An electrical resistivity monitor for the detection of composition changes in Pb-17Li

International Nuclear Information System (INIS)

Hubberstey, P.; Barker, M.G.; Sample, T.

1991-01-01

An electrical resistivity monitor for the detection of composition changes in the lithium-lead eutectic alloy, Pb-17Li, has been developed. A miniature electromagnetic pump is used to sample alloy continuously from a pool or loop system and force it through a capillary section, within which the necessary resistance measurements are made, prior to its return to the bulk source. To calibrate the monitor, detailed resistivity-temperature and resistivity-composition data have been determined for Pb-Li alloys at temperatures from 600 to 800K and compositions from 0 to 20.5 at% Li. The resistivity increases with both temperature and composition; for Pb-17li at 723 K, dρ/dT=0.054x10 -8 ΩmK -1 , and dρ/d[Li]=1.27x10 -8 Ωm(at% Li) -1 . The sensitivity of the monitor is such that changes in composition of as little as ±0.05 at% Li can be detected and its response time is limited soley by the rate of sampling. (orig.)

Corrosion of ferrous alloys exposed to thermally convective Pb-17 at. % Li

International Nuclear Information System (INIS)

Tortorelli, P.F.; DeVan, J.H.

1986-01-01

A type 316 stainless steel thermal convection loop with type 316 stainless steel coupons and a Fe-9 Cr-1 Mo steel loop containing Fe-12 Cr-1 MoVW steel specimens circulated molten Pb-17 at. % Li at a maximum temperature of 500 0 C. Specimens were exposed for greater than 6000 h. Mass loss and surface characterization data were compared for these two alloys. At any particular exposure time, the corrosion of type 316 stainless steel by Pb-17 at. % Li was more severe, and of a different type than that of similarly exposed Fe-12 Cr-1 MoVW steel. The austenitic alloy suffered nonuniform penetration and dissolution by the lead-lithium, whereas the Fe-12 Cr-1 MoVW steel tended to be more uniformly corroded. The presence of a ferritic layer on the type 316 stainless steel, and its susceptibility to spalling during specimen cleaning, were shown to be important in evaluating the data and in comparing corrosion losses for the type types of alloys. A model for the nonuniform penetration of type 316 stainless steel by Pb-17 at. % Li was suggested

Perspectives for tritium recovery from liquid Pb83Li17 alloy

International Nuclear Information System (INIS)

Pierini, G.

1983-01-01

A preliminary analysis has been made on the perspectives of tritium recovery from a blanket constituted by liquid Pb 83 Li 17 alloy, by investigating the Tsub(2g) 2T alloy thermodynamic behaviour and trying to individuate those chemico-physical characteristics of which knowledge is useful for the effective design of the apparatus. In the light of the recent work carried out on the systems, hydrogen and its isotopes with molten Li-Pb eutectics, it has been possible to individuate a fair reliability of the Sieverts constant for the T 2 -Pb 83 Li 17 system. As far as the kinetics of the tritium desorption process from the liquid phase are concerned, the hydrogen (and its isotopes) transfer rates across the liquid interphase have been evaluated. Even though in first approximation the diffusion coefficient of tritium in the molten alloy at 400 0 C has been found. Then, the processing equations of attractive tritium recovery systems have been presented, individuating the fundamental parameters and reporting an example of applied engineering of the desorption system in a packed column. None of these tritium recovery systems seem incompatible at the actual state of the research

The creep-rupture behaviour of the martensitic steel X18CrMoVNb 121 (no.1.4914) in liquid Pb-17 Li at 550 and 6000C

International Nuclear Information System (INIS)

Grundmann, M.; Borgstedt, H.U.; Schirra, M.

1988-01-01

One of the candidate structural materials for the NET blanket is the martensitic steel X18 CrMoVNb 12 1 (no.1.4914). Its compatibility with the molten eutectic Pb-17Li, which might be used as liquid breeder and coolant in a self-cooled liquid metal blanket, should be satisfying even under superimposed mechanical stress. The mechanical high-temperature strength of the steel should not be significantly reduced by the interaction with the liquid metal which is in close contact with the surface of the components of such a blanket. The corrosion behaviour of this steel in flowing Pb-17Li eutectic is also studied, results will be presented at this conference. A certain influence of a liquid metal environment on the creep-rupture behaviour of steels was observed earlier in a study on the mechanical properties of austenitic stainless steel in liquid sodium. Therefore, a test programme was initiated to evaluate the effects of liquid Pb-17Li alloy on the creep strength of the steel no. 1.4914. Liquid lithium environment showed an influence on the fracture of this material in short time tests at moderate temperature

Thermophysical properties of the Li(17)Pb(83) eutectic alloy

International Nuclear Information System (INIS)

Jauch, U.; Haase, G.; Schulz, B.

1986-01-01

Methods of measurements and results for the following properties of Li(17)Pb(83) are presented: density, specific heat, latent heat of fusion, surface energy, thermal conductivity and diffusivity, electrical conductivity and viscosity. The range of the temperature for extrapolation of the physical properties is discussed. (orig.)

Evaporation of lead and lithium from molten Pb-17Li - transport of aerosols

International Nuclear Information System (INIS)

Feuerstein, H.; Graebner, H.; Oschinski, J.; Horn, S.; Bender, S.

1991-01-01

Evaporation of Pb and Li from molten Pb-17Li was investigated between 350 and 800deg C in vacuum, argon and helium covergas. Results were also obtained from other experimental facilities. Similarities were found to observations from sodium cooled reactors. The results show that Pb and Li evaporate independent on each other. The two elements show different behavior along the transport pathway. Deposits of the evaporated metals contained between 0.2 and 98 at% Li. As in the reactor RAPSODIE for sodium, evaporation rates for lithium were smaller in helium than in argon, however evaporation rates of lead were the same in both gases. No aerosol problems will exist with normal blanket operation. Under experimental conditions, aerosol concentrations were in the range of 10 -9 to 10 -6 g/m 3 . Aerosols can easily be trapped with sintered metal filters. (orig.)

Continuous monitoring of the composition of liquid Pb-17Li eutectic using electrical resistivity methods

International Nuclear Information System (INIS)

Hubberstey, P.; Sample, T.; Barker, M.G.

1991-01-01

The composition of liquid Pb-17Li alloys has been continously determined, using an electrical resistivity monitor, during their interaction with nitrogen, oxygen, hydrogen and water vapour. The operation of the monitor depends on the fact that the resistivity of liquid Pb-Li alloys is dependent on their composition. Accurate resistivity-composition isotherms have been derived from resistivity-temperature data for 15 Pb-Li alloys (0 Li -8 Ω m (mol% Li) -1 at 725 K) is such that a change of 0.05 mol% Li in the alloy composition can be measured. The addition of oxygen and water vapour resulted in a decrease in the resistivity of the liquid alloy. Neither nitrogen nor hydrogen had any effect. The observed changes were shown to be consistent with Li 2 O formation. (orig.)

Determination of the diffusion coefficients of iron and chromium in Pb17Li at 500 deg C

International Nuclear Information System (INIS)

Simon, N.; Flament, T.; Terlain, A.

1992-01-01

The diffusion of the dissolved metallic species in a liquid metal towards the boundary layer is one of the elementary steps of the overall mass transfer process induced by thermal gradient. This phenomenon is very probably the limiting step in the mass transfer of martensitic Fe-Cr steels in the presence of Pb17Li liquid eutectic alloy. For estimating diffusion flux, the diffusion coefficients of iron and chromium in Pb17Li are needed but are not known. Consequently these data have been determined in CEA laboratory by measuring metal loss of cylindrical specimens after rotation at 500 deg C in Pb17Li for several hours and applying the first Fick diffusion law in the boundary layer whose the thickness has been previously determined by EISENBERG. After a description of the experimental device, the results are presented and discussed

Dynamic response of INTOR/NET blankets after coolant tube rupture

International Nuclear Information System (INIS)

Klippel, H.T.

1985-01-01

The dynamic response of different water-cooled liquid Li 17 Pb 83 breeder blanket modules has been calculated to study the potential of these modules in case of coolant tube rupture. Numerical calculations with the code PISCES have been carried out taking into account the fluid-structure interaction and the elasto-plastic behaviour of the structural material. The results show that for inert coolant characteristics the proposed conceptual designs for NET and INTOR have sufficient resistance against coolant tube rupture but when taking into account energy release due to chemical reaction of water with LiPb-alloy up to doubling of the wall thickness has to be envisaged to guarantee structural reliability. (orig.)

Neutronics optimization of LiPb-He dual-cooled fuel breeding blanket for the fusion-driven sub-critical system

International Nuclear Information System (INIS)

Zheng Shanliang; Wu Yican

2002-01-01

The concept of the liquid Li 17 Pb 83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR > 1.05) and annual output of 100 kg or more fissile 239 Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimized calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio (Br = Tbr + Fbr) is listed corresponding to different cases

Interaction of hydrogen with Pb83Li17 eutectic alloy

International Nuclear Information System (INIS)

Kumar, Sanjay; Taxak, Manju; Krishnamurthy, N.

2011-01-01

Liquid Metal blankets are attractive candidates for both near-term and long-term fusion applications. Lead-lithium alloy appears to be promising for the use in self cooled breeding blanket, which has inherent simplicity with candidate material liquid lithium serving as both breeder and coolant. The crucial issues in case of lead lithium are the permeation loss of tritium (T) to the coolant and surroundings and the formation of new phase LiH/LiT, which eventually change the physical properties. Present investigation is based on the interaction process of hydrogen with the alloy and the relevant changes in physical properties. (author)

Behavior of radioisotope in liquid neutron irradiated Pb-17Li eutectic

International Nuclear Information System (INIS)

Tebus, V.N.; Aksenov, B.S.; Klabukov, U.G.

1994-01-01

Investigation of radioisotope 210 Po evaporation from liquid neutron irradiated Pb- 17 Li eutectic has been performed by Knudsen method. Equilibrium 210 Po vapor pressures at temperatures 250-700 degrees C were found about 3-4 orders of magnitude less than that for pure Po and were closed to equilibrium vapor pressures of Po-Pb compound. It was proposed Po forms stable Po-Pb compounds in eutectic at temperatures up to 750-800 degrees C. But disintegrates during long storage owing to self irradiation. It was determined Po aerosol transfer with radio gases takes place at the melting period. Contamination is happened also under irradiated eutectic storage at room temperature owing to aggregate recoil characteristic of Po

Manufacturing and characterization of porous SiC for flow channel inserts in dual-coolant blanket designs

International Nuclear Information System (INIS)

Bereciartu, Ainhoa; Ordas, Nerea; Garcia-Rosales, Carmen; Morono, Alejandro; Malo, Marta; Hodgson, Eric R.; Abella, Jordi; Sedano, Luis

2011-01-01

SiC is the primary candidate for the flow channel inserts in dual-coolant blanket concepts. Porous SiC ceramics are attractive candidates for this non-structural application, since they can satisfy the required properties through a low cost manufacturing route, compared to SiC f /SiC. This work shows first results of the manufacturing of porous SiC ceramics prepared with different amounts of Y 2 O 3 and Al 2 O 3 as sintering additives. C powders were used as pore-formers by their burnout during oxidation after sintering. Comparison of microstructure, porosity, flexural strength, thermal and electrical conductivity and corrosion under Pb-15.7Li of porous SiC without and with sintering additives is presented. The addition of 2.5 wt.% of Y 2 O 3 and Al 2 O 3 improves the mechanical properties, and reduces the thermal and electrical conductivity down to reasonable values. Preliminary corrosion tests under Pb-15.7 Li at 500 deg. C show that the absence of a dense coating on porous SiC leads to poor corrosion behavior.

