which total an annual 2000 hours, a substitute aggregate was designed consisting of two biomass (chiefly bark)-fueled grate-firing boilers (4 MW th and 2.5 MW th ). The problems that arose during the first years in the areas of biomass storage, conveyor systems, slagging, emission values, and ash removal have meanwhile been solved satisfactorily. Moreover, the Draukraftwerke intend to use biomass for electricity generation in their coal power stations. (orig.) [de

  8. The Wonderland of Operating the ALICE Experiment

    CERN Document Server

    Augustinus, A; Pinazza, O; Rosinský, P; Lechman, M; Jirdén, L; Chochula, P

    2011-01-01

    ALICE is one of the experiments at the Large Hadron Collider (LHC), CERN, Geneva, Switzerland. Composed of 18 sub-detectors each with numerous subsystems that need to be controlled and operated in a safe and efficient way. The Detector Control System (DCS) is the key to this and has been used by detector experts with success during the commissioning of the individual detectors. During the transition from commissioning to operation, more and more tasks were transferred from detector experts to central operators. By the end of the 2010 datataking campaign, the ALICE experiment was run by a small crew of central operators, with only a single controls operator. The transition from expert to non-expert operation constituted a real challenge in terms of tools, documentation and training. A relatively high turnover and diversity in the operator crew that is specific to the HEP experiment environment (as opposed to the more stable operation crews for accelerators) made this challenge even bigger. Thi...

  9. Operational experience in underwater photogrammetry

    Science.gov (United States)

    Leatherdale, John D.; John Turner, D.

    Underwater photogrammetry has become established as a cost-effective technique for inspection and maintenance of platforms and pipelines for the offshore oil industry. A commercial service based in Scotland operates in the North Sea, USA, Brazil, West Africa and Australia. 70 mm cameras and flash units are built for the purpose and analytical plotters and computer graphics systems are used for photogrammetric measurement and analysis of damage, corrosion, weld failures and redesign of underwater structures. Users are seeking simple, low-cost systems for photogrammetric analysis which their engineers can use themselves.

  10. Development of a Remotely Operated, Field-Deployable Tritium Analysis System for Surface and Ground Water Measurement

    International Nuclear Information System (INIS)

    Hofstetter, K.J.; Cable, P.R.; Noakes, J.E.; Spaulding, J.D.; Neary, M. P.; Wasyl, M.S.

    1996-01-01

    The environmental contamination resulting from decades of testing and manufacturing of nuclear materials for a national defense purposes is a problem now being faced by the United States. The Center for Applied Isotope Studies at the University of Georgia, in cooperation with the Westinghouse Savannah River Company and Packard Instrument Company, have developed a prototype unit for remote, near real time, in situ analysis of tritium in surface and ground water samples

  11. US remote monitoring operational experience

    International Nuclear Information System (INIS)

    Dupree, S.A.; Sonnier, C.S.

    1997-01-01

    Under international partnerships and bilateral agreements with the U.S. Department of Energy, Sandia National Laboratories, other national laboratories, and international partner organizations have emplaced remote monitoring systems in nuclear facilities and laboratories in various parts of the world for the purpose of conducting field trials of remote monitoring. The purpose of the present report is to review the results from these field trials and draw general conclusions regarding the trials. Many thousands of hours of sensor and system operation have been logged, and data have been retrieved from many locations. In virtually all cases the system components have functioned as intended and data have been successfully collected and transmitted for review. Comparisons between front-end-triggered video and time-lapse video have shown that the triggered record has captured all relevant monitored operations at the various nuclear facilities included in the field trials. We believe the utility and functional reliability of remote monitoring for international safeguards has been shown. However, it should be kept in mind that openness and transparency, including some form of short-notice inspections, are likely to be prerequisites to the safeguards implementation of remote monitoring in any State

  12. The LLNL portable tritium processing system

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

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

  13. A tritium target system for μCF

    International Nuclear Information System (INIS)

    Zmeskal, J.; Ackerbauer, P.; Durham, W.B.; Heard, H.C.; Neumann, W.; Bossy, H.

    1990-12-01

    An apparatus has been constructed for the safe handling of tritium as part of a series of muon-catalyzed fusion experiments. The equipment was designed to handle 100 kCi of tritium. The main parts of this system are the oil-free high vacuum and transfer system, and the quadrupole mass analyzer for a direct determination of the target content. The system was used successfully for five continuous periods of operation of over one month each. A new target system was constructed at Lawrence Livermore National Laboratory (LLNL) for ultimate use at Paul Scherrer Institute (PSI) to investigate the high temperature and high pressure region. 9 refs., 4 figs

  14. The modelling of tritium behaviour in the environment

    International Nuclear Information System (INIS)

    Raskob, W.

    1991-01-01

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

  15. Setup and commissioning of a combined water detritiation and isotope separation experiment at the Tritium Laboratory Karlsruhe

    Energy Technology Data Exchange (ETDEWEB)

    Welte, S., E-mail: stefan.welte@kit.edu; Cristescu, I.; Dittrich, H.; Lohr, N.; Melzer, C.; Michling, R.; Plusczyk, C.; Schaefer, P.

    2013-10-15

    Highlights: • Technical scale, tritium compatible liquid phase catalytic exchange (LPCE). • Technical scale, tritium compatible cryogenic distillation. • Combines processing option for tritiated water and isotope separation. -- Abstract: The European union in kind supply for the ITER fuel cycle development consists, among others, of the water detritiation system (WDS) and the isotope separation system (ISS). In order to mitigate the release of tritium to the environment, these systems are combined by feeding hydrogen exhaust from the ISS into the WDS for final processing. Therefore, the WDS is the final tritium barrier before releasing hydrogen (H{sub 2}) exhaust to the environment. The TRENTA 4 scaled prototype facility at TLK is based on combination of the combined electrolysis and catalytic exchange (CECE) process and a cryogenic distillation (CD) process. All components are fully tritium compatible and controlled using a state of the art control system for process automation, backed up by an additional dedicated safety system. The paper will give a detailed overview of the current experimental facility including all process components. Furthermore the paper will present the results of the functional test of the WDS/ISS combination using protium and deuterium, as well the results of the first commissioning runs using HTO of approximately 5 × 10{sup 9} Bq kg{sup −1} activity concentration.

  16. History of 232-F, tritium extraction processing

    International Nuclear Information System (INIS)

    Blackburn, G.W.

    1994-08-01

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

  17. Tritium behavior in the Caisson, a simulated fusion reactor room

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Yamada, Masayuki; Suzuki, Takumi; O'hira, Shigeru; Nakamura, Hirofumi; Shu, Weimin; Yamanishi, Toshihiko; Kawamura, Yoshinori; Isobe, Kanetsugu; Konishi, Satoshi; Nishi, Masataka

    2000-01-01

    In order to confirm tritium confinement ability in the deuterium-tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called 'Caisson', at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak-tight vessel of 12 m 3 , simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code

  18. Simulation of a field scale tritium tracer experiment in a fractured, weathered shale using discrete-fracture/matrix-diffusion and equivalent porous medium models

    Energy Technology Data Exchange (ETDEWEB)

    Stafford, Paige L. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Geological Sciences

    1996-05-01

    Simulations of a tritium tracer experiment in fractured shale saprolite, conducted at the Oak Ridge National Laboratory, were performed using 1D and 2D equivalent porous medium (EPM) and discrete-fracture/matrix-diffusion (DFMD) models. The models successfully reproduced the general shape of the breakthrough curves in down-gradient monitoring wells which are characterized by rapid first arrival, a slow-moving center of mass, and a persistent ``tail`` of low concentration. In plan view, the plume shows a large degree of transverse spreading with the width almost as great as the length. EPM models were sensitive to dispersivity coefficient values which had to be large (relative to the 3.7m distance between the injection and monitoring wells) to fit the tail and transverse spreading. For example, to fit the tail a longitudinal dispersivity coefficient, αL, of 0.8 meters for the 2D simulations was used. To fit the transverse spreading, a transverse dispersivity coefficient, αT, of 0.8 to 0.08 meters was used indicating an αLT ratio between 10 and 1. Transverse spreading trends were also simulated using a 2D DFMD model using a few larger aperture fractures superimposed onto an EPM. Of the fracture networks studied, only those with truncated fractures caused transverse spreading. Simulated tritium levels in all of the cases were larger than observed values by a factor of approximately 100. Although this is partly due to input of too much tritium mass by the models it appears that dilution in the wells, which were not purged prior to sampling, is also a significant factor. The 1D and 2D EPM models were fitted to monitoring data from the first five years of the experiment and then used to predict future tritium concentrations.

  19. Simulation of a field scale tritium tracer experiment in a fractured, weathered shale using discrete-fracture/matrix-diffusion and equivalent porous medium models

    International Nuclear Information System (INIS)

    Stafford, P.L.

    1996-05-01

    Simulations of a tritium tracer experiment in fractured shale saprolite, conducted at the Oak Ridge National Laboratory, were performed using 1D and 2D equivalent porous medium (EPM) and discrete-fracture/matrix-diffusion (DFMD) models. The models successfully reproduced the general shape of the breakthrough curves in down-gradient monitoring wells which are characterized by rapid first arrival, a slow-moving center of mass, and a persistent ''tail'' of low concentration. In plan view, the plume shows a large degree of transverse spreading with the width almost as great as the length. EPM models were sensitive to dispersivity coefficient values which had to be large (relative to the 3.7m distance between the injection and monitoring wells) to fit the tail and transverse spreading. For example, to fit the tail a longitudinal dispersivity coefficient, α L , of 0.8 meters for the 2D simulations was used. To fit the transverse spreading, a transverse dispersivity coefficient, α T , of 0.8 to 0.08 meters was used indicating an α L /α T ratio between 10 and 1. Transverse spreading trends were also simulated using a 2D DFMD model using a few larger aperture fractures superimposed onto an EPM. Of the fracture networks studied, only those with truncated fractures caused transverse spreading. Simulated tritium levels in all of the cases were larger than observed values by a factor of approximately 100. Although this is partly due to input of too much tritium mass by the models it appears that dilution in the wells, which were not purged prior to sampling, is also a significant factor. The 1D and 2D EPM models were fitted to monitoring data from the first five years of the experiment and then used to predict future tritium concentrations

  20. Irradiation behaviour of a tritium breeding material, γ-LiAlO 2- results of two in-pile experiments: ALICE I and ALICE II

    Science.gov (United States)

    Botter, F.; Rasneur, B.; Roth, E.

    1988-11-01

    γ-LiAlO 2 has been studied at CEA as potential breeder material for fusion reactors within the scope of the EEC fusion technology program. Radiation damage was investigated by irradiating unclad aluminate samples in the core of the OSIRIS reactor at Saclay. As part of the international breeder material comparison program named BEATRIX, US samples were irradiated along with those prepared in Saclay; samples of natural 6Li content and 96% enriched ones were irradiated. Shapes were chosen to enable postirradiation examinations (PIE), and microstructures were optimized for tritium release. The ALICE 1 experiment was carried out during 25.7 full power days (FPD), ALICE II lasted 36.3 FPD. Temperatures ranged from 400 to 600°C in the first, from 750 to 850°C in the second ALICE irradiation (sample core temperatures). In both cases the maximum flux on the samples was 2.1 × 10 18n m -2 s -1 fast, and 0.7 × 10 18n m -2 s -2 thermal Power dissipated was up to 100 W/cm 3, higher than the average in most reactor blanket designs by a factor 3 to 10, thus enabling the highest burn-ups to correspond to more than two years of possible operation in a full-scale reactor. In the lower temperature range of irradiation no significant damage was observed. In the higher one shrinkage due to sintering was induced. Whatever the microstructure, the flux and temperature, all samples (but one) not exceeding 5 mm diameter and length were mechanically intact. Above those dimensions cracking, which can be assigned to excessive thermal stress, could be observed. Given anticipated operating conditions of blankets being designed, the behaviour of γ-LiAlO 2 under irradiation is that of a very promising material.

  1. Tritium generation and neutron measurements in Pd-Si under high deuterium gas pressure

    International Nuclear Information System (INIS)

    Claytor, T.N.; Tuggle, D.G.; Menlove, H.O.

