Tritium application: self-luminous glass tube(SLGT)

Energy Technology Data Exchange (ETDEWEB)

Kim, K.; Lee, S.K.; Chung, E.S.; Kim, K.S.; Kim, W.S. [Nuclear Power Lab., Korea Electric Power Research Inst. (KEPRI), Daejeon (Korea); Nam, G.J. [Engineering Information Technology Center, Inst. for Advanced Engineering (IAE), Kyonggi-do (Korea)

2005-07-01

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

ICRF heating and transport of deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

Murakami, M.; Batchelor, D.B.; Bush, C.E.

1994-01-01

This paper describes results of the first experiments utilizing high-power ion cyclotron range of frequency (ICRF) to heat deuterium-tritium (D-T) plasmas in reactor-relevant regimes on the Tokamak Fusion Test Reactor (TFTR). Results from these experiments have demonstrated efficient core, second harmonic, tritium heating of D-T supershot plasmas with tritium concentrations ranging from 6%--40%. Significant direct ion heating on the order of 60% of the input radio frequency (rf) power has been observed. The measured deposition profiles are in good agreement with two-dimensional modeling code predictions. Confinement in an rf-heated supershot is at least similar to that without rf, and possibly better in the electron channel. Efficient electron heating via mode conversion of fast waves to ion Bernstein waves (IBW) has been demonstrated in ohmic, deuterium-deuterium and DT-neutral beam injection plasmas with high concentrations of minority 3 He (n 3 He /n e > 10%). By changing the 3 He concentration or the toroidal field strength, the location of the mode-conversion radius was varied. The power deposition profile measured with rf power modulation showed that up to 70% of the power can be deposited on electrons at an off-axis position. Preliminary results with up to 4 MW coupled into the plasma by 90-degree phased antennas showed directional propagation of the mode-converted IBW. Heat wave propagation showed no strong inward thermal pinch in off-axis heating of an ohmically-heated (OH) target plasma in TFIR

Isotopic scaling of transport in deuterium-tritium plasmas

International Nuclear Information System (INIS)

Scott, S.D.; Murakami, M.; Adler, H.; Chang, Z.; Duong, H.; Grisham, L.R.; Fredrickson, E.D.; Grek, B.; Hawryluk, R.J.; Hill, K.W.; Hosea, J.; Jassby, D.L.; Johnson, D.W.; Johnson, L.C.; Loughlin, M.J.; Mansfield, D.K.; McGuire, K.M.; Meade, D.M.; Mikkelsen, D.M.; Murphy, J.; Park, H.K.; Ramsey, A.T.; Schivell, J.; Skinner, C.H.; Strachan, J.D.; Synakowski, E.J.; Taylor, G.; Thompson, M.E.; Wieland, R.; Zarnstorff, M.C.

1995-01-01

Both global and thermal energy confinement improve in high-temperature supershot plasmas in the Tokamak Fusion Test Reactor (TFTR) when deuterium beam heating is partially or wholly replaced by tritium beam heating. For the same heating power, the tritium-rich plasmas obtain up to 22% higher total energy, 30% higher thermal ion energy, and 20-25% higher central ion temperature. Kinetic analysis of the temperature and density profiles indicates a favorable isotopic scaling of ion heat transport and electron particle transport, with τ Ei (a/2) ∝ left angle A right angle 0.7-0.8 and τ pe (a) ∝ left angle A right angle 0.8 . (orig.)

Synthesis of deuterium and tritium labelled tyrosine

International Nuclear Information System (INIS)

Kanska, M.; Drabarek, S.

1980-01-01

A new method of synthesis of tyrosine labelled with deuterium and tritium in the aromatic ring has been developed. Deuterated and tritiated tyrosine was obtained by isotope exchange between tyrosine and deuterated or tritiated water at elevated temperature in hydrochloric acid medium using K 2 PtCl 4 as a catalyst. For synthesis of tritiated tyrosine 1 Ci HTO was used; the specific activity of the product was 5 mCi/mMol. (author)

ICRF heating and transport of deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

Rogers, J.H.; Schilling, G.; Stevens, J.E.; Taylor, G.; Wilson, J.R.; Bell, M.G.; Budny, R.V.; Bretz, N.L.; Darrow, D.; Fredrickson, E.

1995-02-01

This paper describes results of the first experiments utilizing high-power ion cyclotron range of frequency (ICRF) to heat deuterium-tritium (D-T) plasmas in reactor-relevant regimes on the Tokamak Fusion Test Reactor (TFTR). Results from these experiments have demonstrated efficient core, second harmonic, tritium beating of D-T supershot plasmas with tritium concentrations ranging from 6%-40%. Significant direct ion heating on the order of 60% of the input radio frequency (rf) power has been observed. The measured deposition profiles are in good agreement with two-dimensional modeling code predictions. Energy confinement in an rf-heated supershot is at least similar to that without rf, and possibly better in the electron channel. Efficient electron heating via mode conversion of fast waves to ion Bernstein waves (IBW) has been demonstrated in ohmic, deuterium-deuterium and DT-neutral beam injection plasmas with high concentrations of minority 3 He (n 3He /n e = 15% - 30%). By changing the 3 He concentration or the toroidal field strength, the location of the mode-conversion radius was varied. The power deposition profile measured with rf power modulation indicated that up to 70% of the power can be deposited on electrons at an off-axis position. Preliminary results with up to 4 MW coupled into the plasma by 90-degree phased antennas showed directional propagation of the mode-converted IBW. Analysis of heat wave propagation showed no strong inward thermal pinch in off-axis heating of an ohmically-heated target plasma in TFTR

Long Term Retention of Deuterium and Tritium in Alcator C-Mod

International Nuclear Information System (INIS)

FIORE, C.; LABOMBARD, B.; LIPSCHULTZ, B.; PITCHER, C.S.; SKINNER, C.H.; WAMPLER, WILLIAM R.

1999-01-01

We estimate the total in-vessel deuterium retention in Alcator C-Mod from a run campaign of about 1090 plasmas. The estimate is based on measurements of deuterium retained on 22 molybdenum tiles from the inner wall and divertor. The areal density of deuterium on the tiles was measured by nuclear reaction analysis. From these data, the in-vessel deuterium inventory is estimated to be about 0.1 gram, assuming the deuterium coverage is toroidally symmetric. Most of the retained deuterium is on the walls of the main plasma chamber, only about 2.5% of the deuterium is in the divertor. The D coverage is consistent with a layer saturated by implantation with ions and charge-exchange neutrals from the plasma. This contrasts with tokamaks with carbon plasma-facing components (PFC's) where long-term retention of tritium and deuterium is large and mainly in the divertor due to codeposition with carbon eroded by the plasma. The low deuterium retention in the C-Mod divertor is mainly due to the absence of carbon PFC's in C-Mod and the low erosion rate of Mo

Tritium pellet injector for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Gouge, M.J.; Baylor, L.R.; Combs, S.K.; Fisher, P.W.; Foust, C.R.; Milora, S.L.

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) plasma phase. An existing deuterium pellet injector (DPI) was 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 for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet 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 (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

Laser separation of hydrogen isotopes: Tritium-from-deuterium recovery

International Nuclear Information System (INIS)

Magnotta, F.; Herman, I.P.; Aldridge, F.T.; Maienschein, J.L.

1984-01-01

Single-step enrichment factors exceeding 15,000 have been observed in the removal of tritium-from-deuterium by 12 μm laser multiple-photon dissociation of chloroform. The photochemistry and photophysics of this process is discussed along with prospects for implementation of this method in practical heavy water reactor detritiation. 7 refs., 7 figs., 1 tab

Burn performance of deuterium-tritium, deuterium-deuterium, and catalyzed deuterium ICF targets

International Nuclear Information System (INIS)

Harris, D.B.; Blue, T.E.

1983-01-01

The University of Illinois hydrodynamic burn code, AFBURN, has been used to model the performance of homogeneous D-T, D 2 , and catalyzed deuterium ICF targets. Yields and gains are compared for power-producing targets. AFBURN is a one-dimensional, two-temperature, single-fluid hydrodynamic code with non-local fusion product energy deposition. The initial conditions for AFBURN are uniformly compressed targets with central hot spots. AFBURN predicts that maximum D 2 target gains are obtained for target rhoR and spark rhoR about seven times larger than the target and spark rhoR for maximum D-T target gains, that the maximum D 2 target gain is approximately one third of the maximum D-T target gain, and that the corresponding yields are approximately equal. By recycling tritium and 3 He from previous targets, D 2 target performance can be improved by about 10%. (author)

Assembly of high-areal-density deuterium-tritium fuel from indirectly driven cryogenic implosions.

Science.gov (United States)

Mackinnon, A J; Kline, J L; Dixit, S N; Glenzer, S H; Edwards, M J; Callahan, D A; Meezan, N B; Haan, S W; Kilkenny, J D; Döppner, T; Farley, D R; Moody, J D; Ralph, J E; MacGowan, B J; Landen, O L; Robey, H F; Boehly, T R; Celliers, P M; Eggert, J H; Krauter, K; Frieders, G; Ross, G F; Hicks, D G; Olson, R E; Weber, S V; Spears, B K; Salmonsen, J D; Michel, P; Divol, L; Hammel, B; Thomas, C A; Clark, D S; Jones, O S; Springer, P T; Cerjan, C J; Collins, G W; Glebov, V Y; Knauer, J P; Sangster, C; Stoeckl, C; McKenty, P; McNaney, J M; Leeper, R J; Ruiz, C L; Cooper, G W; Nelson, A G; Chandler, G G A; Hahn, K D; Moran, M J; Schneider, M B; Palmer, N E; Bionta, R M; Hartouni, E P; LePape, S; Patel, P K; Izumi, N; Tommasini, R; Bond, E J; Caggiano, J A; Hatarik, R; Grim, G P; Merrill, F E; Fittinghoff, D N; Guler, N; Drury, O; Wilson, D C; Herrmann, H W; Stoeffl, W; Casey, D T; Johnson, M G; Frenje, J A; Petrasso, R D; Zylestra, A; Rinderknecht, H; Kalantar, D H; Dzenitis, J M; Di Nicola, P; Eder, D C; Courdin, W H; Gururangan, G; Burkhart, S C; Friedrich, S; Blueuel, D L; Bernstein, L A; Eckart, M J; Munro, D H; Hatchett, S P; Macphee, A G; Edgell, D H; Bradley, D K; Bell, P M; Glenn, S M; Simanovskaia, N; Barrios, M A; Benedetti, R; Kyrala, G A; Town, R P J; Dewald, E L; Milovich, J L; Widmann, K; Moore, A S; LaCaille, G; Regan, S P; Suter, L J; Felker, B; Ashabranner, R C; Jackson, M C; Prasad, R; Richardson, M J; Kohut, T R; Datte, P S; Krauter, G W; Klingman, J J; Burr, R F; Land, T A; Hermann, M R; Latray, D A; Saunders, R L; Weaver, S; Cohen, S J; Berzins, L; Brass, S G; Palma, E S; Lowe-Webb, R R; McHalle, G N; Arnold, P A; Lagin, L J; Marshall, C D; Brunton, G K; Mathisen, D G; Wood, R D; Cox, J R; Ehrlich, R B; Knittel, K M; Bowers, M W; Zacharias, R A; Young, B K; Holder, J P; Kimbrough, J R; Ma, T; La Fortune, K N; Widmayer, C C; Shaw, M J; Erbert, G V; Jancaitis, K S; DiNicola, J M; Orth, C; Heestand, G; Kirkwood, R; Haynam, C; Wegner, P J; Whitman, P K; Hamza, A; Dzenitis, E G; Wallace, R J; Bhandarkar, S D; Parham, T G; Dylla-Spears, R; Mapoles, E R; Kozioziemski, B J; Sater, J D; Walters, C F; Haid, B J; Fair, J; Nikroo, A; Giraldez, E; Moreno, K; Vanwonterghem, B; Kauffman, R L; Batha, S; Larson, D W; Fortner, R J; Schneider, D H; Lindl, J D; Patterson, R W; Atherton, L J; Moses, E I

2012-05-25

The National Ignition Facility has been used to compress deuterium-tritium to an average areal density of ~1.0±0.1 g cm(-2), which is 67% of the ignition requirement. These conditions were obtained using 192 laser beams with total energy of 1-1.6 MJ and peak power up to 420 TW to create a hohlraum drive with a shaped power profile, peaking at a soft x-ray radiation temperature of 275-300 eV. This pulse delivered a series of shocks that compressed a capsule containing cryogenic deuterium-tritium to a radius of 25-35 μm. Neutron images of the implosion were used to estimate a fuel density of 500-800 g cm(-3).

Complementarity of deuterium and tritium measurements to study hailstone formation. An example

International Nuclear Information System (INIS)

Jouzel, Jean.

1975-12-01

The complementarity of deuterium and tritium measurement to study hailstone formation is presented and this method is applied to measurements made on three hailstones produced during a storm on August 7th 1971 in the province of Alberta (Canada). Firstly, the isotopic variations of condensed water in a hail cloud are studied on the basis of a cloud model as proposed by Chisholm. The importance of tritium analysis is demonstrated and conditions of validity of the isotopic model proposed by Merlivat, Nief and Roth in the case of deuterium are defined. The tritium measurement method used for analysis of small water samples and carrier out with an Oeschgger gas counter is described. The results connected with three Alberta hailstones are discussed. It is demonstrated particularly that these hailstones were formed during a succession of upward and downward movements, at least two of the latter occuring in the updraft core, and that the distribution of the updraft velocity with altitude was not stable during the storm which is in agreement with its multicellular nature [fr

Neutron yield in experiments with Z-pinches in frozen deuterium filaments

International Nuclear Information System (INIS)

Meierovich, B.E.; Sukhorukov, S.T.

1991-01-01

Experiments on the properties of the neutron yield and stability of dense Z-pinches in frozen deuterium filaments two years ago have still not received a proper explanation. A discharge in a dielectric deuterium filament differs qualitatively in its behavior from one in a metal conductor. The authors assert attempts to interpret the experiment have not taken this fact properly into account. The enhanced stability and substantial neutron yield do not follow from the classical picture of a Bennett equilibrium in the current channel. The most important experimental characteristic of a discharge in frozen deuterium is the dependence of the neutron yield on the current strength at a given stage. The spread in the experimental points is quite large so one naturally prefers t describe the experimental data by means of a physically simple model and not go to a full-scale simulation of all the processes. When the deuterium is fully ionized the electron drift velocity is smaller than the thermal velocity, enabling calculation of the neutron yield by means of a self-similar model of the Z-pinch compression which treats electron degeneracy. To find the neutron yield it suffices to consider only the first half-period of the self-similar oscillations including the state of maximum compression. The subsequent evolution of the current channel, which is associated with radial expansion, does not contribute significantly to the neutron yield

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.

Design of a Multi-Tube Pd-Membrane Module for Tritium Recovery from He in DEMO

Directory of Open Access Journals (Sweden)

Marco Incelli

2016-10-01

Full Text Available Dense self-supported Pd-alloy membranes are used to selectively separate hydrogen and hydrogen isotopes. In particular, deuterium (D and tritium (T are currently identified as the main elements for the sustainability of the nuclear fusion reaction aimed at carbon free power generation. In the fusion nuclear reactors, a breeding blanket produces the tritium that is extracted and purified before being sent to the plasma chamber in order to sustain the fusion reaction. In this work, the application of Pd-alloy membranes has been tested for recovering tritium from a solid breeding blanket through a helium purge stream. Several simulations have been performed in order to optimize the design of a Pd-Ag multi-tube module in terms of geometry, operating parameters, and membrane module configuration (series vs. parallel. The results demonstrate that a pre-concentration stage before the Pd-membrane unit is mandatory because of the very low tritium concentration in the He which leaves the breeding blanket of the fusion reactor. The most suitable operating conditions could be reached by: (i increasing the hydrogen partial pressure in the lumen side and (ii decreasing the shell pressure. The preliminary design of a membrane unit has been carried out for the case of the DEMO fusion reactor: the optimized membrane module consists of an array of 182 Pd-Ag tubes of 500 mm length, 10 mm diameter, and 0.100 mm wall thickness (total active area of 2.85 m2.

Energy cost of negative pion production on deuterium-tritium target

Energy Technology Data Exchange (ETDEWEB)

Kuzminov, V.V. (Petersburg Nuclear Physics Inst., Gatchina, St. Petersburg (Russian Federation)); Petrov, Yu.V. (Petersburg Nuclear Physics Inst., Gatchina, St. Petersburg (Russian Federation)); Shabelski, Yu.M. (Petersburg Nuclear Physics Inst., Gatchina, St. Petersburg (Russian Federation))

1993-12-01

The negative pion production by deuterons (T[sub 0] = 0.8 GeV/nucl.) was calculated for a cylindrical gaseous deuterium-tritium target (the density of DT-mixture is [phi] = 0.5). Revised cross sections of nucleon-nucleus interaction were used in a Monte Carlo simulation and multiple nucleon-nuclei collisions were taken into account. The energy cost of negative pion production is [epsilon][sub [pi][sup -

Wildcat: A commercial deuterium-deuterium tokamak reactor

International Nuclear Information System (INIS)

Evans, K.; Baker, C.C.; Barry, K.M.

1983-01-01

WILDCAT is a conceptual design of a catalyzed deuterium-deuterium tokamak commercial fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing deuterium-tritium (D-T) designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete conceptual design

Physics of high performance deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

McGuire, K.M.; Batha, S.

1996-11-01

During the past two years, deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR) have been used to study fusion power production, isotope effects associated with tritium fueling, and alpha-particle physics in several operational regimes. The peak fusion power has been increased to 10.7 MW in the supershot mode through the use of increased plasma current and toroidal magnetic field and extensive lithium wall conditioning. The high-internal-inductance (high-I i ) regime in TFTR has been extended in plasma current and has achieved 8.7 MW of fusion power. Studies of the effects of tritium on confinement have now been carried out in ohmic, NBI- and ICRF- heated L-mode and reversed-shear plasmas. In general, there is an enhancement in confinement time in D-T plasmas which is most pronounced in supershot and high-I i discharges, weaker in L-mode plasmas with NBI and ICRF heating and smaller still in ohmic plasmas. In reversed-shear discharges with sufficient deuterium-NBI heating power, internal transport barriers have been observed to form, leading to enhanced confinement. Large decreases in the ion heat conductivity and particle transport are inferred within the transport barrier. It appears that higher heating power is required to trigger the formation of a transport barrier with D-T NBI and the isotope effect on energy confinement is nearly absent in these enhanced reverse-shear plasmas. Many alpha-particle physics issues have been studied in the various operating regimes including confinement of the alpha particles, their redistribution by sawteeth, and their loss due to MHD instabilities with low toroidal mode numbers. In weak-shear plasmas, alpha-particle destabilization of a toroidal Alfven eigenmode has been observed

Tritium target manufacturing for use in accelerators

Science.gov (United States)

Bach, P.; Monnin, C.; Van Rompay, M.; Ballanger, A.

2001-07-01

As a neutron tube manufacturer, SODERN is now in charge of manufacturing tritium targets for accelerators, in cooperation with CEA/DAM/DTMN in Valduc. Specific deuterium and tritium targets are manufactured on request, according to the requirements of the users, starting from titanium target on copper substrate, and going to more sophisticated devices. A wide range of possible uses is covered, including thin targets for neutron calibration, thick targets with controlled loading of deuterium and tritium, rotating targets for higher lifetimes, or large size rotating targets for accelerators used in boron neutron therapy. Activity of targets lies in the 1 to 1000 Curie, diameter of targets being up to 30 cm. Special targets are also considered, including surface layer targets for lowering tritium desorption under irradiation, or those made from different kinds of occluders such as titanium, zirconium, erbium, scandium, with different substrates. It is then possible to optimize either neutron output, or lifetime and stability, or thermal behavior.

Hot muonic deuterium and tritium from cold targets

International Nuclear Information System (INIS)

Marshall, G.M.; Beveridge, J.L.; Bailey, J.M.; Beer, G.A.; Knowles, P.E.; Mason, G.R.; Olin, A.; Brewer, J.H.; Forster, B.M.; Huber, T.M.; Pippitt, B.; Jacot-Guillarmod, R.; Schellenberg, L.; Martoff, C.J.; Petitjean, C.

1992-01-01

Experiments are described which use a solid hydrogen layer to form muonic hydrogen isotopes in vacuum. The method relies on transfer of the muon from protium to either a deuteron or a triton. The resulting muonic deuterium or muonic tritium will not immediately thermalize because of the very low elastic cross sections, and may be emitted from the surface of the layer. Measurements which detect decay electrons, muonic x-rays, and fusion products have been used to study the processes. A target has been constructed which exploits muonic atom emission in order to study the energy dependence of transfer and muon molecular formation

Thermonuclear reaction rates in a deuterium-tritium plasma

International Nuclear Information System (INIS)

Beckman, L.

1978-12-01

In a deuterium-tritium plasma six thermonuclear reactions take place between the deuterons, tritons and the 3 He-particles formed in about half of the d-d-reactions. The rate constants for these six reactions have been calculated from the latest evaluations of the reaction cross sections which were available. In some cases, notably the reactions t+t, t+ 3 He and 3 He+ 3 He, the number of published cross section measurements is small, and the uncertainty in the calculated rate constants consequently large. Analytical expressions for the rate constants as functions of the plasma temperature have been set up. (author)

Transport of deuterium, tritium and helium in a tokamak

International Nuclear Information System (INIS)

Potters, J.H.H.M.

1984-02-01

A one-dimensional numerical model for determining steady-state radial profiles of the densities of the particles, including neutrals, in a multispecies toroidal plasma is described. For prescribed temperature profiles, the coupled momentum and particle balances of the ions are solved numerically with a newly developed compact finite difference scheme for a non-equidistant mesh. Neutral densities are obtained by solving the Boltzmann equations, using a collocation method. The model is applied to deuterium-tritium plasmas without and with a helium admixture. For the charged particles, Pfirsch-Schlueter transport, including the highly collisional extension, and either of two anomalous transport models are adopted. For equal densities of deuterons and tritons in the plasma centre, the neutral tritium density in front of the wall is found to be 1.3 to 1.6 times higher than that of deuterium, depending on the plasma density, the temperature profile and the transport model. Secondly, it is found that pumping neutral helium, originating from fusion alpha particles, out of a cold plasma/gas blanket surrounding the hot plasma is not feasible, as the helium gas density, corresponding to a relative abundance of alpha-particles in the plasma core below 10%, is very low. Although depending strongly on the ion transport model and being increased by elastic collisions between neutral helium and charged hydrogen isotopes, the neutral helium enrichment ratio is always much less than unity. (Auth.)

Ultraviolet radiation-damage absorption peak in solid deuterium-tritium. Revision 1

International Nuclear Information System (INIS)

Fearon, E.M.; Tsugawa, R.T.; Souers, P.C.; Poll, J.D.; Hunt, J.L.

1985-01-01

An ultraviolet absorption peak has been seen in solid deuterium-tritium and hydrogen-tritium at a sensor temperature of 5 K. The peak occurs at 3.6 eV and is about 1.5 eV wide. It bleaches out when the temperature is raised to about 10 K but reappears upon cooling and is, therefore, radiation induced. At 5 K, the peak forms on a time scale of minutes and appears to represent part-per-million levels of electron-mass defects. The suggested model is that of a trapped electron, where the peak is the ground state-to-the-conduction band transition. A marked isotope effect is seen between D-T and H-T

Implosion characteristics of deuterium--tritium pellets surrounded by high-density shells

International Nuclear Information System (INIS)

Fraley, G.S.

1976-09-01

The effect of high-density shells on deuterium-tritium pellets imploded by laser energy deposition or other means is investigated. Attention is centered on the inner parts of the pellet where hydrodynamics is the dominant mechanism. The implosions can then be characterized by a pressure boundary condition. Numerical solutions of the implosions are carried out over a wide range of parameters both for solid pellets and pellets with a central void

Spin-polarized hydrogen, deuterium, and tritium : I

International Nuclear Information System (INIS)

Haugen, M.; Ostgaard, E.

1989-01-01

The ground-state energy of spin-polarized hydrogen, deuterium and tritium is calculated by means of a modified variational lowest order constrained-variation method, and the calculations are done for five different two-body potentials. Spin-polarized H is not self-bound according to our theoretical results for the ground-state binding energy. For spin-polarized D, however, we obtain theoretical results for the ground-state binding energy per particle from -0.4 K at an equilibrium particle density of 0.25 σ -3 or a molar volume of 121 cm 3 /mol to +0.32 K at an equilibrium particle density of 0.21 σ -3 or a molar volume of 142 cm 3 /mol, where σ = 3.69 A (1A = 10 -10 m). It is, therefore, not clear whether spin-polarized deuterium should be self-bound or not. For spin-polarized T, we obtain theoretical results for the ground-state binding energy per particle from -4.73 K at an equilibrium particle density of 0.41 σ -3 or a molar volume of 74 cm 3 /mol to -1.21 K at an equilibrium particle density of 0.28 σ -3 or a molar volume of 109 cm 3 /mol. (Author) 27 refs., 9 figs., tab

In situ sampling for pressure tube deuterium concentration

International Nuclear Information System (INIS)

Harrington, A.J.; Kittmer, C.A.

1988-01-01

The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

Synthesis of multidrug resistance modulator LY335979 labeled with deuterium and tritium

International Nuclear Information System (INIS)

Czeskis, B.A.

1997-01-01

DIDEUTERO AND DITRITIOISOTOPOMERS OF THE MULTIDRUG RESISTANCE MODULATOR LY335979 WERE PREPARED BY INITIAL BROMINATION OF 5-HYDROXYQUINOLINE UNDER ACIDIC CONDITIONS FOLLOWED BY MITSUNOBU COUPLING OF 6,8-DIBROMO-5-HYDROXYQUINOLINE WITH (S)-GLYCIDOL. OPENING OF THE RESULTING EPOXIDE WITH DIBENZOSUBERYLPIPERAZINE LY335995 RESULTED IN DIBROMOANALOG OF LY335979, WHICH WAS FINALLY REDUCTIVELY DEBROMINATED WITH DEUTERIUM OR TRITIUM IN THE PRESENCE OF PALLADIUM ON CARBON. (AUTHOR)

Diagnosing radiative shocks from deuterium and tritium implosions on NIF.

Science.gov (United States)

Pak, A; Divol, L; Weber, S; Döppner, T; Kyrala, G A; Kilne, J; Izumi, N; Glenn, S; Ma, T; Town, R P; Bradley, D K; Glenzer, S H

2012-10-01

During the recent ignition tuning campaign at the National Ignition Facility, layered cryogenic deuterium and tritium capsules were imploded via x-ray driven ablation. The hardened gated x-ray imager diagnostic temporally and spatially resolves the x-ray emission from the core of the capsule implosion at energies above ~8 keV. On multiple implosions, ~200-400 ps after peak compression a spherically expanding radiative shock has been observed. This paper describes the methods used to characterize the radial profile and rate of expansion of the shock induced x-ray emission.

[Solid state isotope hydrogen exchange for deuterium and tritium in human gene-engineered insulin].

Science.gov (United States)

Zolotarev, Yu A; Dadayan, A K; Kozik, V S; Gasanov, E V; Nazimov, I V; Ziganshin, R Kh; Vaskovsky, B V; Murashov, A N; Ksenofontov, A L; Haribin, O N; Nikolaev, E N; Myasoedov, N F

2014-01-01

The reaction of high temperature solid state catalytic isotope exchange in peptides and proteins under the action of catalyst-activated spillover hydrogen was studied. The reaction of human gene-engineered insulin with deuterium and tritium was conducted at 120-140° C to produce insulin samples containing 2-6 hydrogen isotope atoms. To determine the distribution of the isotope label over tritium-labeled insulin's amino acid residues, oxidation of the S-S bonds of insulin by performic acid was performed and polypeptide chains isolated; then their acid hydrolysis, amino acid analysis and liquid scintillation counts of tritium in the amino acids were conducted. The isotope label was shown to be incorporated in all amino acids of the protein, with the peptide fragment FVNQHLCGSHLVE of the insulin β-chain showing the largest incorporation. About 45% of the total protein isotope label was incorporated in His5 and His10 of this fragment. For the analysis of isotope label distribution in labeled insulin's peptide fragments, the recovery of the S-S bonds by mercaptoethanol, the enzymatic hydrolysis by glutamyl endopeptidase from Bacillus intermedius and HPLC division of the resulting peptides were carried out. Attribution of the peptide fragments formed due to hydrolysis at the Glu-X bond in the β-chain was accomplished by mass spectrometry. Mass spectrometry analysis data of the deuterium-labeled insulin samples' isotopomeric composition showed that the studied solid state isotope exchange reaction equally involved all the protein molecules. Biological studying of tritium-labeled insulin showed its physiological activity to be completely retained.

Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

International Nuclear Information System (INIS)

Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

1989-01-01

Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm 2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

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

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

Current CTR-related tritium handling studies at ORNL

International Nuclear Information System (INIS)

Watson, J.S.; Bell, J.T.; Clinton, S.D.; Fisher, P.W.; Redman, J.D.; Smith, F.J.; Talbot, J.B.; Tung, C.P.

1976-01-01

The Oak Ridge National Laboratory has a comprehensive program concerned with plasma fuel recycle, tritium recovery from blankets, and tritium containment in fusion reactors. Two studies of most current interest are investigations of cryosorption pumping of hydrogen isotopes and measurements of tritium permeation rates through steam generator materials. Cryosorption pumping speeds have been measured for hydrogen, deuterium, and helium at pressures from 10 -8 torr to 3 x 10 -3 torr. Permeation rates through Incoloy 800 have been shown to be drastically reduced when the low pressure side of permeation tubes are exposed to steam. These results will be important considerations in the design of fusion reactor steam generators

Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

Energy Technology Data Exchange (ETDEWEB)

Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

2015-06-15

In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

Fusion power production from TFTR plasmas fueled with deuterium and tritium

International Nuclear Information System (INIS)

Strachan, J.D.; Adler, H.; Alling, P.

1994-03-01

Peak fusion power production of 6.2 ± 0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2 x 10 17 m -3 without the appearance of either disruptive MHD events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Ploc, R.A.; McRae, G.A.

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable β-Zr that is present as a thin layer encasing the predominant α-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded β-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the β-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Ploc, R.A.; McRae, G.A

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable {beta}-Zr that is present as a thin layer encasing the predominant {alpha}-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded {beta}-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the {beta}-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

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

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.)

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

Compression of a spherically symmetric deuterium-tritium plasma liner onto a magnetized deuterium-tritium target

International Nuclear Information System (INIS)

Santarius, J. F.

2012-01-01

Converging plasma jets may be able to reach the regime of high energy density plasmas (HEDP). The successful application of plasma jets to magneto-inertial fusion (MIF) would heat the plasma by fusion products and should increase the plasma energy density. This paper reports the results of using the University of Wisconsin’s 1-D Lagrangian, radiation-hydrodynamics, fusion code BUCKY to investigate two MIF converging plasma jet test cases originally analyzed by Samulyak et al.[Physics of Plasmas 17, 092702 (2010)]. In these cases, 15 cm or 5 cm radially thick deuterium-tritium (DT) plasma jets merge at 60 cm from the origin and converge radially onto a DT target magnetized to 2 T and of radius 5 cm. The BUCKY calculations reported here model these cases, starting from the time of initial contact of the jets and target. Compared to the one-temperature Samulyak et al. calculations, the one-temperature BUCKY results show similar behavior, except that the plasma radius remains about twice as long near maximum compression. One-temperature and two-temperature BUCKY results differ, reflecting the sensitivity of the calculations to timing and plasma parameter details, with the two-temperature case giving a more sustained compression.

Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

International Nuclear Information System (INIS)

Le-Guillou, Mael

2014-01-01

In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

First results of radiation-driven, layered deuterium-tritium implosions with a 3-shock adiabat-shaped drive at the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Smalyuk, V. A.; Robey, H. F.; Döppner, T.; Jones, O. S.; Milovich, J. L.; Bachmann, B.; Baker, K. L.; Berzak Hopkins, L. F.; Bond, E.; Callahan, D. A.; Casey, D. T.; Celliers, P. M.; Cerjan, C.; Clark, D. S.; Dixit, S. N.; Edwards, M. J.; Haan, S. W.; Hamza, A. V.; Hurricane, O. A.; Jancaitis, K. S. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

2015-08-15

Radiation-driven, layered deuterium-tritium plastic capsule implosions were carried out using a new, 3-shock “adiabat-shaped” drive on the National Ignition Facility. The purpose of adiabat shaping is to use a stronger first shock, reducing hydrodynamic instability growth in the ablator. The shock can decay before reaching the deuterium-tritium fuel leaving it on a low adiabat and allowing higher fuel compression. The fuel areal density was improved by ∼25% with this new drive compared to similar “high-foot” implosions, while neutron yield was improved by more than 4 times, compared to “low-foot” implosions driven at the same compression and implosion velocity.

Deuterium/hydrogen isotope exchange on beryllium and beryllium nitride; Deuterium/Wasserstoff-Isotopenaustausch an Beryllium und Berylliumnitrid

Energy Technology Data Exchange (ETDEWEB)

Dollase, Petra; Eichler, Michael; Koeppen, Martin; Dittmar, Timo; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik (Germany)

2016-07-01

In the fusion experiments JET and ITER, the first wall is made up of beryllium. The use of nitrogen is discussed for radiative cooling in the divertor. This can react with the surface of the first wall to form beryllium nitride (Be{sub 3}N{sub 2}). The hydrogen isotopes deuterium and tritium, which react in the fusion reaction to helium and a neutron, are used as fuel. Since the magnetic confinement of the plasma is not perfect, deuterium and tritium ions are also found on the beryllium wall and can accumulate there. This should be avoided due to the radioactivity of tritium. Therefore the isotope exchange with deuterium is investigated to regenerate the first wall. We investigate the isotopic exchange of deuterium and protium in order to have not to work with radioactive tritium. The ion bombardment is simulated with an ion source. With voltages up to a maximum of 5 kV, deuterium and protic hydrogen ions are implanted in polycrystalline Be and Be{sub 3}N{sub 2}. The samples are then analyzed in situ using X-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS). Subsequently, samples prepared under the same conditions are characterized ex-situ by means of nuclear reaction analysis (NRA). [German] In den Fusionsexperimenten JET und ITER besteht die erste Wand im Hauptraum aus Beryllium (Be). Zur Strahlungskuehlung im Divertor wird der Einsatz von Stickstoff diskutiert. Dieser kann mit der Oberflaeche der ersten Wand zu Berylliumnitrid (Be{sub 3}N{sub 2}) reagieren. Als Brennstoff werden die Wasserstoffisotope Deuterium und Tritium eingesetzt, die in der Fusionsreaktion zu Helium und einem Neutron reagieren. Da der magnetische Einschluss des Plasmas nicht perfekt ist, treffen auch Deuterium- und Tritiumionen auf die Berylliumwand auf und koennen sich dort anreichern. Das soll aufgrund der Radioaktivitaet von Tritium unbedingt vermieden werden. Daher wird zur Regenerierung der ersten Wand der Isotopenaustausch mit Deuterium untersucht. Wir

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

Numerical studies of deuterium-tritium ignition in impact-fusion targets

International Nuclear Information System (INIS)

Zubrin, R.M.; Ribe, F.L.

1989-01-01

A numerical one-dimensional solution of the Euler equations for an imploding spherical tungsten shell with internal deuterium-tritium gas is applied to study impact-fusion dynamics with parameters of fusion reactor relevance. Thermal conduction and radiative energy loss by the plasma are taken into account, as is heating by fusion generated alpha particles. A variety of target sizes and impact velocities are examined, and scaling laws for fusion yields are deduced which define possible parameters for conceptual commercial impact-fusion power reactors. It is found that shell energies and velocities of about 30 MJ and 110 km/s would be satisfactory. A commercial impact-fusion reactor based on such parameters is discussed

Characterization of a deuterium-deuterium plasma fusion neutron generator

Science.gov (United States)

Lang, R. F.; Pienaar, J.; Hogenbirk, E.; Masson, D.; Nolte, R.; Zimbal, A.; Röttger, S.; Benabderrahmane, M. L.; Bruno, G.

2018-01-01

We characterize the neutron output of a deuterium-deuterium plasma fusion neutron generator, model 35-DD-W-S, manufactured by NSD/Gradel-Fusion. The measured energy spectrum is found to be dominated by neutron peaks at 2.2 MeV and 2.7 MeV. A detailed GEANT4 simulation accurately reproduces the measured energy spectrum and confirms our understanding of the fusion process in this generator. Additionally, a contribution of 14 . 1 MeV neutrons from deuterium-tritium fusion is found at a level of 3 . 5%, from tritium produced in previous deuterium-deuterium reactions. We have measured both the absolute neutron flux as well as its relative variation on the operational parameters of the generator. We find the flux to be proportional to voltage V 3 . 32 ± 0 . 14 and current I 0 . 97 ± 0 . 01. Further, we have measured the angular dependence of the neutron emission with respect to the polar angle. We conclude that it is well described by isotropic production of neutrons within the cathode field cage.

Process and device for step by step enrichment of deuterium and/or tritium by isotope exchange

International Nuclear Information System (INIS)

Iniotakis, N.; Decken, C.B. von der.