Influence of a magnetic field on the corrosion of austenitic and martensitic steels by semi-stagnant Pb17Li

International Nuclear Information System (INIS)

Terlain, A.; Dufrenoy, T.

1994-01-01

The influence of a magnetic field on the compatibility of 316L austenitic and 1.4914 martensitic steels with Pb17Li has been studied in conditions simulating the special features of the water-cooled Pb17Li blanket (low Pb17Li velocity, significant radial thermal gradient and short distances between hot and cold zones). In the 420-475 C temperature range, the results show an increase of the corrosion rate in the presence of a magnetic field. This increase is about 50% for 316L steel and 30% for 1.4914 martensitic steel. Moreover the magnetic field induces a loss of symmetry in the deposition process: the amount of recovered deposit is greater in the direction parallel to the magnetic field than in the perpendicular one. ((orig.))

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.))

Corrosion behavior of CLAM steel weldment in flowing liquid Pb-17Li at 480 °C

International Nuclear Information System (INIS)

Chen Xizhang; Shen Zheng; Chen Xing; Lei Yucheng; Huang Qunying

2011-01-01

Highlights: ► The research shows that the CLAM steel weldment have its own corrosion mechanism in liquid Pb-17Li alloy. The basic rule of the corrosion behaviour of weldments is that the coorosion rate decreases obviously with the increasing of exposed time. ► The weight loss of CLAM steel weldment is far higer than the base metal after exposed to Pb-17Li alloy. Corrosion has little effects on elements of weldment sample surfaces. And an easier corrosion area in the weld joint are found. ► A simple presumably corrosion behavior model is established. The model demonstrate that the easier corroded area will be formed when the direction of martensite laths form small-angle with the specimen surface, The easy corrosion area is the martensite lath area lack of Cr and distributes like laths, the cross-section area is 1 μm 2 to 4 μm 2, the existence of the easier corrosion area is one of the reasons that lead to the difference of the corrosion rate. - Abstract: CLAM (China Low Activation Martensitic) steel is considered as one of the candidate structural materials in liquid LiPb blanket concepts. Welding is one of the essential technologies for its practical application, CLAM steel weldment shows a great difference with base metal due to the effect of welding thermal cycle. In order to investigate the corrosion behavior and mechanism of CLAM weldments in liquid Pb-17Li, the experiments were performed by exposing the TIG weldment samples in flowing LiPb at 480 °C. The weight loss test of exposed specimens show that in 500 h, 1000 h dynamic conditions, corrosion resistance of CLAM steel weldment is poor, SEM analysis shows that the thicker martensite lath in weld area lead to higher corrosion amount, EDS results show that the influence of corrosion on surface elements is small, and surface corrosion is even, EDX analysis shows that the penetration of Pb-17Li does not exist in the joint. With the increasing of exposure time, the corrosion rate decreases

Analysis of tritium behaviour and recovery from a water-cooled Pb17Li blanket

International Nuclear Information System (INIS)

Malara, C.; Casini, G.; Viola, A.

1995-01-01

The question of the tritium recovery in water-cooled Pb17Li blankets has been under investigation for several years at JRC Ispra. The method which has been more extensively analysed is that of slowly circulating the breeder out from the blanket units and of extracting the tritium from it outside the plasma vacuum vessel by helium gas purging or vacuum degassing in a suited process apparatus. A computerized model of the tritium behaviour in the blanket units and in the extraction system was developed. It includes four submodels: (1) tritium permeation process from the breeder to the cooling water as a function of the local operative conditions (tritium concentration in Pb17Li, breeder temperature and flow rate); (2) tritium mass balance in each breeding unit; (3) tritium desorption from the breeder material to the gas phase of the extraction system; (4) tritium extraction efficiency as a function of the design parameters of the recovery apparatus. In the present paper, on the basis of this model, a parametric study of the tritium permeation rate in the cooling water and of the tritium inventory in the blanket is carried out. Results are reported and discussed in terms of dimensionless groups which describe the relative effects of the overall resistance on tritium transfer to the cooling water (with and without permeation barriers), circulating Pb17Li flow rate and extraction efficiency of the tritium recovery unit. The parametric study is extended to the recovery unit in the case of tritium extraction by helium purge or vacuum degassing in a droplet spray unit. (orig.)

Manufacturing and characterization of porous SiC for flow channel inserts in dual-coolant blanket designs

Energy Technology Data Exchange (ETDEWEB)

Bereciartu, Ainhoa [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Ordas, Nerea, E-mail: nordas@ceit.es [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Garcia-Rosales, Carmen [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Morono, Alejandro; Malo, Marta; Hodgson, Eric R. [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Abella, Jordi [Institut Quimic de Sarria, University Ramon Llull, Via Augusta 390, 08017 Barcelona (Spain); Sedano, Luis [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain)

2011-10-15

SiC is the primary candidate for the flow channel inserts in dual-coolant blanket concepts. Porous SiC ceramics are attractive candidates for this non-structural application, since they can satisfy the required properties through a low cost manufacturing route, compared to SiC{sub f}/SiC. This work shows first results of the manufacturing of porous SiC ceramics prepared with different amounts of Y{sub 2}O{sub 3} and Al{sub 2}O{sub 3} as sintering additives. C powders were used as pore-formers by their burnout during oxidation after sintering. Comparison of microstructure, porosity, flexural strength, thermal and electrical conductivity and corrosion under Pb-15.7Li of porous SiC without and with sintering additives is presented. The addition of 2.5 wt.% of Y{sub 2}O{sub 3} and Al{sub 2}O{sub 3} improves the mechanical properties, and reduces the thermal and electrical conductivity down to reasonable values. Preliminary corrosion tests under Pb-15.7 Li at 500 deg. C show that the absence of a dense coating on porous SiC leads to poor corrosion behavior.

Corrosion behaviour of Al based tritium permeation barriers in flowing Pb-17Li

International Nuclear Information System (INIS)

Glasbrenner, H.; Konys, J.; Voss, Z.; Wedemeyer, O.

2002-01-01

Tritium permeation barriers on low-activation steels are required in fusion technology in order to reduce the tritium permeation rate through the structural material into the cooling water system. Al-Fe layers with alumina on top can fulfil the required reduction rate. Three techniques were selected to produce such a multi-layered coating system: chemical vapour deposition (CVD) by CEA, hot-dip aluminising (HDA) by FZK and vacuum plasma spraying (VPS) by JRC Ispra. A sufficient corrosion resistance against Pb-17Li attack is also required for the coating. Therefore, the corrosion behaviour of these three coatings on ferritic-martensitic steels was studied in the PICOLO loop of FZK in flowing Pb-17Li at 480 deg. C up to 10 000 h. Corrosion effects could not be found on HDA and VPS coated specimens even up to the longest time of exposure. The total thickness of the two-layered system remained unchanged at around 130 μm for all examined HDA and VPS specimens. In contrast to this, corrosion effects could be inspected on CVD coated specimens

Fusion technology for the production of PbLi eutectic alloys; Obtencion de aleaciones eutecticas PbLi mediante procesos de fusion

Energy Technology Data Exchange (ETDEWEB)

Barrena, M. J.; Gomez de Salazar, J. M.; Quinones, J.; Pascual, L.; Soria, A.

2012-07-01

The development of thermonuclear experimental reactor (ITER), whose objective is to produce energy from nuclear fusion, has raised the study of Pb-Li eutectic alloys, as they have been selected for the manufacture of test blanket modules (TBM). However, during the manufacturing process of the Pb-Li alloys, thermal conditions used result in a loss of litium element, which inhibits the formation of eutectic structures. In this work we have done fusion of pure lead and lithium, evaluating different process parameters to obtain Pb-Li (17 at. %) eutectic alloys. The alloys manufactured were characterized by DSC, SEM-EDX and microhardness tests. From these studies we noted that the used of an induction reactor and the process parameters optimized to obtain Pb-Li alloy allow for completely eutectic ingots and high chemical homogeneity and microstructural. (Author) 26 refs.

Fabricación de carburo de silicio poroso con capa densa para su aplicación en inserciones aislantes en canal para futuros reactores de fusión nuclear.

OpenAIRE

Bereciartu Andrés, A. (Ainhoa); Ordas Mur, N. (Nerea; Garcia-Rosales, C. (Carmen)

2015-01-01

Within the project TECNO_FUS on CONSOLIDER- INGENIO 2010 program, a dual coolant blanket design is developing (DCLL = Dual Coolant Lithium Lead) for DEMO with Pb-15.7Li and He as coolant. It is a ferritic-martensitic steel with low activation as structural metrial cooled by He. The Pb-15.7Li acts as tritium breeder, neutron multiplier and coolant. The Pb-15.7Li outlet temperature has been as high as possible to achieve the highest possible efficiency, without exceeding the m...

Fusion Blanket Coolant Section Criteria, Methodology, and Results

Energy Technology Data Exchange (ETDEWEB)

DeMuth, J. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meier, W. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Jolodosky, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Frantoni, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reyes, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-10-02

The focus of this LDRD was to explore potential Li alloys that would meet the tritium breeding and blanket cooling requirements but with reduced chemical reactivity, while maintaining the other attractive features of pure Li breeder/coolant. In other fusion approaches (magnetic fusion energy or MFE), 17Li- 83Pb alloy is used leveraging Pb’s ability to maintain high TBR while lowering the levels of lithium in the system. Unfortunately this alloy has a number of potential draw-backs. Due to the high Pb content, this alloy suffers from very high average density, low tritium solubility, low system energy, and produces undesirable activation products in particular polonium. The criteria considered in the selection of a tritium breeding alloy are described in the following section.

Progress in construction of liquid metal LiPb experimental loops in China

International Nuclear Information System (INIS)

Zhang, M.; Zhu, Z.; Gao, S.; Song, Y.; Li, C.; Huang, Q.; Wu, Y.

2007-01-01

The activities of FDS series fusion reactors design with liquid tritium breeder blankets have been performed at ASIPP (Institute of Plasma Physics, Chinese Academy of Sciences) for years. In the designs, CLAM (China Low Activation Martensitic steel), is considered as the primary candidate structural material and LiPb eutectic as both tritium breeder and coolant of the blankets. Therefore, researches on LiPb experimental loop and construction of LiPb loop are severely needed in order to carry out experimental study on the compatibility of candidate structural materials for fusion reactors such as CLAM etc., flowing characteristics of LiPb and Magnetohydrodynamic (MHD) effect and so on, which is essential to researches of China liquid LiPb blankets. A lot of work has been done at ASIPP on design, manufacture and experiments for the series LiPb experimental loops i.e. Dragon-I, Dragon-II, Dragon-III and Dragon-IV. Dragon-I is a thermal convection LiPb loop made of SS316L steel and operating at 500 degree C. The first 3000 hour loop operation at 480 degree C for compatibility test on CLAM was done. Dragon-II and Dragon-III are also thermal convection LiPb loops, made of Inconel 600 and SiCr/SiC, and operating at 700 degree C and 1000 degree C, respectively, to obtain corrosion results of materials such as SiCr/SiC composite. Dragon-II has already been built up and under testing. Dragon-III is under construction. Base on requirement for experiments on characteristics of LiPb on its flow, MHD effect and corrosion to materials, Dragon-IV forced convection loop is being designed. The operation temperature ranges from 480 degree C at the cold leg to 700 degree C at the hot leg, the magnetic field is about 2-5T. Experiments and related studies in those loops are underway. (authors)

Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

2014-10-15

A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

Corrosion of austenitic and martensitic stainless steels in flowing 17Li-83Pb alloy

International Nuclear Information System (INIS)

Broc, M.; Flament, T.; Fauvet, P.; Sannier, J.