    1991-01-01

    This paper summarizes some of the methods applicable for low level tritium detection needed in the search for anomalous fusion in metal hydrides. It is also intended to further detail our tritium and neutron results that have been obtained with the Pd-Si-D system, originally presented at earlier workshops. A measure of reproducibility that was not evident in our previous work has been achieved partially due to the better detection sensitivity afforded by the use of low tritium deuterium and partially from the fact that the foil-wafer cells can be made with nearly identical electrical characteristics. This reproducibility has allowed us to narrow the optimum conditions for the experiment. While this experiment is rather different from the ''standard'' electrolytic cell or the Ti gas hydride experiment, similarities exist in that non equilibrium conditions are sought and the tritium generation levels are low and neutron emission is extremely weak. In contrast to many electrochemical cell experiments, the system used in these experiments is completely sealed during operation and uses no electrolyte. The major improvements to the experiment have been the use of vary low tritium deuterium for the hydriding and the replacement of the aluminum neutron counter tubes with ones of stainless steel. These changes have resulted in pronounced improvements to the detection systems since the background tritium level in the gas has been reduced by a factor of 300 and the neutron background has been decreased by a factor of 14. 16 refs., 8 figs., 1 tab

  2. Operator training and the training simulator experience

    International Nuclear Information System (INIS)

    Mills, D.

    The author outlines the approach used by Ontario Hydro to train operators from the day they are hired as Operators-in-Training until they are Authorized Unit First Operators. He describes in detail the use of the simulator in the final year of the authorization program, drawing on experience with the Pickering NGS A simulator. Simulators, he concludes, are important aids to training but by no means all that is required to guarantee capable First Operators

  3. Industry Operating Experience Process at Krsko NPP

    International Nuclear Information System (INIS)

    Bach, B.; Bozin, B.; Cizmek, R.

    2012-01-01

    Experience has shown that number of minor events and near misses, usually without immediate or significant impact to plant safety and reliability, are precursors of significant or severe events due to the same or similar root or apparent cause(s). It is therefore desirable to identify and analyze weaknesses of the precursor problems (events) in order to prevent occurrence of significant events. Theoretically, significant events could be prevented from occurring if the root cause(s) of these precursor problems could be identified and eliminated. The Operating Experience Program identifies such event precursors and by reporting them to the industry, plant specific corrective actions can be taken to prevent events at other operational plants. The intent of the Operating Experience Program is therefore to improve nuclear power plant safety and reliability of the operating nuclear power plants. Each plant develops its own Operating Experience Program in order to learn from the in-house operating experience as well as from the world community of nuclear plants. The effective use of operating experience includes analyzing both plant and industry events in order to identify fundamental weaknesses and then determining appropriate plant-specific actions that will minimize the likelihood of similar events. Learning and applying the lessons from operating experience is an integral part of station safety culture and is encouraged by managers throughout the top plant administrative programs and procedures. Krsko NPP is developed it own Operating Experience Program by using the most relevant INPO/WANO/IAEA guidelines as well as its own knowledge, skills an operating practice. The Operating Experience Program is a part of the Corrective Action Program, which is among top management programs, thus program is strongly encouraged by top management. The purpose of Operating Experience Program is to provide guidance for using, sharing, and evaluating operating experience information

  4. Universal tritium transmitter

    International Nuclear Information System (INIS)

    Cordaro, J. V.; Wood, M.

    2008-01-01

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

  5. Magmatic tritium

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  6. Monitoring and comparison of tritium content in urine

    International Nuclear Information System (INIS)

    Su Feng; Hua Wei; Zheng Chuancheng; Wang Xu; Wen Wanxin

    2009-01-01

    Objective: To ensure the health of staff engaged in tritium, the purpose of experiment is to find out a fast, convenient and reliable sample preparation and measurement methods for such routine monitoring. Methods: We use the conventional distillation decolorization and non-decoloration quenching correction methods dealing with urine sample, and then carried out the urine sample liquid scintillation measurements, statistical analysis between the two measurements. Results: By using above two different methods of sample pretreatment, the results that we measure tritium in urine sample are not obviously different in comparison. Conclusion: The above two different methods can be used for nuclear facilities staff and staff related to conventional tritium detection. However, non-decoloration quenching correction method is simpler and less in time and manpower than the conventional distillation method in operation. It is suitable for a large number of samples prepared, measured, and analyzed in a short period of time. (authors)

  7. Regulatory challenges in using nuclear operating experience

    International Nuclear Information System (INIS)

    2006-01-01

    The fundamental objective of all nuclear safety regulatory bodies is to ensure that nuclear utilities operate their plants in an acceptably safe manner at all times. Learning from experience has been a key element in meeting this objective. It is therefore very important for nuclear power plant operators to have an active programme for collecting, analysing and acting on the lessons of operating experience that could affect the safety of their plants. NEA experts have noted that almost all of the recent, significant events reported at international meetings have occurred earlier in one form or another. Counteractions are usually well-known, but information does not always seem to reach end users, or corrective action programmes are not always rigorously applied. Thus, one of the challenges that needs to be met in order to maintain good operational safety performance is to ensure that operating experience is promptly reported to established reporting systems, preferably international in order to benefit from a larger base of experience, and that the lessons from operating experience are actually used to promote safety. This report focuses on how regulatory bodies can ensure that operating experience is used effectively to promote the safety of nuclear power plants. While directed at nuclear power plants, the principles in this report may apply to other nuclear facilities as well. (author)

  8. 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

    temporal pattern in above or below ground plant parts. Native soil after ∼20 years of operations at SRBT had high initial OBT that persisted through the growing season; little OBT formed in garden plot soil during experiments. High OBT in native soil appeared to be a signature of higher past releases at SRBT. This phenomenon was confirmed in soils obtained at another processing facility in Canada with a similar history. The insights into variation in OBT/HTO ratios found here are of regulatory interest and should be incorporated in assessment models to aid in the design of relevant environmental monitoring programs for OBT. - Highlights: • Measured HTO and OBT at an experimental garden near a tritium processing facility. • Pulsed daytime releases of both HT and HTO, high and low tritium irrigation. • OBT/HTO ratios similar to expected values with high tritium irrigation water. • OBT/HTO ratios very high with natural rainfall or low tritium irrigation water. • Persistence of high OBT in natural soil from historical operations.

  9. Simulator experiments: effects of NPP operator experience on performance

    International Nuclear Information System (INIS)

    Beare, A.N.; Gray, L.H.

    1985-01-01

    Experiments are being conducted on nuclear power plant (NPP) control room training simulators by the Oak Ridge National Laboratory, its subcontractor, General Physics Corporation, and participating utilities. The experiments are sponsored by the Nuclear Regulatory Commission's (NRC) Human Factors and Safeguards Branch, Division of Risk Analysis and Operations, and are a continuation of prior research using simulators, supported by field data collection, to provide a technical basis for NRC human factors regulatory issues concerned with the operational safety of nuclear power plants. During the FY83 research, a simulator experiment was conducted at the control room simulator for a GE boiling water reactor (BWR) NPP. The research subjects were licensed operators undergoing requalification training and shift technical advisors (STAs). This experiment was designed to investigate the effects of (a) senior reactor operator (SRO) experience, (b) operating crew augmentation with an STA and (c) practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator. Methodology and results to date are reported

  10. Estimation of Biological Effects of Tritium.

    Science.gov (United States)

    Umata, Toshiyuki

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  12. Operational experience with the Daresbury accelerator tube

    International Nuclear Information System (INIS)

    Aitken, T.W.; Eastham, D.A.; Joy, T.; Leese, J.M.; Tait, N.R.S.; Thorn, R.

    1986-01-01

    Operational experience with the Daresbury MKI accelerator tube is reviewed with particular attention to conditioning and high voltage performance. The effects of surges and transients on the tube are described and lines of future development are discussed. (orig.)

  13. Organically bound tritium (OBT) formation in rainbow trout (Oncorhynchus mykiss): HTO and OBT-spiked food exposure experiments.

    Science.gov (United States)

    Kim, S B; Shultz, C; Stuart, M; McNamara, E; Festarini, A; Bureau, D P

    2013-02-01

    In order to determine the rate of organically bound tritium (OBT) formation, rainbow trout (Oncorhynchus mykiss) were exposed to tritiated water (HTO) or OBT-spiked food. The HTO (in water) exposure study was conducted using a tritium activity concentration of approximately 7000 Bq/L and the OBT (in food) exposure study was conducted using a tritium activity concentration of approximately 30,000 Bq/L. Fish in both studies were expected to be exposed to similar tritium levels assuming 25% incorporation of the tritiated amino acids found in the food. Four different sampling campaigns of HTO exposure (Day 10, 30, 70, 140) and five different sampling campaigns of OBT-spiked food exposure (Day 9, 30, 70, 100, 140) were conducted to measure HTO and OBT activity concentrations in fish tissues. OBT depuration was also evaluated over a period of 30 days following the 140 d exposure studies. The results suggested that the OBT formation rate was slower when the fish were exposed to HTO compared to when the fish were ingesting OBT. In addition, the results indicated that OBT can bioaccumulate in fish tissues following OBT-spiked food exposure. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

  14. 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)

  15. 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).

  16. Effects of interfering constituents on tritium smears

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  17. Tritium behavior on a cultivated plot in the 1994 chronic HT release experiment at Chalk River

    International Nuclear Information System (INIS)

    Noguchi, H.; Yokoyama, S.; Kinouchi, N.; Murata, M.; Amano, H.; Atarashi, M.; Ichimasa, Y.; Ichimasa, M.

    1995-01-01

    The behavior of HT and HTO in air and surface soil has been studied extensively in the chronic HT release experiment carried out at Chalk River during the summer of 1994. HTO concentrations in air moisture and soil water collected in a cultivated plot showed similar time-variations, increasing rapidly during the first and second days and becoming gradual after the first 3-4 days. The air HTO concentration decreased during and following rainfall but recovered within a day. The rainfall reduced the HTO concentrations in ridge soil water but little in furrows. Time histories of HTO concentrations in air moisture and soil water suggest that the system was near steady-state within a continuous HT release period of 12 days, in spite of the presence of rain during the period. The air HTO concentrations on clear days showed diurnal cycles that were higher during daytime than at night. The experimental field had a very complex soil regime with respect to HT deposits. The deposits to soil surface varied depending on soil conditions. 12 refs., 5 figs

  18. 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

  19. Tritium pellet injector for TFTR

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. Combining risk analysis and operating experience

    International Nuclear Information System (INIS)

    1986-10-01

    In recent years there has been an increasing interest in the systematic utilization of operating experience in the decision making process concerning large industrial facilities. Even before the advent of Probabilistic Safety Assessment (PSA), operating experience had always played an important role in such decisions. Of course, operating experience has always been an input to PSA also; however, as PSA becomes more mature and the quality and quantity of operating experience improve, greater emphasis is now being placed on the use of operating experience to update and validate PSA and thereby provide a more rational basis for decision making. This report outlines the ways in which data are collected, processed using mathematical techniques and utilized in decision making. It is not intended to provide details of the methods and procedures to be used in these areas, but is rather intended as an introduction to these topics and some of the relevant literature. The meeting presentations were divided into three sessions devoted to the following topics: evaluation of nuclear power plants operational experience (5 papers); uncertainties (2 papers); probabilistic safety assessment studies in Member States (7 papers). A separate abstract was prepared for each of these papers

  1. Operational experience of the Marcoule reactors

    International Nuclear Information System (INIS)

    Conte, F.

    1963-01-01

    The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

  2. MIT January Operational Internship Experience 2011

    Science.gov (United States)

    DeLatte, Danielle; Furhmann, Adam; Habib, Manal; Joujon-Roche, Cecily; Opara, Nnaemeka; Pasterski, Sabrina Gonzalez; Powell, Christina; Wimmer, Andrew

    2011-01-01

    This slide presentation reviews the 2011 January Operational Internship experience (JOIE) program which allows students to study operational aspects of spaceflight, how design affects operations and systems engineering in practice for 3 weeks. Topics include: (1) Systems Engineering (2) NASA Organization (3) Workforce Core Values (4) Human Factors (5) Safety (6) Lean Engineering (7) NASA Now (8) Press, Media, and Outreach and (9) Future of Spaceflight.