1984-01-01

Deuterium and/or tritium are bound to steam by isotope exchange after permeation through an exchange wall. Primary and secondary flows are guided by the counterflow principle. The secondary side has a metal oxide as oxidation catalyst. The plant can consist of several enrichment stages. The various flows and parts of the plant are described. (PW) [de

Diffusion of hydrogen, deuterium, and tritium in niobium

International Nuclear Information System (INIS)

Matusiewicz, G.R.

1981-01-01

The diffusion of hydrogen in niobium was investigated over the temperature range 148 to 500 degrees Kelvin, using measurements of the elastic after effect caused by long range diffusion (the Gorsky Effect). Relaxation curves for pure annealed niobium were generally not of the single exponential form expected from the Gorsky Effect theory, but were described well by a sum of two exponential curves with different amplitudes and relaxation times. The effects of oxygen and nitrogen interstitials on the diffusion were studied and were not in agreement with conventional trapping models. Deuterium and tritium diffusion in niobium were also studied, and a non-classical isotope effect was observed. Hydrogen diffusion coefficients in several Nb-Ta alloys were measured, and the diffusivity in all these alloys exhibited a non-Arrhenius temperature dependence. Experimental results were compared to several models for diffusion and trapping. A model is presented which can account for the form of the relaxation curves observed in pure, annealed niobium

Tritium containment of controlled thermonuclear fusion reactor

International Nuclear Information System (INIS)

Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

1979-01-01

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

Deuterium-tritium TFTR plasmas in the high poloidal beta regime

International Nuclear Information System (INIS)

Sabbagh, S.A.; Mauel, M.E.; Navratil, G.A.

1995-03-01

Deuterium-tritium plasmas with enhanced energy confinement and stability have been produced in the high poloidal beta, advanced tokamak regime in TFTR. Confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.46 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasma,s exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Fusion power exceeding 6.7 MW with a fusion power gain Q DT = 0.22 has been produced with reduced alpha particle first orbit loss provided by the increased l i

Synthesis of tritium or deuterium labelled 19-nor-3. cap alpha. -hydroxy-5. cap alpha. -androstan-17-one from nortestosterone

Energy Technology Data Exchange (ETDEWEB)

Protiva, J; Klinotova, E [Karlova Univ., Prague (Czechoslovakia). Prirodovedecka Fakulta; Filip, J [Ustav pro Vyzkum, Vyrobu a Vyuziti Radioisotopu, Prague (Czechoslovakia); Hampl, R [Research Inst. of Endocrinology, Praha (Czechoslovakia)

1982-10-20

Tritium and/or deuterium (5-H) labelled 19-nor-3..cap alpha..-hydroxy-5..cap alpha..-androstan-17-one (norandrosterone) was prepared from nortestosterone in view to use it as a radioligand for radioimmunoassay of the main nortestosterone metabolites. Based upon model experiments using testosterone and deuterium labelling, the following four step procedure was established: nortestosterone was oxidized with pyridine chlorochromate and the resulting 19-nor-4-androsten-3,17-dione was tritiated with tritium gas under catalysis with tris(triphenylphosphine)rhodium chloride to give (4,5..cap alpha..-/sup 3/H)19-nor-5..cap alpha..-androstan-3,17-dione. A selective reduction of the latter compound yielded (5-/sup 3/H)19-nor-3..cap alpha..-hydroxy-5..cap alpha..-androstan-17-one of the molar radioactivity 0.3 TBq (8.15 Ci)/mmol.

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

Strong isotope effects on the charge transfer in slow collisions of He2+ with atomic hydrogen, deuterium, and tritium

NARCIS (Netherlands)

Stolterfoht, N.; Cabrera-Trujillo, R.; Oehrn, Y.; Deumens, E.; Hoekstra, R.; Sabin, J. R.

2007-01-01

Probabilities and cross sections for charge transfer by He2+ impact on atomic hydrogen (H), deuterium (D), and tritium (T) at low collision energies are calculated. The results are obtained using an ab initio theory, which solves the time-dependent Schrodinger equation. For the H target, excellent

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

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.)

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.

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.

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.)

Compact deuterium-tritium neutron generator using a novel field ionization source

Energy Technology Data Exchange (ETDEWEB)

Ellsworth, J. L., E-mail: ellsworth7@llnl.gov; Falabella, S.; Sanchez, J.; Tang, V. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94550 (United States); Wang, H. [Department of Computer Science, Stanford University, Stanford, California 94305 (United States)

2014-11-21

Active interrogation using neutrons is an effective method for detecting shielded nuclear material. A lightweight, lunch-box-sized, battery-operated neutron source would enable new concepts of operation in the field. We have developed at-scale components for a highly portable, completely self-contained, pulsed Deuterium-Tritium (DT) neutron source producing 14 MeV neutrons with average yields of 10{sup 7} n/s. A gated, field ionization ion source using etched electrodes has been developed that produces pulsed ion currents up to 500 nA. A compact Cockcroft-Walton high voltage source is used to accelerate deuterons into a metal hydride target for neutron production. The results of full scale DT tests using the field ionization source are presented.

Specific features of X-ray generation by plasma focus chambers with deuterium and deuterium–tritium fillings

Energy Technology Data Exchange (ETDEWEB)

Dulatov, A. K., E-mail: bogolubov@vniia.ru; Krapiva, P. S.; Lemeshko, B. D.; Mikhailov, Yu. V.; Moskalenko, I. N.; Prokuratov, I. A.; Selifanov, A. N. [All-Russia Research Institute of Automatics (Russian Federation)

2016-01-15

The process of hard X-ray (HXR) generation in plasma focus (PF) chambers was studied experimentally. The radiation was recorded using scintillation detectors with a high time resolution and thermoluminescent detectors in combination with the method of absorbing filters. Time-resolved analysis of the processes of neutron and X-ray generation in PFs is performed. The spectra of HXR emission from PF chambers with deuterium and deuterium–tritium fillings are determined. In experiments with PF chambers filled with a deuterium–tritium mixture, in addition to the HXR pulse with photon energies of up to 200–300 keV, a γ-ray pulse with photon energies of up to 2.5–3.0 MeV is recorded, and a mechanism of its generation is proposed.

Tritium stripping by a catalytic exchange stripper

International Nuclear Information System (INIS)

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

1991-01-01

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

Study of tritium permeation through Peach Bottom Steam Generator tubes

International Nuclear Information System (INIS)

Yang, L.; Baugh, W.A.; Baldwin, N.L.

1977-06-01

The report describes the equipment developed, samples tested, procedures used, and results obtained in the tritium permeation tests conducted on steam generator tubing samples which were removed from the Peach Bottom Unit No. 1 reactor

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)

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)

Tritium transport and control in the FED

International Nuclear Information System (INIS)

Rogers, M.L.

1981-01-01

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

High performance deuterium-tritium plasmas in TFTR

International Nuclear Information System (INIS)

Sabbagh, S.A.; Bell, M.G.

1995-03-01

Plasmas composed of nominally equal concentrations of deuterium and tritium (DT) have been created in TFTR with the goals of producing significant levels of fusion power and of examining the effects of DT fusion alpha particles. Conditioning of the limiter by the injection of lithium pellets has led to an approximate doubling of the energy confinement time, τ E , in supershot plasmas at high plasma current (I p ≤ 2.5 MA) and high heating power (P b ≤ 33 MW). Operation with DT typically results in an additional 20% increase in τ E . In the high poloidal beta, advanced tokamak regime in TFTR, confinement enhancement H triple-bond τ E /τ E ITER-89P > 4 has been obtained in a limiter H-mode configuration at moderate plasma current I p = 0.85 - 1.5 MA. By peaking the plasma current profile, β N dia triple-bond 10 8 tperpendicular > aB 0 /I p = 3 has been obtained in these plasmas, exceeding the β N limit for TFTR plasmas with lower internal inductance, l i . Confinement of alpha particles appears to be classical and losses due to collective effects have not been observed. While small fluctuations in fusion product loss were observed during ELMs, no large loss was detected in DT plasmas

Measurement of high-pressure shock waves in cryogenic deuterium-tritium ice layered capsule implosions on NIF.

Science.gov (United States)

Robey, H F; Moody, J D; Celliers, P M; Ross, J S; Ralph, J; Le Pape, S; Berzak Hopkins, L; Parham, T; Sater, J; Mapoles, E R; Holunga, D M; Walters, C F; Haid, B J; Kozioziemski, B J; Dylla-Spears, R J; Krauter, K G; Frieders, G; Ross, G; Bowers, M W; Strozzi, D J; Yoxall, B E; Hamza, A V; Dzenitis, B; Bhandarkar, S D; Young, B; Van Wonterghem, B M; Atherton, L J; Landen, O L; Edwards, M J; Boehly, T R

2013-08-09

The first measurements of multiple, high-pressure shock waves in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility have been performed. The strength and relative timing of these shocks must be adjusted to very high precision in order to keep the DT fuel entropy low and compressibility high. All previous measurements of shock timing in inertial confinement fusion implosions [T. R. Boehly et al., Phys. Rev. Lett. 106, 195005 (2011), H. F. Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] have been performed in surrogate targets, where the solid DT ice shell and central DT gas regions were replaced with a continuous liquid deuterium (D2) fill. This report presents the first experimental validation of the assumptions underlying this surrogate technique.

Design of deuterium and tritium pellet injector systems for Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Wysor, R.B.; Baylor, L.R.; Bryan, W.E.

1985-01-01

Three pellet injector designs developed by the Oak Ridge National Laboratory (ORNL) are planned for the Tokamak Fusion Test Reactor (TFTR) to reach the goal of a tritium pellet injector by 1988. These are the Repeating Pneumatic Injector (RPI), the Deuterium Pellet Injector (DPI) and the Tritium Pellet Injector (TPI). Each of the pellet injector designs have similar performance characteristics in that they deliver up to 4-mm-dia pellets at velocities up to 1500 m/s with a dsign goal to 2000 m/s. Similar techniques are utilized to freeze and extrude the pellet material. The injector systems incorporate three gun concepts which differ in the number of gun barrels and the method of forming and chambering the pellets. The RPI, a single barrel repeating design, has been operational on TFTR since April 1985. Fabrication and assembly are essentially complete for DPI, and TPI is presently on hold after completing about 80% of the design. The TFTR pellet injector program is described, and each of the injector systems is described briefly. Design details are discussed in other papers at this symposium

Consideration for a tritium removal facility at the Cernavoda Nuclear Power Station

International Nuclear Information System (INIS)

2006-01-01

Full text: A pre-feasibility study considering process options for a Tritium Removal Facility at the Cernavoda Nuclear Power Station has been completed by ICIT and AECL. Three different process options were considered. These three options differ in the front-end process used to transfer tritium from heavy water to deuterium gas. All three options use cryogenic distillation (CD) as a back end process to extract tritium from the deuterium gas stream and concentrate it into a small volume stream of pure DT or T 2 that can be immobilized on a titanium sponge. The first option for the front-end process is Liquid Phase Catalytic Exchange (LPCE). The LPCE column is used to transfer the tritium from the heavy water to a recirculating stream of deuterium gas. The separation of hydrogen isotopes takes place in the cryogenic distillation column. Tritium-depleted deuterium gas from the CD system is fed back to the LPCE column. The cryogenic distillation system concentrates the tritium into a small volume of elemental tritium for storage. Tritiated heavy water that has been purified to remove catalyst poisons is fed to the top of the LPCE column. The heavy water leaving the column is depleted in deuterium. Both existing detritiation plants built to detritiate CANDU reactors (the Darlington TRF in Canada and the Wolsung TRF in Korea) use variations of the LPCE-CD process. The second option uses electrolysis to convert tritiated heavy water into oxygen and tritiated deuterium gas. The deuterium gas is sent to the Cryogenic Distillation system to extract and concentrate the tritium. The tritium depleted deuterium gas is recombined with the electrolytic oxygen to give a tritium-depleted heavy water product. The third option uses a Combined Electrolysis and Catalytic Exchange (CECE) front end. A CECE process concentrates the tritium in the water and, using water electrolysis, converts the concentrated tritium into deuterium gas. An overhead catalytic recombiner converts the

Separation of hydrogen isotopes for tritium waste removal

International Nuclear Information System (INIS)

Wilkes, W.R.

1975-01-01

A distillation cascade for separating hydrogen isotopes was simulated by means of a multicomponent, multistage computer code. A hypothetical test mixture containing equal atomic fractions of protium, deuterium and tritium, equilibrated to high temperature molecular concentrations was used as feed. The results show that a two-column cascade can be used to separate the protium from the tritium. Deuterium appears both in the protium and the tritium product streams. (auth)

Deuterium desorption from tungsten using laser heating

Directory of Open Access Journals (Sweden)

J.H. Yu

2017-08-01

Full Text Available Retention and desorption of hydrogenic species need to be accurately modeled to predict the tritium inventory of next generation fusion devices, which is needed both for tritium fuel recovery and for tritium safety concerns. In this paper, experiments on thermal desorption of deuterium from intrinsic polycrystalline tungsten defects using laser heating are compared to TMAP-7 modeling. The samples during deuterium plasma exposure were at a temperature of 373K for this benchmark study with ion fluence of 0.7–1.0 ×1024Dm−2. Following plasma exposure, a fiber laser (λ= 1100nm heated the samples to peak surface temperatures ranging from ∼500 to 1400K with pulse widths from 10ms to 1s, and 1 to 10 pulses applied to each sample. The remaining deuterium retention was measured using temperature programmed desorption (TPD. Results show that > 95% of deuterium is desorbed when the peak surface temperature reached ∼950K for > 1s. TMAP-7 is used to predict deuterium desorption from tungsten for a range of surface temperatures and heating durations, and is compared to previous work on desorption from beryllium codeposits.

Preparation of titanium-tritium suspension for use in fluorescent tube starters

International Nuclear Information System (INIS)

Sastry, K.V.S.; Kulkarni, B.A.

1997-01-01

Titanium-tritium suspension is useful in the starter glow bulbs of fluorescent tube lights for initial ionisation. This paper describes the method of preparation and conditions for better yield. (author). 1 tab

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

Deuterium- and tritium-labelled compounds. Applications in the life sciences

International Nuclear Information System (INIS)

Atzrodt, Jens; Derdau, Volker; Kerr, William J.; Reid, Marc

2018-01-01

Hydrogen isotopes are unique tools for identifying and understanding biological and chemical processes. Hydrogen isotope labelling allows for the traceless and direct incorporation of an additional mass or radioactive tag into an organic molecule with almost no changes in its chemical structure, physical properties, or biological activity. Using deuterium-labelled isotopologues to study the unique mass-spectrometric patterns generated from mixtures of biologically relevant molecules drastically simplifies analysis. Such methods are now providing unprecedented levels of insight in a wide and continuously growing range of applications in the life sciences and beyond. Tritium ( 3 H), in particular, has seen an increase in utilization, especially in pharmaceutical drug discovery. The efforts and costs associated with the synthesis of labelled compounds are more than compensated for by the enhanced molecular sensitivity during analysis and the high reliability of the data obtained. In this review, advances in the application of hydrogen isotopes in the life sciences are described. (copyright 2018 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

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

Numerical simulation by a random particle method of Deuterium-Tritium fusion reactions in a plasma*

Directory of Open Access Journals (Sweden)

Charles Fréderique

2013-01-01

Full Text Available We propose and we justify a Monte-Carlo algorithm which solves a spatially homogeneous kinetic equation of Boltzmann type that models the fusion reaction between a deuterium ion and a tritium ion, and giving an α particle and a neutron. The proposed algorithm is validated with the use of explicit solutions of the kinetic model obtained by replacing the fusion cross-section by a Maxwellian cross section. On propose et on justifie un algorithme de type Monte-Carlo permettant de résoudre un modèle cinétique homogène en espace de type Boltzmann modélisant la réaction de fusion entre un ion deutérium et un ion tritium, et donnant une particule α et un neutron. L’algorithme proposé est par ailleurs validé via des solutions explicites du modèle cinétique obtenues en remplaçant la section efficace de fusion par une section efficace maxwellienne.

Peculiarities of using mixed deuterium and tritium ion beams of complicated atomic-molecular composition for fast neutron generation

International Nuclear Information System (INIS)

Kir'yanov, G.I.; Syromukov, S.V.

1983-01-01

The neutron yield is calculated depending on deuterium and tritium beam parameters as well as on the target parameters. Cases of target presaturation with hydrogen nuclides and of target stuffing with the ion beam in the process of the system functioning are discussed. It is shown that the neutron yield is approximately three times more in the case with a pure beam compared to the case with a niked beam

Development of tritium technology at the Tritium Systems Test Assembly

International Nuclear Information System (INIS)

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

1982-01-01

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

Fusion Neutronic Source deuterium endash tritium neutron spectrum measurements using natural diamond detectors

International Nuclear Information System (INIS)

Krasilnikov, A.V.; Kaneko, J.; Isobe, M.; Maekawa, F.; Nishitani, T.

1997-01-01

Two natural diamond detectors (NDDs) operating at room temperature were used for Fusion Neutronics Source (FNS) deuterium endash tritium (DT) neutron spectra measurements at different points around the tritium target and for different deuteron beam energies. Energy resolution of both NDDs were measured, with values 1.95% and 2.8%. Due to the higher energy resolution of one of the two NDDs studied it was possible to measure the shape of the DT neutron energy distribution and its broadening due to deuteron scattering inside the target. The influence of pulse pileup on the energy resolution of the combined system (NDD+electronics) at count rates up to 3.8x10 5 counts/s was investigated. A 3.58% energy resolution for the spectrometric system based on NDD and a 0.25 μs shaping time amplifier has been measured at a count rate of 5.7x10 5 counts/s. It is shown that special development of a fast pulse signal processor is necessary for NDD based spectrometry at count rates of approximately 10 6 counts/s. copyright 1997 American Institute of Physics

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

Tritium proof-of-principle pellet injector results

International Nuclear Information System (INIS)

Fisher, P.W.; Fehling, D.T.; Gouge, M.J.; Milora, S.L.

1989-01-01

The tritium proof-of-principle (TPOP) experiment was built by Oak Ridge National Laboratory (ORNL) to demonstrate the feasibility of forming solid tritium pellets and accelerating them to high velocities for fueling future fusion reactors. TPOP used a pneumatic pipe-gun with a 4-mm-i.d. by 1-m-long barrel. Nearly 1500 pellets were fired by the gun during the course of the experiment; about a third of these were tritium or mixtures of deuterium and tritium. The system also contained a cryogenic 3 He separator that reduced the 3 He level to <0.005%. Pure tritium pellets were accelerated to 1400 m/s. Experiments evaluated the effect of cryostat temperature and fill pressure on pellet size, the production of pellets from mixtures of tritium and deuterium, and the effect of aging on pellet integrity. The tritium phase of these experiments was performed at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. About 100 kCi of tritium was processed through the apparatus without incident. 8 refs., 7 figs

In-Situ Imaging and Quantification of Tritium Surface Contamination via Coherent Fiber Bundle

International Nuclear Information System (INIS)

Gentile, Charles A.; Parker, John J.; Zweben, Stewart J.

2001-01-01

Princeton Plasma Physics Laboratory (PPPL) has developed a method of imaging tritium on in-situ surfaces for the purpose of real-time data collection. This method expands upon a previous tritium imaging concept, also developed at PPPL. Enhancements include an objective lens coupled to the entry aperture of a coherent fiber optic (CFO) bundle, and a relay lens connecting the exit aperture of the fiber bundle to an intensifier tube and a charge-coupled device (CCD) camera. The system has been specifically fabricated for use in determining tritium concentrations on first wall materials. One potential complication associated with the development of D-T [deuterium-tritium] fueled fusion reactors is the deposition of tritium (i.e., co-deposited layer) on the surface of the primary wall of the vacuum vessel. It would be advantageous to implement a process to accurately determine tritium distribution on these inner surfaces. This fiber optic imaging device provides a highly practical method for determining the location, concentration, and activity of surface tritium deposition. In addition, it can be employed for detection of tritium ''hot-spots'' and ''hide-out'' regions present on the surfaces being imaged

Deuterium- and tritium-labelled compounds. Applications in the life sciences

Energy Technology Data Exchange (ETDEWEB)

Atzrodt, Jens; Derdau, Volker [Isotope Chemistry and Metabolite Synthesis, Integrated Drug Discovery, Medicinal Chemistry, Frankfurt (Germany); Kerr, William J.; Reid, Marc [Department of Pure and Applied Chemistry, WestCHEM, University of Strathclyde, Glasgow (United Kingdom)

2018-02-12

Hydrogen isotopes are unique tools for identifying and understanding biological and chemical processes. Hydrogen isotope labelling allows for the traceless and direct incorporation of an additional mass or radioactive tag into an organic molecule with almost no changes in its chemical structure, physical properties, or biological activity. Using deuterium-labelled isotopologues to study the unique mass-spectrometric patterns generated from mixtures of biologically relevant molecules drastically simplifies analysis. Such methods are now providing unprecedented levels of insight in a wide and continuously growing range of applications in the life sciences and beyond. Tritium ({sup 3}H), in particular, has seen an increase in utilization, especially in pharmaceutical drug discovery. The efforts and costs associated with the synthesis of labelled compounds are more than compensated for by the enhanced molecular sensitivity during analysis and the high reliability of the data obtained. In this review, advances in the application of hydrogen isotopes in the life sciences are described. (copyright 2018 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

The First Decommissioning of a Fusion Reactor Fueled by Deuterium-Tritium

International Nuclear Information System (INIS)

Gentile, Charles A.; Perry, Erik; Rule, Keith; Williams, Michael; Parsells, Robert; Viola, Michael; Chrzanowski, James

2003-01-01

The Tokamak Fusion Test Reactor (TFTR) at the Plasma Physics Laboratory of Princeton University (PPPL) was the first fusion reactor fueled by a mixture of deuterium and tritium (D-T) to be decommissioned in the world. The decommissioning was performed over a period of three years and was completed safely, on schedule, and under budget. Provided is an overview of the project and detail of various factors which led to the success of the project. Discussion will cover management of the project, engineering planning before the project started and during the field work as it was being performed, training of workers in the field, the novel adaptation of tools from other industry, and the development of an innovative process for the use of diamond wire to segment the activated/contaminated vacuum vessel. The success of the TFTR decommissioning provides a viable model for the decommissioning of D-T burning fusion devices in the future

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%

Neutron generator tube ion source control

International Nuclear Information System (INIS)

Bridges, J.R.

1982-01-01

A system is claimed for controlling the output of a neutron generator tube of the deuterium-tritium accelerator type and having an ion source to produce sharply defined pulses of neutrons for well logging use. It comprises: means for inputting a relatively low voltage input control pulse having a leading edge and a trailing edge; means, responsive to the input control pulse, for producing a relatively high voltage ion source voltage pulse after receipt of the input pulse; and means, responsive to the input control pulse, for quenching, after receipt of the input pulse, the ion source control pulse, thereby providing a sharply time defined neutron output from the generator tube

Process for the extraction of tritium from heavy water

International Nuclear Information System (INIS)

Dombra, A.H.

1984-01-01

The object of the invention is achieved by a process for the extraction of tritium from a liquid heavy water stream comprising: contacting the heavy water with a countercurrent gaseous deuterium stream in a column packed with a water-repellent catalyst such that tritium is transferred by isotopic exchange from the liquid heavy water stream to the gaseous deuterium stream

Results of neutron irradiation of liquid lithium saturated with deuterium

International Nuclear Information System (INIS)

Tazhibayeva, Irina; Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna; Lyublinski, Igor; Vertkov, Alexey; Mazzitelli, Giuseppe

2017-01-01

Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(−144/RT). • The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10 −13 cm −2 s −1 . The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(-144/RT). The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

Tritium in metals: Techniques of preparation

International Nuclear Information System (INIS)

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

1982-08-01

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

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.

Theory of the energy development in a thermonuclear plasma of deuterium and of deuterium-tritium

Energy Technology Data Exchange (ETDEWEB)

Magnac-Valette, D; Lacombe, E; Cuer, P [Particle Physics Laboratory, Strasbourg (France)

1958-07-01

We have studied the evolution of a thermonuclear mixture for concentrations of 10{sup 18} and 10{sup 17} ions per cm{sup 3} and for temperatures of (116 and 1160) x 10{sup 6}K, taking into account the contribution from secondary reactions. It is assumed that no deuterium or tritium are supplied during the evolution time. The temperature is assumed constant and the pinching of the charged species perfect. Neutrons leave the mixture without causing any secondary reactions because of their long mean free path. Integration of the differential equations describing the evolution of the mixture yields the total nuclear power in the plasma, and the power carried off by the neutrons. The calculations were made with the aid of an electronic computer. The initial concentration is unimportant since a scaling of the concentrations changes nothing if the inverse scaling is carried out on the time variable. It is assumed, moreover, that the thermalizing time is negligible in comparison with the mean life of the particles. The calculations were performed using a successive approximation method in which the mesh size was selected such that the error was always less than 10{sup -6}. The results are presented in the paper.

Environmental monitoring for tritium separation facility

International Nuclear Information System (INIS)

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

2001-01-01

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

Shock timing measurements and analysis in deuterium-tritium-ice layered capsule implosions on NIF

Energy Technology Data Exchange (ETDEWEB)

Robey, H. F.; Celliers, P. M.; Moody, J. D.; Sater, J.; Parham, T.; Kozioziemski, B.; Dylla-Spears, R.; Ross, J. S.; LePape, S.; Ralph, J. E.; Dewald, E. L.; Berzak Hopkins, L.; Kroll, J. J.; Yoxall, B. E.; Hamza, A. V.; Landen, O. L.; Edwards, M. J. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Hohenberger, M.; Boehly, T. R. [Laboratory for Laser Energetics, Rochester, New York 14623 (United States); Nikroo, A. [General Atomics, San Diego, California 92196 (United States)

2014-02-15

Recent advances in shock timing experiments and analysis techniques now enable shock measurements to be performed in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility (NIF). Previous measurements of shock timing in inertial confinement fusion implosions [Boehly et al., Phys. Rev. Lett. 106, 195005 (2011); Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] were performed in surrogate targets, where the solid DT ice shell and central DT gas were replaced with a continuous liquid deuterium (D2) fill. These previous experiments pose two surrogacy issues: a material surrogacy due to the difference of species (D2 vs. DT) and densities of the materials used and a geometric surrogacy due to presence of an additional interface (ice/gas) previously absent in the liquid-filled targets. This report presents experimental data and a new analysis method for validating the assumptions underlying this surrogate technique. Comparison of the data with simulation shows good agreement for the timing of the first three shocks, but reveals a considerable discrepancy in the timing of the 4th shock in DT ice layered implosions. Electron preheat is examined as a potential cause of the observed discrepancy in the 4th shock timing.

Shock timing measurements and analysis in deuterium-tritium-ice layered capsule implosions on NIF

Science.gov (United States)

Robey, H. F.; Celliers, P. M.; Moody, J. D.; Sater, J.; Parham, T.; Kozioziemski, B.; Dylla-Spears, R.; Ross, J. S.; LePape, S.; Ralph, J. E.; Hohenberger, M.; Dewald, E. L.; Berzak Hopkins, L.; Kroll, J. J.; Yoxall, B. E.; Hamza, A. V.; Boehly, T. R.; Nikroo, A.; Landen, O. L.; Edwards, M. J.

2014-02-01

Recent advances in shock timing experiments and analysis techniques now enable shock measurements to be performed in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility (NIF). Previous measurements of shock timing in inertial confinement fusion implosions [Boehly et al., Phys. Rev. Lett. 106, 195005 (2011); Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] were performed in surrogate targets, where the solid DT ice shell and central DT gas were replaced with a continuous liquid deuterium (D2) fill. These previous experiments pose two surrogacy issues: a material surrogacy due to the difference of species (D2 vs. DT) and densities of the materials used and a geometric surrogacy due to presence of an additional interface (ice/gas) previously absent in the liquid-filled targets. This report presents experimental data and a new analysis method for validating the assumptions underlying this surrogate technique. Comparison of the data with simulation shows good agreement for the timing of the first three shocks, but reveals a considerable discrepancy in the timing of the 4th shock in DT ice layered implosions. Electron preheat is examined as a potential cause of the observed discrepancy in the 4th shock timing.

Shock timing measurements and analysis in deuterium-tritium-ice layered capsule implosions on NIF

International Nuclear Information System (INIS)

Robey, H. F.; Celliers, P. M.; Moody, J. D.; Sater, J.; Parham, T.; Kozioziemski, B.; Dylla-Spears, R.; Ross, J. S.; LePape, S.; Ralph, J. E.; Dewald, E. L.; Berzak Hopkins, L.; Kroll, J. J.; Yoxall, B. E.; Hamza, A. V.; Landen, O. L.; Edwards, M. J.; Hohenberger, M.; Boehly, T. R.; Nikroo, A.

2014-01-01

Recent advances in shock timing experiments and analysis techniques now enable shock measurements to be performed in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility (NIF). Previous measurements of shock timing in inertial confinement fusion implosions [Boehly et al., Phys. Rev. Lett. 106, 195005 (2011); Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] were performed in surrogate targets, where the solid DT ice shell and central DT gas were replaced with a continuous liquid deuterium (D2) fill. These previous experiments pose two surrogacy issues: a material surrogacy due to the difference of species (D2 vs. DT) and densities of the materials used and a geometric surrogacy due to presence of an additional interface (ice/gas) previously absent in the liquid-filled targets. This report presents experimental data and a new analysis method for validating the assumptions underlying this surrogate technique. Comparison of the data with simulation shows good agreement for the timing of the first three shocks, but reveals a considerable discrepancy in the timing of the 4th shock in DT ice layered implosions. Electron preheat is examined as a potential cause of the observed discrepancy in the 4th shock timing

Results of neutron irradiation of liquid lithium saturated with deuterium

Energy Technology Data Exchange (ETDEWEB)

Tazhibayeva, Irina, E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Lyublinski, Igor [JSC «Red Star», Moscow (Russian Federation); NRNU «MEPhI», Moscow (Russian Federation); Vertkov, Alexey [JSC «Red Star», Moscow (Russian Federation); Mazzitelli, Giuseppe [ENEA, RC Frascati, Frascati (Italy)

2017-04-15

Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(−144/RT). • The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10{sup −13} cm{sup −2} s{sup −1}. The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(-144/RT). The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

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

Catalyzed deuterium fueled tokamak reactors

International Nuclear Information System (INIS)

Southworth, F.H.

1977-01-01

Catalyzed deuterium fuel presents several advantages relative to D-T. These are, freedom from tritium breeding, high charged particle power fraction and lowered neutron energy deposition in the blanket. Higher temperature operation, lower power densities and increased confinement are simultaneously required. However, the present study has developed designs which have capitalized upon the advantages of catalyzed deuterium to overcome the difficulties associated with the fuel while obtaining high efficiency

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)

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

Influence of deuterium content on tensile behavior of Zr-2.5Nb pressure tube material in the temperature range of ambient to 300 degC

International Nuclear Information System (INIS)

Bind, A.K.; Singh, R.N.; Chakravartty, J.K.; Dhandharia, Priyesh; Ghosh, Agnish; More, Nitin S.; Chhatre, A.G.; Vijayakumar, S.

2011-08-01

Tensile properties of autoclaved zirconium-2.5 wt. % niobium pressure tube material were evaluated by uniaxial tension tests at temperatures between 25 and 300 degC and under strain-rates of 1.075 x 10 -4 /s. Six number of Zr-2.5Nb alloy pressure tube spools of length 130 mm were obtained from pressure tube number 19-2557-2. Five spools were polished with abrasive paper to remove the oxide layer. These spools were gaseously charged with controlled amount of deuterium. The target deuterium concentrations were 25, 50, 75, 100 and 200 wppm of hydrogen equivalent. Ten samples were machined by EDM wire cutting from every spool. The tensile specimen axis was oriented along longitudinal direction of the tube. Metallographic examination of the deuterium charged samples suggested that the deuterides were predominantly circumferential deuterides. Analysis of tensile results showed that both yield and ultimate tensile strengths of this alloy decreased monotonically with increasing test temperatures. The tensile ductility decreased marginally with increase in test temperature from ambient to 300 degC. It was also observed that both strength and ductility appear to be unaffected by deuterium content at all temperatures, thereby suggesting that at least up to 200 wppm (Heq.) of deuterium tensile properties are not influenced by deuterium. (author)

Long Term Tritium Trapping in TFTR and JET

International Nuclear Information System (INIS)

Skinner, C.H.; Gentile, C.A.; Young, K.M.; Coad, J.P.; Hogan, J.T.; Penzhorn, R.-D.; Bekris, N.

2001-01-01

Tritium retention in TFTR [Tokamak Fusion Test Reactor] and JET [Joint European Torus] shows striking similarities and contrasts. In TFTR, 5 g of tritium were injected into circular plasmas over a 3.5 year period, mostly by neutral-beam injection. In JET, 35 g were injected into divertor plasmas over a 6 month campaign, mostly by gas puffing. In TFTR, the bumper limiter provided a large source of eroded carbon and a major part of tritium was co-deposited on the limiter and vessel wall. Only a small area of the co-deposit flaked off. In JET, the wall is a net erosion area, and co-deposition occurs principally in shadowed parts of the inner divertor, with heavy flaking. In both machines, the initial tritium retention, after a change from deuterium [D] to tritium [T] gas puffing, is high and is due to isotope exchange with deuterium on plasma-facing surfaces (dynamic inventory). The contribution of co-deposition is lower but cumulative, and is revealed by including periods of D fueling that reversed the T/D isotope exchange. Ion beam analysis of flakes from TFTR showed an atomic D/C ratio of 0.13 on the plasma facing surface, 0.25 on the back surface and 0.11 in the bulk. Data from a JET divertor tile showed a larger D/C ratio with 46% C, 30% D, 20% H and 4% O. Deuterium, tritium, and beryllium profiles have been measured and show a thin less than 50 micron co-deposited layer. Flakes retrieved from the JET vacuum vessel exhibited a high tritium release rate of 2e10 Bq/month/g. BBQ modeling of the effect of lithium on retention in TFTR showed overlapping lithium and tritium implantation and a 1.3x increase in local T retention

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

International Nuclear Information System (INIS)

Asakura, Yamato; Sugiyama, Takahiko; Kawano, Takao; Uda, Tatsuhiko; Tanaka, Masahiro; Tsuji, Naruhito; Katahira, Koji; Iwahara, Hiroyasu

2004-01-01

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

Influence of aging time on residual tritium in Pd beds

International Nuclear Information System (INIS)

Yang Jinshui; Zhang Zhi; Su Yongjun; Jing Wenyong; Du Jie

2012-01-01

The amount of tritium in Pd beds. which were initially loaded at room temperature with tritium at the atomic ratio of T/Pd≈0.65 and aged 1.66 years, 3.47 years and 5.94 years, respectively, was investigated by methods of deuterium exchange, thermal desorption and aqua regia dissolution. Obtained results show that after deuterium exchange and thermal desorption, about 99% of tritium is desorbed from Pd tritide. and the amount of residual tritium become in- creasing significantly as the aged time is increased. which is 3.99 × 10 -7 gT/gPd, 4. 97 × 10 -7 gT/gPd and 1.29 × 10 -6 gT/gPd respectively. The increasing amount of residual tritium could be attributed to the increase of interstitial form of tritium, resulting from increasing interstitial type defects induced by the migration of 3 He atoms in interstitial sites as a function of aged time. (authors)

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

International Nuclear Information System (INIS)

Alvani, C.; Casadio, S.; Mancini, M.R.; Nannetti, C.A.; Avon, J.; Ravel, S.; Pruzzo, G.; Terrosi; Roux, N.; Terlain, A.; and others.

1994-01-01

An irradiation experiment (90 FPD in SILOE reactor) has been designed in order to evaluate the in-situ effect of red-ox power of sweeping gas (helium with 100 vpm of H 2 /H 2 O with relative concentrations varying from pure H 2 to pure H 2 O ) on tritium removal from LiAlO 2 and Li 2 ZrO 3 ; and tritium permeation through AlSl-316L SS tubes with bare and coated surfaces. The conditions and materials explored were selected in order to test possible improvements with respect to critical issues for the 'Breeder Inside Tube' (BIT) blanket concept development. (author) 6 refs.; 4 figs.; 2 tabs

Cross sections for deuterium, tritium, and helium production in p(bar sign)p collisions at √( s )=1.8 TeV

International Nuclear Information System (INIS)

Alexopoulos, T.; Anderson, E. W.; Biswas, N. N.; Bujak, A.; Carmony, D. D.; Erwin, A. R.; Findeisen, C.; Goshaw, A. T.; Gulbrandsen, K.; Gutay, L. J.

2000-01-01

We present the results of a search for the production of light elements in p(bar sign)p collisions at the Fermilab Tevatron collider. Momentum, time of flight, and dE/dx measurements are used to distinguish nuclei from elementary particles. A production ratio for deuterium to hydrogen is calculated and compared to the primordial value of the big bang model. Some evidence for tritium is found and none for helium isotopes. (c) 2000 The American Physical Society

Development of a tritium recovery system from CANDU tritium removal facility

International Nuclear Information System (INIS)

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

2015-01-01

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

Development of a tritium recovery system from CANDU tritium removal facility

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

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

Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

International Nuclear Information System (INIS)

Sayi, Y.S.; Shankaran, P.S.; Yadav, C.S.; Ramanjaneyulu, P.S.; Venugopal, V.; Ramakumar, K.L.; Chhapru, G.C.; Prasad, R.; Jain, H.C.; Sood, D.D.