1988-01-01

With regard to the behaviour of 316 L stainless steel at 400 0 C in flowing anisothermal 17Li-83Pb the mass transfer suffered by this steel appears to be quite important without noticeable influence of constant or cyclic stress. Evaluation made from solution-annealed specimens leads to a corrosion rate of approximately 30 μm yr -1 at steady state to which a depth of 25 μm has to be added to take into account the initial period phenomena. On the other hand, with semi-stagnant 17Li-83Pb at 400 0 C, the mass transfer of 316 L steel appears to be lower and more acceptable after a 3000-h exposure; but long-time kinetics data have to be achieved in order to see if that better behaviour is persistent and does not correspond to a longer incubation period. As for the martensitic steels their corrosion rate at 450 0 C in the thermal convection loop TULIP is constant up to 3000 h and five times lower than that observed for 316 L steel in the same conditions. (orig.)

Feasibility study of LiF-BeF2 and chloride salts as blanket coolants for fusion power reactors

International Nuclear Information System (INIS)

Imamura, Y.

1977-09-01

The feasibility of using molten salts, in particular, nonberyllium-bearing chloride salts, as blanket coolants for Tokamak fusion reactors has been examined for the nucleonic and thermal/hydraulic aspects. It is concluded that the chloride salts, i.e., LiCl--KCl, LiCl--PbCl 2 and LiCl--SnCl 2 , can be used as the blanket coolant for a static lithium metal blanket provided that large blanket thickness can be tolerated, along with the use of U-238 for neutron multiplication in the cases of LiCl--KCl or LiCl--SnCl 2 cooled blankets. However, to make the appraisal complete, the tritium recovery and corrosion problems must be examined extensively, based on data not yet at hand. As for LiF--BeF 2 , it is observed that although the salt mixture can be used for a single fluid blanket with satisfactory nuclear performance, careful attention should be paid to the cooling capability

Transmutation performance analysis on coolant options in a hybrid reactor system design for high level waste incineration

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong-Hee; Siddique, Muhammad Tariq; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

2015-11-15

Highlights: • Waste transmutation performance was compared and analyzed for seven different coolant options. • Reactions of fission and capture showed big differences depending on coolant options. • Moderation effect significantly affects on energy multiplication, tritium breeding and waste transmutation. • Reduction of radio-toxicities of TRUs showed different trend to coolant choice from performance of waste transmutation. - Abstract: A fusion–fission hybrid reactor (FFHR) is one of the most attractive candidates for high level waste transmutation. The selection of coolant affects the transmutation performance of a FFHR. LiPb coolant, as a conventional coolant for a FFHR, has problems such as reduction in neutron economic and magneto-hydro dynamics (MHD) pressure drop. Therefore, in this work, transmutation performance is evaluated and compared for various coolant options such as LiPb, H{sub 2}O, D{sub 2}O, Na, PbBi, LiF-BeF{sub 2} and NaF-BeF{sub 2} applicable to a hybrid reactor for waste transmutation (Hyb-WT). Design parameters measuring performance of a hybrid reactor were evaluated by MCNPX. They are k{sub eff}, energy multiplication factor, neutron absorption ratio, tritium breeding ratio, waste transmutation ratio, support ratio and radiotoxicity reduction. Compared to LiPb, H{sub 2}O and D{sub 2}O are not suitable for waste transmutation because of neutron moderation effect. Waste transmutation performances with Na and PbBi are similar to each other and not different much from LiPb. Even though molten salt such as LiF-BeF{sub 2} and NaF-BeF{sub 2} is good for avoiding MHD pressure drop problem, waste transmutation performance is dropped compared with LiPb.

The solubility of metals in Pb-17Li liquid alloy

International Nuclear Information System (INIS)

Borgstedt, H.U.; Feuerstein, H.

1992-01-01

The solubility data of iron in the eutectic alloy Pb-17Li which were evaluated from corrosion tests in a turbulent flow of the molten alloy are discussed in the frame of solubilities of the transition metals in liquid lead. It is shown that the solubility of iron in the alloy is close to that in lead. This is also the fact for several other alloying elements of steels. A comparison of all known data shows that they are in agreement with generally shown trends for the solubility of the transition metals in low melting metals. These trends indicate comparably high solubilities of nickel and manganese in the liquid metals, lower saturation concentration of vanadium, chromium, iron, and cobalt, and extremely low solubility of molybdenum. (orig.)

Corrosion and surface conditions of EUROFER 97 steel in Pb-17Li at 500 deg C

International Nuclear Information System (INIS)

Zmitko, M.; Splichal, K.; Masarik, V.

2004-01-01

In this work the corrosion behaviour of EUROFER 97 was examined in flowing Pb-17Li at the temperature 500 deg C up to 2500 hours. Surface morphology and chemical composition profiles and weight changes were investigated. Interaction of EUROFER 97 specimens with Pb-17Li melt results in a material dissolution, which is demonstrated by surface morphology and specimen weight changes. The specimen surfaces investigated after 500 and 1000 hours of exposure in Pb-17Li show similar surface appearance in both as-received and polished conditions. The corrosive damage occurs locally and a major part of surface areas is not affected. The exposure after 2500 hours evidences some visible decrease in the surface roughness for both surface conditions. The surface overlapping was observed and industrial tube productions have to avoid such types of defects. A small weight changes after 500 and 1000 hours and a higher weight decrease after 2500 hours were observed. The absolute values of the weight change after 500 and 1000 hours are about one order of magnitude lower than ones of weight changes after 2500 hours exposure. There were no significant differences of weight changes between as-received and polished surface conditions. The weight decrease of about 1 mg/cm 2 after 2500 hours is in a sufficient correlation with the value of about 4 mg/cm 2 evaluated from data of Fe-12Cr-1MoVW steel. The experiments have shown that the surface corrosive attack revealed only after a certain incubation period. During this period the surface layers are relatively stable to a direct attack of the surface by the melt. In the course of exposure time those layers are not further resistant and can influence the dissolutions of steel components. Concentration profiles of steel components near the steel surface were examined by EDX line-scan and point analyses. Under the experimental conditions no considerable profile of Cr and Fe in surface layers, as higher soluble steel components in Pb-17Li, was

A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

International Nuclear Information System (INIS)

Yoo, J. W.; Kim, S. J.; Kim, Y. I.

2000-01-01

A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

Helium Bubbles Cavitation Phenomena in Pb-15.7Li and Potential Impact on Tritium Transport Behaviour in HCLL Breeding Channels

International Nuclear Information System (INIS)

Sedano, L. A.

2007-01-01

COMPU task is devoted to develop a Process Flow Diagram (PFD) modelling tool for DEMO tritium cycle for HCLL and HCPB blanket lines for DEMO. At the actual stage of definition of HCLL blanket design line this global objective requires to progress specifically on the physical reliability of tritium transport assessments at blanket design level. A rough reliability assessment with the identify cation of physical phenomena determining permeation rates into the coolant was tentatively advanced in COMPU Task Deliverable 1. In HCLL design, the tritium diffusion in the alloy under the flow conditions and radiation effects in Pb15.7Li can be theoretically justifies ed as the rate limiting processes for tritium transfer into the coolant. This Deliverable 2 focuses on the analysis of a specific radiation effect: the potential role of helium bubbles in Pb15.7Li, the discussion of its implications on tritium assessment for HCLL design and consequently the analysis of its quantitative impact (as cycle input) on HCLL PFD tritium cycle design. Thus, the contents of this report investigate: (1) the rationality of the consideration on HCLL design of helium bubble cavitation phenomena in irradiated Pb15.7Li channels on the base of fundamental analysis (He solution states in Pb15.7Li) from empirical clues provided by Pb15.7Li irradiation tests, (2) a preliminary rough He-bubble cavitation design assessment and bases for a more precise FEM calculation for helium bubble cavitation phenomena in HCLL blanket channels, (3) the analysis of direct experimental data and numerical developments needed for a precise cavitation assessment and (4) a proposal of the lay-out and general specifications of an integral proof-of-principle Cavitation Experiment (Cevitex) of Helium in Pb15.7Li. (Author) 40 refs

Helium Bubbles Cavitation Phenomena in Pb-15.7Li and Potential Impact on Tritium Transport Behaviour in HCLL Breeding Channels

Energy Technology Data Exchange (ETDEWEB)

Sedano, L. A.

2007-09-27

COMPU task is devoted to develop a Process Flow Diagram (PFD) modelling tool for DEMO tritium cycle for HCLL and HCPB blanket lines for DEMO. At the actual stage of definition of HCLL blanket design line this global objective requires to progress specifically on the physical reliability of tritium transport assessments at blanket design level. A rough reliability assessment with the identify cation of physical phenomena determining permeation rates into the coolant was tentatively advanced in COMPU Task Deliverable 1. In HCLL design, the tritium diffusion in the alloy under the flow conditions and radiation effects in Pb15.7Li can be theoretically justifies ed as the rate limiting processes for tritium transfer into the coolant. This Deliverable 2 focuses on the analysis of a specific radiation effect: the potential role of helium bubbles in Pb15.7Li, the discussion of its implications on tritium assessment for HCLL design and consequently the analysis of its quantitative impact (as cycle input) on HCLL PFD tritium cycle design. Thus, the contents of this report investigate: (1) the rationality of the consideration on HCLL design of helium bubble cavitation phenomena in irradiated Pb15.7Li channels on the base of fundamental analysis (He solution states in Pb15.7Li) from empirical clues provided by Pb15.7Li irradiation tests, (2) a preliminary rough He-bubble cavitation design assessment and bases for a more precise FEM calculation for helium bubble cavitation phenomena in HCLL blanket channels, (3) the analysis of direct experimental data and numerical developments needed for a precise cavitation assessment and (4) a proposal of the lay-out and general specifications of an integral proof-of-principle Cavitation Experiment (Cevitex) of Helium in Pb15.7Li. (Author) 40 refs.

Corrosion and mass transfer of ferrous alloys in Pb-17 at. % Li

International Nuclear Information System (INIS)

Tortorelli, P.F.

1988-01-01

Long term exposures of type 316 stainless and Fe--12Cr--1MoVW steels to thermally convective Pb-17 at. % Li demonstrated the aggressiveness of this environment, the greater corrosion susceptibility of the austenitic stainless steel, the constancy of the Fe--12Cr surface composition, and the applicability of a surface destabilization model. Cold work affected the penetration of type 316 stainless steel. Deposition in the type 316 stainless steel system appeared to be influenced by the effectiveness of nucleation and/or adhesion of deposits. In the Fe--12Cr--1MoVW steel loop, solubility-driven reactions appeared to be the most important process in deposition. 13 refs., 5 figs., 1 tab

Corrosion of martensitic steels in flowing 17Li83Pb alloy

International Nuclear Information System (INIS)

Flament, T.; Fauvet, P.; Hocde, B.; Sannier, J.

1988-01-01

Corrosion of three martensitic steels - 1.4914, HT9 and T91 - in the presence of flowing 17Li83Pb is investigated in thermal convection loops Tulip entirely made of 1.4914 steel. Two 3000-hour tests were performed at maximal temperatures of respectively 450 and 475 0 C with a δT of 60 0 C and an alloy velocity of about 0.08 m.s -1 . In both tests, corrosion is characterized by an homogeneous dissolution of the steel without formation of a corrosion layer. Corrosion rate is constant and very temperature dependent: the sound-metal loss of 1.4914 steel is 22 μm. year -1 at 450 0 C and 40 μm.year -1 at 475 0 C. Behaviours of 1.4914 and HT9 steels are very similar whereas T91 steel is about 20% less corroded

Dynamic corrosion investigations in the eutectic lead-lithium melt Pb-17Li

International Nuclear Information System (INIS)

Frees, G.; Drechsler, G.; Peric, Z.