  3. Operating manual for the critical experiments facility

    International Nuclear Information System (INIS)

    1986-01-01

    The operation of the Critical Experiments Facility (CEF) requires careful attention to procedures in order that all safety precautions are observed. Since an accident could release large amounts of radioactivity, careful operation and strict enforcement of procedures are necessary. To provide for safe operation, detailed procedures have been written for all phases of the operation of this facility. The CEF operating procedures are not to be construed to constitute a part ofthe Technical Specifications. In the event of any discrepancy between the information given herein and the Technical Specifications, limits set forth in the Technical Specifications apply. All normal and most emergency operation conditions are covered by procedures presented in this manual. These procedures are designed to be followed by the operating personnel. Strict adherence to these procedures is expected for the following reasons. (1) To provide a standard, safe method of performing all operations, the procedures were written by reactor engineers experienced in supervising the operation of reactors and were reviewed by an organization with over 30 years of reactor operating experience. (2) To have an up-to-date description of operating techniques available at all times for reference and review, it is necessary that the procedures be written

  4. Operating manual for the critical experiments facility

    Energy Technology Data Exchange (ETDEWEB)

    1986-01-01

    The operation of the Critical Experiments Facility (CEF) requires careful attention to procedures in order that all safety precautions are observed. Since an accident could release large amounts of radioactivity, careful operation and strict enforcement of procedures are necessary. To provide for safe operation, detailed procedures have been written for all phases of the operation of this facility. The CEF operating procedures are not to be construed to constitute a part ofthe Technical Specifications. In the event of any discrepancy between the information given herein and the Technical Specifications, limits set forth in the Technical Specifications apply. All normal and most emergency operation conditions are covered by procedures presented in this manual. These procedures are designed to be followed by the operating personnel. Strict adherence to these procedures is expected for the following reasons. (1) To provide a standard, safe method of performing all operations, the procedures were written by reactor engineers experienced in supervising the operation of reactors and were reviewed by an organization with over 30 years of reactor operating experience. (2) To have an up-to-date description of operating techniques available at all times for reference and review, it is necessary that the procedures be written.

  5. 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

  6. Simulator experiments: effects of NPP operator experience on performance

    International Nuclear Information System (INIS)

    Beare, A.N.; Gray, L.H.

    1984-01-01

    During the FY83 research, a simulator experiment was conducted at the control room simulator for a GE Boiling Water Reactor (BWR) NPP. The research subjects were licensed operators undergoing requalification training and shift technical advisors (STAs). This experiment was designed to investigate the effects of senior reactor operator (SRO) experience, operating crew augmentation with an STA and practice, as a crew, upon crew and individual operator performance, in response to anticipated plant transients. Sixteen two-man crews of licensed operators were employed in a 2 x 2 factorial design. The SROs leading the crews were split into high and low experience groups on the basis of their years of experience as an SRO. One half of the high- and low-SRO experience groups were assisted by an STA. The crews responded to four simulated plant casualties. A five-variable set of content-referenced performance measures was derived from task analyses of the procedurally correct responses to the four casualties. System parameters and control manipulations were recorded by the computer controlling the simulator. Data on communications and procedure use were obtained from analysis of videotapes of the exercises. Questionnaires were used to collect subject biographical information and data on subjective workload during each simulated casualty. For four of the five performance measures, no significant differences were found between groups led by high (25 to 114 months) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have significantly shorter task performance times than crews led by high experience SROs. The presence of the STA had no significant effect on overall team performance in responding to the four simulated casualties. The FY84 experiments are a partial replication and extension of the FY83 experiment, but with PWR operators and simulator

  7. Kayenta advanced series compensation operational experience

    Energy Technology Data Exchange (ETDEWEB)

    1994-01-01

    The world's first three-phase, thyristor-controlled series compensation scheme with continuously variable impedance has been introduced into a transmission system. Energized and dedicated in September 1992, the installation was placed into commercial operation in January 1993 and has provided over one year of operating experience. This paper describes the 230 kV, 330 MVAr (60 Hz) advanced series compensation (ASC) project, located in north-eastern Arizona at Kayenta Substation on the 320 km Glen Canyon-Shiprock transmission line. The paper describes operating experiences, coordination with phase shifting transformer, phase shifter failure, platform power, system disturbances, and future plans.

  8. Tritium Systems Test Facility. Volume II. Appendixes

    International Nuclear Information System (INIS)

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

    1976-10-01

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

  9. TFTR neutral beam D-T gas injection system operational experiences of the first two years

    International Nuclear Information System (INIS)

    Oldaker, M.E.; Lawson, J.E.; Stevenson, T.N.; Kamperschroer, J.H.

    1995-01-01

    The TFTR Neutral Beam Tritium Gas Injection System (TGIS) has successfully performed tritium operations since December 1993. TGIS operation has been reliable, with no leaks to the secondary containment to date. Notable operational problems include throughput leaks on fill, exit and piezoelectric valves. Repair of a TGIS requires replacement of the assembly, involving TFTR downtime and extensive purging, since the TGIS assembly is highly contaminated with residual tritium, and is located within secondary containment. Modifications to improve reliability and operating range include adjustable reverse bias voltage to the piezoelectric valves, timing and error calculation changes to tune the PLC and hardwired timing control, and exercising piezoelectric valves without actually pulsing gas prior to use after extended inactivity. A pressure sensor failure required the development of an open loop piezoelectric valve drive control scheme, using a simple voltage ramp to partially compensate for declining plenum pressure. Replacement of TGIS's have been performed, maintaining twelve system tritium capability as part of scheduled project maintenance activity

  10. Behaviour of tritium in the environment

    International Nuclear Information System (INIS)

    1979-01-01

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

  11. Operating experience and TPA: the Italian perspective

    International Nuclear Information System (INIS)

    Grimaldi, G.

    1990-01-01

    Collection and analysis of operating experience from the Italian plants and utilization of abroad data both to plants in operation and in construction are presented. Some results are also referred, aimed to evidence the role of the international cooperation to safe operation of nuclear plants. The approach to the Trend and Pattern analyses is described as well, and the use of computerized techniques of analysis on personal computer. Finally on going activities are introduced, specifically application of operating experience of plants in operation to small sized reactors and to ones with more intrinsic safety characteristics; review of the reporting system for future application and comparative analysis of the different realization of selected safety systems

  12. 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.)

  13. Tritium loss in molten flibe systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  14. Tritium loss in molten flibe systems

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  15. Experience in operation of heavy water reactors

    International Nuclear Information System (INIS)

    Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

    1999-01-01

    The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

  16. Lifetime and shelf life of sealed tritium-filled plasma focus chambers with gas generator

    Directory of Open Access Journals (Sweden)

    B.D. Lemeshko

    2017-11-01

    Full Text Available The paper describes the operation features of plasma focus chambers using deuterium–tritium mixture. Handling tritium requires the use of sealed, vacuum-tight plasma focus chambers. In these chambers, there is an accumulation of the impurity gases released from the inside surfaces of the electrodes and the insulator while moving plasma current sheath inside chambers interacting with β-electrons generated due to the decay of tritium. Decay of tritium is also accompanied by the accumulation of helium. Impurities lead to a decreased yield of neutron emission from plasma focus chambers, especially for long term operation. The paper presents an option of absorption type gas generator in the chamber based on porous titanium, which allows to significantly increase the lifetime and shelf life of tritium chambers. It also shows the results of experiments on the comparison of the operation of sealed plasma focus chambers with and without the gas generator. Keywords: Plasma focus, Neutron yield, Tritium-filled plasma focus chambers, PACS Codes: 29.25.-v, 52.58.Lq

  17. Management of Tritium in ITER Waste

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  18. 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.)

  19. Engineering studies of tritium recovery from CTR blankets and plasma exhaust

    International Nuclear Information System (INIS)

    Watson, J.S.

    1975-01-01

    Engineering studies on tritium handling problems in fusion reactors have included conceptual and experimental studies of techniques for recovery of tritium bred in the reactor blanket and conceptual designs for recovery and processing of tritium from plasma exhausts. The process requirements and promising techniques for the blanket system depend upon the materials used for the blanket, coolant, and structure and on the operating temperatures. Process requirements are likely to be set in some systems by allowable loss rates to the steam system or by inventory considerations. Conceptual studies have also been made for tritium handling equipment for fueling, recovery, and processing in plasma recycle systems of fusion reactors, and a specific design has been prepared for ''near-term'' Tokamak experiments. (auth)

  20. Tritium processing using metal hydrides

    International Nuclear Information System (INIS)

    Mallett, M.W.

    1986-01-01

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

  1. 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

  2. Tritium Management Loop Design Status

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-01

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

  3. Tokamak fusion reactors with less than full tritium breeding

    International Nuclear Information System (INIS)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed

  4. 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

  5. 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.)

  6. Operating experience insights supporting ageing assessments

    International Nuclear Information System (INIS)

    Nitoi, M.

    2013-01-01

    Be effective in ageing management means looking at the right aspects, with the right techniques, and one of the most effective tool which could be used for that purpose is the analysis of operating experience. The paper has as objective to perform a review of available operating experience, with the aim to provide a better picture about the impact of ageing effects. The IAEA International Reporting System and NRC Licensee Event Reports were chosen as reference databases, both databases being internationally recognized as important sources of information about events occurrences in the nuclear power plants. The ageing related events identified in the selected time window were analyzed in detail, and the contributions of each major degradation mechanisms that have induced the ageing related events (specific to each defined group of components) was represented and discussed. The paper demonstrates the possibility to use operating experience insights in highlighting the ageing effects. (authors)

  7. 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

  8. DOE handbook: Tritium handling and safe storage

    International Nuclear Information System (INIS)

    1999-03-01

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

  9. DOE handbook: Tritium handling and safe storage

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-03-01

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

  10. The SM and MIR reactors operation experience

    International Nuclear Information System (INIS)

    Kuprienko, V.A.; Klinov, A.V.; Svyatkin, M.N.; Shamardin, V.K.

    1995-01-01

    The SM and MIR operation experience show that continuous work on the problem of ageing, in all its aspects, allows for prolongation of the research plant life cycle by several folds as compared to the initial project. The redesigned SM-3 reactor will operate for another 20 years. The similar result is expected from the MIR planned reconstruction which scope will be the topic of future presentations. (orig.)

  11. Nuclear power plant operating experience, 1976

    International Nuclear Information System (INIS)

    1977-11-01

    This report is the third in a series of reports issued annually that summarize the operating experience of U.S. nuclear power plants in commercial operation. Power generation statistics, plant outages, reportable occurrences, fuel element performance, occupational radiation exposure and radioactive effluents for each plant are presented. Summary highlights of these areas are discussed. The report includes 1976 data from 55 plants--23 boiling water reactor plants and 32 pressurized water reactor plants

  12. Operational experience with SLAC's beam containment electronics

    International Nuclear Information System (INIS)

    Constant, T.N.; Crook, K.; Heggie, D.

    1977-03-01

    Considerable operating experience was accumulated at SLAC with an extensive electronic system for the containment of high power accelerated beams. Average beam power at SLAC can approach 900 kilowatts with the potential for burning through beam stoppers, protection collimators, and other power absorbers within a few seconds. Fast, reliable, and redundant electronic monitoring circuits have been employed to provide some of the safeguards necessary for minimizing the risk to personnel. The electronic systems are described, and the design philosophy and operating experience are discussed

  13. LOFT instrumented fuel design and operating experience

    International Nuclear Information System (INIS)

    Russell, M.L.

    1979-01-01

    A summary description of the Loss-of-Fluid Test (LOFT) system instrumented core construction details and operating experience through reactor startup and loss-of-coolant experiment (LOCE) operations performed to date are discussed. The discussion includes details of the test instrumentation attachment to the fuel assembly, the structural response of the fuel modules to the forces generated by a double-ended break of a pressurized water reactor (PWR) coolant pipe at the inlet to the reactor vessel, the durability of the LOFT fuel and test instrumentation, and the plans for incorporation of improved fuel assembly test instrumentation features in the LOFT core

  14. Accelerator/Experiment operations - FY 2006

    Energy Technology Data Exchange (ETDEWEB)

    Brice, S.; Conrad, J.; Denisov, D.; Ginther, G.; Holmes, S.; James, C.; Lee, W.; Louis, W.; Moore, C.; Plunkett, R.; Raja, R.; /Fermilab

    2006-10-01

    This Technical Memorandum (TM) summarizes the Fermilab accelerator and experiment operations for FY 2006. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2006 Run II at the Tevatron Collider, the MiniBooNE experiments running in the Booster Neutrino Beam in neutrino and antineutrino modes, MINOS using the Main Injector Neutrino Beam (NuMI), and SY 120 activities.

  15. Confirmatory experiments for the United States Department of Energy Accelerator Production of Tritium Program: Neutron, triton and radionuclide production by thick targets of lead and tungsten bombarded by 800 MeV protons

    International Nuclear Information System (INIS)

    Lisowski, P.W.; Cappiello, M.; Ullmann, J.L.; Gavron, A.; King, J.D.; Laird, R.; Mayo, D.; Waters, L.; Zoeller, C.; Staples, P.