2009-02-01

Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

Development status of the cryogenic distillation system in Cernavoda Tritium Removal Facility

International Nuclear Information System (INIS)

Draghia, Mirela; Ana, George; Pasca, Gheorghe; Porcariu, Florina

2009-01-01

Full text: The reference design technology for the heavy water detritiation plant of Cernavoda CANDU station is based on combination of Liquid Phase Catalytic Exchange (LPCE) and Cryogenic Distillation (CD) processes. Based on this technology, tritium is transferred from the heavy water to a deuterium stream in the catalyzed isotopic exchange process, LPCE, followed by a final enrichment within the cryogenic distillation cascade. The final step is the tritium storage on metallic hydride. The basic function of the Cryogenic Distillation System (CDS) is the separation of tritium from the tritiated deuterium coming from the LPCE column in the following conditions: - the final product has to be tritium with a concentration of at least 99%; - it must be provided a detritiation factor of at least 100 (the ration between the tritium concentration in the deuterium stream fed to the CD system and the tritium concentration in the returned stream to the LPCE); - the deuterium must be enriched up to 99.995%, by removing the protium; - provisions for safe discharge of the entire inventory of the CD cascade into buffer vessels shall be implemented. To summarize, the present status of the project consists of technical documentation for all the components of CDS, including the P and ID (Pipping and Instrumentation Diagram), preliminary data sheets, technical specifications, drawings for the major components as the buffer vessels, coldbox, etc, and 3D models as well for almost all the components. (authors)

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

Design of a deuterium and tritium-ablator shock ignition target for the National Ignition Facility

International Nuclear Information System (INIS)

Terry, Matthew R.; Perkins, L. John; Sepke, Scott M.

2012-01-01

Shock ignition presents a viable path to ignition and high gain on the National Ignition Facility (NIF). In this paper, we describe the development of the 1D design of 0.5 MJ class, all-deuterium and tritium (fuel and ablator) shock ignition target that should be reasonably robust to Rayleigh-Taylor fluid instabilities, mistiming, and hot electron preheat. The target assumes “day one” NIF hardware and produces a yield of 31 MJ with reasonable allowances for laser backscatter, absorption efficiency, and polar drive power variation. The energetics of polar drive laser absorption require a beam configuration with half of the NIF quads dedicated to launching the ignitor shock, while the remaining quads drive the target compression. Hydrodynamic scaling of the target suggests that gains of 75 and yields 70 MJ may be possible.

Simulation for Synthesis of Tritiated Styrene by Catalyzed Addition of Deuterium

Directory of Open Access Journals (Sweden)

CUI Xiao-jing;KANG Yi;HU Shi-lin

2016-08-01

Full Text Available Tritiated styrene plays an important role in the organic tritium lights, which could be made by selective hydrogenation of phenylacetylene. A simulated reaction of the preparation of tritiated styrene was studied by using deuterium instead of tritium and using the Lindlar catalyst instead of Pd/C catalyst to improve the conversion and selectivity of the reaction. Experiment results showed that stirring speed, temperature and the amount of deuterium were the most important factors to effect the conversion and selectivity of the reaction. The relative stronger stirring speed and higher temperature could improve the conversion rate of the reaction, but could not change the selectivity. When the excessive or less deuterium was added in the reaction, the selectivity was decreased significantly, since the over deuterium promoted the reaction toward ethylbenzene. Lindlar catalyst exhibited higher selectivity toward styrene than Pd/C.

Deuterium and tritium profile through the Vatnajoekull icecap

International Nuclear Information System (INIS)

Arnasson, B.

1974-06-01

During the years 1971-73 a research project was carried out at the Science Institute of the University of Iceland with the financial support of the IAEA. A rotary drill for deep coring in temperate ice has been developed, and a 415 m deep hole was drilled during the summer of 1972 into the temperate accumulation area of the Vatnajoekull glacier, Iceland, at an altitude of 1800 m a.s.l. The core recovery was 99%. The bottom, at 480-500 m depth, was not reached because of a fault in the cable. Detail on the drill is given in: Arnason, Bjoernsson and Theodorsson, J. of Glaciology, 13, 133 (1974). Several volcanic ash layers have been recognized in the ice core and associated with historically known eruptions. This provides the age-depth relationship. Isotopic analyses along the core show that precipitation in the period 1931-1960 is enriched by 5 per mil in deuterium with respect to that in the period 1891-1920, in agreement with the 1degC increase in the mean air temperature occurred from the beginning of the century. Tritium analyses show appreciable isotopic exchange due to water percolation (summer rains and ice melting). The Na + and Cl - content decreases with depth (from 1 μg/g ice to 0.1 μg/g ice). This decrease has been attributed to dissolution of ions in water percolating along the ice crystals. Other studies of the ice core, in progress or planned, include: size and orientation of ice crystals, size and pressure of air bubbles, measurement of F - , SO 4 -- and Hg

Comparison of the recently proposed super-Marx generator approach to thermonuclear ignition with the deuterium-tritium laser fusion-fission hybrid concept by the Lawrence Livermore National Laboratory

International Nuclear Information System (INIS)

Winterberg, F.

2009-01-01

The recently proposed super-Marx generator pure deuterium microdetonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser deuterium-tritium fusion-fission hybrid concept (LIFE). In a super-Marx generator, a large number of ordinary Marx generators charge up a much larger second stage ultrahigh voltage Marx generator from which for the ignition of a pure deuterium microexplosion an intense GeV ion beam can be extracted. Typical examples of the LIFE concept are a fusion gain of 30 and a fission gain of 10, making up a total gain of 300, with about ten times more energy released into fission as compared to fusion. This means the substantial release of fission products, as in fissionless pure fission reactors. In the super-Marx approach for the ignition of pure deuterium microdetonation, a gain of the same magnitude can, in theory, be reached. If feasible, the super-Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of thermonuclear microexplosions

Predicting the Equilibrium Deuterium-Tritium Fuel Layer Thickness Profile in an Indirect-Drive Hohlraum Capsule

International Nuclear Information System (INIS)

Sanchez, Jorge J.; Giedt, Warren H.

2004-01-01

A numerical procedure for calculating the equilibrium thickness distribution of a thin layer of deuterium and tritium on the inner surface of an indirect drive target sphere (∼2.0 mm in diameter) is described. Starting with an assumed uniform thickness layer and with specified thermal boundary conditions, the temperature distribution throughout the capsule and hohlraum (including natural convection in the hohlraum gas) is calculated. Results are used to make a first estimate of the final non-uniform thickness distribution of the layer. This thickness distribution is then used to make a second calculation of the temperature distribution with the same boundary conditions. Legendre polynomial coefficients are evaluated for the two temperature distributions and the two thickness profiles. Final equilibrium Legendre coefficients are determined by linear extrapolation. From these coefficients, the equilibrium layer thickness can be computed

Reduction of weighing errors caused by tritium decay heating

International Nuclear Information System (INIS)

Shaw, J.F.

1978-01-01

The deuterium-tritium source gas mixture for laser targets is formulated by weight. Experiments show that the maximum weighing error caused by tritium decay heating is 0.2% for a 104-cm 3 mix vessel. Air cooling the vessel reduces the weighing error by 90%

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

Distillation plant for tritium enrichment in metallic lithium

International Nuclear Information System (INIS)

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

1984-06-01

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

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

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

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

Tritium permeation barriers in contact with liquid lithium-lead eutectic (Pb-17Li)

International Nuclear Information System (INIS)

Forcey, K.S.; Perujo, A.

1995-01-01

The permeation of deuterium through coated stainless steel tubes containing liquid lithium-lead eutectic (Pb-17Li) has been studied and compared to measurements through tubes without the lithium compound. The measurements form part of an investigation into the effect of a potential tritium breeder material on permeation barriers for fusion reactors. The coatings studied were CVD TiC and Al 2 O 3 and a pack aluminised layer. Without the lithium-lead, the CVD coatings reduced the permeation rate up to 1 order of magnitude, and the aluminised layer up to 2 orders of magnitude. A CVD layer was unaffected by Pb-17Li whilst in the case of the aluminised tube, the lithium-lead completely removed the permeation barrier, presumably by attacking the surface oxide. Furthermore, the aluminised sample presented a large number of cracks and poor adheren ce to the substrate. ((orig.))

Tritium Issues in Next Step Devices

Energy Technology Data Exchange (ETDEWEB)

C.H. Skinner; G. Federici

2001-09-05

Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

Tritium Issues in Next Step Devices

International Nuclear Information System (INIS)

C.H. Skinner; G. Federici

2001-01-01

Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

Overview of light sources powered by tritium

International Nuclear Information System (INIS)

Wu Jian; Lei Jiarong; Liu Wenke

2012-01-01

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

[The Qualitative Analysis of the Amide Derivative of HLDF-6 Peptide and Its Metabolites with the Use of Tritium- and Deuterium-Labeled Derivatives].

Science.gov (United States)

Zolotarev, A; Dadayan, A K; Kost, N V; Voevodina, M E; Sokolov, O Y; Kozik, V S; Shram, S I; Azev, V N; Bocharov, E V; Bogachouk, A P; Lipkin, V M; Myasoedov, N F

2015-01-01

The goal of the study was to elaborate the pharmacokinetics methods of the amide derivative of peptide HLDF-6 (TGENHR-NH2) and its range of nootropic and neuroprotective activity is wide. The hexapeptide 41TGENHR46 is a fragment of the HDLF differentiation factor. It forms the basis for the development of preventive and therapeutic preparations for treating cerebrovascular and neurodegenerative conditions. Pharmacokinetic and molecular mechanisms of the action of the HLDF-6 peptide were studied using tritium- and deuterium-labeled derivatives of this peptide, produced with the use of the high-temperature solid-state catalytic isotope exchange reaction (HSCIE). This reaction was employed to produce the tritium-labeled peptide [3H]TGENHR-NH2 with a molar radioactivity of 230 Ci/mmol and the deuterium-labeled peptide [2H]TGENHR-NH2 with an average deuterium incorporation equal to 10.5 atoms. It was shown by the NMR spectroscopy that the isotope label distribution over the labeled peptide's molecule was uniform, which allowed qualitative analysis ofboth the peptide itself and its fragments in the organism's tissues to be conducted. The newly developed pharmacokinetics method makes it possible to avoid almost completely losses of the peptides under study due to biodegradation during the analysis of tissues. These labeled peptides were used in mice, rats and rabbits to study the pharmacokinetics of the peptide and to calculate the values of its principal pharmacokinetic parameters. Characteristics of its pharmacokinetic profile in the blood were obtained, the hypothesis of pharmacokinetics linearity tested, its metabolism analyzed and its bioavailability value, 34%, calculated. It has been shown that the studied TGENHR-NH2 peptide shows high resistance to hydrolysis in the blood plasma, with dipeptidyl aminopeptidases making the largest contribution to its hydrolysis.

Tritium permeation through iron

International Nuclear Information System (INIS)

Hagi, Hideki; Hayashi, Yasunori

1989-01-01

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

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

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

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

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

Process for the production of high purity deuterium

International Nuclear Information System (INIS)

Arrathoon, R.

1977-01-01

A process for the electrolysis of heavy water which results in the production of high purity deuterium without periodic replenishment of the electrolyte with additional deuterated compounds is defined. Electrolysis is effected through the use of an inexpensive cation-action permselective membrane which is essentially a solid polymer electrolyte and which is capable of automatically separating the evolved deuterium and oxygen gas. This cation-active permselective membrane does not introduce any intrinsic impurities or tritium contamination in the generated deuterium gas, does not require periodic revitalization with deuterated compounds or other chemical compounds, and is characterized by an unusually high electrical efficiency

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)

Deuterium, oxygen-18 and tritium in precipitation, surface and groundwater in the far east of Russia

Energy Technology Data Exchange (ETDEWEB)

Chelnokov, George; Kharitonova, Natalia; Bragin, Ivan; Vasil' eva, Maria [Far East Geological Insitute Rus. Acad. of Sci., 690022, Prospect 100 letya 159, Vladivostok (Russian Federation)

2013-07-01

This is the first report describing the parallel measurement of deuterium (δD), tritium ({sup 3}H), and oxygen-18 (δ{sup 18}O) in precipitation, seawater, surface and groundwater in relation to the Russian Far East. dD and δ{sup 18}O demonstrate that the studied waters have a meteoric origin, and variations are the result of water-rock-gas interactions. All studied waters reveal obvious 'latitudinal' and 'continental' effects: there is a universal decrease in δ{sup 18}O and δD from the south to the north, and from the ocean inland. The background level of {sup 3}H is 20 TU in Amursky region's rivers, 13 TU in Primorsky region's rivers, and 5.5 TU in one of the Kuril Islands. The majority of studied groundwaters have short residence times. (authors)

Separation of tritium from gaseous and aqueous effluent systems

International Nuclear Information System (INIS)

Kobisk, E.H.

1977-01-01

Removal or reduction of tritium content in a wide variety of effluent streams has been extensively studied in the United States. This paper specifically reviews three processes involving tritium separation in the gaseous phase and the aqueous phase. Diffusion through a selective Pd-25Ag alloy membrane at temperatures up to 600 0 C and at pressures up to 700 kg/cm 2 has resulted in successful separation of hydrogen-deuterium mixtures with an associated separation factor of 1.65 (and gives a calculated separation factor for hydrogen-tritium mixtures of 2.0). Use of a single palladium bipolar membrane in an electrolysis system has been found to yield a hydrogen-deuterium separation factor of 4 and a hydrogen-tritium factor of 6 to 11 without the production of gaseous hydrogen. Finally, countercurrent catalytic exchange between tritium-containing hydrogen gas and water has yielded a separation factor of 6.3. The specific advantages of each of these systems will be discussed in terms of their potential applications. In all cases, further investigations are necessary to scale the systems to handle large quantities of feed material in a continuous mode and to minimize energy requirements. Such separative systems must necessarily be cascaded to yield gaseous or aqueous product streams suitable for recycling to the tritium producing systems, for storage or for discharge to the environment. (orig./HP) [de

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

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

The deuterium inventory in ASDEX Upgrade

International Nuclear Information System (INIS)

Mayer, M.; Rohde, V.; Ramos, G; Vainonen-Ahlgren, E.; Likonen, J.; Herrmann, A.; Neu, R.

2007-01-01

The deuterium inventory in ASDEX Upgrade was determined by quantitative ion beam analysis techniques and SIMS for different discharge campaigns between the years 2002 and 2005. ASDEX Upgrade was a carbon dominated machine during this phase. Full poloidal sections of the lower and upper divertor tile surfaces, limiter tiles, gaps between divertor tiles, gaps between inner heat shield tiles and samples from remote areas below the roof baffle and in pump ducts were analysed, thus offering an exhaustive survey of all relevant areas in ASDEX Upgrade. Deuterium is mainly trapped on plasma-exposed surfaces of inner divertor tiles, where about 70% of the retained deuterium inventory is found. About 20% of the inventory is retained at or below the divertor roof baffle, and about 10% is observed in other areas, such as the outer divertor and in gaps between tiles. The long term deuterium retention is 3-4% of the total deuterium input. The obtained results are compared with gas balance measurements, and conclusions about tritium retention in ITER are made

Mechanism and deuterium pickup in Zr-2.5Nb alloy

International Nuclear Information System (INIS)

Ploc, R.A.

1999-12-01

There are approximately 400 Zr-2.5Nb pressure tubes in a CANDU reactor. During operation, the pressure tubes contain heavy water at about 300 deg C, 10.3 NPa with a room-temperature pD of 10.5. Operation of the pressure tube in the environment leads to oxide formation and absorption of deuterium. Excess deuterium absorption leads to precipitation of zirconium deuterides in the metal. A knowledge of how the deuterium passes through the oxide film to enter into the metal is an important step in gaining control over ingress rates. Fresnel fringe imaging of cross-sectioned oxides grown on pressure tubes, combined with tilting in the electron microscope, has revealed the three-dimensional nature of porosity in the oxide films. Two primary types exist, flake and ribbon. The main route for deuterium ingress is via ribbon porosity, as shown by electrochemical impedance spectroscopy. The location of the ribbon porosity is along the boundary between the oxidized α-Zr and β-Zr phases. Modifications to reduce ribbon porosity are possible and this, in turn, leads to significantly lower rates of deuterium absorption and extension of pressure-tube lifetime. (author)

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

Cold versus hot fusion deuterium branching ratios

International Nuclear Information System (INIS)

Fox, H.; Bass, R.

1995-01-01

A major source of misunderstanding of the nature of cold nuclear fusion has been the expectation that the deuterium branching ratios occurring within a palladium lattice would be consistent with the gas-plasma branching ratios. This misunderstanding has led to the concept of the dead graduate student, the 1989's feverish but fruitless search for neutron emissions from cold fusion reactors, and the follow-on condemnation of the new science of cold fusion. The experimental facts are that in a properly loaded palladium lattice, the deuterium fusion produces neutrons at little above background, a greatly less-than-expected production of tritium (the tritium desert), and substantially more helium-4 than is observed in hot plasma physics. The experimental evidence is now compelling (800 reports of success from 30 countries) that cold nuclear fusion is a reality, that the branching ratios are unexpected, and that a new science is struggling to be recognized. Commercialization of some types of cold fusion devices has already begun

Preliminary measurements of tritium, deuterium and oxygen-18 in lakes and groundwater of volcanic Rotorua region, New Zealand

International Nuclear Information System (INIS)

Taylor, C.B.; Freestone, H.J.; Nairn, I.A.

1977-06-01

This report presents an initial survey of the isotopic characteristics of non-hydrothermal waters of Rotorua region. Measurements of deuterium, oxygen and tritium have been made on samples collected between 1970 and 1973. Some major springs emerging close to land-locked lakes have been identified as containing mixtures of lake-derived water and precipitation-derived recharge in the catchments between lake and spring. The stable isotope composition of precipitation-recharged groundwater occupies a fairly narrow range, but is nevertheless seperable into two distinct isotopic families. The circumstances leading to these isotopic differences are not yet fully understood. This isotopic data has been gathered with a view to identifying hydrological problems in the region capable of study by isotopic methods and as essential background data to studies of the region's many hydrothermal systems. (auth.)

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

Deuterium behavior in first-wall materials for nuclear fusion

International Nuclear Information System (INIS)

Bernard, E.

2012-01-01

Plasma-wall interactions play an important part while choosing materials for the first wall in future fusion reactors. Moreover, the use of tritium as a fuel will impose safety limits regarding the total amount present in the tokamak. Previous analyses of first-wall samples exposed to fusion plasma highlighted an in-bulk migration of deuterium (as an analog to tritium) in carbon materials. Despite its limited value, this retention is problematic: contrary to co-deposited layers, it seems very unlikely to recover easily the deuterium retained in such a way. Because of the difficult access to in situ samples, most published studies on the subject were carried out using post-mortem sample analysis. In order to access to the dynamic of the phenomenon and come apart potential element redistribution during storage, we set up a bench intended for simultaneous low-energy ion implantation, reproducing the deuterium interaction with first-wall materials, and high-energy micro beam analysis. Nuclear reaction analysis performed at the micrometric scale (μNRA) allows to characterize deuterium repartition profiles in situ. This analysis technique was confirmed to be non-perturbative of the mechanisms studied. We observed on the experimental data set that the material surface (0-1 μm) display a high and nearly constant deuterium content, with a uniform distribution. On the contrary, in-bulk deuterium (1-11 μm) localizes in preferential trapping sites related to the material microstructure. In-bulk deuterium inventory seems to increase with the incident fluence, in spite of the wide data scattering attributed to the structure variation of studied areas. Deuterium saturation at the surface as well as in-depth migration are instantaneous; in-vacuum storage leads to a small deuterium global desorption. Observations made via μNRA were coupled with results from other characterization techniques. X-ray μtomography allowed to identify porosities as the preferential trapping sites

Theoretical studies on the stopping power of deuterium-tritium mixed with uranium plasmas for α particles

International Nuclear Information System (INIS)

Wang, Zhigang; Fu, Zhen-Guo; Zhang, Ping

2014-01-01

The stopping power of a compressed and highly ionized deuterium-tritium (DT) and uranium (U) plasma for α particles at very high temperatures (T = 5 keV) is examined theoretically with the dimensional continuation method. We show that with increasing density of U, both the magnitude and width of the resonance peak in the stopping power (as a function of the α particle energy), increases because of the ions, while the penetration distance of the α particles decreases. A simple relation of decreasing penetration distance as a function of plasma density is observed, which may be useful for inertial confinement fusion experiments. Moreover, by comparing the results with the case of a DT plasma mixed with beryllium, we find that the effect of a higher Z plasma is stronger, with regard to energy loss as well as the penetration distance of α particles, than that of a lower Z plasma

Stability of Global Alfven Waves (Tae, Eae) in Jet Tritium Discharges

NARCIS (Netherlands)

Kerner, W.; Borba, D.; Huysmans, G. T. A.; Porcelli, F.; Poedts, S.; Goedbloed, J. P.; Betti, R.

1994-01-01

The interaction of alpha-particles in JET tritium discharges with global Alfven waves via inverse Landau damping is analysed. It is found that alpha-particle driven eigenmodes were stable in the PTE1 and should also be stable in a future 50:50 deuterium-tritium mix discharge aiming at Q(DT) = 1,

Point design for deuterium-deuterium compact reversed-field pinch reactors

International Nuclear Information System (INIS)

Dabiri, A.E.; Dobrott, D.R.; Gurol, H.; Schnack, D.D.

1984-01-01

A deuterium-deuterium (D-D) reversed-field pinch (RFP) reactor may be made comparable in size and cost to a deuterium-tritium (D-T) reactor at the expense of high-thermal heat load to the first wall. This heat load is the result of the larger percentage of fusion power in charged particles in the D-D reaction as compared to the D-T reaction. The heat load may be reduced by increasing the reactor size and hence the cost. In addition to this ''degraded'' design, the size may be kept small by means of a higher heat load wall, or by means of a toroidal divertor, in which case most of the heat load seen by the wall is in the form of radiation. Point designs are developed for these approaches and cost studies are performed and compared with a D-T reactor. The results indicate that the cost of electricity of a D-D RFP reactor is about20% higher than a D-T RFP reactor. This increased cost could be offset by the inherent safety features of the D-D fuel cycle

Phenomenological nuclear reaction description in deuterium-saturated palladium and synthesized structure in dense deuterium gas under γ-quanta irradiation

International Nuclear Information System (INIS)

Didyk, A.Yu.; Wisniewski, R.

2012-01-01

The observed phenomena on the changes of chemical compositions in our previous reports allowed us to develop a phenomenological nuclear fusion-fission model with taking into consideration the elastic and inelastic scattering of photoprotons and photoneutrons, heating of surrounding deuterium nuclei, following D-D fusion reactions and fission of middle-mass nuclei by 'hot' protons, deuterons and various-energy neutrons. Such chain processes could produce the necessary number of neutrons, 'hot' deuterons for explanation of the observed experimental results. The developed approach can be a basis for creation of deuterated nuclear fission reactors (DNFR) with high-density deuterium gas and so-called deuterated metals. Also, the developed approach can be used for the study of nuclear reactions in high-density deuterium or tritium gases and deuterated metals

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

Measurement of Tritium Surface Distribution on TFTR Bumper Limiter Tiles

International Nuclear Information System (INIS)

Sugiyama, K.; Tanabe, T.; Skinner, C.H.; Gentile, C.A.

2004-01-01

The tritium surface distribution on graphite tiles used in the Tokamak Fusion Test Reactor (TFTR) bumper limiter and exposed to TFTR deuterium-tritium (D-T) discharges from 1993 to 1997 was measured by the Tritium Imaging Plate Technique (TIPT). The TFTR bumper limiter shows both re-/co-deposition and erosion. The tritium images for all tiles measured are strongly correlated with erosion and deposition patterns, and long-term tritium retention was found in the re-/co-depositions and flakes. The CFC tiles located at erosion dominated areas clearly showed their woven structure in their tritium images owing to different erosion yields between fibers and matrix. Significantly high tritium retention was observed on all sides of the erosion tiles, indicating carbon transport via repetition of local erosion/deposition cycles

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

Investigation of tritium in groundwater at Pickering NGS

International Nuclear Information System (INIS)

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

2001-01-01

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

Tritium handling facility at KMS Fusion Inc

International Nuclear Information System (INIS)

Bowman, C.C.; Vis, V.A.

1990-01-01

The tritium facility at KMS Fusion, Inc. supports the inertial confinement fusion research program. The main function of the facility is to fill glass and polymer Microshell (TM) capsules (small fuel containers) to a maximum pressure of 100 atm with tritium (T 2 ) or deuterium--tritium (DT). The recent upgrade of the facility allows us to fill Microshell capsules to a maximum pressure of 200 atm. A second fill port allows us to run long term fills of Macroshell (TM) capsules (large fuel containers) concurrently. The principle processes of the system are: (1) storage of the tritium as a uranium hydride; (2) pressure intensification using cryogenics; and (3) filling of the shells by permeation at elevated temperatures. The design of the facility was centered around a NRC license limit of 6000 Ci

Scoping studies of tritium handling in a tokamak experimental power reactor

International Nuclear Information System (INIS)

Cherdack, R.; Watson, J.S.; Clinton, S.D.; Fisher, P.W.

1975-01-01

Tritium handling techniques in an experimental fusion power reactor (EPR) are evaluated to determine the requirements of the system and to compare different equipment and techniques for meeting those requirements. Tritium process equipment is needed to (1) evacuate and maintain a vacuum in the plasma vessel and the neutral beam injectors, (2) purify and recycle tritium and deuterium for the plasma fuel cycle, (3) recover tritium from experimental breeding modules, and (4) provide tritium containment and atmospheric cleanup. A development program is outlined to develop and demonstrate the required techniques and equipment and to permit confident design of an EPR for operation by the mid-1980s

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

Tritium inventory and permeation in liquid breeder blankets

International Nuclear Information System (INIS)

Reiter, F.

1990-01-01

This report reviews studies of the transport of hydrogen isotopes in the DEMO relevant water-cooled Pb-17Li blanket to be tested in NET and in a self-cooled blanket which uses Pb-17Li or Flibe as a liquid breeder material and V or Fe as a first wall material. The time dependences of tritium inventory and permeation in these blankets and of deuterium and tritium recycling in the self-cooled blanket are presented and discussed

Deuterium Depth Profile in Neutron-Irradiated Tungsten Exposed to Plasma

International Nuclear Information System (INIS)

Shimada, Masashi; Cao, G.; Hatano, Y.; Oda, T.; Oya, Y.; Hara, M.; Calderoni, P.

2011-01-01

The effect of radiation damage has been mainly simulated using high-energy ion bombardment. The ions, however, are limited in range to only a few microns into the surface. Hence, some uncertainty remains about the increase of trapping at radiation damage produced by 14 MeV fusion neutrons, which penetrate much farther into the bulk material. With the Japan-US joint research project: Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), the tungsten samples (99.99 % pure from A.L.M.T., 6mm in diameter, 0.2mm in thickness) were irradiated to high flux neutrons at 50 C and to 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). Subsequently, the neutron-irradiated tungsten samples were exposed to a high-flux deuterium plasma (ion flux: 1021-1022 m-2s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment (TPE) at the Idaho National Laboratory (INL). First results of deuterium retention in neutron-irradiated tungsten exposed in TPE have been reported previously. This paper presents the latest results in our on-going work of deuterium depth profiling in neutron-irradiated tungsten via nuclear reaction analysis. The experimental data is compared with the result from non neutron-irradiated tungsten, and is analyzed with the Tritium Migration Analysis Program (TMAP) to elucidate the hydrogen isotope behavior such as retention and depth distribution in neutron-irradiated and non neutron-irradiated tungsten.

Current Sandia programs and laboratory facilities for tritium research

International Nuclear Information System (INIS)

Swansiger, W.A.; West, L.A.

1975-01-01

Currently envisioned fusion reactor systems will contain substantial quantities of tritium. Strict control of the overall tritium inventory and environmental safety considerations require an accurate knowledge of the behavior of this isotope in the presence of Controlled Thermonuclear Reactor (CTR) materials. A 14,000 ft 2 laboratory for tritium research is currently under construction at Sandia Laboratories in Livermore. Details about the laboratory in general are provided. Results from studies of hydrogen isotope diffusion in surface-characterized metals will be presented. Details of two permeation systems (one for hydrogen and deuterium, the other for tritium) will be discussed. Data will also be presented concerning the gettering of hydrogen isotopes and application to CTR collector designs. (auth)

Laser ``M'egajoule'' cryogenic target program: from target fabrication to conformation of the deuterium-tritium ice layer

Science.gov (United States)

Collier, Rémy; Durut, Frédéric; Reneaume, Benoît; Chicane, Cédric; Théobald, Marc; Breton, Olivier; Martin, Michel; Fleury, Emmanuel; Vincent-Viry, Olivier; Bachelet, Franck; Jeannot, Laurent; Geoffray, Isabelle; Botrel, Ronan; Dauteuil, Christophe; Hermerel, Cyril; Choux, Alexandre; Bednarczyk, Sophie; Legaie, Olivier

2008-11-01

For the French inertial confinement fusion (ICF) experiments, cryogenic target assemblies (CTAs) for the LMJ program are manufactured and filled at CEA Valduc (Dijon) in the cryogenic targets filling station (IRCC). They will be moved at about 20 K into a transport cryostat for cryogenic targets and will be driven from CEA/Valduc to CEA/CESTA (Bordeaux). Cryogenic targets will then be transferred by several cryogenic grippers on the cryogenic target positioner before shots. The CTA has to meet severe specifications and involves a lot of challenging tasks for its manufacture. To fill CTAs by permeation with deuterium-tritium (DT), the IRCC need to meet strict thermal, mechanical and dimensional specifications. To obtain a good combustion yield, a very homogenous DT ice layer and very smooth roughness at 1.5 K below the DT triple point are also required. This paper deals with the up to date main issues in the different fields of the LMJ cryogenic target program.

Tritium accounting for PHWR plants

International Nuclear Information System (INIS)

Nair, P.S.; Duraisamy, S.

2012-01-01

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

Tritium migration in nuclear desalination plants

International Nuclear Information System (INIS)

Muralev, E.D.

2003-01-01

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

Solid deuterium centrifuge pellet injector

International Nuclear Information System (INIS)

Foster, C.A.

1982-01-01

Pellet injectors are needed to fuel long pulse tokamak plasmas and other magnetic confinement devices. For this purpose, an apparatus has been developed that forms 1.3-mm-diam pellets of frozen deuterium at a rate of 40 pellets per second and accelerates them to a speed of 1 km/s. Pellets are formed by extruding a billet of solidified deuterium through a 1.3-mm-diam nozzle at a speed of 5 cm/s. The extruding deuterium is chopped with a razor knife, forming 1.3-mm right circular cylinders of solid deuterium. The pellets are accelerated by synchronously injecting them into a high speed rotating arbor containing a guide track, which carries them from a point near the center of rotation to the periphery. The pellets leave the wheel after 150 0 of rotation at double the tip speed. The centrifuge is formed in the shape of a centrifugal catenary and is constructed of high strength KEVLAR/epoxy composite. This arbon has been spin-tested to a tip speed of 1 km/s

Solid deuterium centrifuge pellet injector

International Nuclear Information System (INIS)

Foster, C.A.

1983-01-01

Pellet injectors are needed to fuel long pulse tokamak plasmas and other magnetic confinement devices. For this purpose, an apparatus has been developed that forms 1.3-mm-diam pellets of frozen deuterium at a rate of 40 pellets per second and accelerates them to a speed of 1 km/s. Pellets are formed by extruding a billet of solidified deuterium through a 1.3-mm-diam nozzle at a speed of 5 cm/s. The extruding deuterium is chopped with a razor knife, forming 1.3-mm right circular cylinders of solid deuterium. The pellets are accelerated by synchronously injecting them into a high speed rotating arbor containing a guide track, which carries them from a point near the center of rotation to the periphery. The pellets leave the wheel after 150 0 of rotation at double the tip speed. The centrifuge is formed in the shape of a centrifugal catenary and is constructed of high strength Kevlar/epoxy composite. This arbor has been spin-tested to a tip speed of 1 km/s

Assessment of the cryogenic distillation system in Cernavoda tritium removal facility

International Nuclear Information System (INIS)

Pasca, Gheorghe; Draghia, Mirela; Porcariu, Florina; Ana, George

2010-01-01

Full text: This paper aims at presenting an assessment of the Cryogenic Distillation system (CD) in the Cernavoda Tritium Removal Facility (CTRF). The cryogenic distillation system is one of the key components of the CTRF which comprises other systems as: the liquid phase catalytic exchange system, designed to transfer tritium from heavy water to a deuterium stream to be fed into the CD system; the atmosphere detritiation system; the tritium recovery system; the tritium/hydrogen monitoring system; the central interlocking system; the tritium extraction and storage system. Thus, the need to build a tritium separation and recovery system results from economic opportunities offered both by heavy water reuse and tritium production, but, at the same time, it offers an alternative for the storage of tritiated heavy water as radioactive waste. (authors)

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

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

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

Investigation of Workplace-like Calibration Fields via a Deuterium-Tritium (D-T) Neutron Generator.

Science.gov (United States)

Mozhayev, Andrey V; Piper, Roman K; Rathbone, Bruce A; McDonald, Joseph C

2017-04-01

Radiation survey meters and personal dosimeters are typically calibrated in reference neutron fields based on conventional radionuclide sources, such as americium-beryllium (Am-Be) or californium-252 (Cf), either unmodified or heavy-water moderated. However, these calibration neutron fields differ significantly from the workplace fields in which most of these survey meters and dosimeters are being used. Although some detectors are designed to yield an approximately dose-equivalent response over a particular neutron energy range, the response of other detectors is highly dependent upon neutron energy. This, in turn, can result in significant over- or underestimation of the intensity of neutron radiation and/or personal dose equivalent determined in the work environment. The use of simulated workplace neutron calibration fields that more closely match those present at the workplace could improve the accuracy of worker, and workplace, neutron dose assessment. This work provides an overview of the neutron fields found around nuclear power reactors and interim spent fuel storage installations based on available data. The feasibility of producing workplace-like calibration fields in an existing calibration facility has been investigated via Monte Carlo simulations. Several moderating assembly configurations, paired with a neutron generator using the deuterium tritium (D-T) fusion reaction, were explored.

Applications of hydrophobic Pt catalysts in separation of tritium from liquid effluents

International Nuclear Information System (INIS)

Ionita, Gheorghe; Popescu, Irina; Stefanescu, Ioan; Varlam, Carmen

2003-01-01

Hydrophobic Pt catalysts were first prepared and used in deuterium or tritium separation while after their application was extended to chemical reactions occurring in liquid water or saturated humidity environments. Capillary condensing produced at the contact with liquid water or vapors engenders in classical hydrophilic catalysts a decrease in activity what makes them inefficient. Consequently, liquid water 'repealing' catalysts are to be used allowing, at the same time gaseous reactants and reaction products to diffuse to and fro the catalytic active centers. These catalysts were successfully applied in deuterium enrichment and tritium separation based on hydrogen- liquid water isotopic exchange at both pilot and industrial scale. High activity and a prolonged stability were demonstrated and checked in: - detritiation of the heavy water used as both moderator and coolant in CANDU type reactors; removing of tritium from light water recirculated in nuclear fuel reprocessing facilities; removal and recovery of tritium from atmosphere and tritium processing installations. Due to their incontestable advantages the use of these catalysts was recently extended to other chemical processes occurring in the presence of liquid water or in high humidity environment or else when water occurs as a reaction product, such as catalytic hydrogen - oxygen recombination at room temperature or removal of stable organic pollutants from waste waters

Determination of the Tritium Concentration in Deuterium-Tritium Fusion Plasmas from the Jet TTE Campaign

International Nuclear Information System (INIS)

Gatu Johnson, Maria

2004-01-01

This thesis describes the development and implementation of methods for tritium concentration determination for JET fusion plasmas. The usefulness of MPR data in this context is investigated. It is shown that results from MPR spectral analysis can simplify the calculations for neutral beam heated plasmas and that it is essential for calculations for radio frequency heated plasmas. The described methods are applied to pulses from the Trace Tritium Experiment (TTE), staged at JET in October 2003. Results from simple, time resolved analysis using MPR and other public JET data are presented and the assumptions made in the calculations are discussed. The results agree with expectations but would be even more interesting if spatial variations were taken into account

Determination of the Tritium Concentration in Deuterium-Tritium Fusion Plasmas from the Jet TTE Campaign

Energy Technology Data Exchange (ETDEWEB)

Gatu Johnson, Maria

2004-01-01

This thesis describes the development and implementation of methods for tritium concentration determination for JET fusion plasmas. The usefulness of MPR data in this context is investigated. It is shown that results from MPR spectral analysis can simplify the calculations for neutral beam heated plasmas and that it is essential for calculations for radio frequency heated plasmas. The described methods are applied to pulses from the Trace Tritium Experiment (TTE), staged at JET in October 2003. Results from simple, time resolved analysis using MPR and other public JET data are presented and the assumptions made in the calculations are discussed. The results agree with expectations but would be even more interesting if spatial variations were taken into account.