1989-01-01

The Pb-17Li circuit 'PICOLO' was constructed and commissioned in the Institute for Material and Solid State Research II of the Kernforschungszentrum Karlsruhe. This circuit serves for corrosion tests with the martensitic steel 1.4914, which is under discussion for the application as structural material for a fusion reactor blanket. The design and the functions of the circuit are described. The experience which has been gained so far is principally favorable. A temperature of 500deg C seems, however, to be the upper limit for the operation with the structural materials of the circuit. The corrosion of specimens and materials of the components becomes considerably high at higher temperatures; this has been seen to cause the plugging due to the precipitation of corrosion products. The results of the first test series are presented and discussed. (orig.) [de

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

International Nuclear Information System (INIS)

Malara, C.; Viola, A.

1995-01-01

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

Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

International Nuclear Information System (INIS)

Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

1976-04-01

The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

International Nuclear Information System (INIS)

Wong, C.P.C.; Malang, S.; Sawan, M.

2005-01-01

As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li 2 BeF 4 and the low melting point molten salts such as LiBeF 3 and LiNaBeF 4 (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC f /SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R and D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan

Obtención de aleaciones eutécticas PbLi mediante procesos de fusión

Directory of Open Access Journals (Sweden)

Barrena, M. I.

2012-12-01

Full Text Available The development of thermonuclear experimental reactor (ITER, whose objective is to produce energy from nuclear fusion, has raised the study of Pb-Li eutectic alloys, as they have been selected for the manufacture of test blanket modules (TBM. However, during the manufacturing process of the Pb-Li alloys, thermal conditions used result in a loss of litium element, which inhibits the formation of eutectic structures. In this work we have done fusion of pure lead and lithium, evaluating different process parameters to obtain Pb-Li (17 at. % eutectic alloys. The alloys manufactured were characterized by DSC, SEM-EDX and microhardness tests. From these studies we noted that the used of an induction reactor and the process parameters optimized to obtain Pb-Li alloy allow for completely eutectic ingots and high chemical homogeneity and microstructural.

El desarrollo del reactor experimental termonuclear (ITER, cuyo objetivo es la producción de energía a partir de la fusión nuclear, ha suscitado el estudio de las aleaciones eutécticas Pb-Li, ya que éstas han sido seleccionadas para la fabricación de las envolturas regeneradoras del reactor (TBM. Sin embargo, durante el proceso de fabricación de las aleaciones Pb-Li se produce una pérdida de litio, que inhibe la formación de estructuras eutécticas. En el presente trabajo se han realizado fusiones partiendo de plomo y litio puros evaluando diferentes parámetros de proceso para la obtención de aleaciones eutécticas Pb-Li (17 % at.. Las aleaciones obtenidas fueron caracterizadas mediante calorimetría diferencial de barrido, microscopía SEM-EDX y microdureza. De estos estudios podemos señalar, que la utilización de hornos de inducción y los parámetros de proceso optimizados para la obtención de la aleación Pb-Li, permiten obtener lingotes completamente eutécticos y con gran homogeneidad química y microestructural.

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

Energy Technology Data Exchange (ETDEWEB)

Merrill, B.J., E-mail: Brad.Merrill@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Wong, C.P.C. [General Atomics, San Diego, CA 92186-5608 (United States); Cadwallader, L.C. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Abdou, M.; Morley, N.B. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)

2014-10-15

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the {sup 210}Po and {sup 203}Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.

Neutronic performance of two European breeder-inside-tube (BIT) blankets for DEMO: the helium-cooled ceramic LiAlO2 with Be multiplier and the water-cooled liquid Li17Pb

International Nuclear Information System (INIS)

Petrizzi, L.; Rado, V.

1995-01-01

In support of ENEA activity in the European Community Test Programme, neutron analysis has been performed on the two latest blanket designs: helium-cooled ceramic breeder-inside-tube (BIT) (with LiAlO 2 and Be multiplier) and water-cooled liquid Li 17 Pb in cylindrical modules (CM). The powerful MCNP Monte Carlo code was used (version 4.2). A detailed and accurate description of the geometrical model has been performed by inserting the main reactor details and avoiding breeder material dilution inside the modules. The tritium breeding ratio (TBR) performance is low for the solid breeder BIT blanket (with 10 ports 1.011) due mainly to low blanket coverage near the exhaust duct, and this solution should be revised. The CM Li 17 Pb blanket reaches a sufficient TBR (1.059, with ports) to rely on tritium self-sufficiency. Shielding properties, with respect to the toroidal field coils, have been estimated in a simplified model by means of the ANISN code, supplied with a nuclear data library consistent with that used by MCNP. The analysis suggests that a careful shield thickness/composition design should be used to ensure the shielding capability of the whole blanket plus shield system. (orig.)

Estimation of thermophysical properties in the system Li-Pb

International Nuclear Information System (INIS)

Jauch, U.; Schulz, B.

1986-01-01

Based on the phase diagram and the knowledge of thermophysical properties data of alloys and intermetallic compounds in the Li-Pb system, quantitative relationships between several properties and between the properties in solid and liquid state are used: to interpret the results on thermophysical properties in the quasibinary system LiPb-Pb and to estimate unknown properties in the concentration range 100 > Li (at.%) > 50. (orig.)

Compatibility of 316L stainless steel with the liquid alloy Pb17Li

International Nuclear Information System (INIS)

Broc, M.; Fauvet, P.; Flament, T.; Terlain, A.; Sannier, J.

1988-01-01

The behavior of 316L austenitic stainless steel in liquid eutectic lead alloy is investigated. The 316L is a possible structural material for fusion reactors. The obtained results are summarized and compared with other experimental data. The mechanisms which control the corrosion process are discussed. The investigation shows that whatever, the hydraulic flow, the corrosion of 316L stainless steel exposed to Pb17Li is characterized by the formation of a porous ferritic layer. The corrosion kinetics is mainly dependent on temperature, hydraulic flow and metallurgical state of the steel. At 400 0 C in turbulent flow, the corrosion rate at steady state of 316L solution annealed is estimated to 27 microns/year to which a depth of 25 microns has to be added to take into account the initial transient period. From overall available results, dissolution and solid state transformation in case of turbulent flow and diffusion in liquid phase for laminar flow, may be suggested

Compatibility of structural materials with fusion reactor coolant and breeder fluids

International Nuclear Information System (INIS)

DeVan, J.H.

1979-01-01

Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF-BeF 2 , Pb-Li alloys, and solid ceramic compounds such as Li 2 O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies. (orig.)

Alloying of aluminum and its influence on the properties of aluminide coatings: oxidation behavior and the chemical stability in Pb-17Li

International Nuclear Information System (INIS)

Glasbrenner, H.; Peric, Z.; Borgstedt, H.U.

1996-01-01

Electrical insulation of the structural material is necessary to reduce the MHD pressure drop in a self-cooled liquid metal blanket. This coating has to be compatible with liquid Pb-17Li up to 450 C. Specimens with different types of coatings were exposed to static Pb-17Li for 1200 h at 450 C in order to study their compatibility. Iron and a ferritic steel were coated with an aluminide layer by means of an aluminizing process. Iron metal plate was hot dip aluminized at Thyssen, Germany. The preheated sheet was coated for this purpose by exposing for a few seconds to a melt of Al with 10 wt% Si. The ferritic steel, MANET, was immersed into a melt of the same composition. In this case, cold specimens were dipped into the melt at 700 C for up to 10 min. The formation of the required oxide scale on top of the aluminide layer was performed by using two different methods: high temperature oxidation in air and anodic oxidation at room temperature. All the exposed specimens were examined before and after the corrosion experiments. The analytical method used is EDX measurements on the cut of the specimens and metallographical examinations. (orig.)

Construction and initial operation of MHD PbLi facility at UCLA

Energy Technology Data Exchange (ETDEWEB)

Smolentsev, S., E-mail: sergey@fusion.ucla.edu; Li, F.-C.; Morley, N.; Ueki, Y.; Abdou, M.; Sketchley, T.

2013-06-15

Highlights: • New MHD PbLi loop has been constructed and tested at UCLA. • Pressure diagnostics system has been developed and successfully tested. • Ultrasound Doppler velocimeter is tested as velocity diagnostics. • Experiments on pressure drop reduction have been performed. • Experiments on MHD flow in a duct with SiC flow channel insert are underway. -- Abstract: A magnetohydrodynamic flow facility MaPLE (Magnetohydrodynamic PbLi Experiment) that utilizes molten eutectic alloy lead–lithium (PbLi) as working fluid has been constructed and tested at University of California, Los Angeles. The loop operation parameters are: maximum magnetic field 1.8 T, PbLi temperature up to 350 °C, maximum PbLi flow rate with/without a magnetic field 15/50 l/min, maximum pressure head 0.15 MPa. The paper describes the loop itself and its major components, basic operation procedures, experience of handling PbLi, initial loop testing, flow diagnostics and current and near-future experiments. The obtained test results of the loop and its components have demonstrated that the new facility is fully functioning and ready for experimental studies of magnetohydrodynamic, heat and mass transfer phenomena in PbLi flows and also can be used in mock up testing in conditions relevant to fusion applications.

Dual coolant blanket concept

International Nuclear Information System (INIS)

Malang, S.; Schleisiek, K.

1994-11-01

A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

A novel UV-emitting phosphor: LiSr4(BO3)3: Pb2+

International Nuclear Information System (INIS)

Pekgözlü, İlhan

2013-01-01

Pure and Pb 2+ doped LiSr 4 (BO 3 ) 3 materials were prepared by a solution combustion synthesis method. The phase analysis of all synthesized materials were determined using the powder XRD. The synthesized materials were investigated using spectrofluorometer at room temperature. The excitation and emission bands of LiSr 4 (BO 3 ) 3 : Pb 2+ were observed at 284 and 328 nm, respectively. The dependence of the emission intensity on the Pb 2+ concentration for the LiSr 4 (BO 3 ) 3 were studied in detail. It was observed that the concentration quenching of Pb 2+ in LiSr 4 (BO 3 ) 3 is 0.005 mol. The Stokes shifts of LiSr 4 (BO 3 ) 3 : Pb 2+ phosphor was calculated to be 4723 cm –1 . -- Highlights: • A novel UV-emitting phosphor: LiSr 4 (BO 3 ) 3 : Pb 2+ ” synthesized for the first time. • The emission band of LiSr 4 (BO 3 ) 3 : Pb 2+ was observed at 328 nm upon excitation with 284 nm. • LiSr 4 (BO 3 ) 3 : Pb 2+ is a good phosphor for broadband UV application

Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

Energy Technology Data Exchange (ETDEWEB)

Blokhin, D. A.; Mitenkova, E. F. [Nuclear Safety Inst., Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Khorasanov, G. L.; Zemskov, E. A.; Blokhin, A. I. [State Scientific Center, Russian Federation, Inst. of Physics and Power Engineering, Bondarenko Square 1, Obninsk, 249033 (Russian Federation)

2012-07-01

In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

Study of MHD Corrosion and Transport of Corrosion Products of Ferritic/Martensitic Steels in the Flowing PbLi and its Application to Fusion Blanket

Science.gov (United States)