    1994-01-01

    Neutron and Triton Production by 800 MeV Protons: The experiments presented in this report were performed in support of the Accelerator Production of Tritium (APT) project at the Los Alamos Weapons Neutron Research (WNR) facility in order to provide data to benchmark and validate physics simulations used in the APT target/blanket design. An experimental apparatus was built that incorporated many of the features of the neutron source region of the 3 He target/blanket. Those features included a tungsten neutron source, flux traps, neutron moderator, lead backstop, lead multiplying annulus, neutron absorbing blanket and a combination neutron de-coupler and tritium producing gas ( 3 He). The experiments were performed in two separate proton irradiations each with approximately 100 nA-hr of 800 MeV protons. The first irradiation was made with a small neutron moderating blanket, allowing the authors to measure tritium production in the 3 He gas by sampling, and counting the amount of tritium. The second irradiation was performed with a large neutron moderating blanket (light water with a 1% manganese sulfate solution) that allowed them to measure both the tritium production in the central region and the total neutron production. The authors did this by sampling and counting the tritium produced and by measuring the activation of the manganese solution. Results of the three tritium production measurements show large disagreements with each other and therefore with the values predicted using the LAHET-MCNP code system. The source of the discrepancies may lie with the sampling system or adsorption on the tungsten surfaces. The authors discuss tests that may resolve that issue. The data for the total neutron production measurement is much more consistent. Those results show excellent agreement between calculation and experiment

  16. 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

  17. Mapping of tritium in drinking water from various Indian states

    International Nuclear Information System (INIS)

    Shah, Chirag A.; Baburajan, A.; Ravi, P.M.; Tripathi, R.M.

    2015-01-01

    The tritium in fresh water used for drinking purpose across five state of India was analyzed for tritium activity. The tritium data obtained were compared with the monitoring data of tritium in drinking water sources at Tarapur site, which houses a number of nuclear facilities. It is observed that the tritium activity in the water sample from various out station locations were in the range of < 0.48 to 1.33 Bq/l. The tritium value obtained in the drinking water sources at Tarapur was found to be in the range of 0.91 to 3.10 Bq/l. The monitoring of tritium in drinking water from Tarapur and from various out station location indicate that the level is negligible compared to the USEPA limit of 10000 Bq/l and the contribution of operation nuclear facilities to the tritium activity in drinking water source at Tarapur is insignificant. (author)

  18. Transfer and incorporation of tritium in mammals

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  19. 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

  20. Accelerator/Experiment Operations - FY 2015

    Energy Technology Data Exchange (ETDEWEB)

    Czarapata, P. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); et al.

    2015-10-01

    This Technical Memorandum summarizes the Fermilab accelerator and experiment operations for FY 2015. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2015 NOvA, MINOS+ and MINERvA experiments using the Main Injector Neutrino Beam (NuMI), the activities in the SciBooNE Hall using the Booster Neutrino Beam (BNB), and the SeaQuest experiment and Meson Test Beam (MTest) activities in the 120 GeV external Switchyard beam (SY120).

  1. 1997 evaluation of tritium removal and mitigation technologies for Hanford Site wastewaters

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Biyani, R.K.; Duncan, J.B.; Flyckt, D.L.; Mohondro, P.C.; Sinton, G.L.

    1997-01-01

    This report contains results of a biennial assessment of tritium separation technology and tritium nitration techniques for control of tritium bearing wastewaters at the Hanford Site. Tritium in wastewaters at Hanford have resulted from plutonium production, fuel reprocessing, and waste handling operations since 1944. this assessment was conducted in response to the Hanford Federal Facility Agreement and Consent Order

  2. Accelerator/Experiment Operations - FY 2016

    International Nuclear Information System (INIS)

    Blake, A.; Convery, M.; Geer, S.; Geesaman, D.; Harris, D.; Johnson, D.; Lang, K.; McFarland, K.; Messier, M.; Moore, C. D.; Newhart, D.; Reimer, P. E.; Plunkett, R.; Rominsky, M.; Sanchez, M.; Schmidt, J. J.; Shanahan, P.; Tate, C.; Thomas, J.; Donatella Torretta, Donatella Torretta; Matthew Wetstein, Matthew Wetstein

    2016-01-01

    This Technical Memorandum summarizes the Fermilab accelerator and experiment operations for FY 2016. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2016 NOvA, MINOS+ and MINERvA experiments using the Main Injector Neutrino Beam (NuMI), the MicroBooNE experiment and the activities in the SciBooNE Hall using the Booster Neutrino Beam (BNB), and the SeaQuest experiment, LArIAT experiment and Meson Test Beam activities in the 120 GeV external switchyard beam (SY120). Each section was prepared by the relevant authors, and was then edited for inclusion in this summary.

  3. Accelerator/Experiment Operations - FY 2016

    Energy Technology Data Exchange (ETDEWEB)

    Blake, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Convery, M. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Geer, S. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Geesaman, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Harris, D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Johnson, D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Lang, K. [Argonne National Lab. (ANL), Argonne, IL (United States); McFarland, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Messier, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Moore, C. D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Newhart, D. [Fermilab; Reimer, P. E. [Argonne; Plunkett, R. [Fermilab; Rominsky, M. [Fermilab; Sanchez, M. [Iowa State U.; Schmidt, J. J. [Fermilab; Shanahan, P. [Fermilab; Tate, C. [Fermilab; Thomas, J. [University Coll. London; Donatella Torretta, Donatella Torretta [Fermilab; Matthew Wetstein, Matthew Wetstein [Iowa State University

    2016-10-01

    This Technical Memorandum summarizes the Fermilab accelerator and experiment operations for FY 2016. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the FY 2016 NOvA, MINOS+ and MINERvA experiments using the Main Injector Neutrino Beam (NuMI), the MicroBooNE experiment and the activities in the SciBooNE Hall using the Booster Neutrino Beam (BNB), and the SeaQuest experiment, LArIAT experiment and Meson Test Beam activities in the 120 GeV external switchyard beam (SY120). Each section was prepared by the relevant authors, and was then edited for inclusion in this summary.

  4. In-situ tritium recovery from Li2O irradiated in fast neutron flux: BEATRIX-II initial results

    International Nuclear Information System (INIS)

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

    1990-10-01

    The BEATRIX-II experiment in FFTF is an in-situ tritium recovery experiment to evaluate the tritium release characteristics of Li 2 O and its stability under fast neutron irradiation to extended burnups. This experiment includes two specimens: a thin annular specimen capable of temperature transients and a larger temperature gradient specimen. During the first 85 days of the operating cycle of the reactor, the tritium recovery rate of a temperature transient capsule was examined as a function of temperature, gas flow rate, gas composition and burnup. Temperature changes in the range from 525 to 625 degree C resulted in decreasing tritium inventory with increasing temperature. Lower gas flow rates resulted in slightly lower tritium release rates while gas composition changes affected the tritium release rate significantly, more than either flow rate or temperature changes. Three different sweep gases were used: He with 0.1% H 2 , He with 0.01% H 2 , and pure He. Decreasing the amount of hydrogen in the sweep gas decreased the steady-state release rate by as much as a factor of two. A temperature gradient capsule is more prototypic of the conditions expected in a fusion blanket and was designed to provide data that can be used in evaluating the operational parameters of a solid breeder in a blanket environment. The operation of this canister during the first 85 EFPD cycle suggests that Li 2 O is a viable solid breeder material. 9 refs., 5 figs., 3 tabs

  5. Distribution of tritium in a chronically contaminated lake

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  6. Operational experience with CMS Tier-2 sites

    International Nuclear Information System (INIS)

    Gonzalez Caballero, I

    2010-01-01

    In the CMS computing model, more than one third of the computing resources are located at Tier-2 sites, which are distributed across the countries in the collaboration. These sites are the primary platform for user analyses; they host datasets that are created at Tier-1 sites, and users from all CMS institutes submit analysis jobs that run on those data through grid interfaces. They are also the primary resource for the production of large simulation samples for general use in the experiment. As a result, Tier-2 sites have an interesting mix of organized experiment-controlled activities and chaotic user-controlled activities. CMS currently operates about 40 Tier-2 sites in 22 countries, making the sites a far-flung computational and social network. We describe our operational experience with the sites, touching on our achievements, the lessons learned, and the challenges for the future.

  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. Weapons engineering tritium facility overview

    Energy Technology Data Exchange (ETDEWEB)

    Najera, Larry [Los Alamos National Laboratory

    2011-01-20

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

  9. 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

  10. HANARO operation experience in the year 2004

    International Nuclear Information System (INIS)

    Oh, Soo-Youl; Kim, Heonil; Cho, Yeong-Garp; Jun, Byung-Jin

    2006-01-01

    The experiences of the HANARO operation and maintenance in the year 2004 are presented in this article. The operation of HANARO, a 30 MW research reactor operated by the Korea Atomic Energy Research Institute (KAERI), aims at a safe and effective operation to enhance its utilization in various fields of scientific research and industry. Regardless of its importance of the routine operation, this article is devoted to rather unusual matters such as irregular maintenance events and incidents. Since the first criticality in 1995, it has been a long-cherished task to reach the designed power level of 30 MW from the temporarily approved 24 MW. By resolving the concern on the fuel integrity, the designed level could be licensed and, eventually, it was achieved last November. On the other hand, after its 9 years of operation, the mechanical integrity of the heavy water reflector tank was checked. The measurement of the vertical straightness of the tank inner shell indicated its integrity. Meanwhile, the HANARO fuel production facility was completed at the KAERI site, and it will begin to supply centrifugally atomized fuels, instead of conventional comminuted fuels, to HANARO shortly. There were several incidents in 2004, which have all been cleared, including a leak of heavy water, melting of a sample in an irradiation hole for the neutron activation analysis, and a condensation problem in a horizontal beam tube. The progress of and lessons from each incident are presented. The utilization of HANARO is expanding every year and the trend will also continue in 2005. The operation mode has been changed from an 18-day continuous operation and 10-day shutdown (18-10 mode) to the 23-12 mode since the end of 2004, and a further extension is planned to the 30-12 mode. Thanks to this extended operation term, an increased power level and, most importantly, a reliable operation, the HANARO is gaining more and more credit from the end users. (author)

  11. The tritium monitoring requirements of fusion and the status of research

    International Nuclear Information System (INIS)

    Nickerson, S.B.; Gerdingh, R.F.; Penfold, K.

    1982-10-01

    This report is a summary of an investigation into the tritium monitoring requirements of tritium laboratories, D-T burning ignition experiments, and fusion reactors. There is also a summary of the status of research into tritium monitoring and a survey of commercially available tritium monitors

  12. The INEL Tritium Research Facility

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-01-01

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

  13. The INEL Tritium Research Facility

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  14. Disposal of tritium-exposed metal hydrides

    International Nuclear Information System (INIS)

    Nobile, A.; Motyka, T.

    1991-01-01

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

  15. Methane generated from graphite--tritium interaction

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  16. 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)

  17. Handling of tritium-bearing wastes

    International Nuclear Information System (INIS)

    1981-01-01

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

  18. 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

  19. Operational experience of the ATLAS accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Den Hartog, P K; Bogaty, J M; Bollinger, L M; Clifft, B E; Craig, S L; Harden, R E; Markovich, P; Munson, F H; Nixon, J M; Pardo, R C; Phillips, D R; Shepard, K W; Tilbrook, I R; Zinkmann, G P [Argonne National Lab., IL (USA). Physics Div.

    1990-02-01

    The ATLAS accelerator consists of a HVEC model FN tandem accelerator injecting into a linac of independently-phased niobium superconducting resonators. The accelerator provides beams with masses 6 {le} A {le} 127 and with energies ranging up to 20 MeV/A for the lightest ions and 4 MeV/A for the heaviest ions. Portions of the linac have been in operation since 1978 and, over the last decade, more than 35000 h of operating experience have been accumulated. The long-term stability of niobium resonators, and their feasibility for use in heavy-ion accelerators is now well established. (orig.).

  20. Operational experience of extreme wind penetrations

    Energy Technology Data Exchange (ETDEWEB)

    Estanqueiro, Ana [INETI/LNEG - National Laboratory for Energy and Geology, Lisbon (Portugal); Mateus, Carlos B. [Instituto de Meteorologia, Lisboa (Portugal); Pestana, Rui [Redes Energeticas Nacionais (REN), Lisboa (Portugal)

    2010-07-01

    This paper reports the operational experience from the Portuguese Power System during the 2009/2010 winter months when record wind penerations were observed: the instantaneous wind power penetration peaked at 70% of consumption during no-load periods and the wind energy accounted for more than 50% of the energy consumed for a large period. The regulation measures taken by the TSO are presented in the paper, together with the additional reserves operated for added system security. Information on the overall power system behavior under such extreme long-term wind power penetrations will also be addressed. (org.)

  1. Operating procedures: Fusion Experiments Analysis Facility

    Energy Technology Data Exchange (ETDEWEB)

    Lerche, R.A.; Carey, R.W.

    1984-03-20

    The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility.