Vacuum pumping of tritium in fusion power reactors

International Nuclear Information System (INIS)

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

1979-01-01

Compound cryopumps of three different designs will be tested with deuterium-tritium (DT) mixtures under simulated fusion reactor conditions at the Tritium Systems Test Assembly (TSTA) now being constructed at the Los Alamos Scientific Laboratory (LASL). The first of these pumps is already in operation, and its preliminary performance is presented. The supporting vacuum facility necessary to regenerate these fusion facility cryopumps is also described. The next generation of fusion system vacuum pumps may include non-cryogenic or conventional-cryogenic hybrid systems, several of which are discussed

Correlation between abnormal deuterium flux and heat flow in a D/Pd system

International Nuclear Information System (INIS)

Li Xingzhong; Liu Bin; Tian Jian; Wei Qingming; Zhou Rui; Yu Zhiwu

2003-01-01

Deuterium flux through the thin wall of a palladium tube has been studied by monitoring gas pressure and temperature. A high-precision calorimeter (Calvet) was used to detect heat flow when the heater was shut down and the palladium tube was cooling down slowly. At certain temperatures an abnormal deuterium flux appeared. This deuterium flux reached a peak when the temperature of the palladium was decreasing. This abnormal deuterium flux differs from the monotonic feature of a normal diffusive flux and is accompanied by a heat flow

Evaluation of thin CaF2 (Eu) scintillator for detecting tritium

International Nuclear Information System (INIS)

Chiles, M.M.

1986-10-01

The primary objective of this project was to investigate the feasibility of using a CaF 2 (Eu) scintillator for detecting low-energy beta particles from tritium. A proof-of-principle detector was designed for flowing tritium-spiked nitrogen gas across the surface of a thin scintillator, which was optically coupled between two low-noise photomultiplier tubes. Electronics for operating the two photomultiplier tubes in coincidence eliminated most of the tube noise pulses and allowed detection of the small pulses from the low-energy tritium beta particles

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

Pneumatic injector of deuterium macroparticles for TORE-SUPRA tokamak

International Nuclear Information System (INIS)

Vinyar, I.V.; Umov, A.P.; Lukin, A.Ya.; Skoblikov, S.V.; Reznichenko, P.V.; Krasil'nikov, I.A.

2006-01-01

The pneumatic injector for periodic injection of fuel-solid-deuterium pellets into the plasma of the TORE-SUPRA tokamak in a steady-state mode is described. The deuterium pellet injection with an unlimited duration is ensured by a screw extruder in which gaseous deuterium is frozen and squeezed outwards in the form of a rod with a rectangular cross section. A cutter installed on the injector's barrel cuts a cylinder with a diameter of 2 mm and a length of 1.0-3.5 mm out from this rod. The movement of the cutter is controlled by a pulsed electromagnetic drive at a pulse repetition rate of 10 Hz. In the injector's barrel, a compressed gas accelerates a deuterium pellet to a velocity of 100-650 m/s [ru

Ion beam characteristics of the controlatron/zetatron family of the gas filled neutron tubes

Energy Technology Data Exchange (ETDEWEB)

Berg, R.S.; Shope, L.A.; O' Neal, M.L.; Boers, J.E.; Bickes, R.W. Jr.

1981-03-01

A gas filled tube used to produce a neutron flux with the D(T,He/sup 4/)n reaction is described. Deuterium and tritium ions generated in a reflex discharge are extracted and accelerated to 100 keV by means of an accelerator electrode onto a deutero-tritide target electrode. The electrodes are designed to focus the ion beam onto the target. Total tube currents consisting of extracted ions, unsuppressed secondary electrons, and ions generated by interactions with the background gas are typically 100 mA. The characteristics of the extracted ion beam are discussed. Accelerating voltages greater than 50 kV are required to focus the beam through the accelerator aperture for configurations that give beams with the proper energy density onto the target. The perveance of the beam is discussed. Maximum perveance values are 2 to 20 nanopervs. Tube focusing and neutron production characteristics are described.

Tritium in groundwater investigation at the Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

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

2001-01-01

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

Tritium system test assembly control system cost estimate

International Nuclear Information System (INIS)

Stutz, R.A.

1979-01-01

The principal objectives of the Tritium Systems Test Assembly (TSTA), which includes the development, demonstration and interfacing of technologies related to the deuterium--tritium fuel cycle for fusion reactor systems, are concisely stated. The various integrated subsystems comprising TSTA and their functions are discussed. Each of the four major subdivisions of TSTA, including the main process system, the environmental and safety systems, supporting systems and the physical plant are briefly discussed. An overview of the Master Data Acquisition and Control System, which will control all functional operation of TSTA, is provided

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)

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

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.

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)

Demonstration of Fuel Hot-Spot Pressure in Excess of 50 Gbar for Direct-Drive, Layered Deuterium-Tritium Implosions on OMEGA

Science.gov (United States)

Regan, S. P.; Goncharov, V. N.; Igumenshchev, I. V.; Sangster, T. C.; Betti, R.; Bose, A.; Boehly, T. R.; Bonino, M. J.; Campbell, E. M.; Cao, D.; Collins, T. J. B.; Craxton, R. S.; Davis, A. K.; Delettrez, J. A.; Edgell, D. H.; Epstein, R.; Forrest, C. J.; Frenje, J. A.; Froula, D. H.; Gatu Johnson, M.; Glebov, V. Yu.; Harding, D. R.; Hohenberger, M.; Hu, S. X.; Jacobs-Perkins, D.; Janezic, R.; Karasik, M.; Keck, R. L.; Kelly, J. H.; Kessler, T. J.; Knauer, J. P.; Kosc, T. Z.; Loucks, S. J.; Marozas, J. A.; Marshall, F. J.; McCrory, R. L.; McKenty, P. W.; Meyerhofer, D. D.; Michel, D. T.; Myatt, J. F.; Obenschain, S. P.; Petrasso, R. D.; Radha, P. B.; Rice, B.; Rosenberg, M. J.; Schmitt, A. J.; Schmitt, M. J.; Seka, W.; Shmayda, W. T.; Shoup, M. J.; Shvydky, A.; Skupsky, S.; Solodov, A. A.; Stoeckl, C.; Theobald, W.; Ulreich, J.; Wittman, M. D.; Woo, K. M.; Yaakobi, B.; Zuegel, J. D.

2016-07-01

A record fuel hot-spot pressure Phs=56 ±7 Gbar was inferred from x-ray and nuclear diagnostics for direct-drive inertial confinement fusion cryogenic, layered deuterium-tritium implosions on the 60-beam, 30-kJ, 351-nm OMEGA Laser System. When hydrodynamically scaled to the energy of the National Ignition Facility, these implosions achieved a Lawson parameter ˜60 % of the value required for ignition [A. Bose et al., Phys. Rev. E 93, LM15119ER (2016)], similar to indirect-drive implosions [R. Betti et al., Phys. Rev. Lett. 114, 255003 (2015)], and nearly half of the direct-drive ignition-threshold pressure. Relative to symmetric, one-dimensional simulations, the inferred hot-spot pressure is approximately 40% lower. Three-dimensional simulations suggest that low-mode distortion of the hot spot seeded by laser-drive nonuniformity and target-positioning error reduces target performance.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

International Nuclear Information System (INIS)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J.P.

2011-01-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10 21 m -2 s -1 , ion fluence: 4 x 10 25 m -2 ) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Science.gov (United States)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J. P.

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

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

International Nuclear Information System (INIS)

McKone, T.E.

1978-08-01

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

The Effect of Ion Energy and Substrate Temperature on Deuterium Trapping in Tungsten

Science.gov (United States)

Roszell, John Patrick Town

Tungsten is a candidate plasma facing material for next generation magnetic fusion devices such as ITER and there are major operational and safety issues associated with hydrogen (tritium) retention in plasma facing components. An ion gun was used to simulate plasma-material interactions under various conditions in order to study hydrogen retention characteristics of tungsten thus enabling better predictions of hydrogen retention in ITER. Thermal Desorption Spectroscopy (TDS) was used to measure deuterium retention from ion irradiation while modelling of TDS spectra with the Tritium Migration Analysis Program (TMAP) was used to provide information about the trapping mechanisms involved in deuterium retention in tungsten. X-ray Photoelectron Spectroscopy (XPS) and Secondary Ion Mass Spectrometry (SIMS) were used to determine the depth resolved composition of specimens used for irradiation experiments. Carbon and oxygen atoms will be among the most common contaminants within ITER. C and O contamination in polycrystalline tungsten (PCW) specimens even at low levels (˜0.1%) was shown to reduce deuterium retention by preventing diffusion of deuterium into the bulk of the specimen. This diffusion barrier was also responsible for the inhibition of blister formation during irradiations at 500 K. These observations may provide possible mitigation techniques for problems associated with tritium retention and mechanical damage to plasma facing components caused by hydrogen implantation. Deuterium trapping in PCW and single crystal tungsten (SCW) was studied as a function of ion energy and substrate temperature. Deuterium retention was shown to decrease with decreasing ion energy below 100 eV/D+. Irradiation of tungsten specimens with 10 eV/D+ ions was shown to retain up to an order of magnitude less deuterium than irradiation with 500 eV/D+ ions. Furthermore, the retention mechanism for deuterium was shown to be consistent across the entire energy range studied (10-500 e

Development of positron annihilation spectroscopy for investigating deuterium decorated voids in neutron-irradiated tungsten

Energy Technology Data Exchange (ETDEWEB)

Taylor, C.N., E-mail: chase.taylor@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Shimada, M.; Merrill, B.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Akers, D.W. [Experimental Programs, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan)

2015-08-15

The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900 °C, and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium-free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined.

Development of positron annihilation spectroscopy for investigating deuterium decorated voids in neutron-irradiated tungsten

Science.gov (United States)

Taylor, C. N.; Shimada, M.; Merrill, B. J.; Akers, D. W.; Hatano, Y.

2015-08-01

The present work is a continuation of a recent research to develop and optimize positron annihilation spectroscopy (PAS) for characterizing neutron-irradiated tungsten. Tungsten samples were exposed to neutrons in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory and damaged to 0.025 and 0.3 dpa. Subsequently, they were exposed to deuterium plasmas in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory. The implanted deuterium was desorbed through sample heating to 900 °C, and Doppler broadening (DB)-PAS was performed both before and after heating. Results show that deuterium impregnated tungsten is identified as having a smaller S-parameter. The S-parameter increases after deuterium desorption. Microstructural changes also occur during sample heating. These effects can be isolated from deuterium desorption by comparing the S-parameters from the deuterium-free back face with the deuterium-implanted front face. The application of using DB-PAS to examine deuterium retention in tungsten is examined.

Fast ignition upon the implosion of a thin shell onto a precompressed deuterium-tritium ball

Science.gov (United States)

Gus'kov, S. Yu.; Zmitrenko, N. V.

2012-11-01

Fast ignition of a precompressed inertial confinement fusion (ICF) target by a hydrodynamic material flux is investigated. A model system of hydrodynamic objects consisting of a central deuterium-tritium (DT) ball and a concentric two-layer shell separated by a vacuum gap is analyzed. The outer layer of the shell is an ablator, while the inner layer consists of DT ice. The igniting hydrodynamic flux forms as a result of laser-driven acceleration and compression of the shell toward the system center. A series of one-dimensional numerical simulations of the shell implosion, the collision of the shell with the DT ball, and the generation and propagation of thermonuclear burn waves in both parts of the system are performed. Analytic models are developed that describe the implosion of a thin shell onto a central homogeneous ball of arbitrary radius and density and the initiation and propagation of a thermonuclear burn wave induced by such an implosion. Application of the solution of a model problem to analyzing the implosion of a segment of a spherical shell in a conical channel indicates the possibility of fast ignition of a spherical ICF target from a conical target driven by a laser pulse with an energy of 500-700 kJ.

Method of removing tritium in exhaust water in a nuclear equipment

Energy Technology Data Exchange (ETDEWEB)

Iwamoto, T

1976-05-12

A method is claimed to increase the efficiency of removing tritium from waste water through adsorption treatment. Steam is produced by heating waste water containing tritium, and it is passed through a tube filled with an adsorbent such as activated alumina, silica gel or zeolite. When a control limiting value is reached by the concentration of tritium within the steam, the flow of steam is stopped, and the adsorption tube is removed from the path of steam flow. Thereafter, another adsorption tube containing the afore-said adsorbent is provided in the steam flow path, and the steam is then allowed to flow again.

Tritium contamination of concrete walls and floors in tritium-handling laboratory

International Nuclear Information System (INIS)

Kawano, T.; Kuroyanagi, M.; Tabei, T.

2006-01-01

A tritium handling laboratory was constructed at the National Institute for Fusion Science about twenty years ago and it was recently closed down. We completed the necessary work that is legally required in Japan at the laboratory, when the use of radioisotopes is discontinued, involving measurements of radioactive contamination. We mainly used smear and direct-immersion methods for the measurements. In applying the smear method, we used a piece of filter paper to wipe up the tritium staining the surfaces. The filter paper containing the tritium was placed directly into a dedicated vial, a scintillation cocktail was then poured over it, and the tritium was measured with a liquid scintillation counter. With the direct-immersion method, a piece of concrete was placed directly into a vial containing a scintillation cocktail, and the tritium in the concrete was measured with a liquid scintillation counter. As well as these measurements, we investigated water-extraction and heating-cooling methods for measuring tritium contamination in concrete. With the former, a piece of concrete was placed into water in a tube to extract the tritium, the water containing the extracted tritium was then poured into a dedicated vial containing a scintillation cocktail, and the tritium contamination was measured. With the latter, a piece of concrete was placed into a furnace and heated to 800 degrees centigrade to vaporize the tritiated water into flowing dry air. The flowing air was then cooled to collect the vaporized tritiated water in a tube. The collected water was placed in a vial for scintillation counting. To evaluate the direct-immersion method, ratios were determined by dividing the contamination measured with the heating-cooling method by that measured with the direct-immersion method. The average ratio was about 2.5, meaning a conversion factor from contamination obtained with the direct-immersion method to that with the heating-cooling method. We also investigated the

A simple trapping method of exhaled water using an ice-cooled tube to monitor the tritium level in human body

International Nuclear Information System (INIS)

Nogawa, Norio; Makide, Yoshihiro

1994-01-01

A convenient and efficient method is developed for the trapping of water in exhaled air. A bent-V-shaped glass sampling tube was immersed in iced water and exhaled air was introduced into the tube through a plastic straw. The trapping efficiency of exhaled water was equivalent to those with more complex and troublesome methods. Using anywhere available ice, the water in exhaled air can be rapidly collected with this method and the tritium level in the body will be quickly obtained. (author)

Mixed deuterium-tritium neutral beam injection

International Nuclear Information System (INIS)

Ruby, L.; Lewis, M.S.

1989-01-01

An alternative mixed beam neutral beam injector (MNBI) for fusion reactors is proposed that eliminates the conventional isotope separation system (ISS) in the fuel cycle. The principal advantage of the alternative system is a capital and operating cost savings in the fuel cycle, as the ISS employs cryogenic distillation at liquid-hydrogen temperatures to effect a separation of hydrogen isotopes and to eliminate a buildup of normal hydrogen in the recycled fuel. Possible additional advantages of the alternative method involve an improvement in overall safety and a reduction of the amount of tritium in the fuel cycle. The alternative heating system uses an electromagnetic separation in the MNBI to limit the buildup of normal hydrogen. Calculations indicate that an MNBI can be reasonably optimized in the case of an upgraded injection system for the Tokamak Fusion Test Reactor

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Energy Technology Data Exchange (ETDEWEB)

Shimada, Masashi, E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Calderoni, P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Oda, T. [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo 113-8656 (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Sokolov, M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Zhang, K. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Cao, G. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706 (United States); Kolasinski, R. [Hydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, CA 94551 (United States); Sharpe, J.P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10{sup 21} m{sup -2} s{sup -1}, ion fluence: 4 x 10{sup 25} m{sup -2}) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Neutron generator tube ion source control apparatus

International Nuclear Information System (INIS)

Bridges, J.R.

1982-01-01

A pulsed neutron well logging system includes a neutron generator tube of the deuterium-tritium accelerator type and an ion source control apparatus providing extremely sharply time-defined neutron pulses. A low voltage control pulse supplied to an input by timing circuits turns a power FET on via a buffer-driver whereby a 2000 volt pulse is produced in the secondary of a pulse transformer and applied to the ion source of the tube. A rapid fall in this ion source control pulse is ensured by a quenching circuit wherein a one-shot responds to the falling edge of the control pulse and produces a 3 microsecond delay to compensate for the propagation delay. A second one-shot is triggered by the falling edge of the output of the first one-shot and gives an 8 microsecond pulse to turn on the power FET which, via an isolation transformer turns on a series-connected transistor to ground the secondary of the pulse transformer and the ion source. (author)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Tritium permeation losses in HYLIFE-II heat exchanger tubes

International Nuclear Information System (INIS)

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

1990-01-01

Tritium permeation through the intermediate heat exchanger of the HYLIFE-II inertial fusion design concept is evaluated for routine operating conditions. The permeation process is modelled using the Lewis analogy combined with surface recombination. It is demonstrated that at very low driving potentials, permeation becomes proportional to the first power of the driving potential. The model predicts that under anticipated conditions the primary cooling loop will pass about 6% of the tritium entering it to the intermediate coolant. Possible approached to reducing tritium permeation are explored. Permeation is limited by turbulent diffusion transport through the molten salt. Hence, surface barriers with impendance factors typical of present technology can do very little to reduce permeation. Low Flibe viscosity is desirable. An efficient tritium removal system operating on the Flibe before it gets to the intermediate heat exchanger is required. Needs for further research are highlighted. 9 refs., 2 figs., 1 tab

Direct depth distribution measurement of deuterium in bulk tungsten exposed to high-flux plasma

Directory of Open Access Journals (Sweden)

C. N. Taylor

2017-05-01

Full Text Available Understanding tritium retention and permeation in plasma-facing components is critical for fusion safety and fuel cycle control. Glow discharge optical emission spectroscopy (GD-OES is shown to be an effective tool to reveal the depth profile of deuterium in tungsten. Results confirm the detection of deuterium. A ∼46 μm depth profile revealed that the deuterium content decreased precipitously in the first 7 μm, and detectable amounts were observed to depths in excess of 20 μm. The large probing depth of GD-OES (up to 100s of μm enables studies not previously accessible to the more conventional techniques for investigating deuterium retention. Of particular applicability is the use of GD-OES to measure the depth profile for experiments where high deuterium concentration in the bulk material is expected: deuterium retention in neutron irradiated materials, and ultra-high deuterium fluences in burning plasma environment.

Tritium aging effect of LaNi4.9Al0.1Dx on de-deuterating kinetics

International Nuclear Information System (INIS)

Xiong Yifu; Luo Deli; Li Rong

2002-01-01

Kinetics parameters are measured at different aging times of LaNi 4.9 Al 0.1 alloy. The influence of tritium aging on kinetic feature of LaNi 4.9 Al 0.1 alloy is assessed. The results show that reaction rate decreases with aging time, but tritium aging does not change de-deuterating reaction order. De-deuterating reaction orders a (with respect to deuterium pressure) and b (with respect to deuterium content) are 0.5 and 1, respectively. Activation energy (E) increases with aging time. After 1120 d, the reaction rate constant is decreased by two orders of magnitude, activation energy is increased by a factor of 0.3

Transport of recycled deuterium to the plasma core in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Bell, M.G.; Budny, R.V.; Jassby, D.L.; Park, H.; Ramsey, A.T.; Stotler, D.P.; Strachan, J.D.

1997-10-01

The authors report a study of the fueling of the plasma core by recycling in the Tokamak Fusion Test Reactor (TFTR). They have analyzed discharges fueled by deuterium recycled from the limiter and tritium-only neutral beam injection. In these plasmas, the DT neutron rate provides a measure of the deuterium influx into the core plasma. They find a reduced influx with plasmas using lithium pellet conditioning and with plasmas of reduced major (and minor) radius. Modeling with the DEGAS neutrals code shows that the dependence on radius can be related to the penetration of neutrals through the scrape-off layer

Experimental study of curved guide tubes for pellet injection

International Nuclear Information System (INIS)

Combs, S.K.; Baylor, L.R.; Foust, C.R.; Gouge, M.J.; Jernigan, T.C.; Milora, S.L.

1997-01-01

The use of curved guide tubes for transporting frozen hydrogen pellets offers great flexibility for pellet injection into plasma devices. While this technique has been previously employed, an increased interest in its applicability has been generated with the recent ASDEX Upgrade experimental data for magnetic high-field side (HFS) pellet injection. In these innovative experiments, the pellet penetration appeared to be significantly deeper than for the standard magnetic low-field side injection scheme, along with corresponding greater fueling efficiencies. Thus, some of the major experimental fusion devices are planning experiments with HFS pellet injection. Because of the complex geometries of experimental fusion devices, installations with multiple curved guide tube sections will be required for HFS pellet injection. To more thoroughly understand and document the capability of curved guide tubes, an experimental study is under way at the Oak Ridge National Laboratory (ORNL). In particular, configurations and pellet parameters applicable for the DIII-D tokamak and the International Thermonuclear Experimental Reactor (ITER) were simulated in laboratory experiments. Initial test results with nominal 2.7- and 10-mm-diam deuterium pellets are presented and discussed

Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility.

Science.gov (United States)

Döppner, T; Callahan, D A; Hurricane, O A; Hinkel, D E; Ma, T; Park, H-S; Berzak Hopkins, L F; Casey, D T; Celliers, P; Dewald, E L; Dittrich, T R; Haan, S W; Kritcher, A L; MacPhee, A; Le Pape, S; Pak, A; Patel, P K; Springer, P T; Salmonson, J D; Tommasini, R; Benedetti, L R; Bond, E; Bradley, D K; Caggiano, J; Church, J; Dixit, S; Edgell, D; Edwards, M J; Fittinghoff, D N; Frenje, J; Gatu Johnson, M; Grim, G; Hatarik, R; Havre, M; Herrmann, H; Izumi, N; Khan, S F; Kline, J L; Knauer, J; Kyrala, G A; Landen, O L; Merrill, F E; Moody, J; Moore, A S; Nikroo, A; Ralph, J E; Remington, B A; Robey, H F; Sayre, D; Schneider, M; Streckert, H; Town, R; Turnbull, D; Volegov, P L; Wan, A; Widmann, K; Wilde, C H; Yeamans, C

2015-07-31

We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a "high-foot" laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shape closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 10^{16} neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.

Low-level tritium research facility for the University of Toronto Institute for Aerospace Studies

International Nuclear Information System (INIS)

Kherani, N.P.; Shmayda, W.T.

1984-06-01

The objective of the Low-level Tritium Research Facility for the University of Toronto Institute for Aerospace Studies (UTIAS) is to investigate tritium-material interactions and how they differ with respect to protium and deuterium. The tritium laboratory will also be employed to study tritium retention, tritium imaging, and the effect of tritium on diagnostic devices. This report is a preliminary design document of the UTIAS Low-Level Tritium Research Facility including the fundamentals of tritium, a description of the facility, tritium laboratory requirements and the safety analysis of the laboratory. The facility is designed to handle a total elemental tritium inventory of 10 Ci, though it will initially commence operation with 1 Ci and later increased to the maximum value. In the event of an instantaneous emission of the total tritium inventory within the laboratory, the working personnel would be exposed to an airborne tritium concentration less than the maximum permissible. Moreover, with all the safety features included in this design the likelihood of such an accident is very remote. Thus, the tritium laboratory design is intrinsically safe

Tritium supply assessment for ITER and DEMOnstration power plant

International Nuclear Information System (INIS)

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

2013-01-01

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

Tritium supply assessment for ITER and DEMOnstration power plant

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Effects of hydrogen mixture into helium gas on deuterium removal from lithium titanate

Energy Technology Data Exchange (ETDEWEB)

Tsuchiya, Akihito, E-mail: tsuchiya@frontier.hokudai.ac.jp [Laboratory of Plasma Physics and Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628 (Japan); Hino, Tomoaki; Yamauchi, Yuji; Nobuta, Yuji [Laboratory of Plasma Physics and Engineering, Hokkaido University, Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628 (Japan); Akiba, Masato; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193 (Japan)

2013-10-15

Lithium titanate (Li{sub 2}TiO{sub 3}) pebbles were irradiated with deuterium ions with energy of 1.7 keV and then exposed to helium or helium–hydrogen mixed gas at various temperatures, in order to evaluate the effects of gas exposure on deuterium removal from the pebbles. The amounts of residual deuterium in the pebbles were measured by thermal desorption spectroscopy. The mixing of hydrogen gas into helium gas enhanced the removal amount of deuterium. In other words, the amount of residual deuterium after the helium–hydrogen mixed gas exposure at lower temperature was lower than that after the helium gas exposure. In addition, we also evaluated the pebbles exposed to the helium gas with different hydrogen mixture ratio from 0% to 1%, at 573 K. Although the amount of residual deuterium in the pebbles after the exposure decreased with increasing the hydrogen mixture ratio, the implanted deuterium partly remained after the exposure. These results suggest that the tritium inventory may occur at low temperature region in the blanket during the operation.

Relevance of d-D interactions on neutron and tritium production in IFMIF-EVEDA accelerator prototype

International Nuclear Information System (INIS)

Mayoral, A.; Sanz, J.; Sauvan, P.; Lopez, D.; Garcia, M.; Ogando, F.

2011-01-01

In the IFMIF-EVEDA accelerator prototype, deuterium is implanted in the components due to beam losses and in the beam dump, where the beam is stopped. The interaction of the deuterons with the deuterium previously implanted leads to the production of neutrons and tritium, which are important issues for radioprotection and safety analysis. A methodology to assess these production pathways in more realistic approach has been developed. The new tools and their main achievement are: (i) an 'effective diffusivity coefficient' (deduced from available experimental data) that enables simulation of the diffusion phase, and (ii) the MCUNED code (able to handle deuteron transport libraries) allows to simulate the transport-slowdown of deuteron/tritium (to get the concentration profiles) and the neutron/tritium productions from d-Cu and d-D for up to 9 MeV incident deuteron. The results with/without theses tools are presented and their effect on the relevance of d-D sources versus d-Cu is evaluated.

Radiation assisted thermonuclear burn wave dynamics in heavy ion fast ignition of cylindrical deuterium-tritium fuel target

International Nuclear Information System (INIS)

Rehman, S.; Kouser, R.; Nazir, R.; Manzoor, Z.; Tasneem, G.; Jehan, N.; Nasim, M.H.; Salahuddin, M.

2015-01-01

Dynamics of thermonuclear burn wave propagation assisted by thermal radiation precursor in a heavy ion fast ignition of cylindrical deuterium-tritium (DT) fuel target are studied by two dimensional radiation hydrodynamic simulations using Multi-2D code. Thermal radiations, as they propagate ahead of the burn wave, suffer multiple reflections and preheat the fuel, are found to play a vital role in burn wave dynamics. After fuel ignition, the burn wave propagates in a steady state manner for some time. Multiple reflection and absorption of radiation at the fuel-tamper interface, fuel ablation and radial implosion driven by ablative shock and fast fusion rates on the fuel axis, at relatively later times, result into filamentary wave front. Strong pressure gradients are developed and sausage like structures behind the front are appeared. The situation leads to relatively reduced and non-uniform radial fuel burning and burn wave propagation. The fuel burning due to DD reaction is also taken into account and overall fusion energy and fusion power density, due to DT and DD reactions, during the burn wave propagation are determined as a function of time. (authors)

Sensitivity of inertial confinement fusion hot spot properties to the deuterium-tritium fuel adiabat

Energy Technology Data Exchange (ETDEWEB)

Melvin, J.; Lim, H.; Rana, V.; Glimm, J. [Department of Applied Mathematics and Statistics, Stony Brook University, Stony Brook, New York 11794-3600 (United States); Cheng, B.; Sharp, D. H.; Wilson, D. C. [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

2015-02-15

We determine the dependence of key Inertial Confinement Fusion (ICF) hot spot simulation properties on the deuterium-tritium fuel adiabat, here modified by addition of energy to the cold shell. Variation of this parameter reduces the simulation to experiment discrepancy in some, but not all, experimentally inferred quantities. Using simulations with radiation drives tuned to match experimental shots N120321 and N120405 from the National Ignition Campaign (NIC), we carry out sets of simulations with varying amounts of added entropy and examine the sensitivities of important experimental quantities. Neutron yields, burn widths, hot spot densities, and pressures follow a trend approaching their experimentally inferred quantities. Ion temperatures and areal densities are sensitive to the adiabat changes, but do not necessarily converge to their experimental quantities with the added entropy. This suggests that a modification to the simulation adiabat is one of, but not the only explanation of the observed simulation to experiment discrepancies. In addition, we use a theoretical model to predict 3D mix and observe a slight trend toward less mixing as the entropy is enhanced. Instantaneous quantities are assessed at the time of maximum neutron production, determined dynamically within each simulation. These trends contribute to ICF science, as an effort to understand the NIC simulation to experiment discrepancy, and in their relation to the high foot experiments, which features a higher adiabat in the experimental design and an improved neutron yield in the experimental results.

Sensitivity of inertial confinement fusion hot spot properties to the deuterium-tritium fuel adiabat

International Nuclear Information System (INIS)

Melvin, J.; Lim, H.; Rana, V.; Glimm, J.; Cheng, B.; Sharp, D. H.; Wilson, D. C.

2015-01-01

We determine the dependence of key Inertial Confinement Fusion (ICF) hot spot simulation properties on the deuterium-tritium fuel adiabat, here modified by addition of energy to the cold shell. Variation of this parameter reduces the simulation to experiment discrepancy in some, but not all, experimentally inferred quantities. Using simulations with radiation drives tuned to match experimental shots N120321 and N120405 from the National Ignition Campaign (NIC), we carry out sets of simulations with varying amounts of added entropy and examine the sensitivities of important experimental quantities. Neutron yields, burn widths, hot spot densities, and pressures follow a trend approaching their experimentally inferred quantities. Ion temperatures and areal densities are sensitive to the adiabat changes, but do not necessarily converge to their experimental quantities with the added entropy. This suggests that a modification to the simulation adiabat is one of, but not the only explanation of the observed simulation to experiment discrepancies. In addition, we use a theoretical model to predict 3D mix and observe a slight trend toward less mixing as the entropy is enhanced. Instantaneous quantities are assessed at the time of maximum neutron production, determined dynamically within each simulation. These trends contribute to ICF science, as an effort to understand the NIC simulation to experiment discrepancy, and in their relation to the high foot experiments, which features a higher adiabat in the experimental design and an improved neutron yield in the experimental results

Sensitivity of inertial confinement fusion hot spot properties to the deuterium-tritium fuel adiabat

Science.gov (United States)

Melvin, J.; Lim, H.; Rana, V.; Cheng, B.; Glimm, J.; Sharp, D. H.; Wilson, D. C.

2015-02-01

We determine the dependence of key Inertial Confinement Fusion (ICF) hot spot simulation properties on the deuterium-tritium fuel adiabat, here modified by addition of energy to the cold shell. Variation of this parameter reduces the simulation to experiment discrepancy in some, but not all, experimentally inferred quantities. Using simulations with radiation drives tuned to match experimental shots N120321 and N120405 from the National Ignition Campaign (NIC), we carry out sets of simulations with varying amounts of added entropy and examine the sensitivities of important experimental quantities. Neutron yields, burn widths, hot spot densities, and pressures follow a trend approaching their experimentally inferred quantities. Ion temperatures and areal densities are sensitive to the adiabat changes, but do not necessarily converge to their experimental quantities with the added entropy. This suggests that a modification to the simulation adiabat is one of, but not the only explanation of the observed simulation to experiment discrepancies. In addition, we use a theoretical model to predict 3D mix and observe a slight trend toward less mixing as the entropy is enhanced. Instantaneous quantities are assessed at the time of maximum neutron production, determined dynamically within each simulation. These trends contribute to ICF science, as an effort to understand the NIC simulation to experiment discrepancy, and in their relation to the high foot experiments, which features a higher adiabat in the experimental design and an improved neutron yield in the experimental results.

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

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

Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Bell, M.G.; Beer, M.

1997-02-01

Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l i ). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q a ∼ 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l i plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q 0 > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions

Growth acceleration and photosynthesis of the scenedesmus algae and cocconeis algae in deuterium water

International Nuclear Information System (INIS)

Liu Feng; Wang Wenqing

1998-01-01

In order to find new way to treat the radioactive tritium waste water, scenedesmus algae and cocconeis algae are cultured in medium which contains 30% (w) deuterium water. During different time, activities of photosymthesis, absorption spectrum, growth rate and low-temperature fluorescence spectrum are measured. Accelerated growth is found in the deuterium water compared to the normal water. Activities of photosynthesis show the similar result (F v /F m ) to the growth data. It is also concluded from low-temperature fluorescence spectra that algae activities in the deuterium water, which are expressed by PS I/PS II, are more sensitive than those in the normal water

Demonstration of Fuel Hot-Spot Pressure in Excess of 50 Gbar for Direct-Drive, Layered Deuterium-Tritium Implosions on OMEGA.

Science.gov (United States)

Regan, S P; Goncharov, V N; Igumenshchev, I V; Sangster, T C; Betti, R; Bose, A; Boehly, T R; Bonino, M J; Campbell, E M; Cao, D; Collins, T J B; Craxton, R S; Davis, A K; Delettrez, J A; Edgell, D H; Epstein, R; Forrest, C J; Frenje, J A; Froula, D H; Gatu Johnson, M; Glebov, V Yu; Harding, D R; Hohenberger, M; Hu, S X; Jacobs-Perkins, D; Janezic, R; Karasik, M; Keck, R L; Kelly, J H; Kessler, T J; Knauer, J P; Kosc, T Z; Loucks, S J; Marozas, J A; Marshall, F J; McCrory, R L; McKenty, P W; Meyerhofer, D D; Michel, D T; Myatt, J F; Obenschain, S P; Petrasso, R D; Radha, P B; Rice, B; Rosenberg, M J; Schmitt, A J; Schmitt, M J; Seka, W; Shmayda, W T; Shoup, M J; Shvydky, A; Skupsky, S; Solodov, A A; Stoeckl, C; Theobald, W; Ulreich, J; Wittman, M D; Woo, K M; Yaakobi, B; Zuegel, J D

2016-07-08

A record fuel hot-spot pressure P_{hs}=56±7  Gbar was inferred from x-ray and nuclear diagnostics for direct-drive inertial confinement fusion cryogenic, layered deuterium-tritium implosions on the 60-beam, 30-kJ, 351-nm OMEGA Laser System. When hydrodynamically scaled to the energy of the National Ignition Facility, these implosions achieved a Lawson parameter ∼60% of the value required for ignition [A. Bose et al., Phys. Rev. E 93, 011201(R) (2016)], similar to indirect-drive implosions [R. Betti et al., Phys. Rev. Lett. 114, 255003 (2015)], and nearly half of the direct-drive ignition-threshold pressure. Relative to symmetric, one-dimensional simulations, the inferred hot-spot pressure is approximately 40% lower. Three-dimensional simulations suggest that low-mode distortion of the hot spot seeded by laser-drive nonuniformity and target-positioning error reduces target performance.

Tritium Removal by Laser Heating and Its Application to Tokamaks

International Nuclear Information System (INIS)

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

2001-01-01

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

Development of Self-Luminous Glass Tube (SLGT) Manufacturing Technology

International Nuclear Information System (INIS)

Kim, Kwang Sin; Kim, Kyeong Sook; Chung, Eun Su; Song, Kyu Min; Lee, Sook Kyung; Son, Soon Hwan

2005-01-01

Tritium produced from the Wolsong Tritium Removal Facility (WTRF) will be a radioactive waste when it is stored in the vault inside the WTRF, which requires maintenance cost and is a troublesome waste such that it cannot be sent to the radioactive waste disposal facility. However, when tritium is utilized it can be valuable resource for many applications. As a starting point to utilize tritium we tried to domesticate the selfluminous glass tube (SLGT) manufacturing technology. As a hydrogen isotope, tritium has similar chemical properties to hydrogen but slightly different physical properties. Due to its unstable nature, tritium emits beta rays, which are streams of electrons, with 0∼18.6 keV (5.7 keV in average) energies and 12.323 years of a half-life. The energy level of tritium is relatively low and the biological effects of tritium to the human body are not significant, which makes tritium a popular radioactive isotope for use in industries. The electrons in a beta ray collide with phosphor to produce light so that tritium sealed in phosphor coated glass tubes can make the tubes glow without an external supply of energy. To manufacture these SLGTs, 4 core technologies are needed: coating technology, tritium injection technology, laser sealing/cutting technology and tritium handling technology

Tritium research and technology facilities for fusion inside the Bruyeres-le-Chatel Research Center of the French Atomic Energy Commission

International Nuclear Information System (INIS)

Hircq, B.