Saeidi, Sheida

Two important components of a liquid breeder blanket of a fusion power reactor are the liquid breeder/coolant and the steel structure that the liquid is enclosed in. One candidate combination for such components is Lead-Lithium (PbLi) eutectic alloy and advanced Reduced Activation Ferritic/Martensitic (RAFM) steel. The research performed here is aimed at: (1) better understanding of corrosion processes in the system including RAFM steel and flowing PbLi in the presence of a strong magnetic field and (2) prediction of corrosion losses in conditions of a Dual Coolant Lead Lithium (DCLL) blanket, which is at present the key liquid metal blanket concept in the US. To do this, numerical and analytical tools have been developed and then applied to the analysis of corrosion processes. First, efforts were taken to develop a computational suite called TRANSMAG (Transport phenomena in Magnetohydrodynamic Flows) as an analysis tool for corrosion processes in the PbLi/RAFM system, including transport of corrosion products in MHD laminar and turbulent flows. The computational approach in TRANSMAG is based on simultaneous solution of flow, energy and mass transfer equations with or without a magnetic field, assuming mass transfer controlled corrosion and uniform dissolution of iron in the flowing PbLi. Then, the new computational tool was used to solve an inverse mass transfer problem where the saturation concentration of iron in PbLi was reconstructed from the experimental data resulting in the following correlation: CS = e 13.604--12975/T, where T is the temperature of PbLi in K and CS is in wppm. The new correlation for saturation concentration was then used in the analysis of corrosion processes in laminar flows in a rectangular duct in the presence of a strong transverse magnetic field. As shown in this study, the mass loss increases with the magnetic field such that the corrosion rate in the presence of a magnetic field can be a few times higher compared to purely

Mixing and mass transfer in bubble extractor: its application to tritium extraction from 17Li83Pb

International Nuclear Information System (INIS)

Baratti, R.; Polcaro, A.M.; Ricci, P.F.; Viola, A.; Pierini, G.

1986-01-01

The tritium extraction from the liquid alloy 17Li83Pb has been examined taking into consideration the equations related to the design of bubble extractors in order to verify which are the highest tritium recovery efficiencies which can be realized so as to minimize the tritium permeation into the water of the cooling system. For the tritium desorption from alloy, flowing countercurrent to a helium stream in a bubble extractor, the axial dispersion in liquid and gaseous phase and the effect of gas phase expansion caused by reduced hydrostatic head in the extractor are taken into account. Taking into consideration some NET technical specifications, such as the alloy volume in the blanket of 65 m 3 and the tritium generation rate of 78 g.d -1 , the results are presented and discussed

The 10B(n,α)7Li reaction in PWR coolants: calculations of the effect on coolant pH and on decreases in 10B isotopic fractions

International Nuclear Information System (INIS)

Polley, M.V.

1988-07-01

Boron is used as a chemical shim in PWRs for reactivity control and is added in the form of boric acid to the primary coolant. The 10 B(n,α) 7 Li reaction leads to a continuous increase in 7 Li in the primary coolant and to a continuous decrease in 10 B the isotope of boron responsible for control of reactivity. The rate of increase in coolant pH due to 7 Li production is calculated for the Sizewell 'B' PWR to enable judgements to be made on the frequency of sampling and removal of lithium required to maintain the pH of the primary coolant within the desired limits. Calculations are contrasted for the cases of natural boron and 100% 10 B chemical shims, for both a normal cycle and an extended 18 month cycle. Calculations of 10 B depletion over 30 years of operation as a function of the quantity of boron discharged to waste are also presented. 10 B isotopic fractions are calculated for the reactor coolant (RC), boric acid tanks (BATs) and refuelling water storage tank (RWST) assuming rapid mixing of BAT and RC boron for tritium control and other reasons. Such predictions enable assessments of the reactor physics implications of 10 B consumption to be made. (author)

Structural mechanical studies on the 12% Cr-steel X 18CrMoVNb 12 1 (1.4914) in liquid Pb-17Li eutecticum

International Nuclear Information System (INIS)

Grundmann, M.

1990-02-01

The possible application of the martensitic 12%-Cr steel 1.4914 as structural material of the fusion reactor blanket is examinated within this study. The superimposure of mechanical stresses and chemical dissolution processes and its consequences for the material properties of steel 1.4914 is simulated and studied in the experiments. Therefore, tensile, creep, low-cycle fatigue and fatigue crack growth tests in the liquid metal have been performed. The test parameters have been varied in a wide range as well as the external state of the material related to the real operational conditions. The examination of the test results and of the numerous examinations of the material (measurement of hardness, metallographic studies and analyses using the scanning and transmission electron microscope, microprobe and AUGER microprobe, GDOS and classical chemical methods) made it sure that the liquid alloy Pb-17Li has an influence on the mechanical properties of the steel only under certain conditions. Necessary conditions are a high temperature (> 500deg C), a long lasting contact with the liquid metal and a minimum degree of deformation. If these conditions are fulfilled, an earlier end of life time occurs in Pb-17Li compared to gaseous environment (air or argon) due to the loss of material caused by the dissolution processes in the liquid alloy. (orig./MM) [de

Measurement of solubility of hydrogen isotopes in Li-Pb by adsorption and desorption method

International Nuclear Information System (INIS)

Edao, Yuki; Katayama, K.; Fukada, S.

2014-01-01

Measurement of tritium solubility in lithium lead eutectic alloy (Li-Pb) has been performed under the Japan-US collaboration work of 'TITAN'. The present paper reports that results of H and D solubility in Li-Pb which melted in an alumina tube determined by means of a constant volume method, and also reports an experimental apparatus for measurement of tritium solubility in Li-Pb in a tungsten crucible is improved and examined in the STAR facility of the Idaho National Laboratory. It was shown that H solubility in Li-Pb was easily influenced by impurities, interaction with surrounding materials and evaporated Li-Pb. The influences were suggested to be caused by large scattering among the previously reported data on solubility of hydrogen isotopes in Li-Pb. (author)

Solubility of hydrogen isotopes in liquid LiPb

International Nuclear Information System (INIS)

Konishi, S.; Yamamoto, Y.; Noborio, K.; Calderoni, P.; Merrill, B.

2014-01-01

This research was performed mainly in the first half of the task 1-2 of TITAN project to investigate the interaction between hydrogen isotopes and liquid LiPb. Solubility of hydrogen in liquid LiPb was measured under a static condition. Kyoto University provided the first experimental apparatus shipped to Idaho, and Kyushu University succeeded the experiment and further improved. Obtained solubility generally agreed with some previous reports, but varied orders of magnitudes suggesting influence of impurity or other chemical processes. (author)

Fusion-reactor blanket-material safety-compatibility studies

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

1982-11-01

Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO 2 , Li 2 ZrO 3 , Li 2 SiO 3 , Li 4 SiO 4 and LiTiO 3 ) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li 7 Pb 2 alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li 17 Pb 83 alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li 17 Pb 83 alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns

Coupled-channels analyses for 9,11Li + 208Pb fusion reactions with multi-neutron transfer couplings

Science.gov (United States)

Choi, Ki-Seok; Cheoun, Myung-Ki; So, W. Y.; Hagino, K.; Kim, K. S.

2018-05-01

We discuss the role of two-neutron transfer processes in the fusion reaction of the 9,11Li + 208Pb systems. We first analyze the 9Li + 208Pb reaction by taking into account the coupling to the 7Li + 210Pb channel. To this end, we assume that two neutrons are directly transferred to a single effective channel in 210Pb and solve the coupled-channels equations with the two channels. By adjusting the coupling strength and the effective Q-value, we successfully reproduce the experimental fusion cross sections for this system. We then analyze the 11Li + 208Pb reaction in a similar manner, that is, by taking into account three effective channels with 11Li + 208Pb, 9Li + 210Pb, and 7Li + 212Pb partitions. In order to take into account the halo structure of the 11Li nucleus, we construct the potential between 11Li and 208Pb with a double folding procedure, while we employ a Woods-Saxon type potential with the global Akyüz-Winther parameters for the other channels. Our calculation indicates that the multiple two-neutron transfer process plays a crucial role in the 11Li + 208Pb fusion reaction at energies around the Coulomb barrier.

Asymptotic three-particle approach to the Coulomb breakup process {sup 6}Li + {sup 208}Pb → {sup 208}Pb + α + d

Energy Technology Data Exchange (ETDEWEB)

Igamov, S. B., E-mail: igamov@inp.uz [Uzbek Academy of Sciences, Institute of Nuclear Physics (Uzbekistan)

2017-03-15

On the basis of the distorted-wave method, experimental data on the triple-differential cross section for the Coulomb breakup reaction {sup 208}Pb({sup 6}Li, αd){sup 208}Pb are analyzed by employing a correct expression for the final-state {sup 208}Pb–α–d three-particle Coulomb wave function. It is shown that the effect of final-state three-particle Coulomb dynamics can be used to assess the kinematical condition of clean Coulomb breakup processes. New values of the astrophysical S factor for the direct-radiative-capture reaction d(α, γ){sup 6}Li at ultralow energies in the range of 70 ≤ E{sub dα} ≤ 600 keV were extracted from experimental data. The value of S(0) = 1.60 ± 0.17 MeV nb was obtained.

Corrosion experiment in the first liquid metal LiPb loop of China

International Nuclear Information System (INIS)

Huang Qunying; Zhang Maolian; Zhu Zhiqiang; Gao Sheng; Wu Yican; Li Yanfen; Song Yong; Li Chunjing; Kong Mingguang

2007-01-01

The liquid metal LiPb blanket design is one of the most promising designs for future fusion power reactors and under wide research in the world. The first liquid metal LiPb loop in China named DRAGON-I was built in 2005 in order to do research on characteristics of liquid metal LiPb such as its corrosion to structural materials of the blankets and so on. The first corrosion experiment in flowing LiPb with a speed of 0.08 m/s at 480 deg. C for 500 h was done in October 2005 on CLAM (China low activation martensitic) steel and 316L stainless steel for comparison. The weights and compositions, etc. of the specimens before and after corrosion experiment were tested and analyzed, the microstructures of the specimens were also inspected by SEM. The results show that the corrosion of CLAM steel is relatively slight, while that for 316L is obvious and very serious. Further study on corrosion behavior of CLAM for longer time experiment in liquid LiPb at different temperatures and flow speeds will be carried out in the near future

Vaporization of liquid Pb-Li eutectic alloy from 1000K to 1200K - A high temperature mass spectrometric study

Science.gov (United States)

Jain, U.; Mukherjee, A.; Dey, G. K.

2017-09-01

Liquid lead-lithium eutectic will be used as a coolant in fusion reactor blanket loop. Vapor pressure of the eutectic is an important parameter to accurately predict its in-loop behavior. Past measurements of vapor pressure of the eutectic relied on indirect methods. In this paper, we report for the first time the in-situ vaporization behavior of the liquid alloy between 1042 and 1176 K by Knudsen effusion mass spectrometry (KEMS). It was seen that the vaporization occurred by independent evaporation of lead and lithium. No complex intermetallic vapor was seen in the mass spectra. The partial pressures and enthalpy of vaporization of Pb and Li were evaluated directly from the measured ion intensities formed from the equilibrium vapor over the alloy. The activity of Li over a temperature range of 1042-1176 K was found to be 4.8 × 10-5 to that of pure Li, indicating its very low activity in the alloy.