  2. Operating procedures: Fusion Experiments Analysis Facility

    International Nuclear Information System (INIS)

    Lerche, R.A.; Carey, R.W.

    1984-01-01

    The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility

  3. Preheat operating experiences at the FFTF

    International Nuclear Information System (INIS)

    Tucker, W.R.

    1978-01-01

    The rather extensive test program performed on the FFTF preheat control system resulted in successful sodium fill of one secondary heat transport loop on July 2, 1978. The data obtained during testing and the attendant operating experience gained resulted in some design changes and provided the information necessary to fully characterize system performance. Temperature excursions and deviations from preset limits of only a minor nature were encountered during preheat for sodium fill. The addition of the rate alarm feature was beneficial to operation of the preheat system and allowed early detection and correction of impending excursions

  4. Conceptual design of tritium treatment facility

    International Nuclear Information System (INIS)

    Tachikawa, Katsuhiro

    1982-01-01

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

  5. Tritium Systems Test Facility. Volume I

    International Nuclear Information System (INIS)

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

    1976-10-01

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

  6. Development of a compact tritium activity monitor and first tritium measurements

    Energy Technology Data Exchange (ETDEWEB)

    Röllig, M., E-mail: marco.roellig@kit.edu; Ebenhöch, S.; Niemes, S.; Priester, F.; Sturm, M.

    2015-11-15

    Highlights: • We report about experimental results of a new tritium activity monitoring system using the BIXS method. • The system is compact and easy to implement. It has a small dead volume of about 28 cm{sup 3} and can be used in a flow-through mode. • Gold coated surfaces are used to improve significantly count rate stability of the system and to reduce stored inventory. - Abstract: To develop a convenient tool for in-line tritium gas monitoring, the TRitium Activity Chamber Experiment (TRACE) was built and commissioned at the Tritium Laboratory Karlsruhe (TLK). The detection system is based on beta-induced X-ray spectrometry (BIXS), which observes the bremsstrahlung X-rays generated by tritium decay electrons in a gold layer. The setup features a measuring chamber with a gold-coated beryllium window and a silicon drift detector. Such a detection system can be used for accountancy and process control in tritium processing facilities like the Karlsruhe Tritium Neutrino Experiment (KATRIN). First characterization measurements with tritium were performed. The system demonstrates a linear response between tritium partial pressure and the integral count rate in a pressure range of 1 Pa up to 60 Pa. Within 100 s measurement time the lower detection limit for tritium is (143.63 ± 5.06) · 10{sup 4} Bq. The system stability of TRACE is limited by a linear decrease of integral count rate of 0.041 %/h. This decrease is most probably due to exchange interactions between tritium and the stainless steel walls. By reducing the interaction surface with stainless steel, the decrease of the integral count rate was reduced to 0.008 %/h. Based on the first results shown in this paper it can be concluded that TRACE is a promising complement to existing tritium monitoring tools.

  7. Design and operational experience with the off-gas cleaning system of the Seibersdorf incinerator plant

    International Nuclear Information System (INIS)

    Patek, P.R.M.

    1983-01-01

    After a description of the design and the construction principles of the incinerator building, the furnace and its attached auxiliary devices are explained. The incinerator is layed out for low level wastes. It has a vertical furnace, operates with discontinuous feeding for trashes with heat-values between 600 and 10,000 kcal/kg waste. The maximum throughput amounts to 40 kg/h. The purification of the off-gas is guaranteed by a multistage filter system: 2 stages with ceramic candles, an electrostatic filter and a HEPA-filter system. The control of the off-gas cleaning is carried out by a stack instrumentation, consisting of an aerosol-, gas-, iodine- and tritium-monitor; the building is surveyed by doserate and aerosolmonitors. Finally the experiences of the first year of operation and the main problems in running the plant are described. (author)

  8. Design and operational experience with the off-gas cleaning system of the Seibersdorf incinerator plant

    International Nuclear Information System (INIS)

    Patek, P.

    1982-05-01

    After a description of the design and the construction principles of the incinerator building, the furnace and its attached auxilary devices are explained. The incinerator is layed out for low level wastes. It has a vertical furnace, operates with discontinuous feeding for trashes with heat-values between 600 and 10000 kcal/kg waste. The maximum throughput ammounts 40 kg/h. The purification of the off-gas is guaranteed by a multistage filter system: 2 stages with ceramic candles, an electrostatic filter and a HEPA-filter system. The control of the off-gas cleaning is carried out by a stack instrumentation, consisting of an aerosol-, gas-, iodine- and tritium-monitor; the building is surveilled by doserate- and aerosolmonitors. Finally the experiences of the first year of operation and the main problems in running the plant are described. (Author) [de

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

    International Nuclear Information System (INIS)

    Muller, M.E.

    1978-01-01

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

  10. 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

  11. 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

  12. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    Dixon, F.; Gow, R.S.

    1978-01-01

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  13. Introduction to Wolsong Tritium Removal Facility (WTRF)

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. Little tritium goes a long way

    International Nuclear Information System (INIS)

    Albright, D.; Taylor, T.B.

    1988-01-01

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

  15. Large-scale distribution of tritium in a commercial product

    International Nuclear Information System (INIS)

    Combs, F.; Doda, R.J.

    1979-01-01

    Tritium enters the environment from various sources including nuclear reactor operations, weapons testing, natural production, and from the manufacture, use and ultimate disposal of commercial products containing tritium. A recent commercial application of tritium in the United States of America involves the backlighting of liquid crystal displays (LCD) in digital electronic watches. These watches are distributed through normal commercial channels to the general public. One million curies (MCi) of tritium were distributed in 1977 in this product. This is a significant quantity of tritium compared with power reactor-produced tritium (3MCi yearly) or with naturally produced tritium (6MCi yearly). This is the single largest commercial application involving tritium to date. The final disposition of tritium from large quantities of this product, after its useful life, must be estimated by considering the means of disposal and the possibility of dispersal of tritium concurrent with disposal. The most likely method of final disposition of this product will be disposal in solid refuse; this includes burial in land fills and incineration. Burial in land fills will probably contain the tritium for its effective lifetime, whereas incineration will release all the tritium gas (as the oxide) to the atmosphere. The use and disposal of this product will be studied as part of an environmental study that is at present being prepared for the U.S. Nuclear Regulatory Commission. (author)

  16. Tritium migration studies at the Nevada Test Site

    International Nuclear Information System (INIS)

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

    1993-05-01

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

  17. Operational Experience with the ATLAS Pixel Detector

    CERN Document Server

    Djama, Fares; The ATLAS collaboration

    2017-01-01

    Run 2 of the LHC collider sets new challenges to track and vertex reconstruction because of its higher energy, pileup and luminosity. The ATLAS tracking performance relies critically on the Pixel Detector. Therefore, in view of Run 2, the ATLAS collaboration has constructed the first 4-layer pixel detector in Particle Physics by installing a new pixel layer, called Insertable B-Layer (IBL). Operational experience and performance of the 4-layer Pixel Detector during Run 2 are presented.

  18. Operational Experience from Solar Thermal Energy Projects

    Science.gov (United States)

    Cameron, C. P.

    1984-01-01

    Over the past few years, Sandia National Laboratories were involved in the design, construction, and operation of a number of DOE-sponsored solar thermal energy systems. Among the systems currently in operation are several industrial process heat projects and the Modular Industrial Solar Retrofit qualification test systems, all of which use parabolic troughs, and the Shenandoah Total Energy Project, which uses parabolic dishes. Operational experience has provided insight to both desirable and undesirable features of the designs of these systems. Features of these systems which are also relevant to the design of parabolic concentrator thermal electric systems are discussed. Other design features discussed are system control functions which were found to be especially convenient or effective, such as local concentrator controls, rainwash controls, and system response to changing isolation. Drive systems are also discussed with particular emphasis of the need for reliability and the usefulness of a manual drive capability.

  19. Continuous Air Monitor Operating Experience Review

    International Nuclear Information System (INIS)

    Cadwallader, L.C.; Bruyere, S.A.

    2008-01-01

    Continuous air monitors (CAMs) are used to sense radioactive particulates in room air of nuclear facilities. CAMs alert personnel of potential inhalation exposures to radionuclides and can also actuate room ventilation isolation for public and environmental protection. This paper presents the results of a CAM operating experience review of the DOE Occurrence Reporting and Processing System (ORPS) database from the past 18 years. Regulations regarding these monitors are briefly reviewed. CAM location selection and operation are briefly discussed. Operating experiences reported by the U.S. Department of Energy and in other literature sources were reviewed to determine the strengths and weaknesses of these monitors. Power losses, human errors, and mechanical issues cause the majority of failures. The average 'all modes' failure rate is 2.65E-05/hr. Repair time estimates vary from an average repair time of 9 hours (with spare parts on hand) to 252 hours (without spare parts on hand). These data should support the use of CAMs in any nuclear facility, including the National Ignition Facility and the international ITER experiment

  20. Operating experience with the DRAGON High Temperature Reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.A.; Capp, P.D.

    2002-01-01

    The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

  1. 14 CFR 121.434 - Operating experience, operating cycles, and consolidation of knowledge and skills.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 3 2010-01-01 2010-01-01 false Operating experience, operating cycles, and... Qualifications § 121.434 Operating experience, operating cycles, and consolidation of knowledge and skills. (a... position, the operating experience, operating cycles, and the line operating flight time for consolidation...

  2. Tritium burning in inertial electrostatic confinement fusion facility

    Energy Technology Data Exchange (ETDEWEB)

    Ohnishi, Masami, E-mail: onishi@kansai-u.ac.jp [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Yamamoto, Yasushi; Osawa, Hodaka [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Hatano, Yuji; Torikai, Yuji [Hydrogen Isotope Science Center, University of Toyama, Gofuku, Toyama 930-8555 (Japan); Murata, Isao [Faculty of Engineering Environment and Energy Department, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Masuda, Kai [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hotta, Eiki [Interdisciplinary Graduate School of Science and Engineering, Tokyo Institute of Technology, 4259 Nagatsuda-cho, Midori-ku, Yokohama 226-8503 (Japan)

    2016-11-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10{sup 8}(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  3. Tritium burning in inertial electrostatic confinement fusion facility

    International Nuclear Information System (INIS)

    Ohnishi, Masami; Yamamoto, Yasushi; Osawa, Hodaka; Hatano, Yuji; Torikai, Yuji; Murata, Isao; Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki; Masuda, Kai; Hotta, Eiki

    2016-01-01

    Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10"8(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

  4. 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

  5. Operational experience feedback with precursor analysis

    International Nuclear Information System (INIS)

    Koncar, M.; Ferjancic, M.; Muehleisen, A.; Vojnovic, D.

    2003-01-01

    Experience of practical operation is a valuable source of information for improving the safety and reliability of nuclear power plants. Operational experience feedback (Olef) system manages this aspect of NPP operation. The traditional ways of investigating operational events, such as the root cause analysis (RCA), are predominantly qualitative. RCA as a part of the Olef system provides technical guidance and management expectations in the conduct of assessing the root cause to prevent recurrence, covering the following areas: conditions preceding the event, sequence of events, equipment performance and system response, human performance considerations, equipment failures, precursors to the event, plant response and follow-up, radiological considerations, regulatory process considerations and safety significance. The root cause of event is recognized when there is no known answer on question 'why has it happened?' regarding relevant condition that may have affected the event. At that point the Olef is proceeding by actions taken in response to events, utilization, dissemination and exchange of operating experience information and at the end reviewing the effectiveness of the Olef. Analysis of the event and the selection of recommended corrective/preventive actions for implementation and prioritization can be enhanced by taking into account the information and insights derived from Pasa-based analysis. A Pasa based method, called probabilistic precursor event analysis (PPE A) provides a complement to the RCA approach by focusing on how an event might have developed adversely, and implies the mapping of an operational event on a probabilistic risk model of the plant in order to obtain a quantitative assessment of the safety significance of the event PSA based event analysis provides, due to its quantitative nature, appropriate prioritization of corrective actions. PPEA defines requirements for PSA model and code, identifies input requirements and elaborates following

  6. Tritium ions in the Source and Transport Section (STS) of KATRIN

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Manuel [Karlsruhe Institute of Technology (Germany); Collaboration: KATRIN-Collaboration

    2016-07-01

    The KArlsruhe TRItium Neutrino (KATRIN) experiment aims at the model independent measurement of the electron neutrino mass. It is designed for a neutrino mass sensitivity of 0.2 eV (90% CL) after three years of measurement time. KATRIN measures the end point of the tritium beta decay spectrum using a MAC-E filter and a Windowless Gaseous Tritium Source (WGTS). While neutral tritium gas molecules are pumped through the WGTS, the decay electrons are guided to the detector with a magnetic field. Tritium ions, however, also leave the WGTS following the magnetic field lines. For KATRIN measurements it is imperative to prevent tritium ions from reaching the detector or the spectrometers, where they could decay and cause an indistinguishable background. Ion blocking measures are implemented by electric blocking potentials and electric dipoles to drift out trapped ions. Their effective operation will be tested during KATRIN commissioning measurements: The ion flux between STS and spectrometers can be measured with the Forward Beam Monitor (FBM). It offers a manipulator arm to introduce a detector into the flux tube. For ion detection, a Faraday Cup for the FBM is being designed and constructed.