1990-01-01

Because of a large tritium experience in the Bruyeres-le-Chatel Research Center (Atomic Energy Commission-FRANCE), new activities could be undertaken in 1986 inside the European Fusion Technology Program, especially tritium studies within the frame work of the Next European Torus. After presenting the general tritium research program which concerns the Torus Exhaust Gas Processing (deuterium-tritium purification and storage) and involved materials (weldability of tritium-helium containing steels and corrosion of steels by tritiated water), major obtained results are given before describing the associated equipments. (orig.)

Safety and environmental advantages of using tritium-lean targets for inertial fusion

International Nuclear Information System (INIS)

Arzeni, S.; Latkowski, J. F.; Logan, B. G.; Meier, W. R.; Moir, R. W.; Perkins, L. J.; Sanz, J.

1999-01-01

While traditional inertial fusion energy target designs typically use equimolar portions of deuterium and tritium and have areal densities (ρr) of ∼ 3 g/cm 2 , significant safety and environmental (S and E) advantages may be obtained through the use of high-density (ρr ∼ 10 g/cm 2 ) targets with tritium components as low as 0.5%. Such targets would absorb much of the neutron energy within the target and could be self-sufficient from a tritium breeding point of view. Tritium self-sufficiency within the target would free target chamber designers from the need to use lithium-bearing blanket materials, while low inventories within each target would translate into low inventories in target fabrication facilities. Absorption of much of the neutron energy within the target, the extremely low tritium inventories, and the greatly moderated neutron spectrum, make ''tritium-lean'' targets appear quite attractive from an S and E perspective

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

International Nuclear Information System (INIS)

Konishi, Satoshi; Hayashi, Takumi; Inoue, Masahiko

1993-03-01

JAERI designed, fabricated, and installed the JAERI Fuel Cleanup System (J-FCU) as a subsystem of simulated fusion fuel loop at the TSTA. The main function of the J-FCU is to purify and to recover hydrogen isotopes from simulated plasma exhaust while exhausting tritium free impurities. After a lot of deuterium tests, a first tritium test of the J-FCU was performed with one gram of tritium at the TSTA on June 1991. Main purpose of this test was to evaluate the total integrity and function of the J-FCU system with a DT mixture. Through this test, the J-FCU was operated well and its function with tritium was demonstrated. This report describes the detail test results of the J-FCU first tritium test and discuss its functions by stand-alone mode. Residual tritium inventory of the J-FCU system was also discussed. (author)

Preparation of paraherquamide labeled with deueterium or tritium

Energy Technology Data Exchange (ETDEWEB)

Blizzard, T.A.; Rosegay, A.; Mrozik, H.; Fisher, M.H. (Merck Sharp and Dohme Research Labs., Rahway, NJ (United States))

1990-04-01

Deprotonation of paraherquamide (1a) at C-24 and subsequent deuteration (or tritiation) is described. The procedure afforded 24-[sup 2]H-paraherquamide (1b) with 66% deuterium incorporation at C-24. Modification of the deuteration procedure to allow the introduction of tritium resulted in the preparation of 24-[sup 3]H-paraherquamide (1c) with specific activity 3.7Ci/mmol. (author).

Tritium concentration monitor

International Nuclear Information System (INIS)

Shono, Kosuke.

1991-01-01

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

Develop of omni-tritium sample preparation device

International Nuclear Information System (INIS)

Tian Junhua; Zheng Min; Zhang Dong

2008-06-01

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

Tritium production, management and its impact on safety for a D-3He fusion reactor

International Nuclear Information System (INIS)

Sze, D.K.; Herring, S.; Sawan, M.

1991-11-01

About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

Tritium interactions of potential importance to fusion reactor systems: technology requirements

International Nuclear Information System (INIS)

Wilkes, W.R.

1976-01-01

The tritium technology requirements created by the controlled thermonuclear research program to develop a demonstration fusion power reactor by the year 2000 are reviewed. It is found that the majority of the technological advances which are needed to ensure adequate tritium containment in a tritium breeding power reactor need to be demonstrated on a pilot scale by approximately 1983, so that they may be incorporated into EPR-II, the second of two planned experimental power reactors. The most important advances include development of containment materials with permeabilities to tritium well below measured values for stainless steel; large scale, low inventory deuterium-tritium separation systems; and improved monitoring and assay systems. There are less critical requirements for information about the effects of tritium and helium on the mechanical properties of materials, the effects of tritium on biological systems, and data on physical and chemical properties of tritium. Substantial progress needs to be made on these problems early enough to permit possible solutions to be tested on EPR-I. In addition, major improvements in tritium handling equipment are required for EPR-I. Those technological problems for which solutions have not yet been demonstrated by EPR-II must be solved by 1989 if they are to be assured successful application in the demonstration reactor

Stack and area tritium monitoring systems for the tokamak fusion test reactor (TFTR)

International Nuclear Information System (INIS)

Pearson, G.G.; Meixler, L.D.; Sirsingh, R.A.P.

1992-01-01

This paper reports on the TFTR Tritium Stack and Area Monitoring Systems which have been developed to provide the required level of reliability in a cost effective manner consistent with the mission of the Tritium Handling System on TFTR. Personnel protection, environmental responsibility, and tritium containing system integrity have been the considerations in system design. During the Deuterium-Tritium (D-T) experiments on TFTR, tritium will be used for the first time as one of the fuels. Area monitors provide surveillance of the air in various rooms at TFTR. Stack monitors monitor the air at the TFTR test site that is exhausted through the HVAC systems, from the room exhaust stacks and the tritium systems process vents. The philosophies for the implementation of the Stack and Area Tritium Monitoring Systems at TFTR are to use hardwired controls wherever personnel protection is involved, and to take advantage of modern intelligent controllers to provide a distributed system to support the functions of tracking, displaying, and archiving concentration levels of tritium for all of the monitored areas and stacks

Analysis of LaNi4.25Al0.75 (LANA.75) tritide after five years of tritium exposure

International Nuclear Information System (INIS)

Wermer, J.R.; Holder, J.S.; Mosley, W.C.

1993-01-01

Tritium aging studies have shown that LaNi 4.25 Al 0.75 (LANA .75) tritide storage material undergoes significant degradation with tritium aging. After 5.4 years of dormant storage at full stoichiometry, which is considered a worst-case condition for this material, the performance is still acceptable for SRS tritium processing applications. The isotherms change, decreasing the desorption pressures, increasing the isotherm plateau slopes, and decreasing the total storage capacity. Eventually, the material will degrade with time to the point where it may no longer be useful for tritium processing applications. At the end of life, the tritium heel can be exchanged with protium or deuterium to produce a final material containing very little tritium

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.

Advanced design of the Mechanical Tritium Pumping System for JET DTE2

International Nuclear Information System (INIS)

Giegerich, T.; Bekris, N.; Camp, P.; Day, Chr.; Gethins, M.; Lesnoy, S.; Luo, X.; Müller, R.; Ochoa, S.; Pfeil, P.; Smith, R.; Strobel, H.; Stump, H.

2016-01-01

For tritium processing in JET during the next Deuterium-Tritium-Experiment (DTE2), a fully tritium compatible and continuously working vacuum pumping system has been developed. This pump train will be used as roughing pump to cover a pressure regime between 10 −1 Pa and ambient pressure. Therefore, a two-stage liquid ring pump in combination with a booster vapor diffusion pump will be applied. In this paper, a close-to-final design of the pumps is being described. Finite element (FEM) simulation results of components where high mechanical stresses due to thermal gradients are expected are presented. Furthermore, the final design of the control and data acquisition system is shown and explained.

Advanced design of the Mechanical Tritium Pumping System for JET DTE2

Energy Technology Data Exchange (ETDEWEB)

Giegerich, T., E-mail: thomas.giegerich@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bekris, N. [EUROfusion Program Management Unit (PMU), ITER Physics Department, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Camp, P. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Day, Chr. [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gethins, M.; Lesnoy, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Luo, X.; Müller, R.; Ochoa, S.; Pfeil, P. [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Smith, R. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Strobel, H.; Stump, H. [Karlsruhe Institute of Technology (KIT), Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2016-11-01

For tritium processing in JET during the next Deuterium-Tritium-Experiment (DTE2), a fully tritium compatible and continuously working vacuum pumping system has been developed. This pump train will be used as roughing pump to cover a pressure regime between 10{sup −1} Pa and ambient pressure. Therefore, a two-stage liquid ring pump in combination with a booster vapor diffusion pump will be applied. In this paper, a close-to-final design of the pumps is being described. Finite element (FEM) simulation results of components where high mechanical stresses due to thermal gradients are expected are presented. Furthermore, the final design of the control and data acquisition system is shown and explained.

Deuterium trapping in tungsten

Science.gov (United States)

Poon, Michael

Tungsten is one of the primary material candidates being investigated for use in the first-wall of a magnetic confinement fusion reactor. An ion accelerator was used to simulate the type of ion interaction that may occur at a plasma-facing material. Thermal desorption spectroscopy (TDS) was the primary tool used to analyze the effects of the irradiation. Secondary ion mass spectroscopy (SIMS) was used to determine the distribution of trapped D in the tungsten specimen. The tritium migration analysis program (TMAP) was used to simulate thermal desorption profiles from the D depth distributions. Fitting of the simulated thermal desorption profiles with the measured TDS results provided values of the D trap energies. Deuterium trapping in single crystal tungsten was studied as a function of the incident ion fluence, ion flux, irradiation temperature, irradiation history, and surface impurity levels during irradiation. The results show that deuterium was trapped at vacancies and voids. Two deuterium atoms could be trapped at a tungsten vacancy, with trapping energies of 1.4 eV and 1.2 eV for the first and second D atoms, respectively. In a tungsten void, D is trapped as atoms adsorbed on the inner walls of the void with a trap energy of 2.1 eV, or as D2 molecules inside the void with a trap energy of 1.2 eV. Deuterium trapping in polycrystalline tungsten was also studied as a function of the incident fluence, irradiation temperature, and irradiation history. Deuterium trapping in polycrystalline tungsten also occurs primarily at vacancies and voids with the same trap energies as in single crystal tungsten; however, the presence of grain boundaries promotes the formation of large surface blisters with high fluence irradiations at 500 K. In general, D trapping is greater in polycrystalline tungsten than in single crystal tungsten. To simulate mixed materials comprising of carbon (C) and tungsten, tungsten specimens were pre-irradiated with carbon ions prior to D

Deuterium trapping in tungsten

International Nuclear Information System (INIS)

Poon, M.

2004-01-01

Tungsten is one of the primary material candidates being investigated for use in the first-wall of a magnetic confinement fusion reactor. An ion accelerator was used to simulate the type of ion interaction that may occur at a plasma-facing material. Thermal desorption spectroscopy (TDS) was the primary tool used to analyze the effects of the irradiation Secondary ion mass spectroscopy (SIMS) was used to determine the distribution of trapped D in the tungsten specimen. The tritium migration analysis program (TMAP) was used to simulate thermal desorption profiles from the D depth distributions. Fitting of the simulated thermal desorption profiles with the measured TDS results provided values of the D trap energies. . Deuterium trapping in single crystal tungsten was studied as a function of the incident ion fluence, ion flux, irradiation temperature, irradiation history, and surface impurity levels during irradiation The results show that deuterium was trapped at vacancies and voids. Two deuterium atoms could be trapped at a tungsten vacancy, with trapping energies of 1.4 eV and 1.2 eV for the first and second D atoms, respectively. In a tungsten void, D is trapped as atoms adsorbed on the inner walls of the void with a trap energy of 2.1 eV, or as D 2 molecules inside the void with a trap energy of 1.2 eV. . Deuterium trapping in polycrystalline tungsten was also studied as a function of the incident fluence, irradiation temperature, and irradiation history. Deuterium trapping in polycrystalline tungsten also occurs primarily at vacancies and voids with the same trap energies as in single crystal tungsten; however, the presence of grain boundaries promotes the formation of large surface blisters with high fluence irradiations at 500 K. In general, D trapping is greater in polycrystalline tungsten than in single crystal tungsten. To simulate mixed materials comprising of carbon (C) and tungsten, tungsten specimens were pre-irradiated with carbon ions prior to D

Deuterium trapping in tungsten

Energy Technology Data Exchange (ETDEWEB)

Poon, M

2004-07-01

Tungsten is one of the primary material candidates being investigated for use in the first-wall of a magnetic confinement fusion reactor. An ion accelerator was used to simulate the type of ion interaction that may occur at a plasma-facing material. Thermal desorption spectroscopy (TDS) was the primary tool used to analyze the effects of the irradiation Secondary ion mass spectroscopy (SIMS) was used to determine the distribution of trapped D in the tungsten specimen. The tritium migration analysis program (TMAP) was used to simulate thermal desorption profiles from the D depth distributions. Fitting of the simulated thermal desorption profiles with the measured TDS results provided values of the D trap energies. . Deuterium trapping in single crystal tungsten was studied as a function of the incident ion fluence, ion flux, irradiation temperature, irradiation history, and surface impurity levels during irradiation The results show that deuterium was trapped at vacancies and voids. Two deuterium atoms could be trapped at a tungsten vacancy, with trapping energies of 1.4 eV and 1.2 eV for the first and second D atoms, respectively. In a tungsten void, D is trapped as atoms adsorbed on the inner walls of the void with a trap energy of 2.1 eV, or as D{sub 2} molecules inside the void with a trap energy of 1.2 eV. . Deuterium trapping in polycrystalline tungsten was also studied as a function of the incident fluence, irradiation temperature, and irradiation history. Deuterium trapping in polycrystalline tungsten also occurs primarily at vacancies and voids with the same trap energies as in single crystal tungsten; however, the presence of grain boundaries promotes the formation of large surface blisters with high fluence irradiations at 500 K. In general, D trapping is greater in polycrystalline tungsten than in single crystal tungsten. To simulate mixed materials comprising of carbon (C) and tungsten, tungsten specimens were pre-irradiated with carbon ions prior to D

Chromatographic measurement of hydrogen isotopic and permanent gas impurities in tritium

International Nuclear Information System (INIS)

Warner, D.K.; Kinard, C.; Bohl, D.C.

1976-01-01

This paper describes a gas chromatograph that was designed for dedicated analysis of hydrogen isotopic and permanent gas impurities in tritium and tritium-deuterium mixtures. The instrument that was developed substantially improved the accuracy and precision of hydrogen isotopic analysis in the 20 ppM to one mole percent range as compared with other analytical methods. Several unique design features of the instrument were required due to the radiation and isotopic exchange properties of the tritium in the samples; descriptions of these features are presented along with details of the complete chromatographic system. The experimental procedures used to calibrate the detector and statistically evaluate its performance are given, and the sources of analytical error are cited. The limitations of the present system are also discussed

Tritium Systems Test Assembly: design for major device fabrication review

International Nuclear Information System (INIS)

Anderson, J.L.; Sherman, R.H.

1977-06-01

This document has been prepared for the Major Device Fabrication Review for the Tritium Systems Test Assembly (TSTA). The TSTA is dedicated to the development, demonstration, and interfacing of technologies related to the deuterium-tritium fuel cycle for fusion reactor systems. The principal objectives for TSTA are: (a) demonstrate the fuel cycle for fusion reactor systems; (b) develop test and qualify equipment for tritium service in the fusion program; (c) develop and test environmental and personnel protective systems; (d) evaluate long-term reliability of components; (e) demonstrate long-term safe handling of tritium with no major releases or incidents; and (f) investigate and evaluate the response of the fuel cycle and environmental packages to normal, off-normal, and emergency situations. This document presents the current status of a conceptual design and cost estimate for TSTA. The total cost to design, construct, and operate TSTA through FY-1981 is estimated to be approximately $12.2 M

The introduction of tritium label into natural and modified prostaglandins

International Nuclear Information System (INIS)

Shevchenko, V.P.; Bezuglov, V.V.; Nagayev, I.Y.; Myasoedov, N.F.

1989-01-01

Studies on the role of the nature of both heterogeneous catalysts and the solvent on the reduction selectively of 5,6-double bonds showed that the largest yield could be obtained by using the Lindlar catalyst and ethyl acetate. The use of different isotopes of hydrogen in the protium-deuterium-tritium series markedly decreased the hydrogenation reaction rate, but the selectivity of the process practically remained unaltered. Homogeneous catalysts were also used in the production of natural tritium-labelled prostaglandins and of their fluorine and deoxy analogues. The label was introduced by selective hydrogenation in the presence of (Ph 3 P) 3 RhCl and dioxane as solvent. Different ways have been studied of tritium-label introduction into prostaglandins modified at the carboxyl group. The synthesis of similar preparations was performed either by selective dehalogenation in the presence of heterogeneous catalysts treated with quinoline or triethylamine, or by condensation of prostaglandins at the carboxyl group by tritium-labelled amino acid. (author). 4 refs.; 1 fig

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

Helium and deuterium permeability in O-rings

International Nuclear Information System (INIS)

Lakner, J.F.

1976-01-01

To obtain more information on gas permeation through elastomeric O-rings, studies were performed on Parker Seal Company O-rings, Nos. 2-113, 2-006, 3-904, and 3-906, all made of a nitrile rubber. Also included in the tests was a valve packing (Autoclave Engineers) encased in AE Valve 20A-2142. Permeation experiments were run usually in duplicate to 82.7 MPa (12,000 psi) with helium and deuterium at room temperature. The data are extrapolated to give values for tritium

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

International Nuclear Information System (INIS)

Quanci, J.F.

1989-01-01

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

Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Nakaya, Hiroyuki, E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, Hideaki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Katayama, Kazunari [Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-koen, Kasuga 8168580 (Japan); Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan)

2015-10-15

Highlights: • Tritium production by a high-temperature gas-cooled reactor was studied. • The loading method considering tritium outflow suppression was estimated. • A reactor with 600 MWt produced 400–600 g of tritium for 180 days. • A possibility that tritium outflow can be sufficiently suppressed was shown. - Abstract: Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al{sub 2}O{sub 3} cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300 MW electrical nominal capacity (GTHTR300) with 600 MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600 g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

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)

Tritium removal: a preliminary evaluation of several getters

International Nuclear Information System (INIS)

Schoenfelder, C.W.; West, L.A.

1975-11-01

The removal of hydrogen isotopes from flowing gas streams is an important aspect of CTR technology for both decontamination and tritium recovery from plasma exhausts. Several getters have been evaluated for their tritium scrubbing potential at the parts per billion level. Measurements of total capacity and dynamic response have been made for barium, erbium, palladium dispersed on molecular sieve, General Electric H-36 (zirconium alloy), Union Carbide Y-993 (PdMnO 2 ), Societa Apparecchi Electtrici e Scientifici Getters ST101 (Zr--Al), ST171, and ST181, and a Sandia developed organic material, dimerized phenyl propargyl ether (DPPE). Preliminary flow studies were conducted by passing mixtures of either hydrogen or deuterium diluted with argon through packed beds containing the getter and periodically sampling the effluent with a gas chromatograph sensitive to 500 ppB H 2 . The results of this work, similar flow experiments using tritium and total capacity measurements are presented in the text

Safety aspects of tritium in ICF reactors with internally-breeding targets

International Nuclear Information System (INIS)

Ragheb, M.; Miley, G.H.; University of Illinois, Urbana, IL)

1985-01-01

The LOTRIT inertial confinement reactor concept employs a deuterium burning target with a DT spark trigger core. This eliminates the need for tritium breeding in a blanket, and leads to a minimization of the tritium inventory and of the possibility of metal fire hazards if lead is used instead of lithium for first wall protection. The active fuel inventory in the fuel cycle and blanket per MJ of energy produced is only 5 percent of the DT case. The most significant reduction in the total tritium inventory is in the target manufacture and storage areas, and is about 1.8% of the DT case per unit of fusion energy produced. If the goal is to reduce the risk from tritium releases from fusion reactors to below that of fission reactors, it is estimated that the tritium releases must be maintained at 0.13-5.0 Ci/day. Attaining these values will be costly, technologically difficult and will constrain the design options in DTbased systems, but may be within the realm of systems using the LOTRIT concept

Deuterium diffusion along the three principal directions in anisotropic Zr–2.5Nb

Energy Technology Data Exchange (ETDEWEB)

Khatamian, D., E-mail: khatamiand@aecl.ca

2013-12-15

Highlights: •Diffusivity of deuterium in a Zr–2.5Nb plate was measured. •The plate had similar microstructure to that of a CANDU pressure tube. •Diffusion coefficients in the radial direction were 24% less than in the longitudinal direction. •These findings are in close agreement with the theoretical estimates made in the literature. -- Abstract: Pressure tubes of cold-worked Zr–2.5Nb material are used in the CANDU (CANada Deuterium Uranium is a registered trademark of Atomic Energy of Canada Ltd.) reactors to contain the fuel bundles and the heavy water coolant. The pressure tube microstructure consists of α-Zr platelets, with an aspect ratio of 1:5:50 in the radial, transverse and longitudinal directions, surrounded by a thin layer of β-Zr. The diffusivity of hydrogen in β-Zr is higher than in α-Zr. As a result, the diffusivity of hydrogen in the pressure tube material is enhanced compared to unalloyed α-Zr. Due to the anisotropic microstructure of the pressure tube with respect to the β-Zr network, the diffusivity of hydrogen is thought to be different in the three principal pressure tube directions. Measurements made using specimens machined from an 8 mm thick Zr–2.5Nb plate, with a microstructure similar to that of a CANDU pressure tube, suggest the difference between the diffusion constants for deuterium in different directions of an as-installed pressure tube may be as high as 25%.

A Visual Detection System for Determining Tritium Surface Deposition Employing Phosphor Coated Materials

International Nuclear Information System (INIS)

Gentile, C.A.; Skinner, C.H.; Young, K.M.; Zweben, S.J.

1999-01-01

A method for visually observing tritium deposition on the surface of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles is being investigated at the Princeton Plasma Physics Laboratory. A green phosphor (P31, zinc sulfide: copper) similar to that used in oscilloscope screens with a wavelength peak of 530 nm was positioned on the surface of a TFTR D-T tile. The approximately 600 gram tile, which contains approximately 1.5 Ci of tritium located on the top approximately 1-50 microns of the surface, was placed in a two liter lexan chamber at Standard Temperature and Pressure (STP). The phosphor plates and phosphor powder were placed on the surface of the tile which resulted in visible light being observed, the consequence of tritium betas interacting with the phosphor. This technique provides a method of visually observing varying concentrations of tritium on the surface of D-T carbon tiles, and may be employed (in a calibrated system) to obtain quantitative data

The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

International Nuclear Information System (INIS)

Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse.

1988-03-01

We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs

A convenient calibrator for tritium survey meters

International Nuclear Information System (INIS)

Cutler, J.M.; Janmohamed, S.R.; Surette, R.A.; Wood, M.J.

1993-06-01

Staff in Ontario Hydro's CANDU nuclear power stations were having difficulty calibrating their tritium-in-air survey meters. Chalk River Laboratories suggested using a standard silica gel collector loaded with tritiated water vapour as a radiation source. The collector is a polyethylene tube with a length of rubber tubing attached to each end. The plastic tube is filled with dry silica gel, and cotton wool is stuffed into each end to retain the gel. To load the source with tritiated vapour, a bubbler containing tritiated water is connected to the collector until the gel is completely pink. The end tubes of the source are then connected to each other to seal the source. To use the source, the rubber tubes are connected to the intake and exhaust fittings of the survey meter to be calibrated. No tritium vapour is released during loading or calibration. This source not only verifies that the ion chamber and electrometer are working, but also checks whether the air sample is reaching the detector. It is more direct and credible than a gamma source, and is inexpensive and convenient to construct and calibrate at the power station. 5 figs

Tritium enrichment by thermal diffusion. I. Calculation of an installation for measuring natural tritium; Enrichissement du tritium par diffusion thermique. - I. Calcul d'une installation destinee a la mesure du tritium naturel

Energy Technology Data Exchange (ETDEWEB)

Shimizu, M; Ravoire, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-07-01

The natural content of tritium is so low that its measurement generally requires a preliminary enrichment. The thermal diffusion on hydrogen is studied as an enrichment method. The installation studied comprises two stages of columns of the hot-wire type, together with a device for transferring the tritium from the water sample into the hydrogen in the columns using catalytic exchange. A complete mathematical treatment for the operation of such a unit has been made and programmed for the IBM 7094 computer. An optimization has been effected by means of this program. It is shown that for similar performances, less hydrogen is retained in the case of hot-wire type columns than in the case of columns composed of concentric tubes. (authors) [French] La teneur naturelle du tritium est si faible que sa mesure demande generalement un enrichissement prealable. On etudie la diffusion thermique sur l'hydrogene comme moyen d'enrichissement. L'installation que l'on etudie comprend deux etages de colonnes du type fil chaud, et un dispositif de transfert du tritium de l'echantillon d'eau dans l'hydrogene des colonnes par echange catalytique. Un traitement mathematique complet du fonctionnement d'un tel ensemble a ete etabli et programme sur machine IBM 7094. Une optimisation a ete faite a l'aide du programme. On montre egalement qu'a performances egales, la retenue d'hydrogene est plus faible dans le cas des colonnes du type fil chaud que dans le cas des colonnes du type tubes concentriques. (auteurs)

Deuterium transport and trapping in polycrystalline tungsten

International Nuclear Information System (INIS)

Anderl, R.A.; Holland, D.F.; Longhurst, G.R.; Pawelko, R.J.; Trybus, C.L.; Sellers, C.H.

1992-01-01

This paper reports that deuterium permeation studies for polycrystalline tungsten foil have been conducted to provide data for estimating tritium transport and trapping in tungsten-clad divertors proposed for advanced fusion-reactor concepts. Based on a detailed transmission electron microscopy (TEM) microstructural characterization of the specimen material and on analyses of permeation data measured at temperatures ranging form 610 to 823 K for unannealed and annealed tungsten foil (25 μm thick), the authors note the following key results: deuterium transport in tungsten foil is dominated by extensive trapping that varies inversely with prior anneal temperatures of the foil material, the reduction in the trapped fraction correlates with a corresponding elimination of a high density of dislocations in cell-wall structures introduced during the foil fabrication process, trapping behavior in these foils can be modelled using trap energies between 1.3 eV and 1.5 eV and trap densities ranging from 1 x 10 -5 atom fraction

Tritium and radon risks for humans

International Nuclear Information System (INIS)

Mauna, Traian; Mauna, Andriesica

2008-01-01

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

Tritium labeling by thermally generated tritons

International Nuclear Information System (INIS)

Morimoto, H.; Williams, P.G.; Saljoughian, M.

1988-06-01

The predominant effect of thermal atom irradiation on solid molecules is saturation of their aromatic functions. Only low level of tritium exchange is observed for aliphatic solids. In contrast, liquids whose frozen surface can be rendered somewhat mobile at appropriate temperatures exhibit more exchange than addition. The rank order of effectiveness of several metals in promoting exchange/addition appears similar to the rank order for heterogeneous catalytic hydrogenation. 3 refs., 8 figs

Effect of the self-pumped limiter concept on the tritium fuel cycle

International Nuclear Information System (INIS)

Finn, P.A.; Sze, D.K.; Hassanein, A.

1988-01-01

The self-pumped limiter concept for impurity control of the plasma of a fusion reactor has a major impact on the design of the tritium systems. To achieve a sustained burn, conventional limiters and divertors remove large quantities of unburnt tritium and deuterium from the plasma which must be then recycled using a plasma processing system. The self-pumped limiter which does not remove the hydrogen species, does not require any plasma processing equipment. The blanket system and the coolant processing systems acquire greater importance with the use of this unconventional impurity control system. 3 refs., 2 figs

Tritium enrichment by thermal diffusion. I. Calculation of an installation for measuring natural tritium; Enrichissement du tritium par diffusion thermique. - I. Calcul d'une installation destinee a la mesure du tritium naturel

Energy Technology Data Exchange (ETDEWEB)

Shimizu, M.; Ravoire, J. [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

1966-07-01

The natural content of tritium is so low that its measurement generally requires a preliminary enrichment. The thermal diffusion on hydrogen is studied as an enrichment method. The installation studied comprises two stages of columns of the hot-wire type, together with a device for transferring the tritium from the water sample into the hydrogen in the columns using catalytic exchange. A complete mathematical treatment for the operation of such a unit has been made and programmed for the IBM 7094 computer. An optimization has been effected by means of this program. It is shown that for similar performances, less hydrogen is retained in the case of hot-wire type columns than in the case of columns composed of concentric tubes. (authors) [French] La teneur naturelle du tritium est si faible que sa mesure demande generalement un enrichissement prealable. On etudie la diffusion thermique sur l'hydrogene comme moyen d'enrichissement. L'installation que l'on etudie comprend deux etages de colonnes du type fil chaud, et un dispositif de transfert du tritium de l'echantillon d'eau dans l'hydrogene des colonnes par echange catalytique. Un traitement mathematique complet du fonctionnement d'un tel ensemble a ete etabli et programme sur machine IBM 7094. Une optimisation a ete faite a l'aide du programme. On montre egalement qu'a performances egales, la retenue d'hydrogene est plus faible dans le cas des colonnes du type fil chaud que dans le cas des colonnes du type tubes concentriques. (auteurs)

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%

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Tritium decontamination of TFTR carbon tiles employing ultra violet light

International Nuclear Information System (INIS)

Shu, W.M.; Ohira, S.; Gentile, C.A.; Oya, Y.; Nakamura, H.; Hayashi, T.; Iwai, Y.; Kawamura, Y.; Konishi, S.; Nishi, M.F.; Young, K.M.

2001-01-01

Tritium decontamination on the surface of Tokamak Fusion Test Reactor (TFTR) bumper limiter tiles used during the Deuterium-Deuterium (D-D) phase of TFTR operations was investigated employing an ultra violet light source with a mean wavelength of 172 nm and a maximum radiant intensity of 50 mW/cm 2 . The partial pressures of H 2 , HD, C and CO 2 during the UV exposure were enhanced more than twice, compared to the partial pressures before UV exposure. In comparison, the amount of O 2 decreased during the UV exposure and the production of a small amount of O 3 was observed when the UV light was turned on. Unlike the decontamination method of baking in air or oxygen, the UV exposure removed hydrogen isotopes from the tile to vacuum predominantly in forms of gases of hydrogen isotopes. The tritium surface contamination on the tile in the area exposed to the UV light was reduced after the UV exposure. The results show that the UV light with a wavelength of 172 nm can remove hydrogen isotopes from carbon-based tiles at the very surface

Ignition of deuterium based fuel cycles in a high beta system

International Nuclear Information System (INIS)

Hirano, K.

1987-01-01

A steady state self-consistent plasma modeling applied to a system having close to unity, such as FRC or like, is found to be quite effective in solving the problems independently of any anomalous process and proves the existence of ignited state of deuterium based fuel cycles. The temperature ranges that the plasma falls into ignited state are obtained as a function of relative feeding rates of tritium and 3 He to deuterium's. We find pure DD cycle will not ignite so that 3 He or/and tritium must be added as catalyzer to achieve ignition. Standing on the points to construct a cleaner system yielding smaller amount of 14 MeV neutrons and to burn the fuel in steady state for long periods of time, we have confirmed superiority of the complex composed of the master reactor of 3 He-Cat.D cycle (catalyzed DD cycle reinjecting only fusion produced 3 He) and the satellite reactor of 3 He enriched D 3 He cycle. In case storage of tritium for 3 He by β - decay is turned out not to be allowed environmentally, we may utilize conventional catalyzed DD cycle although 14 MeV neutron yields will be increased by 35 % over the complex. It is demonstrated that advanced fuel cycle reactors can be very simple in constructions and compact in size such that the field strength and the plasma volume of the order of JT-60's may be enough for 1000 MW power plant. (author)

Tritium recovery and separation from CTR plasma exhausts and secondary containment atmospheres

International Nuclear Information System (INIS)

Forrester, R.C. III; Watson, J.S.

1975-01-01

Recent experimental successes have generated increased interest in the development of thermonuclear reactors as power sources for the future. This paper examines tritium containment problems posed by an operating CTR and sets forth some processing schemes currently being evaluated at the Oak Ridge National Laboratory. An appreciation of the CTR tritium management problem can best be realized by recalling that tritium production rates for various fission reactors range from 2 x 10 4 to 9 x 10 5 Ci/yr per 1000 MW(e). Present estimates of tritium production in a CTR blanket exceed 10 9 Ci/yr for the same level of power generation, and tritium process systems may handle 10 to 20 times that amount. Tritium's high permeability through most materials of construction at high temperatures makes secondary containment mandatory for most piping. Processing of these containment atmospheres will probably involve conversion of the tritium to a nonpermeating form (T 2 O) followed by trapping on conventional beds of desiccant material. In a similar fashion, all purge streams and process fluid vent gases will be subjected to tritium recovery prior to atmospheric release. Two tritium process systems will be required, one to recover tritium produced by breeding in the blanket and another to recover unburned tritium in the plasma exhaust. Plasma exhaust processing will be unconventional since the exhaust gas pressure will lie between 10 -3 and 10 -6 torr. Treatment of this gas stream will entail the removal of small quantities of protium and helium from a much larger deuterium-tritium mixture which will be recycled. (U.S.)

Thermally induced outdiffusion studies of deuterium in ceramic breeder blanket materials after irradiation

Energy Technology Data Exchange (ETDEWEB)

González, Maria, E-mail: maria.gonzalez@ciemat.es [LNF-CIEMAT, Materials for Fusion Group, Madrid (Spain); Carella, Elisabetta; Moroño, Alejandro [LNF-CIEMAT, Materials for Fusion Group, Madrid (Spain); Kolb, Matthias H.H.; Knitter, Regina [Karlsruhe Institute of Technology, Institute for Applied Materials (IAM-WPT), Karlsruhe (Germany)

2015-10-15

Highlights: • Surface defects in Lithium-based ceramics are acting as trapping centres for deuterium. • Ionizing radiation affects the deuterium sorption and desorption processes. • By extension, the release of the tritium produced in a fusion breeder will be effective. - Abstract: Based on a KIT–CIEMAT collaboration on the radiation damage effects of light ions sorption/desorption in ceramic breeder materials, candidate materials for the ITER EU TBM were tested for their outgassing behavior as a function of temperature and radiation. Lithium orthosilicate based pebbles with different metatitanate contents and pellets of the individual oxide components were exposed to a deuterium atmosphere at room temperature. Then the thermally induced release of deuterium gas was registered up to 800 °C. This as-received behavior was studied in comparison with that after exposing the deuterium-treated samples to 4 MGy total dose of gamma radiation. The thermal desorption spectra reveal differences in deuterium sorption/desorption behavior depending on the composition and the induced ionizing damage. In these breeder candidates, strong desorption rate at approx. 300 °C takes place, which slightly increases with increasing amount of the titanate second phase. For all studied materials, ionizing radiation induces electronic changes disabling a number of trapping centers for D{sub 2} adsorption.

Measurement of HTO-HT in atmosphere using the Draeger (Drager) tube

International Nuclear Information System (INIS)

Huang, Z.; Harris, S.J.; Giorgio, P.R.

1993-01-01

It is clearly that, incorporation with the liquid scintillation counting, the Drager tube measuring system for Hydrogen allows the measurement of tritium contamination in atmosphere, e.g. HTO/HT (tritiated water vapour/tritium gas) simultaneously, at levels well below those derived air concentrations (DACs) prescribed by The International Commission on Radiological Protection (ICRP) and the UK Regulations. The novel tritium measuring system of Drager tube, in liquid phase or gel phase, can be used as a cheap, convenient and quick method of tritium detection

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

Gaseous Tritium Light Sources in armament and watches industries

International Nuclear Information System (INIS)

Amme, Marcus; Siegenthaler, Roger

2015-01-01

The industrial application of Tritium gas enclosed in glass tubes is a modern way illuminating instruments and items wherever instant and independent readability is prerequisite. The GTLS (Gaseous Tritium Light Sources) technology follows the principle of radiation-induced luminescence and supersedes the luminous radioactive paints and their hazards such as particles erasure or heavy isotope use. Enclosure of tritium in glass is a demanding micro technology process and work needs to be performed in controlled areas due to handling of open sources. The storage and transport of the Tritium is done via licensed B(U)-containers coming from heavy water reactor sites, and disposal of radioactive Tritium wastes has to be compliant with national and international regulations for transport and waste management.

Development of organic tritium light technology at Ontario Hydro

International Nuclear Information System (INIS)

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

1992-01-01

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

Measurement of Tritium Activity in Plants by Ice Extraction Method

International Nuclear Information System (INIS)

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

2014-01-01

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

Conceptual design of an emergency tritium clean-up system

International Nuclear Information System (INIS)

Muller, M.E.

1978-01-01

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

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

International Nuclear Information System (INIS)

Quanci, J.F.

1989-01-01

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

Tritium permeation and recovery for the Flibe/He blanket design

International Nuclear Information System (INIS)

Moir, R.