Effect of nature convection on heat transfer in the liquid LiPb blanket for FDS-II

Energy Technology Data Exchange (ETDEWEB)

Wang Hongyan; Chen Hongli [Huaibei Coal Industry Teachers Coll. (China). Dept. of Physics; Zhou Tao [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics

2007-07-01

The He-cooled liquid LiPb tritium breeder (SLL) blanket concept is one of options of the blanket design of the fusion power reactor (FDS-II). The SLL blanket could be developed relatively easily with lower LiPb outlet temperature and slower LiPb flow velocity that allows the utilization of relatively mature material technology. The velocity of the liquid LiPb in the blanket is very slowly only in order to extract tritium. The magnetohydrodynamic (MHD) flow and heat transfer become very complex resulting from the differential heating of walls of the channels, especially adjacent to the First Wall (FW), and internal heat sources inside of the liquid LiPb. It is necessary to analyse the effect of the buoyancy-driven LiPb MHD flow on heat transfer in the channels with electrically and thermally conducting walls adjacent to the FW. The nature convection of the liquid LiPb, due to thermal diffusion, in the poloidal channel adjacent to the FW in the presence of the strong magnetic field of the SLL blanket has been considered and studied. The specially numerical MHD code based on the computational fluid dynamic software has been developed for analysis of the buoyancy-driven MHD flow. The properties of buoyantly convective flows have been investigated for various thermal boundary conditions. The numerical analysis was performed for the effect of nature convection on heat transfer of the liquid LiPb MHD flow in the poloidal channel in the SLL blanket. For the strong temperature gradient in the blanket and internal heat flux of Liquid LiPb, the three-dimensional temperature distributions of the LiPb, the FW and other walls have been given. Finally, The effect of the ratio of MHD buoyancy on the heat transfer characteristics of the LiPb flow have been calculated and presented. (orig.)

Disentangling the transfer and breakup contributions for the inclusive 8 Li + 208 Pb reaction

International Nuclear Information System (INIS)

Moro, A.M.; Crespo, R.; Garcia M, H.; Aguilera, E.F.; Martinez Q, E.; Gomez C, J.; Nunes, F.M.

2003-01-01

An analysis of the 8 Li + 208 Pb reaction at energies around the Coulomb barrier is presented. The study is focused on the elastic and one-neutron removal channels. For the elastic scattering, an optical model analysis of the experimental data is performed. The observed 7 Li is interpreted as the superposition of the one-neutron transfer reaction, 208 Pb ( 8 Li, 7 Li) 209 Pb, and the breakup reaction. The separate contribution of each one of these processes has been calculated within the DWBA formalism. The sum of both contributions explains adequately the experimental angular distribution of 7 Li. (Author)

Triple-differential cross section of the 208Pb(6Li, αd)208 Pb Coulomb breakup and astrophysical S-factor of the d(α,γ)6 Li reaction at extremely low energies

International Nuclear Information System (INIS)

Igamov, S.B.; Yarmukhamedov, R.

1999-10-01

A method of calculation of the triple-differential cross section of the 208 Pb( 6 Li, αd) 208 Pb Coulomb breakup at astrophysically relevant energies E of the relative motion of the breakup fragments, taking into account the three-body (α - d - 208 Pb) Coulomb effects and the contributions from the E1- and E2- multipoles, including their interference, has been proposed. The new results for the astrophysical S-factor of the direct radiative capture d(α, γ) 6 Li reaction at E ≤ 250 keV have been obtained. It is shown that the experimental triple-differential cross section of the 208 Pb( 6 Li, αd) 208 Pb Coulomb breakup can also be used to give information about the value of the modulus squared of the nuclear vertex constant for the virtual 6 Li → α + d. (author)

Construction and initial operation of MHD PbLi facility at UCLA

International Nuclear Information System (INIS)

Kunugi, T.; Yokomine, T.; Ueki, Y.; Smolentsev, S.; Li, F.-C.; Sketchley, T.; Abdou, M.A.; Yuki, K.

2014-01-01

We review current accomplishments in Task 1-3 'Flow Control and Thermofluid Modeling' of the Japan-US 'TITAN' collaboration program. Our task focuses on experimental activities and also computer modeling of magnetohydrodynamic flows and heat and mass transfer of electrically conducting fluids under conditions relevant to fusion blankets. Since our task started, major efforts were taken to design, construct and test a new magnetohydrodynamic lead-lithium (PbLi) loop at UCLA, to accumulate the PbLi handling technology, and to develop a high-temperature ultrasonic Doppler velocimetry and a differential-pressure measurement system for PbLi flows. In the present paper, the loop construction, the electromagnetic pump performance test, our on-going experiments with the constructed loop are described. (author)

Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

2012-07-01

The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

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

International Nuclear Information System (INIS)

Xie, Bo

2009-04-01

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

Study on hydrogen isotope behavior in Pb-Li forced convection flow with permeable wall

Energy Technology Data Exchange (ETDEWEB)

Yoshimura, Ryosuke; Fukada, Satoshi, E-mail: sfukada@nucl.kyushu-u.ac.jp; Muneoka, Taiki; Kinjo, Mao; Katayama, Kazunari

2016-12-15

Highlights: • Transient- and steady-state hydrogen permeation from Li-Pb forced convection flow through permeable tube to outside Ar purge gas was investigated at 600 °C. • It was found that the overall permeation rates were limited by diffusion in the Li-Pb boundary layer developed from the flow inlet. • The effect of the boundary layer was correlated in terms of mass transfer coefficient. The values of mass transfer coefficients at 600 °C were compared with those of 400 °C and 500 °C obtained beforehand. - Abstract: Transient- and steady-state hydrogen permeation from Li-Pb forced convection flow in a permeable tube to outside Ar purge gas was investigated between 400–600 °C. The values of the steady-state permeation rate increased with the increase of the Li-Pb flow rate. It was found that the overall permeation rates were limited by diffusion in a Li-Pb boundary layer developed from flow inlet. The effect of the boundary layer was correlated in terms of the mass-transfer coefficient. The values of the mass-transfer coefficient at 600 °C were compared with those of 400 °C and 500 °C obtained beforehand. Judged from these data of mass-transfer coefficients, it can be predicted that the effect of boundary layer varies with the increase of Li-Pb flow rate at different temperature conditions.

Voc enhancement of a solar cell with doped Li+-PbS as the active layer

Science.gov (United States)

Chávez Portillo, M.; Alvarado Pulido, J.; Gallardo Hernández, S.; Soto Cruz, B. S.; Alcántara Iniesta, S.; Gutiérrez Pérez, R.; Portillo Moreno, O.

2018-06-01

In this report, we investigate the fabrication of solar cells obtained by chemical bath technique, based on CdS as window layer and PbS and PbS-Li+-doped as the active layer. We report open-circuit-voltage Voc values of ∼392 meV for PbS and ∼630 meV for PbSLi+-doped, a remarkable enhanced in the open circuit voltage is shown for solar cells with doped active layer. Li+ ion passivate the dangling bonds in PbS-metal layer interface in consequence reducing the recombination centers.

High voltage and high specific capacity dual intercalating electrode Li-ion batteries

Science.gov (United States)

West, William C. (Inventor); Blanco, Mario (Inventor)

2010-01-01

The present invention provides high capacity and high voltage Li-ion batteries that have a carbonaceous cathode and a nonaqueous electrolyte solution comprising LiF salt and an anion receptor that binds the fluoride ion. The batteries can comprise dual intercalating electrode Li ion batteries. Methods of the present invention use a cathode and electrode pair, wherein each of the electrodes reversibly intercalate ions provided by a LiF salt to make a high voltage and high specific capacity dual intercalating electrode Li-ion battery. The present methods and systems provide high-capacity batteries particularly useful in powering devices where minimizing battery mass is important.

Formation of Al{sub 2}O{sub 3}/FeAl coatings on a 9Cr-1Mo steel, and corrosion evaluation in flowing Pb-17Li loop

Energy Technology Data Exchange (ETDEWEB)

Majumdar, Sanjib, E-mail: sanjib@barc.gov.in [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Paul, Bhaskar [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Chakraborty, Poulami [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Kishor, Jugal; Kain, Vivekanand [High Temperature Materials Development Section, Materials Processing & Corrosion Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Dey, Gautam Kumar [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Materials Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India)

2017-04-01

Iron aluminide coating layers were formed on a ferritic martensitic grade 9Cr-1Mo (P 91) steel using pack aluminizing process. The formation of different aluminide compositions such as orthorhombic-Fe{sub 2}Al{sub 5}, B2-FeAl and A2-Fe(Al) on the pack chemistry and heat treatment conditions have been established. About 4–6 μm thick Al{sub 2}O{sub 3} scale was formed on the FeAl phase by controlled heat treatment. The corrosion tests were conducted using both the FeAl and Al{sub 2}O{sub 3}/FeAl coated specimens in an electro-magnetic pump driven Pb-17Li Loop at 500 °C for 5000 h maintaining a flow velocity of 1.5 m/s. The detailed characterization studies using scanning electron microscopy, back-scattered electron imaging and energy dispersive spectrometry revealed no deterioration of the coating layers after the corrosion tests. Self-healing oxides were formed at the cracks generated in the aluminide layers during thermal cycling and protected the base alloy (steel) from any kind of elemental dissolution or microstructural degradation. - Highlights: •Al{sub 2}O{sub 3}/FeAl coating produced on P91 steel by pack aluminizing and heat treatment. •Corrosion tests of coated steel conducted in flowing Pb-17Li loop at 500 °C for 5000 h. •Coating was protective against molten metal corrosion during prolonged exposure. •Self-healing protective oxides formed in the cracks generated in aluminide layers.

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Overview of EU activities on DEMO liquid metal breeder blanket

International Nuclear Information System (INIS)

Giancarli, L.; Proust, E.; Malang, S.; Reimann, J.; Perujo, A.

1994-01-01

The present paper gives an overview of both design and experimental activities within the European Union (EU) concerning the development of liquid metal breeder blankets for DEMO. After several years of studies on breeding blankets, two blanket concepts are presently considered, both using the eutectic Pb-17Li: the dual-coolant concept and the water-cooled concept. The analysis of such concepts has permitted to identify the experimental areas where further data are required. Tritium control and MHD-issues are, at present, the activities on which is devoted the greatest effort within the EU. (authors). 4 figs., 4 tabs., 39 refs

Small scale lithium-lead/water-interaction studies

International Nuclear Information System (INIS)

Kranert, O.; Kottowski, H.

1991-01-01

One current concept in fusion blanket design is to utilize water as the coolant and liquid lithium-lead as the breeding/neutron multiplier material. Considering the complex design of the blanket module, it is likely that a water leakage into the liquid alloy may occur due to a tube rupture provoking an intolerable pressure increase in the blanket module. The pressure increase is caused by the combined chemical and thermohydraulic reaction of lithium-lead with water. Experiments which simulate such a transient event are necessary to obtain information which is important for the blanket module design. The interaction has been investigated by conducting small-scale experiments at various injection pressures, alloy- and coolant temperatures. Besides using eutectic Li 17 Pb 83 , Li 7 Pb 2 , lithium and lead have been used. Among other results, the experiments indicate increasing chemical reaction with increasing lithium concentration. At the same time, the chemical reaction inhibits violent thermohydaulic reactions due to the attenuating effect of the hydrogen produced. The preliminary epxerimental results from Li 17 Pb 83 and Li 7 Pb 2 reveal that the pressure- and temperature transients caused by the chemical and thermohydraulic reactions lie within technically manageable limits. (orig.)

Liquid Li-Pb-Bi, a new tritium breeder

International Nuclear Information System (INIS)

Rogers, A.G.; Benedict, B.L.; Clemmer, R.G.