  7. Tritium separation from light and heavy water by bipolar electrolysis

    International Nuclear Information System (INIS)

    Ramey, D.W.; Petek, M.; Taylor, R.D.; Kobisk, E.H.; Ramey, J.; Sampson, C.A.

    1979-10-01

    Use of bipolar electrolysis with countercurrent electrolyte flow to separate hydrogen isotopes was investigated for the removal of tritium from light water effluents or from heavy water moderator. Deuterium-tritium and protium-tritium separation factors occurring on a Pd-25% Ag bipolar electrode were measured to be 2.05 to 2.16 and 11.6 to 12.4 respectively, at current densities between 0.21 and 0.50 A cm -2 , and at 35 to 90 0 C. Current densities up to 0.3 A cm -2 have been achieved in continuous operation, at 80 to 90 0 C, without significant gas formation on the bipolar electrodes. From the measured overvoltage at the bipolar electrodes and the electrolyte conductivity the power consumption per stage was calculated to be 3.0 kwh/kg H 2 O at 0.2 A cm -2 and 5.0 kwh/kg H 2 O at 0.5 A cm -2 current density, compared to 6.4 and 8.0 kwh/kg H 2 O for normal electrolysis. A mathematical model derived for hydrogen isotope separation by bipolar electrolysis, i.e., for a square cascade, accurately describes the results for protium-tritium separation in two laboratory scale, multistage experiments with countercurrent electrolyte flow; the measured tiritum concentration gradient through the cascade agreed with the calculated values

  8. Long-term investigation of biosphere contamination by tritium

    International Nuclear Information System (INIS)

    Trnovec, T.; Kollar, J.; Tatara, M.; Chorvat, D.

    1974-03-01

    An apparatus was designed and built for isotope enrichment by electrolysis of water samples (taken in several localities in the vicinity of the Jaslovske Bohunice nuclear power plant) and a method was elaborated of measuring tritium using liquid scintillators, serving the determination of natural tritium concentrations. Operating experience showed that the degree of enrichment may easily be controlled and that the reproducibility of the enrichment coefficient value is conditional on the skill of personnel handling the apparatus. The apparatus constraints include a limited capacity of isotope enrichment (given by the number of electrolytic columns), demands on time, and sensitivity to secondary contamination. In addition to isotope enrichment of samples prior to measurement, also the feasibility of direct determination of natural tritium concentration without previous enrichment was tested. Tests were carried out of commercial products by Packard, INSTA-GEL and MONOPHASE-40. It was verified that the above method may be used in direct measuring tritium levels of several hundred TU. The preparation of a representative background sample was found to be the main problem involved in the type of determination described. The detection limit was mainly determined by the measurement statistics. (B.S.)

  9. 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)

  10. 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

  11. 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

  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. Experience from operating germanium detectors in GERDA

    Science.gov (United States)

    Palioselitis, Dimitrios; GERDA Collaboration

    2015-05-01

    Phase I of the Germanium Detector Array (GERDA) experiment, searching for the neutrinoless double beta (0νββ) decay of 76Ge, was completed in September 2013. The most competitive half-life lower limit for the 0νββ decay of 76Ge was set (T-0ν1/2 > 2.1 · 1025 yr at 90% C.L.). GERDA operates bare Ge diodes immersed in liquid argon. During Phase I, mainly refurbished semi-coaxial high purity Ge detectors from previous experiments were used. The experience gained with handling and operating bare Ge diodes in liquid argon, as well as the stability and performance of the detectors during GERDA Phase I are presented. Thirty additional new enriched BEGe-type detectors were produced and will be used in Phase II. A subgroup of these detectors has already been used successfully in GERDA Phase I. The present paper gives an overview of the production chain of the new germanium detectors, the steps taken to minimise the exposure to cosmic radiation during manufacturing, and the first results of characterisation measurements in vacuum cryostats.

  14. Experience from operating germanium detectors in GERDA

    International Nuclear Information System (INIS)

    Palioselitis, Dimitrios

    2015-01-01

    Phase I of the Germanium Detector Array (GERDA) experiment, searching for the neutrinoless double beta (0νββ) decay of 76 Ge, was completed in September 2013. The most competitive half-life lower limit for the 0νββ decay of 76 Ge was set (T- 0ν 1/2 > 2.1 · 10 25 yr at 90% C.L.). GERDA operates bare Ge diodes immersed in liquid argon. During Phase I, mainly refurbished semi-coaxial high purity Ge detectors from previous experiments were used. The experience gained with handling and operating bare Ge diodes in liquid argon, as well as the stability and performance of the detectors during GERDA Phase I are presented. Thirty additional new enriched BEGe-type detectors were produced and will be used in Phase II. A subgroup of these detectors has already been used successfully in GERDA Phase I. The present paper gives an overview of the production chain of the new germanium detectors, the steps taken to minimise the exposure to cosmic radiation during manufacturing, and the first results of characterisation measurements in vacuum cryostats. (paper)

  15. 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

  16. Issues Associated with Tritium Legacy Materials

    International Nuclear Information System (INIS)

    Mills, Michael

    2008-01-01

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

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

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

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

  18. Present status of tritium research activities at universities in Japan

    International Nuclear Information System (INIS)

    Kawamura, K.

    1983-01-01

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

  19. Laser Fusion: status, future, and tritium control

    International Nuclear Information System (INIS)

    Coyle, P.E.

    1978-11-01

    At Livermore the 10 kJ, 20 to 30 TW Shiva facility is now operational and producing regular new fusion results. Design work has begun on a 200 to 300 TW laser designed to carry the program through the first breakeven demonstration experiments in the mid-1980's. Confidence in reaching this goal is based on the significant progress we have made in state-of-the-art, high-power Nd:glass laser technology, in experimental laser fusion and laser plasma interaction physics, and in theoretical and analytical computer codes which reliably model and predict experimental results. For all of these experiments, a variety of fusion targets are being fabricated in the laboratory, and the control and handling of tritium is now a regular and routine part of ongoing inertial fusion experiments. Target design with gains of about 1000 have been studied and the means to mass produce such pellets at low cost are also being developed

  20. The application of integrated safety management principles to the Tritium Extraction Facility project

    International Nuclear Information System (INIS)

    Hickman, M.O.; Viviano, R.R.

    2000-01-01

    The DOE has developed a program that is accomplishing a heightened safety posture across the complex. The Integrated Safety Management (ISM) System (ISMS) program utilizes five core functions and seven guiding principles as the basis for implementation. The core functions define the work scope, analyze the hazards, develop and implement hazard controls, perform the work, and provide feedback for improvement. The guiding principles include line management responsibility, clear roles and responsibilities, competence per responsibilities, identification of safety standards/requirements, tailored hazard control, balanced priorities, and operations authorization. There exists an unspecified eighth principle, that is, worker involvement. A program requiring the direct involvement of the employees who are actually performing the work has been shown to be quite an effective method of communicating safety requirements, controlling work in a safe manner, and reducing safety violations and injuries. The Tritium Extraction Facility (TEF) projects, a component of the DOE's Commercial Light Water Reactor Tritium Production program, has taken the ISM principles and core functions and applied them to the project's design. The task of the design team is to design a facility and systems that will meet the production requirements of the DOE tritium mission as well as a design that minimizes the workers' exposure to adverse safety situations and hazards/hazardous materials. During the development of the preliminary design for the TEF, design teams consisted of not only designers but also personnel who had operational experience in the existing tritium and personnel who had operational experience in the existing tritium and personnel who had specialized experience from across the DOE complex. This design team reviewed multiple documents associated with the TEF operation in order to identify and document the hazards associated with the tritium process. These documents include hazards

  1. The tritium and the controlled fusion reactors

    International Nuclear Information System (INIS)

    Leger, D.; Rouyer, J.L.

    1986-04-01

    It is shown how tritium is used how it is circulating in a fusion reactor. The great functions of tritium circuits are detailed: reprocessing of burnt gases, reprocessing of gases coming from neutral injectors, reprocessing from gaseous wastes, detritiation of cooling fluids. Current technologic developments are quoted. Then tritium confinement and containment, in normal or accidental situations, are displayed. Limitation devices of effluents and release for normal operating (noticeably the reprocessing systems of atmosphere) and safety and protection systems in case of accident are described [fr

  2. Tritium behaviour in ceramic breeder blankets

    International Nuclear Information System (INIS)

    Miller, J.M.

    1989-01-01

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

  3. Operational experience of the ATLAS accelerator

    International Nuclear Information System (INIS)

    Den Hartog, P.K.; Bogaty, J.M.; Bollinger, L.M.

    1989-01-01

    The ATLAS accelerator consists of a HVEC model FN tandem accelerator injecting into a linac of independently-phased niobium superconducting resonators. The accelerator provides beams with masses from 6≤A≤127 and with energies ranging up to 20 MeV/A for the lightest ions and 4 MeV/A for the heaviest ions. Portions of the linac have been in operation since 1978 and, over the last decade, more than 35,000 hours of operating experience have been accumulated. The long-term stability of niobium resonators, and their feasibility for use in heavy-ion accelerators is now well established. 11 refs., 3 figs., 1 tab

  4. Operating experiences with 1 MW steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Sano, A; Kanamori, A; Tsuchiya, T

    1975-07-01

    1 MW steam generator, which was planned as the first stage of steam generator development in Power Reactor and Nuclear Fuel Corp. (PNC) in Japan, is a single-unit, once-through, integrated shell and tube type with multi-helical coil tubes. It was completed in Oarai Engineering Center of PNC in March of 1971, and the various performance tests were carried out up to April, 1972. After the dismantle of the steam generator for structural inspection and material test, it was restored with some improvements. In this second 1 MW steam generator, small leak occurred twice during normal operation. After repairing the failure, the same kind of performance tests as the first steam generator were conducted in order to verify the thermal insulation effect of argon gas in downcomer zone from March to June, 1974. In this paper the above operating experiences were presented including the outline of some performance test results. (author)

  5. Selection of operations staff, qualifications and experience

    International Nuclear Information System (INIS)

    Gutmann, H.

    1977-01-01

    Requirements and suggestions have been made by authorities and various organisations in a number of countries which define necessary experience and training for the various groups of nuclear power plant personnel. For two countries, the USA and the FRG, a comparison has been made which shows that there is only a slight deviation, taking into account the different education systems. With the example of the Biblis nuclear power plant the training on the job is described. Especially the production or operation department is looked at in more detail. The training is split up into several parts: a general part, such as nuclear physics, reactor physics and engineering, reactor safety, radiation protection and so on and a plant related part, such as arrangement and mode of operation of the plant under normal and accident conditions, license conditions and so on. (orig.) [de

  6. Operational experience with the ATLAS Pixel Detector

    CERN Document Server

    Ince, T; The ATLAS collaboration

    2012-01-01

    The ATLAS Pixel Detector is the innermost element of the ATLAS experiment at the Large Hadron Collider at CERN, providing high-resolution measurements of charged particle tracks in the high radiation environment close to the collision region. This capability is vital for the identification and measurement of proper decay times of long-lived particles such as b-hadrons, and thus vital for the ATLAS physics program. The detector provides hermetic coverage with three cylindrical layers and three layers of forward and backward pixel detectors. It consists of approximately 80 million pixels that are individually read out via chips bump-bonded to 1744 n-in-n silicon substrates. In this paper, results from the successful operation of the Pixel Detector at the LHC will be presented, including monitoring, calibration procedures, timing optimization and detector performance. The detector performance is excellent: 96.2% of the pixels are operational, noise occupancy and hit efficiency exceed the design specification, an...