1984-10-01

This study assumes tritium to be a gas dissolved in molten salt, with TF formation suppressed. Tritium permeates readily through the hot steel tubes of the reactor and steam generator and will leak into the steam system at the rate of about one gram per day in the absence of special permeation barriers, assuming that 1% of the helium coolant flow rate is processed for tritium recovery at 90% efficiency per pass. Tritiated water in the steam system is a personnel hazard at concentration levels well below one part per million and this level would soon be reached without costly isotopic processing. Alternatively, including a combination of permeation barriers on reactor and steam generator tubes and molten salt processing is estimated to reduce the leak rate into the steam system by over two orders of magnitude. For the option with the lowest estimated leak rate, 55 Ci/d, it may be possible to purge the steam system continuously to prevent tritiated water buildup. At best, isotopic separation of dilute tritiated water may not be necessary and for higher leak-rate options the isotopic processing rate can be reduced. The proposed permeation barrier for the reactor tubes is a 10 μm layer of tungsten which, in principle, will reduce tritium blanket permeation by a factor of about 300 below the bare-steel rate

Methods of removal of tritium from aqueous effluent: a review of international research and development

International Nuclear Information System (INIS)

Segal, M.G.

1988-01-01

Tritium is formed in thermal nuclear reactors both by neutron activation of elements such as deuterium and lithium and by ternary fission in the fuel. It is a weak beta-emitter with a short half-life, 12.3 years, and its radiological significance in reactor discharges is very low. In heavy-water-cooled and -moderated reactors, such as the CANDU stations, the tritium concentration in the moderator is sufficiently high to cause a potential hazard to operators, and so a major research and development programme has been carried out on processes to remove the tritium. Detritiation of light water has also been the subject of major R and D effort world-wide, because reprocessing operations can generate significant quantities of tritium in liquid waste, and high concentrations of tritium may arise in some aqueous streams in fusion reactors. This Report presents a review of the methods that have been proposed, studied and developed for removal of tritium from light and heavy water: the principles of individual methods are discussed, and the current status of their development is reviewed. (author)

Scanning drift tube measurements of electron transport parameters in different gases: argon, synthetic air, methane and deuterium

International Nuclear Information System (INIS)

Korolov, I; Vass, M; Donkó, Z

2016-01-01

Measurements of transport coefficients of electrons in a scanning drift tube apparatus are reported for different gases: argon, synthetic air, methane and deuterium. The experimental system allows the spatio-temporal development of the electron swarms (‘swarm maps’) to be recorded and this information, when compared with the profiles predicted by theory, makes it possible to determine the ‘time-of-flight’ transport coefficients: the bulk drift velocity, the longitudinal diffusion coefficient and the effective ionization coefficient, in a well-defined way. From these data, the effective Townsend ionization coefficient is determined as well. The swarm maps provide, additionally, direct, unambiguous information about the hydrodynamic/non-hydrodynamic regimes of the swarms, aiding the selection of the proper regions applicable for the determination of the transport coefficients. (paper)

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

Process and device for decontamination of the waste gas of the fuel circuit of a fusion reactor from tritium and/or deuterium in waste gas containing them in chemically bound form

International Nuclear Information System (INIS)

Penzhorn, R.D.; Glugla, M.

1987-01-01

The invention concerns a process and a device for the decontamination of the wate gases of the fuel circuit of a fusion reactor from tritum and/or deuterium in waste gas containing them in chemically bound form, in which the waste gas is taken over an oxidation catalyst and then over a hot metal bed, tritium and/or deuterium is released from its compounds, separated from the waste gas and is returned to the fuel circuit. The process is intended to prevent losses of tritum and/or deuterium by permeation and the high loading of the hot metal getter materials, as occurs in the previously known corresponding process, and to avoid the formation of nitrogen oxides. This is achieved by: a) The catalytic oxidation reaction being carried out at a temperature of 200 0 C to 300 0 C. b) The gas mixture then being brought into contact with a hot metal bed at 200 0 C to 300 0 C to remove the remaining O 2 and for the selective conversion of the proportion of water into the hydrogen isotope. c) The gas mixture being brought into contact with a diaphragm made of palladium or a palladium-silver alloy at 400 0 C to 450 0 C to decompose the ammonia, all the released hydrogen isotope being passed through the diaphragm, separated from the remaining waste gas flow and removed. (orig.) [de

TMIST-2 Transport Test

International Nuclear Information System (INIS)

Glen R Longhurst

2008-01-01

In anticipation of the TMIST-2 experiment in the Advanced Test Reactor, there was a need to determine if the tritium that is expected to be observed at the outlet of the experiment would be seen or if it may be lost on its way from the experiment in the core to the measurement station. To assist in resolving that issue, a bench-scale experiment was conducted in the Idaho National Laboratory's Safety and Tritium Applied Research (STAR) facility using deuterium and a mass spectrometer in lieu of tritium with ion chambers, bubblers, and scintillation counting. The experiment replicated the concentration of the hydrogen isotope, the flow rates anticipated, and the residence times. It was found that there was initial uptake on tubing walls, presumably due to oxidation of the hydrogen isotopes to water and adsorption or isotopic exchange, but that saturates relatively quickly, and once saturated, the concentration of deuterium at the outlet of the tubing system was essentially the same as it was at the experiment inlet under the conditions modeled in the experiment

Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

International Nuclear Information System (INIS)

Shimada, Masashi; Hara, Masanori; Otsuka, Teppei; Oya, Yasuhisa; Hatano, Yuji

2015-01-01

Three tungsten samples irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to deuterium plasma (ion fluence of 1 × 10 26 m −2 ) at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy was performed with a ramp rate of 10 °C min −1 up to 900 °C, and the samples were annealed at 900 °C for 0.5 h. These procedures were repeated three times to uncover defect-annealing effects on deuterium retention. The results show that deuterium retention decreases approximately 70% for at 500 °C after each annealing, and radiation damages were not annealed out completely even after the 3rd annealing. TMAP modeling revealed the trap concentration decreases approximately 80% after each annealing at 900 °C for 0.5 h

Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

Science.gov (United States)

Shimada, Masashi; Hara, Masanori; Otsuka, Teppei; Oya, Yasuhisa; Hatano, Yuji

2015-08-01

Three tungsten samples irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to deuterium plasma (ion fluence of 1 × 1026 m-2) at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy was performed with a ramp rate of 10 °C min-1 up to 900 °C, and the samples were annealed at 900 °C for 0.5 h. These procedures were repeated three times to uncover defect-annealing effects on deuterium retention. The results show that deuterium retention decreases approximately 70% for at 500 °C after each annealing, and radiation damages were not annealed out completely even after the 3rd annealing. TMAP modeling revealed the trap concentration decreases approximately 80% after each annealing at 900 °C for 0.5 h.

Defect annealing and thermal desorption of deuterium in low dose HFIR neutron-irradiated tungsten

Energy Technology Data Exchange (ETDEWEB)

Shimada, Masashi, E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Hara, Masanori [Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan); Otsuka, Teppei [Kyushu University, Interdisciplinary Graduate School of Engineering Science, Higashi-ku, Fukuoka (Japan); Oya, Yasuhisa [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan); Hatano, Yuji [Hydrogen Isotope Research Center, University of Toyama, Toyama (Japan)

2015-08-15

Three tungsten samples irradiated at High Flux Isotope Reactor at Oak Ridge National Laboratory were exposed to deuterium plasma (ion fluence of 1 × 10{sup 26} m{sup −2}) at three different temperatures (100, 200, and 500 °C) in Tritium Plasma Experiment at Idaho National Laboratory. Subsequently, thermal desorption spectroscopy was performed with a ramp rate of 10 °C min{sup −1} up to 900 °C, and the samples were annealed at 900 °C for 0.5 h. These procedures were repeated three times to uncover defect-annealing effects on deuterium retention. The results show that deuterium retention decreases approximately 70% for at 500 °C after each annealing, and radiation damages were not annealed out completely even after the 3rd annealing. TMAP modeling revealed the trap concentration decreases approximately 80% after each annealing at 900 °C for 0.5 h.

Current status for applications of hydrophobic platinum catalysts in tritium removal from nuclear effluents

International Nuclear Information System (INIS)

Vagner, Irina; Ionita, Gheorghe; Varlam, Carmen

2008-01-01

Full text: Based on the long experience of the authors, in the preparation, testing and evaluation of the performances of hydrophobic catalysts, and based on the reviewed references, this paper presents up-to-date R and D results on the preparation methods and applications of the hydrophobic catalysts, in deuterium and tritium separation. The objectives of the paper are: 1. to provide a database for selection of the most appropriate catalyst and catalytic packing for above mentioned processes; 2. to evaluate the potentiality of hydrophobic Pt-catalysts in the deuterium and tritium separation; 3. to assess and find a new procedure for preparation of a new improved hydrophobic catalyst. The merits of the hydrophobic catalysts are shown in comparison to hydrophilic catalysts. As results of the review some general conclusions about the applications of hydrophobic catalysts in environmental field are as follow: 1. the hydrophobic Pt-catalysts packed in the trickle bed reactors showed a high catalytic activity and long stability; 2. the utilization of the hydrophobic Pt-catalysts for tritium removal from liquid and gaseous effluent in nuclear field was entirely confirmed on industrial scale; 3. the extension of the utilization of the hydrophobic Pt-catalysts to other new processes, which take place in presence of liquid water or high humidity, like VOCs oxidation from wastewater or H 2 -O 2 catalytic recombination, are subject to testing

Tritium pellet injector design for tokamak fusion test reactor

International Nuclear Information System (INIS)

Fisher, P.W.; Baylor, L.R.; Bryan, W.E.

1985-01-01

A tritium pellet injector (TPI) system has been designed for the Tokamak Fusion Test Reactor (TFTR) Q approx. 1 phase of operation. The injector gun utilizes a radial design with eight independent barrels and a common extruder to minimize tritium inventory. The injection line contains guide tubes with intermediate vacuum pumping stations and fast valves to minimize propellant leakage to the torus. The vacuum system is designed for tritium compatibility. The entire injector system is contained in a glove box for secondary containment protection against tritium release. Failure modes and effects have been analyzed, and structural analysis has been performed for most intense predicted earthquake conditions. Details of the design and operation of this system are presented in this paper

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Neutron spectroscopy measurements of 14 MeV neutrons at unprecedented energy resolution and implications for deuterium-tritium fusion plasma diagnostics

Science.gov (United States)

Rigamonti, D.; Giacomelli, L.; Gorini, G.; Nocente, M.; Rebai, M.; Tardocchi, M.; Angelone, M.; Batistoni, P.; Cufar, A.; Ghani, Z.; Jednorog, S.; Klix, A.; Laszynska, E.; Loreti, S.; Pillon, M.; Popovichev, S.; Roberts, N.; Thomas, D.; Contributors, JET

2018-04-01

An accurate calibration of the JET neutron diagnostics with a 14 MeV neutron generator was performed in the first half of 2017 in order to provide a reliable measurement of the fusion power during the next JET deuterium-tritium (DT) campaign. In order to meet the target accuracy, the chosen neutron generator has been fully characterized at the Neutron Metrology Laboratory of the National Physical Laboratory (NPL), Teddington, United Kingdom. The present paper describes the measurements of the neutron energy spectra obtained using a high-resolution single-crystal diamond detector (SCD). The measurements, together with a new neutron source routine ‘ad hoc’ developed for the MCNP code, allowed the complex features of the neutron energy spectra resulting from the mixed D/T beam ions interacting with the T/D target nuclei to be resolved for the first time. From the spectral analysis a quantitative estimation of the beam ion composition has been made. The unprecedented intrinsic energy resolution (<1% full width at half maximum (FWHM) at 14 MeV) of diamond detectors opens up new prospects for diagnosing DT plasmas, such as, for instance, the possibility to study non-classical slowing down of the beam ions by neutron spectroscopy on ITER.

Implantation driven permeation behavior of deuterium through pure tungsten

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hirofumi E-mail: nakamura@tpl.tokai.jaeri.go.jp; Hayashi, Takumi; Nishi, Masataka; Arita, Makoto; Okuno, Kenji

2001-09-01

Implantation driven permeation behavior of deuterium through pure tungsten has been investigated to estimate the amount of tritium permeation through its barrier in a thermo-nuclear fusion device. The permeation experiments were performed on pure tungsten foil of 25 {mu}m thickness under conditions of incident flux of 1.9x10{sup 18}-1.1x10{sup 19} D{sup +}/m{sup 2}s, incident ion energy of 200-2000 eV, and specimen temperature of 512-660 K. As a result of this steady-state permeation experiment, the rate-determining process of deuterium permeation was found to be controlled by diffusion at both implanted and permeated sides. On the other hand, transient permeation was strongly affected by trap effect in the specimen. Simulation analysis using TMAP code on transient permeation behavior indicates the existence of a trap site with a trap energy of nearly 1eV and with a trap density of over several ten's ppm in tungsten.

Implantation driven permeation behavior of deuterium through pure tungsten

International Nuclear Information System (INIS)

Nakamura, Hirofumi; Hayashi, Takumi; Nishi, Masataka; Arita, Makoto; Okuno, Kenji

2001-01-01

Implantation driven permeation behavior of deuterium through pure tungsten has been investigated to estimate the amount of tritium permeation through its barrier in a thermo-nuclear fusion device. The permeation experiments were performed on pure tungsten foil of 25 μm thickness under conditions of incident flux of 1.9x10 18 -1.1x10 19 D + /m 2 s, incident ion energy of 200-2000 eV, and specimen temperature of 512-660 K. As a result of this steady-state permeation experiment, the rate-determining process of deuterium permeation was found to be controlled by diffusion at both implanted and permeated sides. On the other hand, transient permeation was strongly affected by trap effect in the specimen. Simulation analysis using TMAP code on transient permeation behavior indicates the existence of a trap site with a trap energy of nearly 1eV and with a trap density of over several ten's ppm in tungsten

Mixing rules for and effects of other hydrogen isotopes and of isotopic swamping on tritium recovery and loss to biosphere from fusion reactors

International Nuclear Information System (INIS)

Pendergrass, J.H.

1978-01-01

Efficient recovery of bred and unburnt tritium from fusion reactors, and control of its migration within reactors and of its escape into the biosphere are essential for self-sufficient fuel cycles and for public, plant personnel, and environmental protection. Tritium in fusion reactors will be mixed with unburnt deuterium and protium introduced by (n,p) reactions and diffusion into coolant loops from steam cycles. Rational design for tritium recovery and escape prevention must acknowledge this fact. Consequences of isotopic admixture are explored, mixing rules for projected fusion reactor dilute-solution conditions are developed, and a rule of thumb regarding their effects on tritium recovery methods is formulated

Tritium depth profiling in carbon by accelerator mass spectrometry

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 measurements by accelerator mass spectrometry have been performed at the facility installed at the Rossendorf 3 MV Tandetron. In order to achieve a uniform erosion at the target surface inside a commercial Cs ion sputtering source and to avoid edge effects, the samples were mechanically scanned and the signals were recorded only during sputtering at the centre of the sputtered area. The sputtered negative ions were mass analysed by the injection magnet of the Tandetron. Hydrogen and deuterium profiles were measured with the Faraday cup between the injection magnet and the accelerator, while the tritium was counted after the accelerator 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 study on the primary requirement for the safety of the Wolsong tritium removal facility

International Nuclear Information System (INIS)

Hwang, K. H.; Lee, K. J.; Jeong, C. W.

2001-01-01

Owing to the using a heavy water as a moderator and a coolant in Heavy water reactor, A large mount of tritium is produced due to a reaction of deuterium with neutron in the reactor and some of tritium is released to the environment. In Wolsong, 4 units (CANDU-600 type) Heavy water reactor is in operation. And the generated amount of tritium is increased with the increase of operational year of the Wolsong nuclear reactor. Decommissioning of the Wolsong unit 1 is expected to start at 2013. Before 2013, to reduce the workers internal radiation doses and environmental release of tritium, Tritium Removal Facility (TRF) is required and should be operated. Wolsong TRF (WTRF) is under developing stage by Korea Electric Power Corporation(KEPCO)and scheduled to start operation about 2006. Once the facility begins operation it can be contributed to the greatly reduction of tritium release to the environment and worker's expose. In this situation, study about the safety assessment method and regulatory requirement is essential for safety insurance of WTRF. And this helps the safety acquirement, successful operation and reliance of WTRF

Recovery of tritium from CANDU reactors, its storage and monitoring of its migration in the environment

International Nuclear Information System (INIS)

Holtslander, W.J.; Osborne, R.V.

1979-07-01

Tritium is produced in CANDU heavy water reactors mainly by neutron activation of deuterium. The typical production rate is 2.4 kCi per megawatt-year (89 TBq. per megawatt-year. In Pickering Generating Station the average concentration of tritium in the moderators has reached 16 Ci.kg -1 (0.6 TBq.kg -1 ) and in coolants, 0.5 Ci.kg -1 (0.02 TBq.kg -1 ). Concentrations will continue to increase towards an equilibrium determined by the production rate, the tritium decay rate and heavy water replacement. Tritium removal methods that are being considered for a pilot plant design are catalytic exchange of DTO with D 2 and electrolysis of D 2 O/DTO to provide feed for cryogenic distillation of D 2 /DT/T 2 . Storage methods for the removed tritium - as elemental gas, as metal hydrides and in cements - are also being investigated. Transport of tritiated wastes should not be a particularly difficult problem in light of extensive experience in transporting tritiated heavy water. Methods for determining the presence of tritium in the environment of any tritium handling facility are well established and have the capability of measuring concentrations of tritium down to current ambient values. (author)

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

Labelling of steroid 3-O-sulfates by tritium and their binding to guinea pig cortical cell membranes

Czech Academy of Sciences Publication Activity Database

Elbert, Tomáš; Slaninová, Jiřina; Chodounská, Hana; Šaman, David

2007-01-01

Roč. 50, 5/6 (2007), s. 574-575 ISSN 0362-4803 R&D Projects: GA AV ČR IAA4055305 Institutional research plan: CEZ:AV0Z40550506 Keywords : tritium * deuterium * tritiated steroid 3-O-sulfates * palladium Subject RIV: CC - Organic Chemistry Impact factor: 1.142, year: 2007

Fusion reactor start-up without an external tritium source

Energy Technology Data Exchange (ETDEWEB)

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

2016-02-15

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

A uranium bed with ceramic body for tritium storage

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

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

A Hydrogen Exchange Method Using Tritium and Sephadex: Its Application to Ribonuclease*

Science.gov (United States)

Englander, S. Walter

2012-01-01

A new method for measuring the hydrogen exchange of macromolecules in solution is described. The method uses tritium to trace the movement of hydrogen, and utilizes Sephadex columns to effect, in about 2 minutes, a separation between tritiated macromolecule and tritiated solvent great enough to allow the measurement of bound tritium. High sensitivity and freedom from artifact is demonstrated and the possible value of the technique for investigation of other kinds of colloid-small molecule interaction is indicated. Competition experiments involving tritium, hydrogen, and deuterium indicate the absence of any equilibrium isotope effect in the ribonuclease-hydrogen isotope system, though a secondary kinetic isotope effect is apparent when ribonuclease is largely deuterated. Ribonuclease shows four clearly distinguishable kinetic classes of exchangeable hydrogens. Evidence is marshaled to suggest the independently measurable classes II, III, and IV (in order of decreasing rate of exchange) to represent “random-chain” peptides, peptides involved in α-helix, and otherwise shielded side-chain and peptide hydrogens, respectively. PMID:14075117

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

International Nuclear Information System (INIS)

Fuetterer, M.A.; Raepsaet, X.; Proust, E.; Leger, D.

1994-01-01

Tritium control issues such as the development of permeation barriers and the choice of the coolant and purge-gas chemistry are of crucial importance for solid breeder blankets. In order to quantify these problems for the helium-cooled ceramic breeder-inside-tube (BIT) blanket concept, the tritium leakage into the coolant was evaluated and the consequent tritium losses into the steam circuit were determined. The results indicate that under certain specified conditions the total tritium release from the coolant can be limited to approximately 10 Ci/d, but only on the assumption that experimental data for tritium permeation barriers can be attained under realistic operating conditions. An experimental study on the impact of the gas chemistry on tritium losses is proposed. (author) 8 refs.; 2 figs

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

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.)

Advanced pressure tube sampling tools

International Nuclear Information System (INIS)

Wittich, K.C.; King, J.M.

2002-01-01

Deuterium concentration is an important parameter that must be assessed to evaluate the Fitness for service of CANDU pressure tubes. In-reactor pressure tube sampling allows accurate deuterium concentration assessment to be made without the expenses associated with fuel channel removal. This technology, which AECL has developed over the past fifteen years, has become the standard method for deuterium concentration assessment. AECL is developing a multi-head tool that would reduce in-reactor handling overhead by allowing one tool to sequentially sample at all four axial pressure tube locations before removal from the reactor. Four sets of independent cutting heads, like those on the existing sampling tools, facilitate this incorporating proven technology demonstrated in over 1400 in-reactor samples taken to date. The multi-head tool is delivered by AECL's Advanced Delivery Machine or other similar delivery machines. Further, AECL has developed an automated sample handling system that receives and processes the tool once out of the reactor. This system retrieves samples from the tool, dries, weighs and places them in labelled vials which are then directed into shielded shipping flasks. The multi-head wet sampling tool and the automated sample handling system are based on proven technology and offer continued savings and dose reduction to utilities in a competitive electricity market. (author)

Permeation behavior of deuterium implanted into beryllium

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hirofumi; Hayashi, Takumi; O' hira, Shigeru; Nishi, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-09-01

Study on Implantation Driven Permeation (IDP) behavior of deuterium through pure beryllium was investigated as a part of the research to predict the tritium permeation through the first wall components ITER (International Thermonuclear Experimental Reactor). The permeation experiments were carried out with two beryllium specimens, one was an unannealed specimen and the other was that annealed at 1173 K. The permeation flux was measured as a function of specimen temperature and incident ion flux. Surface analysis of specimen was also carried out after the permeation experiment. Permeation was observed only with the annealed specimen and no significant permeation was observed with unannealed specimen under the present experimental condition (maximum temperature: 685 K, detection limit: 1x10{sup 13} D atoms/m{sup 2}s). It could be attributed that the intrinsic lattice defects, which act as diffusion preventing site, decreased with the specimen annealing. Based on the result of steady and transient permeation behavior and surface analysis, it was estimated that the deuterium permeation implanted into annealed beryllium was controlled by surface recombination due to the oxide layer on the surface of the permeated side. (author)

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.

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

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

Tritium diffusion in nonmetallic solids of interest for fusion reactors. Final report

International Nuclear Information System (INIS)

Elleman, T.

1979-01-01

Tritium diffusion measurements have been conducted in Al 2 O 3 , BeO, Y 2 O 3 , SiC, B 4 C, Si 3 N 4 and pyrolytic carbon as a basis for evaluating these materials as potential tritium barriers in fusion reactors. Deuterium solubility measurements were conducted with Al 2 O 3 , SiC and pyrolytic carbon to establish the pressure and temperature dependence of solubility and to identify solubility ranges. Hydrogen permeability measurements on commercially available Al 2 O 3 and SiC materials were used as a check on calculated permeabilities and to provide data on hydrogen permeation rates in polycrystalline materials. The diffusion, solubility and permeation results are presented and discussed in terms of fusion reactor applications

Tritium recovery from co-deposited layers using 193-nm laser

Science.gov (United States)

Shu, W. M.; Kawakubo, Y.; Nishi, M. F.

Recovery of tritium from co-deposited layers formed in deuterium-tritium plasma operations of the TFTR (Tokamak Fusion Test Reactor) was investigated by the use of an ArF excimer laser operating at the wavelength of 193 nm. At the laser energy density of 0.1 J/cm2, a transient spike of the tritium-release rate was observed at initial irradiation. Hydrogen isotopes were released in the form of hydrogen-isotope molecules during the laser irradiation in vacuum, suggesting that tritium can be recovered readily from the released gases. In a second experiment, hydrogen (tritium) recovery from the co-deposited layers on JT-60 tiles that had experienced hydrogen-plasma operations was investigated by laser ablation with a focused beam of the excimer laser. The removal rate of the co-deposited layers was quite low when the laser energy density was smaller than the ablation threshold (1.0 J/cm2), but reached 1.1 μm/pulse at the laser energy density of 7.6 J/cm2. The effective absorption coefficient in the co-deposited layers at the laser wavelength was determined to be 1.9 μm-1. The temperature of the surface during the irradiation at the laser energy density of 0.5 J/cm2 was measured on the basis of Planck's law of radiation, and the maximum temperature during the irradiation decreased from 3570 K at the initial irradiation to 2550 K at the 1000th pulse of the irradiation.

A description of an ultra high vacuum device for permeation measurements

International Nuclear Information System (INIS)

Gautsch, O.; Hodapp, G.

1984-01-01

The permeability of deuterium through austenitic stainless steel membranes is determined by means of an UHV-apparatus as a function of temperature, membrane thickness, and deuterium upstream driving pressure. The upstream side of the membrane is in contact with the liquid alloy of the eutectic Pb 83 Li 17 in order to represent the situation of a tritium breeding blanket of a fusion reactor. By means of the experimental results the amounts of tritium which permeate across the walls of the cooling tubes into the coolant will be determined. The purpose of this work is to describe the apparatus and the methods used for the determination of permeabilities. Also some preliminary results of permeation flux measurements are given

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.)

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Effect of coexistent hydrogen isotopes on tracer diffusion of tritium in alpha phase of group-V metal-hydrogen systems

International Nuclear Information System (INIS)

Sakamoto, Kan; Hashizume, Kenichi; Sugisaki, Masayasu

2009-01-01

Tracer diffusion coefficients of tritium in the alpha phase of group-V metal-hydrogen systems, α-MH(D)xTy (M=V and Ta; x>>y), were measured in order to clarify the effects of coexistent hydrogen isotopes on the tritium diffusion behavior. The hydrogen concentration dependence of such behavior and the effects of the coexistent hydrogen isotopes (protium and deuterium) were determined. The results obtained in the present (for V and Ta) and previous (for Nb) studies revealed that tritium diffusion was definitely dependent on hydrogen concentration but was not so sensitive to the kind of coexistent hydrogen isotopes. By summarizing those data, it was found that the hydrogen concentration dependence of the tracer diffusion coefficient of tritium in the alpha phase of group-V metals could be roughly expressed by a single empirical curve. (author)

Use of binary alloys of the lanthanides for tritium recovery from CTR blankets

International Nuclear Information System (INIS)

Carstens, D.H.W.

1978-01-01

Liquid binary alloys of the lanthanide metals have been proposed as getters of tritium from breeder blankets of controlled thermonuclear reactors. Because of the high stability of the lanthanide hydrides at reactor temperatures (500--1000 0 C), these alloys should prove highly efficient in this application and a series of experiments designed to test this applicability are summarized here. Sieverts' experiments using deuterium were carried out on a series of alloys of La and Ce. For eutectics of the approximate composition Ln 5 M where Ln is La or Ce and M is an iron-group metal, it was found that the deuteriding capacities and the equilibrium pressures were close to those of the parent metal. Experiments measuring the extraction rate of low-level tritium from helium streams using La 5 . 25 Ni were carried out. The tritium was rapidly gettered down to about 10 ppM and more slowly over periods of 1--2 h to below 0.1 ppM

In situ measurement of tritium permeation through stainless steel

Science.gov (United States)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2013-06-01

The TMIST-2 irradiation experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory to evaluate tritium permeation through Type 316 stainless steel (316 SS). The interior of a 316 SS seamless tube specimen was exposed to a 4He carrier gas mixed with a specified quantity of tritium (T2) to yield partial pressures of 0.1, 5, and 50 Pa at 292 °C and 330 °C. In situ tritium permeation measurements were made by passing a He-Ne sweep gas over the outer surface of the specimen to carry the permeated tritium to a bubbler column for liquid scintillation counting. Results from in situ permeation measurements were compared with predictions based on an ex-reactor permeation correlation in the literature. In situ permeation data were also used to derive an in-reactor permeation correlation as a function of temperature and pressure over the ranges considered in this study. In addition, the triton recoil contribution to tritium permeation, which results from the transmutation of 3He to T, was also evaluated by introducing a 4He carrier gas mixed with 3He at a partial pressure of 1013 Pa at 330 °C. Less than 3% of the tritium resulting from 3He transmutation contributed to tritium permeation.

Pebble bed blanket design for deuterium burning tandem mirror reactors

International Nuclear Information System (INIS)

Grotz, S.P.; Dhir, V.K.

1983-01-01

The UCLA tandem mirror reactor, SATYR, was developed around the capability of tandem mirrors with thermal barriers to burn deuterium at reasonable efficiency levels. The pebble bed concept has been incorporated into our blanket design for the following reasons: 1) Large area-to-volume ratio for purposes of heat removal; 2) Large volume of structure for high thermal capacity thus increasing the safety margin during off-normal incidents; 3) Relatively inexpensive manufacturing costs because of large acceptable tolerances and lack of exotic materials (i.e., lithium). A simplified stress analysis of the blanket module was performed to optimize and simplify the design. The pre-specified stress intensity limitations used were based upon a 30-year predicted lifetime for each module. Along with stress analysis of the vessel a detailed thermal hydraulic analysis of the pebble bed has been completed. Parameters affecting the pebble bed design are fluidization velocity, pressure drop, heat transfer coefficient, thermally induced stress in the spheres and spatial variation of the power density. Although reasonable gross thermal efficiencies of the 2 designs has been achieved (28% for H 2 O and 39% for He) the high net recirculating power fraction for heating and neutral beams results in relatively low net plant efficiencies (21% and 27%). The results show that a blanket can be designed with good thermal efficiency and a relative-ly simple configuration. However, application of this concept to the high Q deuterium-tritium fuel cycle would have difficulties resulting from the need for continuous removal of the tritium. (orig./HP)

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

Science.gov (United States)

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

2016-08-01

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

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

Analysis of the time dependence of the tritium concentration in the Embalse Rio Tercero lake

International Nuclear Information System (INIS)

Lopez, F.O.; Bruno, H.A.

1998-01-01

In natural uranium and heavy water reactors, tritium is produced mainly as the activation product of the deuterium in the moderator and cooling medium. About 75% of the liquid effluents discharged by nuclear power plants in Argentina correspond to tritium. In the case of the Embalse nuclear power plant, the liquid effluents are discharged into the Rio Tercero reservoir. As its water is used for drinking, 98% of the dose received by the critical group is due to these discharges. A simple mathematical model was developed which predicts the variation in the tritium concentration in the reservoir. It is a complete mixture type model and the entry parameters are the lake volume, entrance volume and discharge volume. The model was solved by means of a Runge-Kutta method of second order. The chosen method is a modified Euler. A good correlation is observed when the values obtained by means of the numeric resolution of the developed model are compared with the values obtained by the tritium measurement made during the 1996 and 1997 environmental monitoring program. (author) [es

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

International Nuclear Information System (INIS)

Retevoi, Carmen; Balteanu, Ovidiu Ioan

1999-01-01

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

Tritium retention in S-65 beryllium after 100 eV plasma exposure

Energy Technology Data Exchange (ETDEWEB)

Causey, R.A. [Sandia National Labs., Livermore, CA (United States); Longhurst, G.R. [Idaho National Engineering Laboratories, Idaho Falls, 83415 (United States); Harbin, W. [Los Alamos National Laboratories, Los Alamos, NM 87545 (United States)

1997-02-01

The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 x 10{sup 17} (D+T)/cm{sup 2} s up to about 3 x 10{sup 18} (D+T)/cm{sup 2} s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H{sub 2} gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 x 10{sup 17} (D+T)/cm{sup 2} at 373 K to a low of 1 x 10{sup 16} (D+T)/cm{sup 2} at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C=0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters. (orig.).

Tritium retention in S-65 beryllium after 100 eV plasma exposure

Science.gov (United States)

Causey, Rion A.; Longhurst, Glen R.; Harbin, Wally

1997-02-01

The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 × 10 17 ( D + T)/ cm2s up to about 3 × 10 18 ( D + T)/ cm2s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H 2 gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 × 10 17 ( D + T)/ cm2 at 373 K to a low of 1 × 10 16 ( D + T)/ cm2 at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C = 0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters.

Management of tritium exposures for professionally exposed workers at Cernavoda 1 NPP

Energy Technology Data Exchange (ETDEWEB)

Chitu, Catalina; Simionov, Vasile [CNE-PROD Cernavoda NPP, No. 1, Medgidiei Str. Cernavoda 905200 (Romania)

2004-07-01

Operating experience to date of CANDU reactors has indicated that the major contributor to the internal dose of professionally exposed workers is the tritiated heavy water (DTO). CANDU reactors are both moderated and cooled by heavy water (D{sub 2}O). Tritium is produced in CANDU reactors by neutron reactions with deuterium, boron, and lithium and by ternary fission. Even small leaks from these systems can produce important contaminations with tritiated water vapours of the air in the reactor building and thus increased individual and collective internal doses. Professionally exposed workers are subject to a combination of acute and chronic tritium exposure and HTO dosimetry program at Cernavoda NPP is based on multiple sample results. The routine urine bioassay program performs the monitoring and dosimetry functions for DTO. A specialized laboratory using Liquid Scintillation Spectrometry methods currently determines tritium activities in urine samples. The frequency of biological samples submission depends on the tritium concentration in the last sample. Dose assignments resulting from routinely measured weekly and monthly urinary levels of tritium oxide are based on the method of linear interpolation unless it is known that there has been no exposure between samples (vacation). All information about these doses is stored into a dedicated electronic database and used to make periodical reports and to ensure that the legal and administrative individual and annual limits are not exceeded. A chronic unprotected exposure to small tritium dose rate (< 50{mu}Sv/h) may lead to internal doses that exceed the intervention level. In case of acute exposure an increased daily water intake combined with a proper medical intervention could reduce the effective half time of tritium 2-3 times. (authors)

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

International Nuclear Information System (INIS)

Fuetterer, M.A.; Raepsaet, X.; Proust, E.

1994-01-01

The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab

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

Energy Technology Data Exchange (ETDEWEB)

Fuetterer, M A; Raepsaet, X; Proust, E

1994-12-31

The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab.

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

Energy Technology Data Exchange (ETDEWEB)

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

2002-11-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Tritium Assay and Dispensing of KEPRI Tritium Lab

International Nuclear Information System (INIS)

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

2009-01-01

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

Heterogeneous Catalysis: Deuterium Exchange Reactions of Hydrogen and Methane

Science.gov (United States)

Mirich, Anne; Miller, Trisha Hoette; Klotz, Elsbeth; Mattson, Bruce

2015-01-01

Two gas phase deuterium/hydrogen exchange reactions are described utilizing a simple inexpensive glass catalyst tube containing 0.5% Pd on alumina through which gas mixtures can be passed and products collected for analysis. The first of these exchange reactions involves H[subscript 2] + D[subscript 2], which proceeds at temperatures as low as 77…

Production of ultrapure D-T gas by removal of molecular tritium by selective adsorption

International Nuclear Information System (INIS)

Maienschein, J.L.; Hudson, R.S.; Tsugawa, R.T.; Fearon, E.M.; Souers, P.C.; Collins, G.W.

1992-01-01

Production of molecular deuterium-tritium (D-T) with very low molecular tritium (T 2 ) is necessary for application as a nuclear spin polarized fuel. Selective adsorption of hydrogen isotopes on zeolites or alumina can provide the separation needed to produce D-T with very low T 2 . Use of an absorption column at 20-25 K offers low inventory, compact size, and rapid operation, in comparison with conventional separation techniques such as cryogenic distillation or thermal diffusion. In this paper, the authors discuss principles of absorption, and describe a calculational model of the absorption column and operational implications revealed by it. The authors show experimental proof-of-principle data for removal of T 2 from D-T with an adsorption column operated at 23 K

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-03-15

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

Tritium Decontamination of TFTR D-T Graphite Tiles Employing Ultra Violet Light and a Nd:YAG Laser

International Nuclear Information System (INIS)

Gentile, C.A.; Skinner, C.H.; Young, K.M.; Ciebiera, L.

1999-01-01

The use of an ultra violet (UV) light source (wavelength = 172 nm) and a Nd:YAG Laser for the decontamination of the Tokamak Fusion Test Reactor (TFTR) deuterium-tritium (D-T) tiles will be investigated at the Princeton Plasma Physics Laboratory (PPPL). The development of this form of tritium decontamination may be useful for future D-T burning fusion devices which employ carbon plasma-facing components on the first wall. Carbon tiles retain hydrogen isotopes, and the in-situ tritium decontamination of carbon can be extremely important in maintaining resident in-vessel tritium inventory to a minimum. A test chamber has been designed and fabricated at PPPL. The chamber has the ability to be maintained under vacuum, be baked to 200 *C, and provides sample ports for gas analyses. Tiles from TFTR that have been exposed to D-T plasmas will be placed within the chamber and exposed to either an UV light source or the ND:YAG Laser. The experiment will determine the effectiveness of these two techniques for the removal of tritium. In addition, exposure rates and scan times for the UV light source and/or Nd:YAG Laser will be determined for tritium removal optimization from D-T tiles

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.