1981-01-01

In light of their potential utility as tritium breeder-blanket materials, a study was conducted to identify and characterize low-melting phases in the lithium-lead-bismuth system. It is found that a low-melting ternary phase field did in fact exist, e.g., compositions with less than or equal to 20 atom percent lithium and Pb/Bi = 0.773 melted at or below 140 0 C. In addition, the qualitative reactivity of Li-Bi-Pb alloys with water was tested, and although minimal evidence of exothermic chemical reaction was observed, a physical vapor explosion did occur in one of the tests

Radiogenic Lead with Dominant Content of 208Pb: New Coolant and Neutron Moderator for Innovative Nuclear Facilities

Directory of Open Access Journals (Sweden)

A. N. Shmelev

2011-01-01

Full Text Available As a rule materials of small atomic weight (light and heavy water, graphite, and so on are used as neutron moderators and reflectors. A new very heavy atomic weight moderator is proposed—radiogenic lead consisting mainly of isotope 208Pb. It is characterized by extremely low neutron radiative capture cross-section (0.23 mbarn for thermal neutrons, i.e., less than that for graphite and deuterium and highest albedo of thermal neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in a fast reactor. This can increase safety of the fast reactors and reduce as well requirements pertaining to the fuel fabrication technology. Radiogenic lead with high 208Pb content as a liquid-metal coolant of fast reactors helps to achieve a favorable (negative reactivity coefficient on coolant temperature. It is noteworthy that radiogenic lead with high 208Pb content may be extracted from thorium (as well as thorium-uranium ores without isotope separation. This has been confirmed experimentally by the investigations performed at San Paulo University, Brazil.

Testing of porous SiC with dense coating under relevant conditions for Flow Channel Insert application

Energy Technology Data Exchange (ETDEWEB)

Ordás, N., E-mail: nordas@ceit.es [CEIT and Tecnun (University of Navarra), Manuel de Lardizábal 15, 20018 San Sebastián (Spain); Bereciartu, A.; García-Rosales, C. [CEIT and Tecnun (University of Navarra), Manuel de Lardizábal 15, 20018 San Sebastián (Spain); Moroño, A.; Malo, M.; Hodgson, E.R. [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Abellà, J.; Colominas, S. [Institut Químic de Sarrià, University Ramon Llull, Via Augusta 390, 08017 Barcelona (Spain); Sedano, L. [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain)

2014-10-15

Highlights: • Porous SiC coated by CVD with a dense coating was developed for Flow Channel Inserts (FCI) in dual-coolant blanket concept. • Porous SiC was obtained following the sacrificial template technique, using Al{sub 2}O{sub 3} and Y{sub 2}O{sub 3} as sintering additives. • Flexural strength, thermal and electrical conductivity, and microstructure of uncoated and coated porous SiC are presented. • Adhesion of coating to porous SiC and its corrosion behavior under Pb-17.5Li at 700 °C are shown. - Abstract: Thermally and electrically insulating porous SiC ceramics are attractive candidates for Flow Channel Inserts (FCI) in dual-coolant blanket concepts thanks to its relatively inexpensive manufacturing route. To prevent tritium permeation and corrosion by Pb-15.7 a dense coating has to be applied on the porous SiC. Despite not having structural function, FCI must exhibit sufficient mechanical strength to withstand strong thermal gradients and thermo-electrical stresses during operation. This work summarizes the results on the development of coated porous SiC for FCI. Porous SiC was obtained following the sacrificial template technique, using Al{sub 2}O{sub 3} and Y{sub 2}O{sub 3} as sintering additives and a carbonaceous phase as pore former. Sintering was performed in inert gas at 1850–1950 °C during 15 min to 3 h, followed by oxidation at 650 °C to eliminate the carbonaceous phase. The most promising bulk materials were coated with a ∼30 μm thick dense SiC by CVD. Results on porosity, bending tests, thermal and electrical conductivity are presented. The microstructure of the coating, its adhesion to the porous SiC and its corrosion behavior under Pb-17.5Li are also shown.

Conceptual design of power conversion system for a fusion power reactor with self-cooled LiPb-blanket. EFDA Task TW2-TRP-PPCS12 - Deliverable 4

International Nuclear Information System (INIS)

Vieider, Gottfried

2002-05-01

For FPRs with self-cooled LiPb-blanket and He-cooled first wall and divertor a conceptual design of the power conversion system is developed with emphasis on component feasibility, safety, reliability and thermal efficiency. The resulting power conversion system with a steam turbine is based on proven technology for Na- and He-cooled fission reactors and is assessed to yield an overall net thermal plant efficiency of ∼40 % provided the high primary coolant temperatures of ∼700 deg C can be achieved. The required complexity of the five linked cooling systems can be expected to influence plant cost and reliability

Neutron flux measurement with 6Li and 7Li dual glass scintillators by γ compensation method

International Nuclear Information System (INIS)

Ji Changsong; Zhang Shulan; Zhang Shuheng

1998-01-01

Based on the characteristics of 6 Li glass scintillator which is sensitive to both neutron and gamma rays, and 7 Li glass scintillator which is sensitive to gamma rays only, a new method of detecting weak neutron flux under interference of strong gamma radiation has been investigated by mans of 6 Li- 7 Li dual glass scintillator gamma compensation method. The result of neutron flux measurement by above-mentioned method with an error of about 1% when the gamma ray interference is up to 18.7% has been obtained

Structural and thermodynamic similarities of phases in the Li-Tt (Tt = Si, Ge) systems: redetermination of the lithium-rich side of the Li-Ge phase diagram and crystal structures of Li17Si4.0-xGex for x = 2.3, 3.1, 3.5, and 4 as well as Li4.1Ge.

Science.gov (United States)

Zeilinger, Michael; Fässler, Thomas F

2014-10-28

A reinvestigation of the lithium-rich section of the Li-Ge phase diagram reveals the existence of two new phases, Li17Ge4 and Li4.10Ge (Li16.38Ge4). Their structures are determined by X-ray diffraction experiments of large single crystals obtained from equilibrated melts with compositions Li95Ge5 and Li85Ge15. Excess melt is subsequently removed through isothermal centrifugation at 400 °C and 530 °C, respectively. Li17Ge4 crystallizes in the space group F4[combining macron]3m (a = 18.8521(3) Å, V = 6700.1(2) Å(3), Z = 20, T = 298 K) and Li4.10Ge (Li16.38Ge4) in Cmcm (a = 4.5511(2) Å, b = 22.0862(7) Å, c = 13.2751(4) Å, V = 1334.37(8) Å(3), Z = 16, T = 123 K). Both phases are isotypic with their Si counterparts and are further representative of the Li17Pb4 and Li4.11Si structure types. Additionally, the solid solutions Li17Si4-xGex follows Vegard's law. A comparison of the GeLin coordination polyhedra shows that isolated Ge atoms are 13- and 14-coordinated in Li17Ge4, whereas in Li16.38Ge4 the Ge atoms possess coordination numbers 12 and 13. Regarding the thermodynamic stability, Li16.38Ge4 is assigned a high-temperature phase existing between ∼400 °C and 627 °C, whereas Li17Ge4 decomposes peritectically at 520-522 °C. Additionally, the decomposition of Li16.38Ge4 below ∼400 °C was found to be very sluggish. These findings are manifested by differential scanning calorimetry, long-term annealing experiments and the results from melt equilibration experiments. Interestingly, the thermodynamic properties of the lithium-rich tetrelides Li17Tt4 and Li4.1Tt (Li16.4Tt4) are very similar (Tt = Si, Ge). Besides Li15Tt4, Li14Tt6, Li12Tt7, and LiTt, the title compounds are further examples of isotypic tetrelides in the systems Li-Tt.

Protection of Lithium (Li) Anodes Using Dual Phase Electrolytes

Energy Technology Data Exchange (ETDEWEB)

Mikhaylik, Yuriy [Sion Power Corporation, Tucson, AZ (United States)

2014-09-30

Sion Power focused on metallic lithium anode protection, employing the Dual-Phase Electrolyte approach. The objective of this project was to develop a unique electrolyte providing two liquid phases having good Li+ conductivity, self-partitioning and immiscibility, serving separately the cathode and anode electrodes. This Dual-Phase Electrolyte was combined with thin film multi-layer, physical barrier membranes developed partially under a separate ARPA-E funded project. All these protective structures were stabilized by externally applied pressure. This strategy was used for Li-S cells. The development directly addressed cell safety, particularly higher thermal stability, while also allowing higher energies and cycle life. Safety tests showed that 100% of cells with Dual-Phase Electrolyte were intact and did not exhibit thermal runaway up to 178 °C and thus met the project objective of increasing the runaway temperature to >165°C. Cells also passed cycling at USABC Dynamic Stress Test conditions developed for Electric Vehicle applications and generated specific energy > 300 Wh/kg.

Fusion-driven sub-critical dual-cooled waste transmutation blanket: design and analysis

International Nuclear Information System (INIS)

Wang Weihua; Wu Yican; Ke Yan; Kang Zhicheng; Wang Hongyan; Huang Qunying

2003-01-01

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis

7 CFR 7.17 - Dual office.

Science.gov (United States)

2010-01-01

... 7 Agriculture 1 2010-01-01 2010-01-01 false Dual office. 7.17 Section 7.17 Agriculture Office of... STATE, COUNTY AND COMMUNITY COMMITTEES § 7.17 Dual office. (a) County committee membership. A member of... any other county office employee. (b) Community committee membership. A member of the community...

Reducing Pb poisoning in birds and Pb exposure in game meat consumers: the dual benefit of effective Pb shot regulation.

Science.gov (United States)

Mateo, Rafael; Vallverdú-Coll, Núria; López-Antia, Ana; Taggart, Mark A; Martínez-Haro, Monica; Guitart, Raimon; Ortiz-Santaliestra, Manuel E

2014-02-01

The use of lead (Pb) ammunition in the form of shot pellets has been identified as a Pb exposure risk in wildlife and their human consumers. We explore the hypothesis that Pb shot ban enforcement reduces the risk of avian Pb poisoning as well as Pb exposure in game meat consumers. We assessed compliance with a partial ban on Pb shot commencing in 2003 by examination of 937 waterbirds harvested by hunters between 2007 and 2012 in the Ebro delta (Spain). Prevalence of Pb shot ingestion was determined, as were Pb concentrations in liver and muscle tissue to evaluate the potential for Pb exposure in game meat consumers. Hunted birds with only embedded Pb shot (no steel) declined from 26.9% in 2007-08 to meat (0.1μg/g wet weight) in the 2008-09 season, when Pb shot ingestion prevalence was also at a minimum (5.1%). Effective restrictions in Pb ammunition use have a dual benefit since this reduces Pb exposure for game meat consumers due to embedded ammunition as well as reducing Pb poisoning in waterbirds. Copyright © 2013 Elsevier Ltd. All rights reserved.