  7. Operational experience of the ATLAS Pixel detector

    CERN Document Server

    Hirschbuehl, D; The ATLAS collaboration

    2011-01-01

    The ATLAS Pixel Detector is the innermost detector of the ATLAS experiment at the Large Hadron Collider at CERN, providing high-resolution measurements of charged particle tracks in the high radiation environment close to the collision region. This capability is vital for the identification and measurement of proper decay times of long-lived particles such as b-hadrons, and thus vital for the ATLAS physics program. The detector provides hermetic coverage with three cylindrical layers and three layers of forward and backward pixel detectors. It consists of approximately 80 million pixels that are individually read out via chips bump-bonded to 1744 n-in-n silicon substrates. In this talk, results from the successful operation of the Pixel Detector at the LHC will be presented, including monitoring, calibration procedures, timing optimization and detector performance. The detector performance is excellent: 97,5% of the pixels are operational, noise occupancy and hit efficiency exceed the design specification, an...

  8. Operational experience of the ATLAS Pixel Detector

    CERN Document Server

    Marcisovsky, M; The ATLAS collaboration

    2011-01-01

    The ATLAS Pixel Detector is the innermost detector of the ATLAS experiment at the Large Hadron Collider at CERN, providing high-resolution measurements of charged particle tracks in the high radiation environment close to the collision region. This capability is vital for the identification and measurement of proper decay times of long-lived particles such as b-hadrons, and thus vital for the ATLAS physics program. The detector provides hermetic coverage with three cylindrical layers and three layers of forward and backward pixel detectors. It consists of approximately 80 million pixels that are individually read out via chips bump-bonded to 1744 n-in-n silicon substrates. In this talk, results from the successful operation of the Pixel Detector at the LHC will be presented, including monitoring, calibration procedures, timing optimization and detector performance. The detector performance is excellent: 97,5% of the pixels are operational, noise occupancy and hit efficiency exceed the design specification, an...

  9. The LHC test string first operational experience

    CERN Document Server

    Bézaguet, Alain-Arthur; Casas-Cubillos, J; Coull, L; Cruikshank, P; Dahlerup-Petersen, K; Faugeras, Paul E; Flemsæter, B; Guinaudeau, H; Hagedorn, Dietrich; Hilbert, B; Krainz, G; Kos, N; Lavielle, D; Lebrun, P; Leo, G; Mathewson, A G; Missiaen, D; Momal, F; Parma, Vittorio; Quesnel, Jean Pierre; Richter, D; Riddone, G; Rijllart, A; Rodríguez-Mateos, F; Rohmig, P; Saban, R I; Schmidt, R; Serio, L; Skiadelli, M; Suraci, A; Tavian, L; Walckiers, L; Wallén, E; Van Weelderen, R; Williams, L; McInturff, A

    1996-01-01

    CERN operates the first version of the LHC Test String which consists of one quadrupole and three 10-m twin aperture dipole magnets. An experimental programme aiming at the validation of the LHC systems started in February 1995. During this programme the string has been powered 100 times 35 of which at 12.4 kA or above. The experiments have yielded a number of results some of which, like quench recovery for cryogenics, have modified the design of subsystems of LHC. Others, like controlled helium leaks in the cold bore and quench propagation bewteen magnets, have given a better understanding on the evolution of the phenomena inside a string of superconducting magnets cooled at superfluid helium temperatures. Following the experimental programme, the string will be powered up and powered down in one hour cycles as a fatigue test of the structure thus simulating 20 years of operation of LHC.

  10. Small sodium valve design and operating experience

    International Nuclear Information System (INIS)

    McGough, C.B.

    1974-01-01

    The United States Liquid Metal Fast Breeder Reactor program (LMFBR) includes an extensive program devoted to the development of small sodium valves. This program is now focused on the development and production of valves for the Fast Flux Test Facility (FFTF) now under construction near Richland, Washington. Other AEC support facilities, such as various test loops located at the Liquid Metal Engineering Center (LMEC), Los Angeles, California, and at the Hanford Engineering Development Laboratory (HEDL), Richland, Washington, also have significant requirements for small sodium valves, and valves similar in design to the FFTF valves are being supplied to these AEC laboratories for use in their critical test installations. A principal motivation for these valve programs, beyond the immediate need to provide high-reliability valves for FFTF and the support facilities, is the necessity to develop small valve technology for the Clinch River Breeder Reactor Plant (CRBRP). FFTF small sodium valve design and development experience will be directly applied to the CRBRP program. Various test programs have been, and are being, conducted to verify the performance and integrity of the FFTF valves, and to uncover any potential problems so that they can be corrected before the valves are placed in service in FFTF. The principal small sodium valve designs being utilized in current U.S. programs, the test and operational experience obtained to date on them, problems uncovered, and future development and testing efforts being planned are reviewed. The standards and requirements to which the valves are being designed and fabricated, the valve designs in current use, valve operators, test and operating experience, and future valve development plans are summarized. (U.S.)

  11. Operating experience with BWR nuclear power

    International Nuclear Information System (INIS)

    Bonsdorf, Magnus von.

    1986-01-01

    The two-unit nuclear power station in Olkiluoto on the western coast of Finland produces about 20 per cent of the electricity consumption of the country. The first unit, TVO-I was first connected to the national grid in September 1978 and TVO-II in February 1980. The original rated power output of each unit was 660 MWe, corresponding to the thermal power of 2000 MW from the reactor. Technical modifications allowed the power to be uprated by 8%. The operating statistics (load factors etc.) are given and the outage experience discussed. The radiological history shows very low radioactivity and dose levels have been maintained at the plant. (UK)

  12. Operational Experience with the ATLAS Pixel Detector

    CERN Document Server

    Lantzsch, Kerstin; The ATLAS collaboration

    2016-01-01

    Run 2 of the LHC is providing new challenges to track and vertex reconstruction with higher energies, denser jets and higher rates. Therefore the ATLAS experiment has constructed the first 4-layer Pixel detector in HEP, installing a new Pixel layer, also called Insertable B-Layer (IBL). In addition the Pixel detector was refurbished with new service quarter panels to recover about 3% of defective modules lost during run 1 and a new optical readout system to readout the data at higher speed while reducing the occupancy when running with increased luminosity. The commissioning, operation and performance of the 4-layer Pixel Detector will be presented.

  13. Operating experience with gamma ray irradiators

    International Nuclear Information System (INIS)

    Fraser, F.M.; Ouwerkerk, T.

    1980-01-01

    The experience of Atomic Energy of Canada, Limited (AECL) with radioisotopes dates back to the mid-1940s when radium was marketed for medical purposes. Cobalt-60 came on the scene in 1949 and within a few years a thriving business in cancer teletherapy machines and research irradiators was developed. AECL's first full-scale cobalt-60 gamma ray sterilizer for medical products was installed in 1964. AECL now has over 50 plants and 30 million curies in service around the world. Sixteen years of design experience in cobalt-60 sources, radiation shielding, safety interlock systems, and source pass mechanisms have made gamma irradiators safe, reliable, and easy to operate. This proven technology is being applied in promising new fields such as sludge treatment and food preservation. Cesium-137 is expected to be extensively utilized as the gamma radiation source for these applications

  14. Operational Experience with the ATLAS Pixel Detector

    CERN Document Server

    Djama, Fares; The ATLAS collaboration

    2017-01-01

    Run-2 of the LHC is providing new challenges to track and vertex reconstruction imposed by the higher collision energy, pileup and luminosity that are being delivered. The ATLAS tracking performance relies critically on the Pixel Detector, therefore, in view of Run-2 of LHC, the ATLAS experiment has constructed the first 4-layer Pixel detector in HEP, installing a new Pixel layer, also called Insertable B-Layer (IBL). Pixel detector was refurbished with a new service quarter panel to recover about 3% of defective modules lost during run-1 and an additional optical link per module was added to overcome in some layers the readout bandwidth limitation when LHC will exceed the nominal peak luminosity by almost a factor of 3. The key features and challenges met during the IBL project will be presented, as well as its operational experience and Pixel Detector performance in LHC.

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

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.

    1991-01-01

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

  16. Tritium Mitigation/Control for Advanced Reactor System

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-03-31

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

  17. [Mechanism of tritium persistence in porous media like clay minerals].

    Science.gov (United States)

    Wu, Dong-Jie; Wang, Jin-Sheng; Teng, Yan-Guo; Zhang, Ke-Ni

    2011-03-01

    To investigate the mechanisms of tritium persistence in clay minerals, three types of clay soils (montmorillonite, kaolinite and illite) and tritiated water were used in this study to conduct the tritium sorption tests and the other related tests. Firstly, the ingredients, metal elements and heat properties of clay minerals were studied with some instrumental analysis methods, such as ICP and TG. Secondly, with a specially designed fractionation and condensation experiment, the adsorbed water, the interlayer water and the structural water in the clay minerals separated from the tritium sorption tests were fractionated for investigating the tritium distributions in the different types of adsorptive waters. Thirdly, the location and configuration of tritium adsorbed into the structure of clay minerals were studied with infrared spectrometry (IR) tests. And finally, the forces and mechanisms for driving tritium into the clay minerals were analyzed on the basis of the isotope effect of tritium and the above tests. Following conclusions have been reached: (1) The main reason for tritium persistence in clay minerals is the entrance of tritium into the adsorbed water, the interlayer water and the structural water in clay minerals. The percentage of tritium distributed in these three types of adsorptive water are in the range of 13.65% - 38.71%, 0.32% - 5.96%, 1.28% - 4.37% of the total tritium used in the corresponding test, respectively. The percentages are different for different types of clay minerals. (2) Tritium adsorbed onto clay minerals are existed in the forms of the tritiated hydroxyl radical (OT) and the tritiated water molecule (HTO). Tritium mainly exists in tritiated water molecule for adsorbed water and interlayer water, and in tritiated hydroxyl radical for structural water. (3) The forces and effects driving tritium into the clay minerals may include molecular dispersion, electric charge sorption, isotope exchange and tritium isotope effect.

  18. 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

  19. 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

  20. Operating experience of steam generator test facility

    International Nuclear Information System (INIS)

    Sureshkumar, V.A.; Madhusoodhanan, G.; Noushad, I.B.; Ellappan, T.R.; Nashine, B.K.; Sylvia, J.I.; Rajan, K.K.; Kalyanasundaram, P.; Vaidyanathan, G.

    2006-01-01

    Steam Generator (SG) is the vital component of a Fast Reactor. It houses both water at high pressure and sodium at low pressure separated by a tube wall. Any damage to this barrier initiates sodium water reaction that could badly affect the plant availability. Steam Generator Test Facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test sodium heated once through steam generator of 19 tubes similar to the PFBR SG dimension and operating conditions. The facility is also planned as a test bed to assess improved designs of the auxiliary equipments used in Fast Breeder Reactors (FBR). The maximum power of the facility is 5.7 MWt. This rating is arrived at based on techno economic consideration. This paper covers the performance of various equipments in the system such as Electro magnetic pumps, Centrifugal sodium pump, in-sodium hydrogen meters, immersion heaters, and instrumentation and control systems. Experience in the system operation, minor modifications, overall safety performance, and highlights of the experiments carried out etc. are also brought out. (author)

  1. Neutronics experiments for uncertainty assessment of tritium breeding in HCPB and HCLL blanket mock-ups irradiated with 14 MeV neutrons

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Pillon, M.; Villari, R.; Fischer, U.; Klix, A.; Leichtle, D.; Kodeli, I.; Pohorecki, W.