A software tool for evaluation of hydrogen ingress in CANDU pressure tubes

International Nuclear Information System (INIS)

Mihalache, Maria; Vasile, Radu; Deaconu, Mariea

2009-01-01

The prediction of hydrogen isotopes concentration into the body and in the rolled joints of operating pressure tubes as a function of reactor hot hours is very important in many fitness-for-service assessments and end of life estimates. The rolled joints are high stress zones with potential for delayed hydride cracking. Predictive models for assessing the long-term deuterium ingress in both body and rolled joint of the pressure tubes have been implemented in a software tool, ROHID, developed in INR-Pitesti. ROHID is a PC-based Windows application with a user-friendly interface that predicts the equivalent hydrogen ingress for Zr-2.5Nb pressure tubes. It uses colour-coded reactor core maps to display the predicted deuterium concentration as a function of time for selected axial locations. Plots of deuterium versus axial location and time for individual pressure tubes are also available. Also, the software tool can predict the exceeding of hydrogen terminal solid solubility (HTSS) from hydrides during precipitation and dissolving processes as a function of time and axial location. (authors)

Tritium monitor calibration at Los Alamos National Laboratory

International Nuclear Information System (INIS)

Bjork, C.J.; Aikin, D.J.; Houlton, T.W.

1997-08-01

Tritium in air is monitored at Los Alamos National Laboratory (LANL) with air breathing instruments based on ionization chambers. Stack emissions are continuously monitored from sample tubes which each connect to a Tritium bubble which differentially collects HTO and HT. A set of glass vials of glycol capture the HTO. The HT is oxidized with a palladium catalyst and the resultant HTO is captured in a second set of vials of glycol. The glycol is counted with a liquid scintillation counter. All calibrations are performed with tritium containing gas. The Radiation Instrumentation and Calibration (RIC) Team has constructed and maintains two closed loop gas handling systems based on femto TECH model U24 tritium ion chamber monitors: a fixed system housed in a fume hood and a portable system mounted on two two wheeled hand trucks. The U24 monitors are calibrated against tritium in nitrogen gas standards. They are used as standard transfer instruments to calibrate other ion chamber monitors with tritium in nitrogen, diluted with air. The gas handling systems include a circulation pump which permits a closed circulation loop to be established among the U24 monitor and typically two to four other monitors of a given model during calibration. Fixed and portable monitors can be calibrated. The stack bubblers are calibrated in the field by: blending a known concentration of tritium in air within the known volume of the two portable carts, coupled into a common loop; releasing that gas mixture into a ventilation intake to the stack; collecting oxidized tritium in the bubbler; counting the glycol; and using the stack and bubbler flow rates, computing the bubbler's efficiency. Gas calibration has become a convenient and quality tool in maintaining the tritium monitors at LANL

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.)

Simulating tritium retention in tungsten with a multiple trap model in the TMAP code

International Nuclear Information System (INIS)

Merrill, Brad J.; Shimada, Masashi; Humrickhouse, Paul W.

2013-01-01

Accurately predicting the quantity of tritium retained in plasma facing components is a key safety issue for licensing future fusion power reactors. Retention of tritium in the lattice damage caused when high energy neutrons collide with atoms in the structural material of the reactor's plasma facing components (PFCs) is an area of ongoing experimental research at the Idaho National Laboratory (INL) under the US/Japan TITAN collaboration. Recent experiments with the Tritium Plasma Experiment (TPE), located in the INL's Safety and Tritium Applied Research (STAR) facility, demonstrate that this damage can only be simulated by computer codes like the Tritium Migration Analysis Program (TMAP) if one assumes that the lattice damage produced by these neutrons results in multiple types of hydrogen traps (energy wells) within the material, each possessing a different trap energy and density. Previous attempts to simulate the quantity of deuterium released from neutron irradiated TPE tungsten targets indicated that at least six different traps are required by TMAP to model this release. In this paper we describe a recent extension of the TMAP trap site model to include as many traps as required by the user to simulate retention of tritium in neutron damaged tungsten. This model has been applied to data obtained for tungsten irradiated to a damage level of 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) after exposure to a plasma in TPE. (author)

Energetics of semi-catalyzed-deuterium, light-water-moderated, fusion-fission toroidal reactors

International Nuclear Information System (INIS)

Jassby, D.L.; Towner, H.H.; Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.

1978-07-01

The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U 3 Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10 14 cm -3 s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor

A Computational Study of A Lithium Deuteride Fueled Electrothermal Plasma Mass Accelerator

OpenAIRE

Gebhart III, Gerald Edward

2013-01-01

Future magnetic fusion reactors such as tokamaks will need innovative, fast, deep-fueling systems to inject frozen deuterium-tritium pellets at high speeds and high repetition rates into the hot plasma core. There have been several studies and concepts for pellet injectors generated, and different devices have been proposed. In addition to fueling, recent studies show that it may be possible to disrupt edge localized mode (ELM) formation by injecting pellets or gas into the fusion plasma. The...

Detection of helium-containing paramagnetic complex at decay of tritium incorporated in HTO and DTO

International Nuclear Information System (INIS)

Legasov, V.A.; Usatyj, A.F.; Ibragimov, R.A.; Myasoedov, N.F.

1979-01-01

EPR technique was used to study the paramagnetic centers appearing during long-term storage of frozen (77 K) non-deoxygenated light (H 2 O) and heavy (D 2 O) water containing 3.5% vol. highly active HTO (29 ci/cm 3 ), so that the resulting activity of the sample (0.1 ml) was about 100 mci. For comparison, the same samples containing no HTO but irradiated by Co-60 rays were studied under identical conditions. A schematic of the processes involved in tritium decay in DTO (or HTO) in the aqueous matrix frozen at low temperatures was suggested

Method for nondestructive fuel assay of laser fusion targets

Science.gov (United States)

Farnum, Eugene H.; Fries, R. Jay

1976-01-01

A method for nondestructively determining the deuterium and tritium content of laser fusion targets by counting the x rays produced by the interaction of tritium beta particles with the walls of the microballoons used to contain the deuterium and tritium gas mixture under high pressure. The x rays provide a direct measure of the tritium content and a means for calculating the deuterium content using the initial known D-T ratio and the known deuterium and tritium diffusion rates.

Cold fusion produces more tritium than neutrons

International Nuclear Information System (INIS)

Rajagopalan, S.R.

1989-01-01

The results of the major cold fusion experiments performed in various laboratories of the world and attempts to explain them are reviewed in brief. Particular reference is made to the experiments carried out in the Bhabha Atomic Research Centre (BARC), Bombay. In BARC experiments, it is found that tritium is the primary product of cold fusion. Author has put forward two hypothetical pictures of D-D fusion. (1) When a metal like Pd or Ti is loaded with D 2 , a crack forms. Propogation of such a crack accelerates deuterons which bombard Pd D 2 /D held by Pd or Ti leading to neutron capture or tritium formation with the release of protons and energy. The released protons might transfer its energy to some other deuteron and a chain reaction is started. This chain reaction terminates when a substantial portion of D in the crack tip is transmuted. This picture explains fusion reaction bursts and the random distribution of reaction sites, but does not explain neutron emission. (2) The deuterons accelerated by a propogating crack may hit a Pd/Ti nucleus instead of a deuterium nucleus and may transmute Pd/Ti. (M.G.B.). 18 refs

Use of hydrophobic Pt-catalysts in tritium removal from effluents

International Nuclear Information System (INIS)

Gheorghe, Ionita; Popescu, Irina; Stefanescu, Ioan; Steflea, Dumitru; Varlam, Carmen

2002-01-01

Based on the long experience of the authors, in the preparation, testing and evaluation of the performances of hydrophobic catalysts, and based on the reviewed references, this paper presents up-to-date R and D activities on the application of the hydrophobic catalysts in tritium removal from nuclear effluents. Tritium removal from the heavy water reactor and nuclear reprocessing plant, the cleanup of atmosphere and gaseous effluents by hydrogen-oxygen recombination, removal of oxygen dissolved in water are presented and discussed. Unlike the conventional hydrophilic catalysts, the hydrophobic catalysts keep a high catalytic activity and stability, even under the direct contact to liquid water or in presence of saturated humidity. A large diversity of catalyst types (over 100 catalysts) was prepared and tested in order to make them feasible for such processes. The objectives of the review are: - to provide a database for selection of the most appropriate catalyst and catalytic packing for above mentioned processes; - the designing and operation of reactor packed with hydrophobic catalysts; - to evaluate the potentiality of hydrophobic Pt-catalysts in the present and future applications. The most important results are the following: - the hydrophobic Pt-catalysts packed in the trickle bed or separated bed reactors, showed a high catalytic activity and long stability; - the utilization of the hydrophobic Pt-catalysts for the hydrogen isotopes (tritium and deuterium) separation and for hydrogen-oxygen recombination in nuclear field was entirely confirmed on industrial scale; - the improvement of the inner geometry of the reactors and of the composition of mixed catalytic packing as well as the evaluation of performances of separation processes constitute a major contribution of the authors; - the extension of the utilization of the hydrophobic Pt-catalysts in the oxidation of volatile organic compounds from wastewater; - the removal of dissolved oxygen, and deuterium

Comprehensive sets of 124Xe(n ,γ )125Xe and 124Xe(n ,2 n )123Xe cross-section data for assessment of inertial-confinement deuterium-tritium fusion plasma

Science.gov (United States)

Bhike, Megha; Fallin, B.; Gooden, M. E.; Ludin, N.; Tornow, W.

2015-01-01

Measurements of the neutron radiative-capture cross section of 124Xe have been performed for the first time for neutron energies above 100 keV. In addition, data for the 124Xe(n ,2 n )123Xe reaction cross section have been obtained from threshold to 14.8 MeV to cover the entire energy range of interest, while previous data existed only at around 14 MeV. The results of these measurements provide the basis for an alternative and sensitive diagnostic tool for investigating properties of the inertial confinement fusion plasma in deuterium-tritium (DT) capsules at the National Ignition Facility located at Lawrence Livermore National Laboratory. Here, areal density ρ R (density × radius) of the fuel, burn asymmetry, and fuel-ablator mix are of special interest. The 124Xe(n ,γ )125Xe reaction probes the down-scattered neutrons, while the 124Xe(n ,2 n )123Xe reaction provides a measure of the 14 MeV direct neutrons.

Analysis of a global database containing tritium in precipitation

Energy Technology Data Exchange (ETDEWEB)

Buckley, R. L. [Savannah River Site (SRS), Aiken, SC (United States); Rabun, R. L. [Savannah River Site (SRS), Aiken, SC (United States); Heath, M. [Savannah River Site (SRS), Aiken, SC (United States)

2016-02-17

The International Atomic Energy Agency (IAEA) directed the collection of tritium in water samples from the mid-1950s to 2009. The Global Network of Isotopes in Precipitation (GNIP) data examined the airborne movement of isotope releases to the environment, with an objective of collecting spatial data on the isotope content of precipitation across the globe. The initial motivation was to monitor atmospheric thermonuclear test fallout through tritium, deuterium, and oxygen isotope concentrations, but after the 1970s the focus changed to being an observation network of stable hydrogen and oxygen isotope data for hydrologic studies. The GNIP database provides a wealth of tritium data collections over a long period of time. The work performed here primarily examined data features in the past 30 years (after much of the effects of above-ground nuclear testing in the late 1950s to early 1960s decayed away), revealing potentially unknown tritium sources. The available data at GNIP were reorganized to allow for evaluation of trends in the data both temporally and spatially. Several interesting cases were revealed, including relatively high measured concentrations in the Atlantic and Indian Oceans, Russia, Norway, as well as an increase in background concentration at a collector in South Korea after 2004. Recent data from stations in the southeastern United States nearest to the Savannah River Site do not indicate any high values. Meteorological impacts have not been considered in this study. Further research to assess the likely source location of interesting cases using transport simulations and/or literature searches is warranted.

In situ measurement of tritium permeation through stainless steel

Energy Technology Data Exchange (ETDEWEB)

Luscher, Walter G., E-mail: walter.luscher@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); Senor, David J., E-mail: david.senor@pnnl.gov [Pacific Northwest National Laboratory, 902 Battelle Blvd, Richland, WA 99352 (United States); Clayton, Kevin K., E-mail: kevin.clayton@inl.gov [Idaho National Laboratory, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States); Longhurst, Glen R., E-mail: glenlonghurst@suu.edu [Idaho National Laboratory, 2525 Fremont Ave., Idaho Falls, ID 83415 (United States)

2013-06-15

Highlights: ► In situ tritium permeation measurements collected over broad pressure range. ► Test conditions relevant to 316 SS in commercial light water reactors. ► Comparisons between in- and ex-reactor measurements provided. ► Correlation between tritium permeation, temperature, and pressure developed. -- Abstract: The TMIST-2 irradiation experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory to evaluate tritium permeation through Type 316 stainless steel (316 SS). The interior of a 316 SS seamless tube specimen was exposed to a {sup 4}He carrier gas mixed with a specified quantity of tritium (T{sub 2}) to yield partial pressures of 0.1, 5, and 50 Pa at 292 °C and 330 °C. In situ tritium permeation measurements were made by passing a He–Ne sweep gas over the outer surface of the specimen to carry the permeated tritium to a bubbler column for liquid scintillation counting. Results from in situ permeation measurements were compared with predictions based on an ex-reactor permeation correlation in the literature. In situ permeation data were also used to derive an in-reactor permeation correlation as a function of temperature and pressure over the ranges considered in this study. In addition, the triton recoil contribution to tritium permeation, which results from the transmutation of {sup 3}He to T, was also evaluated by introducing a {sup 4}He carrier gas mixed with {sup 3}He at a partial pressure of 1013 Pa at 330 °C. Less than 3% of the tritium resulting from {sup 3}He transmutation contributed to tritium permeation.

Study of the trace tritium content in deuterium plasmas of the JET tokamak based on neutron emission spectroscopy measurements

Energy Technology Data Exchange (ETDEWEB)

Ringmar, David

2001-02-01

This thesis describes a study of the evolution of the trace tritium content in the JET tokamak. The study is based on measurements of the neutron emission, which were performed with the magnetic proton recoil (MPR) spectrometer. Data analysis procedures used to extract the results are described in some detail. The thesis also describes a simplified theoretical model to calculate the absolute tritium concentration with a comparison to the experimental results. The present study covers the time period 1996-2000 and the evolution of neutron emission spectroscopy (NES) results are compared with information from related diagnostic sources, and used to discuss the important issue of how tritium is retained in the JET tokamak.

Study of the trace tritium content in deuterium plasmas of the JET tokamak based on neutron emission spectroscopy measurements

International Nuclear Information System (INIS)

Ringmar, David

2001-02-01

This thesis describes a study of the evolution of the trace tritium content in the JET tokamak. The study is based on measurements of the neutron emission, which were performed with the magnetic proton recoil (MPR) spectrometer. Data analysis procedures used to extract the results are described in some detail. The thesis also describes a simplified theoretical model to calculate the absolute tritium concentration with a comparison to the experimental results. The present study covers the time period 1996-2000 and the evolution of neutron emission spectroscopy (NES) results are compared with information from related diagnostic sources, and used to discuss the important issue of how tritium is retained in the JET tokamak

Titanium tritide radioisotope heat source development: palladium-coated titanium hydriding kinetics and tritium loading tests

International Nuclear Information System (INIS)

Van Blarigan, Peter; Shugard, Andrew D.; Walters, R. Tom

2012-01-01

We have found that a 180 nm palladium coating enables titanium to be loaded with hydrogen isotopes without the typical 400-500 C vacuum activation step. The hydriding kinetics of Pd coated Ti can be described by the Mintz-Bloch adherent film model, where the rate of hydrogen absorption is controlled by diffusion through an adherent metal-hydride layer. Hydriding rate constants of Pd coated and vacuum activated Ti were found to be very similar. In addition, deuterium/tritium loading experiments were done on stacks of Pd coated Ti foil in a representative-size radioisotope heat source vessel. The experiments demonstrated that such a vessel could be loaded completely, at temperatures below 300 C, in less than 10 hours, using existing department-of-energy tritium handling infrastructure.

Method for nondestructive fuel assay of laser fusion targets

International Nuclear Information System (INIS)

Farnum, E.H.; Fries, R.J.

1976-01-01

A method is described for nondestructively determining the deuterium and tritium content of laser fusion targets by counting the x rays produced by the interaction of tritium beta particles with the walls of the microballoons used to contain the deuterium and tritium gas mixture under high pressure. The x rays provide a direct measure of the tritium content and a means for calculating the deuterium content using the initial known D-T ratio and the known deuterium and tritium diffusion rates

CTR related tritium research at LASL

International Nuclear Information System (INIS)

Anderson, J.L.; Carstens, D.H.W.; Alire, R.M.

1975-01-01

The solubility and diffusion coefficients of H 2 in Li contained in a cylinder of Nb/1 percent Zr were measured in the temperature range 805-905 0 C at a pressure of 575 Pa. Appropriate corrections for the solubility of H 2 in Nb/1 percent Zr were made. As expected, a finite solid cylinder diffusion model adequately delineates the time-dependent consumption of H 2 in Li. The diffusion coefficient was found to vary from 3.88 x 10 -5 to 6.92 x 10 -5 cm 2 /sec and the activation energy was estimated to be 14.7 kcal/mole. The use of low melting eutectic mixtures containing Y, La, and Ce for extracting tritium from molten lithium is being investigated. Eutectic mixtures being investigated include the 16 percent in La, 12 percent Fe in Ce and 16 percent Co in Ce. At 875 0 K the molar distribution coefficients for tritium between these eutectic mixtures and lithium are less than or equal to 100. The temperature coefficients for these distributions are also being investigated. The pressure-composition-temperature diagram for the La 5 Ni-deuterium system is being studied. A ΔH of --39.7 kcal/mole D 2 was measured for the composition D/La 1.5. (auth)

Validation of effective half-life of tritium and computation of dose at Madras Atomic Power Station

International Nuclear Information System (INIS)

Moolya, L.L.; Kannan, R.K.; Sreekumaran Nair, B.; Mohandas, P.G.; Rajan, R.

2001-01-01

In Pressurised Heavy Water Reactors (PHWRs) tritium is produced by the activation of deuterium present in the moderator and coolant system. Hence, the inventory of tritium in PHWR is high and it contributes about 20 percent of the Station Dose. Tritium enters body by inhalation, ingestion and skin absorption. Dose received by an individual depends on its elimination rate from the body and hence the effective half-life. It is found that the effective half-life varies from person to person. The effective half-life also depends on the weather condition. The variation of effective half-life is taken care of in dose estimation by average method. However, in case of dose calculation for single uptake cases, there is a need to take fixed effective half-life. ICRP has taken the effective half-life as 10 days. The effective half-life is taken as 6 days in Indian conditions based on the studies conducted earlier in MAPS and elsewhere. Now with available data for about 16 years a fresh study was conducted to validate the effective half-life. This paper gives the study of effective half-life of tritium and its variation during different seasons of the year in MAPS. This paper also gives methods used for computing dose due to tritium. (author)

Deuterium isotope separation

International Nuclear Information System (INIS)

Benson, S.W.

1979-01-01

Deuterium-containing molecules are separated and enriched by exposing commercially available ethylene, vinyl chloride, 1,2-dichloroethane or propylene to the radiation of tuned infrared lasers to selectively decompose these compounds into enriched molecular products containing deuterium atoms. The deuterium containing molecules can be easily separated from the starting material by absorption, distillation or other simple chemical separation techniques and methods. After evaporation such deuterium containing molecules can be burned to form water with an enriched deuterium content or pyrolyzed to form hydrogen gas with an enriched deuterium content. (author)

Tritium handling and vacuum considerations for the STARFIRE commercial tokamak reactor

International Nuclear Information System (INIS)

Finn, P.A.; Clemmer, R.G.; Maroni, V.A.; Dillow, C.

1979-01-01

Tritium processing and vacuum pumping requirements were analyzed for the STARFIRE commercial fusion reactor design. It was found that vacuum pumps having a helium capture probability of 0.5 (total helium pump speed 1.2 x 10 4 m 3 /s) in combination with the proposed STARFIRE limiter-vacuum concept is sufficient to achieve plasma impurity control and, simultaneously, high fractional burnup (11%). The high fractional burnup and minimum fuel recycle time result in a very low fuel cycle tritium inventory, approx. 1300 g. A Lean-T burn method that can further reduce the fuel cycle inventory by 30 to 50% is discussed. D 2 O is proposed as a first wall coolant from considerations of plasma contamination (due to hydrogen isotope permeation through coolant tubes) and enrichment of recycled tritium from the coolant circuit. Tritium recovery from solid breeders, under realistic structural and breeder materials constraints, appears to represent a formidable task. The tritium inventory in the solid breeder is estimated to be as high as 10 kg, which would make the blanket the largest single hold-up point for tritium in the plant

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.

Tritium conductivity and isotope effect in proton-conducting perovskites

International Nuclear Information System (INIS)

Mukundan, R.; Brosha, E.L.; Birdsell, S.A.; Costello, A.L.; Garzon, F.H.; Willms, R.S.

1999-01-01

The tritium ion conductivities of SrZr 0.9 Yb 0.1 O 2.95 and BaCe 0.9 Yb 0.1 O 2.95 have been measured by ac impedance analysis. The high tritium conductivity of these perovskites could potentially lead to their application as an electrochemical membrane for the recovery of tritium from tritiated gas streams. The conductivities of these perovskites, along with SrCe 0.95 Yb 0.05 O 2.975 , were also measured in hydrogen- and deuterium-containing atmospheres to illustrate the isotope effect. For the strontium zirconate and barium cerate samples, the impedance plot consists of two clearly resolved arcs, a bulk and a grain boundary arc, in the temperature range 50--350 C. However, for the strontium cerate sample, the clear resolution of the bulk conductivity was not possible and only the total conductivity was measurable. Thus, the isotope effect was clearly established only for the strontium zirconate and barium cerate samples. The decrease in bulk conductivity with increasing isotope mass was found to be a result of an increase in the activation energy for conduction accompanied by a decrease in the pre-exponential factor. Since the concentration of the mobile species (H+, D+, or T+) should remain relatively constant at T < 350 C, this increase in activation energy is directly attributable to the increased activation energy for the isotope mobility

Measurement of anomalous neutron from deuterium/solid system

International Nuclear Information System (INIS)

Zhu Rongbao; Wang Xiaozhong; Lu Feng; Luo Longjun; He Jianyu; Ding Dazhao; Menlove, H.O.

1991-01-01

A series of experiments on both D 2 O electrolysis and thermal cycle of deuterium absorbed Ti Turnings are designed to examine the anomalous phenomena in Deuterium/Solid System. A neutron detector containing 16 BF 3 tubes with a detection limit of 0.38 n/s for two hour counting is used for electrolysis experiments. No neutron counting rate statistically higher than detection limit is observed from Fleischmann and Pons type experiments. An HLNCC-II neutron detector equipped with 18 3 He tubes and JSR-11 shift register unit with a detection limit of 0.20 n/s for a two hour run are employed to study the neutron signals in D 2 gas experiments. Ten batches of dry fusion samples are tested, among them, seven batches with neutron burst signals occur roughly at the temperature from -100 degrees centigrade to near room temperature. In the first four runs of a typical sample batch, seven neutron bursts are observed with neutron numbers from 15 to 482, which are 3 and 75 times, respectively, higher than the uncertainty of background. However, no bursts happened for H 2 dummy samples running in-between and afterwards and for sample batch after certain runs

Initial deuterium pellet experiments on FTU

International Nuclear Information System (INIS)

Snipes, J.A.

1993-01-01

Initial experiments have been performed with the Single Pellet INjector (SPIN) on FTU. SPIN is a two-stage cryogenic deuterium pellet injector capable of injection,a pellets with velocities up to 2.5 km/s. The nominal pellet mass for these experiments was approximately 1 x 10 20 atoms. These initial pellet experiments concentrated on studying pellet penetration under a variety of plasma conditions to compare with code predictions and to examine toroidal particle transport. The principal diagnostics used were two fast (∼1 μsec) photomultiplier tubes at nearly opposite toroidal locations with H α (D α ) interference filters (λ = 656 nm), a microwave cavity for pellet mass and velocity, a vertical array of soft x ray diodes without filters looking down onto the pellet, a DCN interferometer for electron density profiles, and a Michelson ECE system for electron temperature profiles. The time integral of the absolutely calibrated fast H α signal appears to give reasonable agreement with the expected pellet mass. Toroidal transport of deuterium ions from the pellet to nearly the opposite side of the tokamak agrees with calculated thermal deuterium velocities near the plasma edge. Comparison of the experimental results with code calculations using the Neutral Gas Shielding model show good agreement for the post-pellet electron temperature and density profiles and the H α profiles in some cases. Calculated penetration distances agree within 20%

Japanese university program on tritium radiobiology and environmental tritium

International Nuclear Information System (INIS)

Okada, Shigefumi

1989-01-01

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

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

International Nuclear Information System (INIS)

Nakamura, Hirofumi; Nishi, Masataka

2004-01-01

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

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

The influence of material variables on corrosion and deuterium uptake of Zr-2.5Nb alloy during irradiation

Energy Technology Data Exchange (ETDEWEB)

McDougall, G.M.; Urbanic, V.F. [Chalk River Technicians Lab., AECL, Ontario (Canada)

2002-07-01

Current CANDU 2 reactors use Zr-2.5Nb pressure tubes that are extruded at 1088 K, cold-drawn 27%, and autoclaved at 673 K for 24 h. This results in a metastable, two-phase microstructure consisting of elongated {alpha}-Zr grains surrounded by a network of {beta}-Zr filaments. To develop a mathematical model of corrosion and deuterium ingress in pressure tubes, we have considered the impact of variables including: fast neutron flux, temperature, and the as fabricated microstructure and its evolution during irradiation. Small specimens of Zr-2.5Nb are being exposed under CANDU water chemistry conditions in the Halden Boiling Water Reactor. The experiments involve fast neutron fluxes (E {>=} 1.05 MeV) of 0, 1.7, and 4.5 x 10{sup 17} n x m{sup -2} - s{sup -1}, and temperatures of 523 and 598 K. Specimens have been prepared from pressure tube materials representative of all current CANDU reactors, materials subject to thermal decomposition of the {beta}-Zr phase, and tubes extruded over a range of conditions. Results from the first three years of the Halden test program are summarized. At both 523 and 598 K, tubes made of {beta}-quenched material exhibit lower oxidation rates than those made from non-{beta}-quenched materials. In short-term out-of-flux exposures at 523 K, three non-{beta}-quenched tubes appear to show linear oxidation kinetics. Similar behavior is not observed in tests conducted out-of-flux at 598 K, or in-flux at either temperature. At 598 K, {beta}-quenched tubes exhibit significantly lower deuterium pickup rates than non-{beta}-quenched tubes. When tested at 598 K, thermally aged specimens show declining oxidation and deuterium pickup rates with increasing {beta}-Zr phase decomposition. At 523 K, the impact of thermal aging was less significant. Preliminary results from an 'extrusion variable test' suggest that tubes fabricated according to the current CANDU specification show the best corrosion resistance. (authors)

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.

Hydrodynamics and burn of optimally imploded deuterium-tritium spheres

International Nuclear Information System (INIS)

Mason, R.J.; Morse, R.L.

1975-01-01

The phenomenology of optimized laser-driven DT sphere implosions leading to efficient thermonuclear burn is reviewed. The optimal laser deposition profile for spheres is heuristically derived. The performance of a 7.5 μg sphere, exposed to its optimal 5.3 kJ pulse, is scrutinized in detail. The timing requirements for efficient central ignition of propagating burn in the sphere are carefully explored. The difficulties stemming from superthermal electron production and thermal flux limitation are discussed. The hydro-burn performance of spheres is characterized as a function of the pulse energy, peak power, time scale, pulse exponent, wavelength, and on the degree of flux limitation. The optimal pulse parameters are determined for spheres with masses ranging from 40 ng to 250 μg, requiring from 50 J to 150 kJ of input energy, and the corresponding optimal performance levels are calculated. Discussion is given to the hydro-burn performance of new structured fusion targets, in which the DT is contained as a gas or frozen as an ice shell inside a high Z pusher-tamper layer

Review of the knowledge available to date on the effects of tritium exposure on health and the environment in Canada - a tool to guide regulatory compliance monitoring

International Nuclear Information System (INIS)

Thompson, P.A.; Hamlat, M.S.; Lane, R.; Mihok, S.; Reinhardt, P.; Bundy, K.

2011-01-01

The use of tritium in CANDU (Canadian Deuterium-Uranium) reactors, in industry to produce self-luminescent lights and paints, in oil and gas exploration, in hospitals for diagnostic tests and radio-therapeutics, and in research makes the control of tritium releases generated by these activities particularly important in Canada. Releases are regulated and carefully monitored by the Canadian Nuclear Safety Commission (CNSC). Some special interest and citizen groups, however, claim that the scientific uncertainty regarding the effects of tritium on health and on the environment is such that regulation of the facilities releasing or using tritium may be inadequate. In response to these concerns, the CNSC asked its staff to initiate the 'Tritium Studies' project. As part of the project, the environmental fate of tritium and its health effects were studied through direct field measurements and the review of the latest scientific literature on the subject. The project made it possible to conclude that the tritium radiation protection measures and regulatory mechanisms are adequate in protecting the health and safety of Canadians. (authors)

Development of a new cellular solid breeder for enhanced tritium production

International Nuclear Information System (INIS)

Sharafat, Shahram; Williams, Brian; Ghoniem, Nasr; Ghoniem, Adam; Shimada, Masashi; Ying, Alice

2016-01-01

Highlights: • A new cellular solid breeder is presented with 2 to 3× the thermal conductivity and substantially higher density (∼90%) compared with pebble beds. • The cellular solid breeder contains an internal network of interconnected open micro-channels (∼50 –100 μm diam.) for efficient tritium release. • Cellular breeders are made by melt-infiltrating Li-based ceramic materials into an open-cell carbon foam followed by removal of the foam. • High temperature (750 °C and 40 °C/mm) cyclic compression tests demonstrated good structural integrity (no cracking) and low Young’s modulus of of <5 GPa. • Deuterium absorption–desorption release rates were comparable with those from pebble beds with similar characteristic T-diffusion lengths. - Abstract: A new high-performance cellular solid breeder is presented that has several times the thermal conductivity and is substantially denser compared with sphere-packed breeder beds. The cellular breeder is fabricated using a patented process of melt-infiltrating ceramic breeder material into an open-cell carbon foam. Following solidification the carbon foam is removed by oxidation. This process results in a near 90% dense robust freestanding breeder in a block configuration with an internal network of open interconnected micro-channels for tritium release. The network of interconnected micro-channels was investigated using X-ray tomography. Aside from increased density and thermal conductivity relative to pebble beds, high temperature sintering is eliminated and thermal durability is increased. Cellular breeder morphology, thermal conductivity, specific heat, porosity levels, high temperature mechanical properties, and deuterium charging-desorption rates are presented.

Development of a new cellular solid breeder for enhanced tritium production

Energy Technology Data Exchange (ETDEWEB)

Sharafat, Shahram, E-mail: sharams@gmail.com [University of California Los Angeles, 420 Westwood Pl., Los Angeles, CA 90095-1587 (United States); Williams, Brian [Ultramet, Pacoima, CA 91331-2210 (United States); Ghoniem, Nasr [University of California Los Angeles, 420 Westwood Pl., Los Angeles, CA 90095-1587 (United States); Ghoniem, Adam [Digital Materials Solutions, Inc., Westwood, CA 90024 (United States); Shimada, Masashi [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Ying, Alice [University of California Los Angeles, 420 Westwood Pl., Los Angeles, CA 90095-1587 (United States)

2016-11-01

Highlights: • A new cellular solid breeder is presented with 2 to 3× the thermal conductivity and substantially higher density (∼90%) compared with pebble beds. • The cellular solid breeder contains an internal network of interconnected open micro-channels (∼50 –100 μm diam.) for efficient tritium release. • Cellular breeders are made by melt-infiltrating Li-based ceramic materials into an open-cell carbon foam followed by removal of the foam. • High temperature (750 °C and 40 °C/mm) cyclic compression tests demonstrated good structural integrity (no cracking) and low Young’s modulus of of <5 GPa. • Deuterium absorption–desorption release rates were comparable with those from pebble beds with similar characteristic T-diffusion lengths. - Abstract: A new high-performance cellular solid breeder is presented that has several times the thermal conductivity and is substantially denser compared with sphere-packed breeder beds. The cellular breeder is fabricated using a patented process of melt-infiltrating ceramic breeder material into an open-cell carbon foam. Following solidification the carbon foam is removed by oxidation. This process results in a near 90% dense robust freestanding breeder in a block configuration with an internal network of open interconnected micro-channels for tritium release. The network of interconnected micro-channels was investigated using X-ray tomography. Aside from increased density and thermal conductivity relative to pebble beds, high temperature sintering is eliminated and thermal durability is increased. Cellular breeder morphology, thermal conductivity, specific heat, porosity levels, high temperature mechanical properties, and deuterium charging-desorption rates are presented.

Technical solutions for tritium removal from Cernavoda NPP heavy water systems

International Nuclear Information System (INIS)

Barariu, Gheorghe; Panait, Adrian

2002-01-01

In CANDU nuclear plants 2400 KCi/GW(e) - year tritium is generated. At a CANDU - 600 reactor similar to Cernavoda NPP Unit 1, 1500 KCi/year of tritium is generated 95% being in the D 2 O moderator, which can achieve a radioactivity level of 80 - 100 Ci/kg. Tritium in heavy water contributes with 30 - 50% to the doses received by operation personnel and with 20% to the radioactivity released to the environment. The extraction of tritium heavy water at CANDU reactors implies the following possibilities: - the radioactivity level reduction in the operation area; - the maintenance and repair cost reduction due to reduction of personnel protection measures and increased labor productivity; - the increase of NPP utilization factor by shutdown time reduction for maintenance and repair; - tritium concentration reduction from technological systems, ensuring thus the possibility of redesigning the systems in order to lower the cost of investment; - profitable use of extracted tritium. Technical measures provided by AECL project for CANDU 600 at Cernavoda make possible to satisfy the current standards concerning tritium concentration in the operation area atmosphere of 5 x 10 -6 Ci/m 3 . The regulations recommend that the radioactivity level should be maintained as low as possible in conformity with ALARA principles. Also, it is possible that norms will become more restrictive in the future, so the tritium removal technology is a good preventive measure which may become very necessary. The methods, which currently reached the industrial or pilot stages, are based on catalyzed chemical exchange, the heavy water electrolysis, and deuterium distillation. They are known as: VPCE - Vapour Phase Catalytic Exchange; LPCE - Liquid Phase Catalytic Exchange; DE - Direct Electrolysis; CD - Cryogenic Distillation. As transfer processes the catalyzed chemical exchange and heavy water electrolysis are used while concentration of tritium gas is done by cryogenic distillation. At present the

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.

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

Tritium transport in the water cooled Pb-17Li blanket concept of DEMO

International Nuclear Information System (INIS)

Reiter, F.; Tominetti, S.; Perujo, A.

1992-01-01

The code TIRP has been used to calculate the time dependence of tritium inventory and tritium permeation into the coolant and into the first wall boxes in the water cooled Pb-17Li blanket concept of DEMO. The calculations have been performed for the martensitic steel MANET and the austenitic steel AISI 316L as blanket structure materials, for water or helium cooling and for convective or no motion of the liquid breeder in the blanket. Tritium inventories are rather low in blankets with MANET structure and higher in those with AISI 316L structure. Tritium permeation rates are too high in both blankets. Further calculations on tritium inventory and permeation are therefore presented for blankets with TiC permeation barriers of 1 μm thickness on various surfaces of the blanket structure and for blankets with any permeation barriers in function of their thickness, tritium diffusivities, tritium surface recombination rates and atomic densities. These last calculations have been performed for a blanket with coatings on the outer surfaces of the blanket and with a tritium residence time of 10 4 s and for a blanket with coatings on both sides of the cooling tubes and stagnant Pb-17Li in the blanket. The second case for a blanket with MANET structure presents a very interesting solution for tritium recovery by permeation into and pumping from the first wall boxes. (orig.)

The Separation of Hydrogen Tritium and Tritium Hydride by Gas Chromatography; Separation de l'hydrogene, du tritium et de l'hydrure de tritium par chromatographic en phase gazeuse; Razdelenie vodoroda, tritiya i gidrida tritiya posredstvom gazovoj khromatografii; Separacion del hidrogeno, tritio e hidruro de tritio por cromatografia de gases

Energy Technology Data Exchange (ETDEWEB)

Smith, H A; Carter, Jr, E H [University of Tennessee, Knoxville, TN (United States)

1962-01-15

Now that successful separation of hydrogen, deuterium and hydrogen deuteride has been achieved by gas chromatography, similar studies are being made dealing with mixtures of hydrogen, tritium and tritium hydride. Since tritium is used in tracer quantities the usual katharometer cannot be employed for its detection. This difficulty has been overcome by providing immediately following the katharometer a vibrating reed electrometer equipped with a high resistance leak which allows continuous monitoring of the activity of any tritium or tritium hydride emerging from the column by means of synchronized recorders. Separation of such mixtures has been tested with columns packed with palladium on silica, silica, alumina, and alumina coated with chromium oxide or ferric oxide. No effective separation was obtained with the palladium on silica column. Good separation was achieved with the plain silica column where hydrogen was employed as the carrier gas, but helium failed to elute the isotopes. Satisfactory results were obtained with the coated, partially deactivated alumina packing when helium or neon was the carrier gas, but the best separation was found with a column packing of uncoated activated alumina. Calibration with helium-tritium mixtures of known activity plus equilibrated hydrogen-tritium mixtures also of known activity allows quantitative estimation of tritium and tritium hydride. (author) [French] La separation de l'hydrogene, du deuterium et du deuterure d'hydrogene par chromatographic en phase gazeuse ayant ete realisee, on procede maintenant a des etudes semblables sur des melanges d'hydrogene de tritium et d'hydrure de tritium. Comme le tritium n'est present qu'en quantites infimes, on ne peut utiliser le catharometre ordinaire pour le detecter. On a surmonte cette difficulte en faisant suivre le catharometre d'un electrometre a lame vibrante, muni d'une fuite haute resistance, qui permet de mesurer, a l'aide d'enregistreurs synchronises, l'activite de

Deuterium in-vessel retention characterisation through the use of particle balance on Tore Supra

International Nuclear Information System (INIS)

Bucalossi, J.; Brosset, C.; Pegourie, B.; Tsitrone, E.; Dufour, E.; Eckedahl, A.; Geraud, A.; Goniche, M.; Gunn, J.; Loarer, T.; Monier-Garbet, P.; Vallet, J.C.; Vartanian, S.