    2012-01-01

    Two neutronics experiments have been carried out at 14 MeV neutron sources on mock-ups of the helium cooled pebble bed (HCBP) and the helium cooled lithium lead (HCLL) variants of ITER test blanket modules (TBMs). These experiments have provided an experimental validation of the calculations of the tritium production rate (TPR) in the two blanket concepts and an assessment of the uncertainties due to the uncertainties on nuclear data. This paper provides a brief summary of the HCPB experiment and then focuses in particular on the final results of the HCLL experiment. The TPR has been measured in the HCLL mock-up irradiated for long times at the Frascati 14 MeV Neutron Generator (FNG). Redundant and well-assessed experimental techniques have been used to measure the TPR by different teams for inter-comparison. Measurements of the neutron and gamma-ray spectra have also been performed. The analysis of the experiment, carried out by the MCNP code with FENDL-2.1 and JEFF-3.1.1 nuclear data libraries, and also including sensitivity/uncertainty analysis, shows good agreement between measurements and calculations, within the total uncertainty of 5.9% at 1σ level. (paper)

  2. Neutronics experiments and analyses in preparation of DT operations at JET

    Energy Technology Data Exchange (ETDEWEB)

    Villari, R., E-mail: rosaria.villari@enea.it [Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, ENEA, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Batistoni, P.; Angelone, M. [Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, ENEA, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Catalan, J.P. [Universidad Nacional de Educación a Distancia, Madrid (Spain); Colling, B.; Croft, D. [Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Fischer, U. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, 76344 Karlsruhe (Germany); Flammini, D. [Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, ENEA, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Klix, A. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, 76344 Karlsruhe (Germany); Loreti, S. [Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, ENEA, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Lilley, S. [Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Moro, F. [Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, ENEA, Via E. Fermi 45, 00044 Frascati (Roma) (Italy); Naish, J.; Packer, L. [Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Pereslavtsev, P. [Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, 76344 Karlsruhe (Germany); Popovichev, S. [Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom); Sauvan, P. [Universidad Nacional de Educación a Distancia, Madrid (Spain); Syme, B. [Culham Science Centre, Abingdon, OX14 3DB Oxon (United Kingdom)

    2016-11-01

    Highlights: • Neutronics experiments are in preparation at JET for DTE-2 campaign. • Shutdown dose rate benchmark experiments will be performed to validate ITER relevant tools. • Advanced D1S, MCR2S, R2Smesh and R2SUNED are the European tools under validation. • Satisfying results were obtained in the last DD benchmark experiment but efforts to reduce the uncertainties were needed. • The current computational and experimental efforts seem promising to improve the benchmark accuracy. - Abstract: In the frame of the WPJET3-DT Technology project within the EUROfusion Consortium program, neutronics experiments are in preparation for the future deuterium–tritium campaign on JET (DTE2). The experiments will be conducted with the purpose to validate the neutronics codes and tools used in ITER, thus reducing the related uncertainties and the associated risks in the machine operation. This paper summarizes the status of previous shutdown dose rate benchmarks experiments and analyses performed at JET and focuses on the computational and experimental efforts conducted in preparation of the future DTE2 experiments. In particular, preliminary calculations and studies to select detectors and positions aimed to reduce uncertainties in the shutdown dose rate experiment are presented and discussed.

  3. Neutronics experiments and analyses in preparation of DT operations at JET

    International Nuclear Information System (INIS)

    Villari, R.; Batistoni, P.; Angelone, M.; Catalan, J.P.; Colling, B.; Croft, D.; Fischer, U.; Flammini, D.; Klix, A.; Loreti, S.; Lilley, S.; Moro, F.; Naish, J.; Packer, L.; Pereslavtsev, P.; Popovichev, S.; Sauvan, P.; Syme, B.

    2016-01-01

    Highlights: • Neutronics experiments are in preparation at JET for DTE-2 campaign. • Shutdown dose rate benchmark experiments will be performed to validate ITER relevant tools. • Advanced D1S, MCR2S, R2Smesh and R2SUNED are the European tools under validation. • Satisfying results were obtained in the last DD benchmark experiment but efforts to reduce the uncertainties were needed. • The current computational and experimental efforts seem promising to improve the benchmark accuracy. - Abstract: In the frame of the WPJET3-DT Technology project within the EUROfusion Consortium program, neutronics experiments are in preparation for the future deuterium–tritium campaign on JET (DTE2). The experiments will be conducted with the purpose to validate the neutronics codes and tools used in ITER, thus reducing the related uncertainties and the associated risks in the machine operation. This paper summarizes the status of previous shutdown dose rate benchmarks experiments and analyses performed at JET and focuses on the computational and experimental efforts conducted in preparation of the future DTE2 experiments. In particular, preliminary calculations and studies to select detectors and positions aimed to reduce uncertainties in the shutdown dose rate experiment are presented and discussed.

  4. Quick management of accidental tritium exposure cases

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  5. Quick management of accidental tritium exposure cases.

    Science.gov (United States)

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

    2012-07-01

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

  6. The Drop Tower Bremen -Experiment Operation

    Science.gov (United States)

    Könemann, Thorben; von Kampen, Peter; Rath, Hans J.

    The idea behind the drop tower facility of the Center of Applied Space Technology and Micro-gravity (ZARM) in Bremen is to provide an inimitable technical opportunity of a daily access to short-term weightlessness on earth. In this way ZARM`s european unique ground-based microgravity laboratory displays an excellent economic alternative for research in space-related conditions at low costs comparable to orbital platforms. Many national and international ex-perimentalists motivated by these prospects decide to benefit from the high-quality and easy accessible microgravity environment only provided by the Drop Tower Bremen. Corresponding experiments in reduced gravity could open new perspectives of investigation methods and give scientists an impressive potential for a future technology and multidisciplinary applications on different research fields like Fundamental Physics, Astrophysics, Fluid Dynamics, Combus-tion, Material Science, Chemistry and Biology. Generally, realizing microgravity experiments at ZARM`s drop tower facility meet new requirements of the experimental hardware and may lead to some technical constraints in the setups. In any case the ZARM Drop Tower Operation and Service Company (ZARM FAB mbH) maintaining the drop tower facility is prepared to as-sist experimentalists by offering own air-conditioned laboratories, clean rooms, workshops and consulting engineers, as well as scientific personal. Furthermore, ZARM`s on-site apartment can be used for accommodations during the experiment campaigns. In terms of approaching drop tower experimenting, consulting of experimentalists is mandatory to successfully accomplish the pursued drop or catapult capsule experiment. For this purpose there will be a lot of expertise and help given by ZARM FAB mbH in strong cooperation to-gether with the experimentalists. However, in comparison to standard laboratory setups the drop or catapult capsule setup seems to be completely different at first view. While defining a

  7. 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)

  8. 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.

  9. 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

  10. 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)

  11. Risks of tritium and their mitigation

    International Nuclear Information System (INIS)

    Ichimasa, Y.; Shiba, H.; Ichimasa, M.; Chikuuti, M.; Akita, Y.

    1992-01-01

    In this study, the effects of an antibacterial drug, norfloxacin, and an antibiotic, clindamycin, on in vivo oxidation of tritium gas in rats were investigated. Wistar strain male rats were used. They were provided with a standard diet, water ad libitum, and maintained in glass metabolic cages of approximately 20 liters capacity. The air flow and temperature were controlled. To investigate the availability of norfloxacin and clindamycin on the inhibition effects of the oxidation of tritium gas, two types of the experiments were conducted one was that, before the exposure to tritium gas for 2 hours, norfloxacin or clindamycin was administrated to rats three times a day for 4 days, and the other was administration of a drug after tritium gas exposure. After the exposure to tritium gas, blood, the liver, urine and feces samples were collected from rats and the radioactivity of them was determined after combustion using a sample oxidizer. In the case of norfloxacin, tritium concentration in rat body decreased one fifth of that in non-treated rats. On the other hand, administration of clindamycin shortened the biological half-life of tritium in urine to three fifth of that of non-treated rats. (author)

  12. Experimental Tritium Cleanup System availability analysis from 1984 to 1992

    International Nuclear Information System (INIS)

    Cadwallader, L.C.; Taylor, G.L.

    1993-05-01

    This report gives the availability percentage of the Experimental Tritium Cleanup System (ETC) at the Tritium Systems Test Assembly (TSTA), which is a fusion research and technology facility at the Los Alamos National Laboratory. The component failure reports, the numbers of components, and operating times or demands are all given in this report. Sample calculations of the failure rates obtained from these data are given in the appendices. While future fusion experiments might use different or more advanced means to detritiate room air, the analysis of this system gives a data point for an actual detritiation system. Such a data point can be extrapolated for comparison with fault tree results on system designs, or can be used in a Bayesian failure rate analysis for estimating reliability of a new type of system. The nine years of testing operations on TSTA's ETC result in a reasonable average availability value of 92% for the maximal tritium release event. The failure rates for new systems are expected to be lower than for the TSTA ETC, since improvements will be made in the design of the room air detritiation system based on the TSTA system experiences. Nonetheless, these TSTA data should be useful for future fusion reactor design work and safety assessment tasks

  13. 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.)

  14. 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)

  15. Tritium release from lithium titanate, a low-activation tritium breeding material

    International Nuclear Information System (INIS)

    Kopasz, J.P.; Miller, J.M.; Johnson, C.E.

    1994-01-01

    The goals for fusion power are to produce energy in as safe, economical, and environmentally benign a manner as possible. To ensure environmentally sound operation low-activation materials should be used where feasible. The ARIES Tokamak Reactor Study has based reactor designs on the concept of using low-activation materials throughout the fusion reactor. For the tritium breeding blanket, the choices for low activation tritium breeding materials are limited. Lithium titanate is an alternative low-activation ceramic material for use in the tritium breeding blanket. To date, very little work has been done on characterizing the tritium release for lithium titanate. We have thus performed laboratory studies of tritium release from irradiated lithium titanate. The results indicate that tritium is easily removed from lithium titanate at temperatures as low as 600 K. The method of titanate preparation was found to affect the tritium release, and the addition of 0.1% H 2 to the helium purge gas did not improve tritium recovery. ((orig.))

  16. Tritium means of detection and of protection; Le tritium moyens de detection et de protection

    Energy Technology Data Exchange (ETDEWEB)

    Sutra-Fourcade, Y [Commissariat a l' Energie Atomique, Marcoule (France). Centre d' Etudes Nucleaires

    1967-07-01

    The report is an attempt to correlate present data concerning tritium, especially from the health physics points of view. The various detection and measurement methods are reviewed in turn: measurement of tritium in the atmosphere, in liquids and on surfaces. The operation of various types of apparatus is analyzed and the sensitivity limits deduced from laboratory tests are given. Otter sections are devoted to the means of protection which can be used against inhalation of tritium (ventilation, protective clothing) and to calculations of the changes in atmospheric pollution in a given place and of the time spent in a contaminated zone. The last part deals with the decontamination of equipment contaminated with tritium. (author) [French] Le rapport represente un essai de synthese des connaissances actuelles sur le tritium, essentiellement du point de vue de la radioprotection. Les differents moyens de detection et de mesure sont successivement passes en revue: mesure du tritium dans l'atmosphere, dans les liquides, sur les surfaces. Le fonctionnement de differents types d'appareils est analyse et les limites de sensibilite sont donnees d'apres les essais effectues en laboratoire. D'autres paragraphes sont consacres aux moyens de protection contre l'inhalation du tritium (ventilation, vetements de protection), a des calculs d'evolution de pollution atmospherique dans les locaux et de temps de presence en atmosphere contaminee. La derniere partie se rapporte a la de contamination de materiel contamine par du tritium. (auteur)

  17. Tritium concentrations in natural waters in Japan before use of a large quantity of tritium on its fusion program

    International Nuclear Information System (INIS)

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

    1989-01-01

    To clarify environmental tritium levels in Japan before use of a large quantity of tritium on its fusion program, the authors analyzed the tritium concentrations in various water samples, such as rain, river, lake, coastal sea and deep sea waters in Japan. The tritium concentrations in rain water were high at higher latitude. The definite differences of the tritium concentrations due to the weather conditions or seasons were not observed. The average tritium concentration in river water was 51.5 pCi/l in 1982 and that in lake water was 63.5 pCi/l in 1983. The vertical profiles of the tritium concentrations in the representative lakes were almost homogeneous except surface water. The average tritium concentrations in coastal seawater were about 20 pCi/l in both 1982 and 1983. The tendency of the increased tritium level with latitude as reported in literature was not observed by these experiments. Tritium levels in natural water in small isolated islands were lower than those at other places. In the Japan Sea, it was recognized that tritium was distributed down to around 2000 m in depth. This means that the more active vertical mixing of water masses than that in the Pacific Ocean is taking place. (author)

  18. Small sodium valve design and operating experience

    International Nuclear Information System (INIS)

    Abramson, R.; Elie, X.; Vercasson, M.; Nedelec, J.

    1974-01-01

    Conventionally, valves for sodium pipes smaller than 125 mm in diameter are called ''small sodium valves''. However, this limit should rather be considered as the lower limit o ''large sodium valves''. In fact, both the largest sizes of small valves and the smallest of large valves can be found in the range of 125-300 mm in diameter. Thus what is said about small valves also applies, for a few valve types, above the 125 mm limit. Sodium valves are described here in a general manner, with no manufacturing details except when necessary for understanding valve behavior. Operating experience is pointed out wherever possible. Finally, some information is given about ongoing or proposed development plans. (U.S.)

  19. Sixty-fold electrolytic batch enrichment for the assay of environmental tritium

    International Nuclear Information System (INIS)

    Gugelmann, A.A.; Brenninkmeijer, C.A.M.; Jalil, A.

    1977-01-01

    In Part 1 the general principles for the evaluation of low-level counting systems are outlined. Part 2 reviews current practices and experiences in tritium enrichment and counting systems. Part 3 describes a new approach to achieve high enrichment factors in a simple batch process. Sixty-fold tritium enrichment is achieved with an eighty-fold volume reduction. Specially designed large surface electrolysis cells are operated initially in series at 13A constant current, then for the final rundown at a constant voltage at 3.5V the cells are connected in parallel. (auth.)

  20. 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

    Highli