2007-01-01

Fuel retention inside plasma facing components will be a crucial issue not only in fusion reactors of the future, but also in ITER. The estimation of the fraction of the fuel which remains trapped inside the vessel is quite a difficult task. Particle balance analysis provides information for the whole vacuum chamber as a function of time and can be use to monitor the tritium in-vessel retention in real-time. On Tore Supra with a careful choice and position of pressure sensors, proper calibration procedures, the accuracy of the balance is around 10%. Particle balance analysis have been performed on many long pulse discharges and deuterium in-vessel retention has been found to be a constant around 5 x 10 20 D/s after several minutes of plasma. The evolution of the retention rate with plasma parameters indicates that deuterium bulk implantation and diffusion could dominate codeposition with carbon atoms. Particle balance is a powerful tool that should be implemented in ITER

Hypervelocity Launching and Frozen Fuels as a Major Contribution to Spaceflight

Science.gov (United States)

Cocks, F. H.; Harman, C. M.; Klenk, P. A.; Simmons, W. N.

Acting as a virtual first stage, a hypervelocity launch together with the use of frozen hydrogen/frozen oxygen propellant, offers a Single-Stage-To-Orbit (SSTO) system that promises an enormous increase in SSTO mass-ratio. Ram acceleration provides hypervelocity (2 km/sec) to the orbital vehicle with a gas gun supplying the initial velocity required for ram operation. The vehicle itself acts as the center body of a ramjet inside a launch tube, filled with gaseous fuel and oxidizer, acting as an engine cowling. The high acceleration needed to achieve hypervelocity precludes a crew, and it would require greatly increased liquid fuel tank structural mass if a liquid propellant is used for post-launch vehicle propulsion. Solid propellants do not require as much fuel- chamber strengthening to withstand a hypervelocity launch as do liquid propellants, but traditional solid fuels have lower exhaust velocities than liquid hydrogen/liquid oxygen. The shock-stability of frozen hydrogen/frozen oxygen propellant has been experimentally demonstrated. A hypervelocity launch system using frozen hydrogen/frozen oxygen propellant would be a revolutionary new development in spaceflight.

Isotope exchange reactions on ceramic breeder materials and their effect on tritium inventory

Energy Technology Data Exchange (ETDEWEB)

Nishikawa, M; Baba, A [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering; Kawamura, Y; Nishi, M

1998-03-01

Though lithium ceramic materials such as Li{sub 2}O, LiAlO{sub 2}, Li{sub 2}ZrO{sub 3}, Li{sub 2}TiO{sub 3} and Li{sub 4}SiO{sub 4} are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials has not been fully understood. The isotope exchange reaction rate between hydrogen isotopes in the purge gas and tritium on the surface of breeding materials have not been quantified yet, although helium gas with hydrogen or deuterium is planned to be used as the blanket purge gas in the recent blanket designs. The mass transfer coefficient representing the isotope exchange reaction between H{sub 2} and D{sub 2}O or that between D{sub 2} and H{sub 2}O in the ceramic breeding materials bed is experimentally obtained in this study. Effects of isotope exchange reactions on the tritium inventory in the bleeding blanket is discussed based on data obtained in this study where effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reactions are considered. The way to estimate the tritium inventory in a Li{sub 2}ZrO{sub 3} blanket used in this study shows a good agreement with data obtained in such in-situ experiments as MOZART, EXOTIC-5, 6 and TRINE experiments. (author)

Deuterium and heavy water

International Nuclear Information System (INIS)

Vasaru, G.; Ursu, D.; Mihaila, A.; Szentgyorgyi, P.

1975-01-01

This bibliography on deuterium and heavy water contains 3763 references (1932-1974) from 43 sources of information. An author index and a subject index are given. The latter contains a list of 136 subjects, arranged in 13 main topics: abundance of deuterium , catalysts, catalytic exchange, chemical equilibria, chemical kinetics, deuterium and heavy water analysis, deuterium and heavy water properties, deuterium and heavy water separation, exchange reactions, general review, heavy water as moderator, isotope effects, synthesis of deuterium compounds

Innovation in pressure tube life assessment

International Nuclear Information System (INIS)

Guler, B.; Kalenchuk, D.; Celovsky, A.

2003-01-01

The hydrogen equivalent concentration and the rate of hydrogen ingress (in particular, deuterium) in pressure tubes are important parameters that must be assessed to determine the fitness-for-service of CANDU reactors. This paper presents the latest refinement in a process referred to as 'Pressure Tube Sampling', which is the only fully qualified and proven method that allows accurate determination of both the hydrogen equivalent concentration and the rate deuterium ingress without performing an expensive fuel channel removal. Pressure Tube Sampling has evolved over the past fifteen years during which over 2,300 samples have been obtained from CANDU reactors around the globe. In-reactor sampling is the standard method for determining the hydrogen equivalent concentrations and deuterium ingress rates in CANDU reactors. Over the past fifteen years, continual improvements in the Pressure Tube Sampling process have resulted in: the capability to obtain circumferential and axial samples, reduced 'on-face' time, reduced cost, reduced dose to workers, and improved analysis accuracy. Most recently, the new Multi-Head Sampling Tool (MHST) has been developed that continues this trend by using one tool to sample at all four axial pressure tube locations in a single visit to the fuel channel, thereby further improving efficiency. In 2001 October, the MHST was successfully deployed at Wolsong 1 by AECL for Korea Hydro and Nuclear Power. The tool was delivered using their Advanced Delivery Machine (ADM) and a total of sixteen samples were obtained from four channels. A significant saving in time was achieved with a rate of one channel (four samples) being sampled every 2 1/2 hours. For a typical 10-channel campaign, this could equate to a 2 to 3 days time/saving, which is significant in terms of outage schedule, cost, and worker dose. This paper provides a description of some of the latest innovations, with specific details on site application, performance, and end results

Study of peculiarities of hydrogen isotopes mixture permeation through low activated steel F82H

International Nuclear Information System (INIS)

Kenzhin, Ye.A.; Tazhibayeva, I.L; Kulsartov, T.V.; Shestakov, V.P.; Chikhray, Ye.V.; Afanasev, S.E.; Zheldak, Yu.L.

2003-01-01

Full text: The problem of diffusion tritium leakage through blanket materials of future fusion device makes some constructive difficulties concerned with protection of personnel and environment and also with losses of tritium, which is planned to be used in the same device. One of the little-studied problems in the tritium leakage process in Fusion Power Plant is that in fact tritium will penetrate through materials while other hydrogen isotopes are present. These are deuterium and hydrogen which always are present in metals. Therefore, for evaluation of tritium leakage in future Fusion Power Plant under such conditions it is necessary to have experimental data about permeation of these hydrogen isotopes through the structure materials.One of proposed structure materials of fusion reactor blanket is low activated steel F82H. The experiment results on evaluation of .hydrogen, deuterium and its mixture interaction parameters with steel F82H are shown in this work. The tests were carried out within temperature range 273-973 K under inlet hydrogen pressure of 100-2000 Pa. Diffusivity, deuterium and hydrogen permeation constants for low activated steel F82H was determined from experiment results. Those experimental results were used for created phenomenology model which describes hydrogen isotope penetration through tube sample from hydrogen isotopes mixture. That model was used so determining the ratios of desorption rates (D-D, D-H, H-H) on outlet side of sample. Using of so obtained results, we can correctly evaluate, the titanium leakage from blanket of fusion machine which will be constructed using low activated steel F82H

Deuterium ingress at rolled joints in Embalse nuclear power plant

International Nuclear Information System (INIS)

Ramos Nervi, J. E.; Schroeter, F.

2013-01-01

Deuterium ingress model at the Rolled Joint has been extensively used for CANDU Nuclear Power Plants Operators in the Life Management of the Pressure Tubes. The importance of understanding the model is vital to avoid delayed hydride cracking at the Rolled Joint. This work reports the first step on develop the model presented on literature to be used in Argentinean CANDU 6, Embalse Nuclear Power Plant. (author)

Frozen ammonia micropellet generator for Baseball II-T

International Nuclear Information System (INIS)

Denhoy, B.S.

1975-01-01

A ''startup'' plasma at the center of the Baseball II-T magnet was studied. This startup plasma will be used as a target for high energy neutral beams to achieve the required build-up. The target plasma will be created by irradiating a solid pellet with a laser beam. Although a deuterium pellet would be superior because of purity, the development of an ammonia pellet was undertaken because it requires a simpler technology. The ammonia target plasma is physically acceptable for the initial experiment. A frozen ammonia pellet, about 100 μm in diameter, will be irradiated with 300-J CO 2 laser, to produce a density of about 10 13 cm -3 and about 1 kV temperature

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

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

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

Design of a Tritium-in-air-monitor using field programmable gate arrays

International Nuclear Information System (INIS)

McNelles, Phillip; Lu, Lixuan

2015-01-01

Field Programmable Gate Arrays (FPGAs) have recently garnered significant interest for certain applications within the nuclear field. Some applications of these devices include Instrumentation and Control (I and C) systems, pulse measurement systems, particle detectors and health physics purposes. In CANada Deuterium Uranium (CANDU) nuclear power plants, the use of heavy water (D2O) as the moderator leads to the increased production of Tritium, which poses a health risk and must be monitored by Tritium-In-Air Monitors (TAMs). Traditional TAMs are mostly designed using microprocessors. More recent studies show that FPGAs could be a potential alternative to implement the electronic logic used in radiation detectors, such as the TAM, more effectively. In this paper, an FPGA-based TAM is designed and constructed in a laboratory setting using an FPGA-based cRIO system. New functionalities, such as the detection of Carbon-14 and the addition of noble gas compensation are incorporated into a new FPGA-based TAM. Additionally, all of the standard functions included in the original microprocessor-based TAM, such as tritium detection, gamma compensation, pump and air flow control, and background and thermal drift corrections were also implemented. The effectiveness of the new design is demonstrated through simulations as well as laboratory testing on the prototype system. (author)

Determination of the tritium content in the reactor heavy water; Odredjivanje porasta kolicine tritijuma u reaktorskoj teskoj vodi

Energy Technology Data Exchange (ETDEWEB)

Ribnikar, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-01-15

Quantity of tritium was measured in the heavy water, in the heavy water vapour and radiolyzed deuterium from the helium cover gas of the RA reactor. It has been shown that isotopic equilibrium D{sub 2}O+DT{r_reversible}DTO+D{sub 2} exists and that it is catalyzed by irradiation. Small quantities of ammonium in the reactor cover gas are inhibiting the migration of tritium from the heavy water. Izmerena ja sadrzina tritijuma u tecnoj teskoj vodi, njenoj pari i radiolizovanom deuterijumu iz helijumske atmosfere reaktora RA. Pokazano je da postoji izotopska ravnoteza D{sub 2}O+DT{r_reversible}DTO+D{sub 2}, koja je katalizovana zracenjem. Male kolicine amonijaka reaktorske atmosfere deluju u smislu otezavanja migracije tritijuma iz teske vode (author)

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.

Suppression of hydrogenated carbon film deposition by scavenger techniques and their application to the tritium inventory control of fusion devices

International Nuclear Information System (INIS)

Tabares, F.L.; Tafalla, D.; Tanarro, I.; Herrero, V.J.; Islyaikin, A.; Maffiotte, C.

2002-01-01

The well-known radical and ion scavenger techniques of application in amorphous hydrogenated carbon film deposition studies are investigated in relation to the mechanism of tritium and deuterium co-deposition in carbon-dominated fusion devices. A particularly successful scheme results from the injection of nitrogen into methane/hydrogen plasmas for conditions close to those prevailing in the divertor region of present fusion devices. A complete suppression of the a-C : H film deposition has been achieved for N 2 /CH 4 ratios close to one in methane (5%)/hydrogen DC plasma. The implications of these findings in the tritium retention control in future fusion reactors are addressed. (author). Letter-to-the-editor

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

International Nuclear Information System (INIS)

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

1975-04-01

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

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

Tritium Aspects of Fueling and Exhaust Pumping in Magnetic Fusion Energy

Energy Technology Data Exchange (ETDEWEB)

Baylor, Larry R. [ORNL; Meitner, Steven J. [ORNL

2017-04-01

Magnetically confined fusion plasmas generate energy from deuterium-tritium (DT) fusion reactions that produce energetic 3.5 MeV alpha particles and 14 MeV neutrons. Since the DT fusion reaction rate is a strong function of plasma density, an efficient fueling source is needed to maintain high plasma density in such systems. Energetic ions in fusion plasmas are able to escape the confining magnetic fields at a much higher rate than the fusion reactions occur, thus dictating the fueling rate needed. These lost ions become neutralized and need to be pumped away as exhaust gas to be reinjected into the plasma as fuel atoms.The technology to fuel and pump fusion plasmas has to be inherently compatible with the tritium fuel. An ideal holistic solution would couple the pumping and fueling such that the pump exhaust is directly fed back into pellet formation without including impurity gases. This would greatly reduce the processing needs for the exhaust. Concepts to accomplish this are discussed along with the fueling and pumping needs for a DT fusion reactor.

Analysis of the separation of protium from blanket tritium-product streams

International Nuclear Information System (INIS)

Misra, B.; Maroni, V.A.

1981-07-01

The case is considered in which the blanket product stream has been purified to the point where only protium, tritium, and a small quantity of deuterium remain. A cryogenic distillation cascade concept developed specifically to handle this enrichment problem is shown. The concept is based on a series of distillation columns and equilibrators capable of producing a protium-rich stream containing less than 1000 appm T and a tritium-rich stream containing less than 2000 appm H. It is envisioned that both of these streams could be blended with streams of comparable composition in the mainstream position of the fuel cycle without further processing. The computational analysis of the cascade was based on a fixed arrangement of columns and equilibrators and a fixed number of theoretical plates per columns, since these features are less easily varied in an actual system than reflux ratios and flow rates. In order to test the flexibility of this conceptual enruchment system to adjust to variations of the H/T ratio in the feed, H/T values of 0.333, 1.00, and 3.00 were investigated

Continuous monitoring of tritium in water

International Nuclear Information System (INIS)

Descours, S.; Guerin, P.

1980-02-01

TRYDYN is a detector studied for continuous monitoring of tritium in water. Its sensitivity of approximately 10 -3 μCi/milliliter (one third of the maximum permissible tritium concentration for the population at large) makes it ideal for radiological protection applications (effluents flowing in process drains, sewers, etc ...). The effluent is filtered and then passed through a transparent flowcell contaIning plastic scintillator beads held between two photomultiplier tubes (PMTs). The cell's geometry and scintillator's geometry are designed to maximize measuring efficiency. Background is minimized by a 50 millimeter thick lead shielding and electronic circuitry of the same type as employed with liquid scintillators. This effluent purification unit can operate continuously for 8 days without manual intervention, the scintillator can operate for 6 months with a loss of sensitivity of less of 10%. The response time of the TRIDYN is less than 30 minutes for a concentration of 3.10 -3 μCi/milliliter [fr

Water containing deuterium electrolysis to obtain gaseous hydrogen isotope in a high state of purity

International Nuclear Information System (INIS)

Bellanger, Gilbert

1992-01-01

In this paper, the basic concept is to prepare hydrogen in a high state of purity by electrolysing water using a palladium cathode. During electrolysis, hydrogen is at first adsorbed at the palladium surface, and next it diffuses through it till opposite face of its entry where it is desorbed; thus permitting to regain it in a very pure state for storage. The method can be used from water containing deuterium. To improve hydrogen adsorption, surface effect of palladium must be studied. It was found that heat treatment of palladium improved the hydrogen permeation flux. The diffusivity of hydrogen is controlled by Fick and Sieverts equations in which temperature has a significant influence on permeation rates. Anyway, hydrogen desorption does not cause any difficulty. In a second part, we have studied the isotopic separation factor using water containing deuterium. We remarked in fact that it depends on current density, overpotential, diffusivity of hydrogen and deuterium and isotopic composition of electrolyte as expected. In the last part, we realized an original electrolysis model in a glove-box in which are taken into account the results given before and also the technology components employed in processes involving the use of tritium. (author) [fr

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Advances in associated-particle sealed-tube neutron probe diagnostics for substance detection

International Nuclear Information System (INIS)

Rhodes, E.; Dickerman, C.E.; Frey, M.

1995-01-01

The development and investigation of a small associated-particle sealed-tube neutron generator (APSTNG) shows potential to allow the associated-particle diagnostic method to be moved out of the laboratory into field applications. The APSTNG interrogates the inspected object with 14-MeV neutrons generated from the deuterium-tritium reaction and detects the alpha-particle associated with each neutron inside a cone encompassing the region of interest. Gamma-ray spectra of resulting neutron reactions identify many nuclides. Flight-times determined from detection times of the gamma-rays and alpha-particles can yield a separate coarse tomographic image of each identified nuclide, from a single orientation. Chemical substances are identified by comparing relative spectral line intensities with ratios of elements in reference compounds. The high-energy neutrons and gamma-rays penetrate large objects and dense materials. Generally no collimators or radiation shielding are needed. Proof-of-concept laboratory experiments have been successfully performed for simulated nuclear, chemical warfare, and conventional munitions. Most recently, inspection applications have been investigated for radioactive waste characterization, presence of cocaine in propane tanks, and uranium and plutonium smuggling. Based on lessons learned with the present APSTNG system, an advanced APSTNG tube (along with improved high voltage supply and control units) is being designed and fabricated that will be transportable and rugged, yield a substantial neutron output increase, and provide sufficiently improved lifetime to allow operation at more than an order of magnitude increase in neutron flux

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.)

IN-SITU TRITIUM BETA DETECTOR

International Nuclear Information System (INIS)

Berthold, J.W.; Jeffers, L.A.

1998-01-01

The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye. Based on the results achieved, it is premature to initiate Phase 2 and commit to a prototype

IN-SITU TRITIUM BETA DETECTOR

Energy Technology Data Exchange (ETDEWEB)

J.W. Berthold; L.A. Jeffers

1998-04-15

The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye. Based on the results achieved, it is premature to initiate Phase 2 and commit to a prototype

Measurement of extremely (2) H-enriched water samples by laser spectrometry: application to batch electrolytic concentration of environmental tritium samples.

Science.gov (United States)

Wassenaar, L I; Kumar, B; Douence, C; Belachew, D L; Aggarwal, P K

2016-02-15

Natural water samples artificially or experimentally enriched in deuterium ((2) H) at concentrations up to 10,000 ppm are required for various medical, environmental and hydrological tracer applications, but are difficult to measure using conventional stable isotope ratio mass spectrometry. Here we demonstrate that off-axis integrated cavity output (OA-ICOS) laser spectrometry, along with (2) H-enriched laboratory calibration standards and appropriate analysis templates, allows for low-cost, fast, and accurate determinations of water samples having δ(2) HVSMOW-SLAP values up to at least 57,000 ‰ (~9000 ppm) at a processing rate of 60 samples per day. As one practical application, extremely (2) H-enriched samples were measured by laser spectrometry and compared to the traditional (3) H Spike-Proxy method in order to determine tritium enrichment factors in the batch electrolysis of environmental waters. Highly (2) H-enriched samples were taken from different sets of electrolytically concentrated standards and low-level (tritium samples, and all cases returned accurate and precise initial low-level (3) H results. The ability to quickly and accurately measure extremely (2) H-enriched waters by laser spectrometry will facilitate the use of deuterium as a tracer in numerous environmental and other applications. For low-level tritium operations, this new analytical ability facilitated a 10-20 % increase in sample productivity through the elimination of spike standards and gravimetrics, and provides immediate feedback on electrolytic enrichment cell performance. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

Simulation of tritium behavior after intended tritium release in ventilated room

International Nuclear Information System (INIS)

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

2001-01-01

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

Deuterium exchange reaction in a trickle bed packed with a mixture of hydrophobic catalyst and hydrophilic packings

International Nuclear Information System (INIS)

Seungwoo Paek; Heui-Joo Choi; DO-Hee Ahn; Kwang-Rag Kim; Minsoo Lee; Sung-Paal Yim; Hongsuk Chung

2006-01-01

Full text of publication follows: The isotopic exchange reaction between hydrogen and water on the platinum supported catalysts provides a useful step for separating hydrogen isotopes such as deuterium and tritium. The CECE (Combined Electrolysis Catalytic Exchange) with a hydrophobic catalyst is a very effective method to remove small quantities of tritium from light or heavy waste water streams because of its high separation factor and mild operating conditions. The CECE column is composed of an electrolysis cell and a liquid phase catalytic exchange column. This paper deals with the experiments for the hydrogen isotopic exchange reaction in a trickle bed reactor packed with a hydrophobic catalyst in order to develop the catalytic column of the CECE. Hydrophobic Pt/SDBC catalyst which has been developed for the LPCE column of WTRF (Wolsong Tritium Removal Facility) was tested in a trickle bed reactor. The catalyst column was packed with a mixture of hydrophobic catalyst and hydrophilic packing (Dixon gauze ring) to improve liquid distribution and vapor/liquid transfer area. An experimental apparatus was built for the test of the catalyst at various temperatures and gas velocities. The catalyst was packed wet into the column and water was injected at the top through a liquid distributor and trickled through a catalyst mixture. Hydrogen gas passed up the column and deuterium was transferred to water stream flowing counter currently. The temperature of the column was controlled to maintain at 60 deg. C using water jackets around the reactor and equilibrator, a feed waster heater, and a circulation water heater. A metal bellows pump was used to circulate the hydrogen gas at the typical flow rate of 60 LPM.The reactor pressure was controlled to maintain at 135 kPa (abs) by a water column. Gas samples were drawn off from the top and bottom of the column. The difference in deuterium concentration between the inlet and outlet gas samples was analyzed using Gas

Design of a Cryogenic Distillation Column for JET Water Detritiation System for Tritium Recovery

International Nuclear Information System (INIS)

Parracho, A.I.; Camp, P.; Dalgliesh, P.; Hollingsworth, A.; Lefebvre, X.; Lesnoj, S.; Sacks, R.; Shaw, R.; Smith, R.; Wakeling, B.

2015-01-01

A Water Detritiation System (WDS) is currently being designed and manufactured to be installed in the Active Gas Handling System (AGHS) of JET, currently the largest magnetic fusion experiment in the world. JET has been designed and built to study fusion operating conditions with the plasma fuelling done by means of a deuterium-tritium gas mixture. AGHS is a plant designed and built to safely process gas mixtures and impurities containing tritium recovered from the JET torus exhaust gases. Tritium is removed from these gas mixtures and recycled. Tritium depleted gases are sent to Exhaust Detritiation System (EDS) for final tritium removal prior to discharge into the environment. In EDS, tritium and tritiated species are catalytically oxidized into water, this tritiated water is then adsorbed onto molecular sieve beds (MSB). After saturation the MSBs are heated and the water is desorbed and collected for tritium recovery. The WDS facility is designed to recover tritium from water with an average activity of 1.9 GBq/l, and is able to process water with activities of 85 GBq/l and higher. Tritiated water is filtered and supplied to the electrolyser where the water is converted into gaseous oxygen and tritiated hydrogen. The hydrogen stream is first purified by selective diffusion through membranes of palladium alloy and then is fed to two cryogenic distillation columns (CD). These operate in parallel or in series depending on the water activity. In the CD columns, hydrogen isotopes containing tritium are recovered as the bottom product and hydrogen, the top product, is safely discarded to a stack. The CD columns are foreseen to have a throughput between 200 and 300 mole/h of hydrogen isotopes vapour and they operate at approximately ≈21.2K and 105 kPa. The design of the CD columns will be presented in this work. This work has been carried out within the framework of the Contract for the Operation of the JET Facilities and has received funding from the European Union�

Deuterium-depleted water

International Nuclear Information System (INIS)

Stefanescu, Ion; Steflea, Dumitru; Saros-Rogobete, Irina; Titescu, Gheorghe; Tamaian, Radu

2001-01-01

Deuterium-depleted water represents water that has an isotopic content smaller than 145 ppm D/(D+H) which is the natural isotopic content of water. Deuterium depleted water is produced by vacuum distillation in columns equipped with structured packing made from phosphor bronze or stainless steel. Deuterium-depleted water, the production technique and structured packing are patents of National Institute of Research - Development for Cryogenics and Isotopic Technologies at Rm. Valcea. Researches made in the last few years showed the deuterium-depleted water is a biological active product that could have many applications in medicine and agriculture. (authors)

Gas-driven permeation of deuterium through tungsten and tungsten alloys

Energy Technology Data Exchange (ETDEWEB)

Buchenauer, Dean A., E-mail: dabuche@sandia.gov [Sandia National Laboratories, Energy Innovation Department, Livermore, CA 94550 (United States); Karnesky, Richard A. [Sandia National Laboratories, Energy Innovation Department, Livermore, CA 94550 (United States); Fang, Zhigang Zak; Ren, Chai [University of Utah, Department of Metallurgical Engineering, Salt Lake City, UT 84112 (United States); Oya, Yasuhisa [Shizuoka University, Graduate School of Science, Shizuoka (Japan); Otsuka, Teppei [Kyushu University, Department of Advanced Energy Engineering Science, Fukuoka (Japan); Yamauchi, Yuji [Hokkaido University, Third Division of Quantum Science and Engineering, Faculty of Engineering, Sapporo (Japan); Whaley, Josh A. [Sandia National Laboratories, Energy Innovation Department, Livermore, CA 94550 (United States)

2016-11-01

Highlights: • We have designed and performed initial studies on a high temperature gas-driven permeation cell capable of operating at temperatures up to 1150 °C and at pressures between 0.1–1 atm. • Permeation measurements on ITER grade tungsten compare well with past studies by Frauenfelder and Zahkarov in the temperature range from 500 to 1000 °C. • First permeation measurements on Ti dispersoid-strengthened ultra-fine grained tungsten show higher permeation at 500 °C, but very similar permeation with ITER tungsten at 1000 °C. Diffusion along grain boundaries may be playing a role for this type of material. - Abstract: To address the transport and trapping of hydrogen isotopes, several permeation experiments are being pursued at both Sandia National Laboratories (deuterium gas-driven permeation) and Idaho National Laboratories (tritium gas- and plasma-driven tritium permeation). These experiments are in part a collaboration between the US and Japan to study the performance of tungsten at divertor relevant temperatures (PHENIX). Here we report on the development of a high temperature (≤1150 °C) gas-driven permeation cell and initial measurements of deuterium permeation in several types of tungsten: high purity tungsten foil, ITER-grade tungsten (grains oriented through the membrane), and dispersoid-strengthened ultra-fine grain (UFG) tungsten being developed in the US. Experiments were performed at 500–1000 °C and 0.1–1.0 atm D{sub 2} pressure. Permeation through ITER-grade tungsten was similar to earlier W experiments by Frauenfelder (1968–69) and Zaharakov (1973). Data from the UFG alloy indicates marginally higher permeability (< 10×) at lower temperatures, but the permeability converges to that of the ITER tungsten at 1000 °C. The permeation cell uses only ceramic and graphite materials in the hot zone to reduce the possibility for oxidation of the sample membrane. Sealing pressure is applied externally, thereby allowing for elevation

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

International Nuclear Information System (INIS)

Lei, Jiarong; Du, Yang; Yang, Yong

2008-01-01

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

New extruder-based deuterium feed system for centrifuge pellet injection

International Nuclear Information System (INIS)

Combs, S.K.; Foust, C.R.

1997-01-01

The pellet injection systems for the next-generation fusion devices (such as the International Thermonuclear Experimental Reactor) and future fusion reactors will have to provide deuterium-tritium fueling for much longer pulse lengths (up to ∼1000s) than present applications (typically limited to less than several seconds). Thus, a prototype pellet feed system for centrifuge pellet injection has been developed and used in long-pulse (>100s) tests at the Oak Ridge National Laboratory (ORNL). The new apparatus has two key components: (1) a cryogenic deuterium extruder and (2) an electromagnetic pellet punch mechanism. For maximum testing flexibility, the prototype is equipped with several other active components that allow remote adjustments, including precise positioning of the punch and the capability to index through eight different pellet lengths. The new feed system was designed to mate with an existing centrifuge accelerator facility at ORNL, and experiments in the facility were carried out to document the performance and reliability of the new feed system. With 2.3-mm-diam deuterium pellets and a catenary-shaped accelerator (∼1.2mdiam), the prototype feed system was found to be capable of placing up to ∼90% of the punched pellets in the proper time/space window for pickup and acceleration by the high-speed rotating (∼50Hz) arbor. For these operating parameters, the pellet nominal speed was ∼430m/s, and maximum pellet feed rates of 10 pellets/s and greater were tested. In this article the equipment is briefly described, and the experimental test results are summarized. Also, issues affecting overall pellet delivery efficiency are discussed. copyright 1997 American Institute of Physics

A geochemical and hydrological investigation of groundwater recharge in the Roswell basin of New Mexico: summary of results and updated listing of tritium determinations

International Nuclear Information System (INIS)

Gross, G.W.; Hoy, R.N.

1980-04-01

Different approaches were used to study recharge and flow patterns in the Roswell, New Mexico, artesian basin. Isotope determination for tritium, deuterium, and oxygen-18 were made as a function of time and space. Observation well levels, springflow, and precipitation were analyzed by stochastic/numerical approaches. Also, a hydrogeologic survey was made of representative springs in the recharge zone on the basin western flank. An updated listing of tritium activity in precipitation, springs, surface runoff, and subsurface water from over 120 sampling sites in the basin covers 117 pages of the report. Substantial deep leakage contributions from the basin western flank must be included to account for the basin groundwater budget

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.)

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

International Nuclear Information System (INIS)

Foerstel, H.

1997-01-01

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

Associated-particle sealed-tube neutron probe: Detection of explosives, contraband, and nuclear materials

International Nuclear Information System (INIS)

Rhodes, E.; Dickerman, C.E.

1996-01-01

Continued research and development of the APSTNG shows the potential for practical field use of this technology for detection of explosives, contraband, and nuclear materials. The APSTNG (associated-particle sealed-tube generator) inspects the item to be examined using penetrating 14-MeV neutrons generated by the deuterium-tritium reaction inside a compact accelerator tube. An alpha detector built into the sealed tube detects the alpha-particle associated with each neutron emitted in a cone encompassing the volume to be inspected. Penetrating high-energy gamma-rays from the resulting neutron reactions identify specific nuclides inside the volume. Flight-times determined from the detection times of gamma-rays and alpha-particles separate the prompt and delayed gamma-ray spectra and allow a coarse 3-D image to be obtained of nuclides identified in the prompt spectrum. The generator and detectors can be on the same side of the inspected object, on opposite sides, or with intermediate orientations. Thus, spaces behind walls and other confined regions can be inspected. Signals from container walls can be discriminated against using the flight-time technique. No collimators or shielding are required, the neutron generator is relatively small, and commercial-grade electronics are employed. The use of 14-MeV neutrons yields a much higher cross-section for detecting nitrogen than that for systems based on thermal-neutron reactions alone, and the broad range of elements with significant 14-MeV neutron cross-sections extends explosives detection to other elements including low-nitrogen compounds, and allows detection of many other substances. Proof-of-concept experiments have been successfully performed for conventional explosives, chemical warfare agents, cocaine, and fissionable materials

Hydrogen and deuterium transport and inventory parameters in a Cu-0.65Cr-0.08Zr alloy for fusion reactor applications

International Nuclear Information System (INIS)

Serra, E.; Perujo, A.

1998-01-01

A time dependent permeation method has been used to measure the transport and inventory parameters of hydrogen and deuterium in the precipitation hardened (PH) Cu-0.65 wt% Cr-0.08 wt% Zr alloy (ELBRODUR HF). The measurements cover the temperature range from 553 to 773 K using a hydrogen or deuterium pressure in the range 1-100 kPa. The permeabilities, diffusivities and Sieverts' constant values for ELBRODUR obtained in this work were about 2 time higher, 2 times lower and between 3 and 10 time higher respectively to that of Cu (F. Reiter et al., 1993) in the temperature range studied. Using the measured diffusivities and Sieverts' constant values for H 2 and D 2 and a quantum-mechanical model, an extrapolation for tritium is also presented. (orig.)

Effect of deposited tungsten on deuterium accumulation in beryllium in contact with atomic deuterium

Energy Technology Data Exchange (ETDEWEB)

Sharapov, V.M.; Gavrilov, L.E. [Institute of Physical Chemistry, Russian Academy of Sciences, Moscow (Russian Federation); Kulikauskas, V.S.

1998-01-01

Usually ion or plasma beam is used for the experiment with beryllium which simulates the interaction of plasma with first wall in fusion devices. However, the use of thermal or subthermal atoms of hydrogen isotopes seems to be useful for that purpose. Recently, the authors have studied the deuterium accumulation in beryllium in contact with atomic deuterium. The experimental setup is shown, and is explained. By means of elastic recoil detection (ERD) technique, it was shown that in the exposure to D atoms at 740 K, deuterium is distributed deeply into the bulk, and is accumulated up to higher concentration than the case of the exposure to molecular deuterium. The depth and concentration of deuterium distribution depend on the exposure time, and those data are shown. During the exposure to atomic deuterium, oxide film grew on the side of a sample facing plasma. In order to understand the mechanism of deuterium trapping, the experiment was performed using secondary ion mass spectrometry (SIMS) and residual gas analysis (RGA). The influence that the tungsten deposit from the heated cathode exerted to the deuterium accumulation in beryllium in contact with atomic deuterium was investigated. These results are reported. (K.I.)

Determination of the Deuterium Abundances in Water from 156 to 10,000 ppm by SIFT-MS

Czech Academy of Sciences Publication Activity Database

Španěl, Patrik; Shestivska, Violetta; Chippendale, T. W. E.; Smith, D.

2011-01-01

Roč. 22, č. 1 (2011), s. 179-186 ISSN 1044-0305 Institutional research plan: CEZ:AV0Z40400503 Keywords : deuterium abundance * total body water * selected ion flow tube mass spectrometry Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 4.002, year: 2011

Operation of the repeating pneumatic injector on TFTR and design of an 8-shot deuterium pellet injector

International Nuclear Information System (INIS)

Combs, S.K.; Milora, S.L.; Foust, C.R.

1985-01-01

The repeating pneumatic hydrogen pellet injector, which was developed at the Oak Ridge National Laboratory (ORNL), has been installed and operated on the Tokamak Fusion Test Reactor (TFTR). The injector combines high-speed extruder and pneumatic acceleration technologies to propel frozen hydrogen isotope pellets repetitively at high speeds. The pellets are transported to the plasma in an injection line that also serves to minimize the gas loading on the torus; the injection line incorporates a fast shutter valve and two stages of guide tubes with intermediate vacuum pumping stations. A remote, stand-alone control and data acquisition system is used for injector and vacuum system operation. In early pellet fueling experiments on TFTR, the injector has been used to deliver deuterium pellets at speeds ranging from 1.0 to 1.5 km/s into plasma discharges. First, single large (nominal 4-mm-dia) pellets provided high densities in TFTR (1.8 x 10 14 cm -3 on axis); after conversion to smaller (nominal 2.7-mm-dia) pellets, up to five pellets were injected at 0.25-s intervals into a plasma discharge, giving a line-averaged density of 1 x 10 14 cm -3 . Operating characteristics and performance of the injector in initial tests on TFTR are presented

Phase 1 Final Report for In-Situ Tritium Beta Detector

Energy Technology Data Exchange (ETDEWEB)

Berthold, J.W.; Jeffers, L.A.

1998-04-15

The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye.

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-08-14

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

Heating and transport in TFTR D-T plasmas

International Nuclear Information System (INIS)

Zarnstorff, M.C.; Scott, S.D.

1994-01-01

The confinement and heating of supershot plasmas are significantly enhanced with tritium beam injection relative to deuterium injection in TFTR. The global energy confinement and local thermal transport are analyzed for deuterium and tritium fueled plasmas to quantify their dependence on the average mass of the hydrogenic ions. The radial profiles of the deuterium and tritium densities are determined from the DT fusion neutron emission profile

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)