New dual emission fluorescent sensor for pH and Pb(II) based on bis(napfthalimide) derivative

International Nuclear Information System (INIS)

Pina-Luis, Georgina; Martínez-Quiroz, Marisela; Ochoa-Terán, Adrián; Santacruz-Ortega, Hisila; Mendez-Valenzuela, Eduardo

2013-01-01

This paper describes a novel dual emission bis-1,8-naphthalimide sensor for selective determination of pH and Pb 2+ ions. The influence of the variability in the backbone that links the two fluorophores (naphthalimides) as a function of pH and metal ions was studied by UV–visible and fluorescence spectroscopy. Compounds 1(a–d) with different length alkyl linkers (CH 2 ) n (n=1, 2, 4 and 6) showed no excimer formation in aqueous solution. Fluorescence emission of these derivatives varied in a narrow range of pH (5–8) and was only slightly influenced by the addition of metal ions in CH 3 CN solutions. However, derivative 1e with amino-containing spacer (CH 2 –NH–CH 2 ) showed excimer emission in aqueous solution, a wide response to pH (2.5–9.5) and fluorescence enhancement with selective behavior towards metal ions. The pH sensor based in derivative 1e has a sufficient selectivity for practical pH monitoring in the presence of Li + , Na + , K + , Cs + , Ca 2+ , Mg 2+ , Ba 2+ , Cu 2+ , Pb 2+ , Ni 2+ , Zn 2+ and Cd 2+ . The coordination chemistry of these complexes was studied by UV–Vis, fluorescence and 1 H NMR. This chemosensor displayed high selectivity fluorescence enhancement toward Pb 2+ ions in the presence of the metals ions mentioned in CH 3 CN solutions. Competitive assays show that a 1-fold of metal cations in each case, compared with Pb 2+ ions, results in less than ±5% fluorescence intensity changes. Linear calibration up to 1×10 −5 M for Pb(II) ions (R=0.9968) was obtained and detection limit resulted of 5.0×10 −8 M. - Highlights: ► A novel dual emission bis-1,8-naphthalimide sensor for pH and Pb 2+ ions is synthetized. ► The excimer formation depends on the spacer that links the two naphthalimide groups. ► Bis(naphthalimide) with amino-containing spacer showed a wide selective response to pH. ► This chemosensor displayed a selective fluorescence enhancement effect towards Pb 2+ ions. ► Mechanism for the fluorescence OFF�

The concentration of the coolant 7Li in Kozloduy Nuclear Power Plant operating with potassium hydroxide as an alkalizing reagent (possible impact on the occurrence of axial offset anomaly)

International Nuclear Information System (INIS)

Dobrevski, I.D.; Minkova, K.F.; Ivanova, R.A.

2003-01-01

The phenomenon of axial offset anomaly (AOA) has occurred in a number of PWRs operating with extended fuel cycles and high boiling duty cores. Up to now AOA has been observed in PWRs operating with lithium hydroxide and the alkalizing reagent used for pH adjustment in boric acid water solutions. Since AOA is connected with the LiBO 2 precipitation in porous corrosion product deposits on the fuel cladding surfaces, we could presume that the replacement of lithium hydroxide with potassium hydroxide will avoid AOA. Nowadays there is a lack of observed AOA in VVER, i.e., a lack of formation of lithium metaborate (LiBO 2 ) deposits on the fuel element surfaces by coolant alkalization with potassium hydroxide. Nevertheless, the concentrations of 7 Li appear in the coolant, as a product of the neutron reaction with boron: 10 B (n,α) → 7 Li (n, α). As a consequence the possibility it is not excluded of LiBO 2 formation in VVERs with potassium hydroxide water chemistry. The aim of this study is to inform the reader about the development of the concentration of the coolant lithium concentration during the fuel cycles of VVERs and to discuss the possibility of LiBO 2 formation under VVER operation conditions. (orig.)

7Li(18O, 17N8Be reaction and the 17N + 8Be-potential

Directory of Open Access Journals (Sweden)

A. T. Rudchik

2010-12-01

Full Text Available Angular distributions of the 7Li(18O, 17N8Be reaction were measured for the transitions to the ground states of 8Be and 17N and excited states of 17N at the energy Elab(18O = 114 MeV. The data were analyzed with coupled-reaction-channels method for one- and two-step transfers of nucleons and clusters. In the analysis, the 7Li + 18O potential de-duced in the analysis of the elastic 7Li + 18O-scattering data as well as shell-model spectroscopic amplitudes of trans-ferred nucleons and clusters were used. Parameters of the 8Be + 17N potential were deduced using the reaction data. Contributions of different one- and two-step transfers in the 7Li(18O, 17N8Be reaction cross-section was studied.

New dual emission fluorescent sensor for pH and Pb(II) based on bis(napfthalimide) derivative

Energy Technology Data Exchange (ETDEWEB)

Pina-Luis, Georgina, E-mail: gpinaluis@yahoo.com [Centro de Graduados e Investigacion en Quimica, Instituto Tecnologico de Tijuana, AP 1166, Tijuana 22500, BC (Mexico); Martinez-Quiroz, Marisela; Ochoa-Teran, Adrian [Centro de Graduados e Investigacion en Quimica, Instituto Tecnologico de Tijuana, AP 1166, Tijuana 22500, BC (Mexico); Santacruz-Ortega, Hisila [Departamento de investigacion en Polimeros y Materiales, Universidad de Sonora, Hermosillo, Sonora 83000 (Mexico); Mendez-Valenzuela, Eduardo [Centro de Graduados e Investigacion en Quimica, Instituto Tecnologico de Tijuana, AP 1166, Tijuana 22500, BC (Mexico)

2013-02-15

This paper describes a novel dual emission bis-1,8-naphthalimide sensor for selective determination of pH and Pb{sup 2+} ions. The influence of the variability in the backbone that links the two fluorophores (naphthalimides) as a function of pH and metal ions was studied by UV-visible and fluorescence spectroscopy. Compounds 1(a-d) with different length alkyl linkers (CH{sub 2}){sub n} (n=1, 2, 4 and 6) showed no excimer formation in aqueous solution. Fluorescence emission of these derivatives varied in a narrow range of pH (5-8) and was only slightly influenced by the addition of metal ions in CH{sub 3}CN solutions. However, derivative 1e with amino-containing spacer (CH{sub 2}-NH-CH{sub 2}) showed excimer emission in aqueous solution, a wide response to pH (2.5-9.5) and fluorescence enhancement with selective behavior towards metal ions. The pH sensor based in derivative 1e has a sufficient selectivity for practical pH monitoring in the presence of Li{sup +}, Na{sup +}, K{sup +}, Cs{sup +}, Ca{sup 2+}, Mg{sup 2+}, Ba{sup 2+}, Cu{sup 2+}, Pb{sup 2+}, Ni{sup 2+}, Zn{sup 2+} and Cd{sup 2+}. The coordination chemistry of these complexes was studied by UV-Vis, fluorescence and {sup 1}H NMR. This chemosensor displayed high selectivity fluorescence enhancement toward Pb{sup 2+} ions in the presence of the metals ions mentioned in CH{sub 3}CN solutions. Competitive assays show that a 1-fold of metal cations in each case, compared with Pb{sup 2+} ions, results in less than {+-}5% fluorescence intensity changes. Linear calibration up to 1 Multiplication-Sign 10{sup -5} M for Pb(II) ions (R=0.9968) was obtained and detection limit resulted of 5.0 Multiplication-Sign 10{sup -8} M. - Highlights: Black-Right-Pointing-Pointer A novel dual emission bis-1,8-naphthalimide sensor for pH and Pb{sup 2+} ions is synthetized. Black-Right-Pointing-Pointer The excimer formation depends on the spacer that links the two naphthalimide groups. Black-Right-Pointing-Pointer Bis

Analysis on Radioactive Waste Transmutation in Light Water cooled Hyb-WT

International Nuclear Information System (INIS)

Hong, Seonghee; Kim, Myung Hyun

2014-01-01

A feasibility of realization is much higher in FFHR compared with pure fusion. A combination of plasma fusion source for neutrons with a subcritical reactor at the blanket side has much higher capability in transmutation of waste as well as reactor safety compared with fission reactor options. Fusion-Fission Hybrid Reactor (FFHR) uses various coolants depending on the purpose. It is important that coolant being used should be suitable to reactor purpose, because reactor performance and the design constraints may change depending on the coolant. There are basically two major groups of coolants for FFHR. One group of coolant does not contain Li. They are Na, Pb-Bi, H 2 O and D 2 O. The other group contains Li for tritium breeding. They are Li, LiPb, LiSN, FLIBE and FLiNaBe. Currently, the issue in FFHR is its implication for radioactive waste transmutation (FFHR for WT). Because radioactive wastes of spent nuclear fuel (SNF) are transmuted using fusion neutron source. Therefore a suitable coolant should be used for effective waste transmutation. . In FFHR for WT, LiPb coolant is being used mainly because of tritium production in Li and high neutron economic through reaction in Pb. However different coolants use such as Na, Pb-Bi are used in fast reactors and accelerator driven systems (ADS) having same purpose. In this study, radioactive waste transmutation performance of various coolants mentioned above will be compared and analyzed. Through this study, the coolants are judged primarily for their support to waste transmutation disregarding their limitation to reactor design and tritium breeding capability. First, performance of the light water coolant regarding radioactive waste transmutation was analyzed among various coolants mentioned above. In this paper, performance of radioactive waste transmutation can be known depending on different volume fractions (54.53, 60.27, 97.94vol.%) of the light water. Light water dose required fusion power lower than LiPb due to

A comparative neutronic analysis of 150MWe TRU burner according to the coolant alteration

International Nuclear Information System (INIS)

Yoo, J. W.; Kim, S. J.; Kim, Y. I.

2000-01-01

A comparative neutronic analysis has been conducted for the small TRU burner according to their coolant material. The use of Pb-Bi coolant gave a low burnup reactivity swing and negative or less positive coolant void coefficient with harder neutron spectrum. By a lower burnup reactivity swing and higher conversion ratio of Pb-Bi cooled core, the total amount of TRU consumption was found to be small compared with Na cooled core despite of the higher MA consumption ratio of Pb-Bi cooled core. However, Pb-Bi cooled reactor have a lager margin in the coolant void coefficient, so that a variable MA composition can be loaded in the core. Accordingly, even though the Pb-Bi cooled TRU burner has not effectiveness on TRU burning in the same geometry and material condition, a flexible MA loading is envisaged to result in 10 times larger MA burning amount, still preserving a low coolant void worth

Complex titanates Sr_1_-_xPb_xLi_2Ti_6O_1_4 (0≤x≤1) as anode materials for high-performance lithium-ion batteries

International Nuclear Information System (INIS)

Qian, Shangshu; Yu, Haoxiang; Yan, Lei; Li, Peng; Lin, Xiaoting; Wu, Yaoyao; Long, Nengbing; Shui, Miao; Shu, Jie

2016-01-01

Highlights: • Sr_1_-_xPb_xLi_2Ti_6O_1_4 (0≤x≤1) is prepared by a simple solid state reaction. • Sr_0_._5Pb_0_._5Li_2Ti_6O_1_4 exhibits enhanced lithium storage capability. • Sr_0_._5Pb_0_._5Li_2Ti_6O_1_4 can deliver a capacity of 141.8 mAh g"−"1 at 700 mA g"−"1. • In-situ XRD is performed to study the reversibility of Sr_1_-_xPb_xLi_2Ti_6O_1_4. - Abstract: With the Pb doping content at Sr-site increasing, a series of Sr_1_-_xPb_xLi_2Ti_6O_1_4 (x = 0, 0.25, 0.50, 0.75, 1.0) are synthesized by a simple solid-state reaction. It is found that the reversible capacity and rate capability experience a parabolic course from SrLi_2Ti_6O_1_4 to PbLi_2Ti_6O_1_4. Among all the as-prepared samples, Sr_0_._5Pb_0_._5Li_2Ti_6O_1_4 shows the best cycling and rate properties. It delivers an initial charge capacity of 163.2 mAh g"−"1 at 100 mA g"−"1 with the capacity retention of 96.08% after 100 cycles. In addition, it can also deliver a reversible capacity of 141.8 mAh g"−"1 at 700 mA g"−"1. The superior electrochemical properties of Sr_0_._5Pb_0_._5Li_2Ti_6O_1_4 are attributed to the reduced charge transfer resistance and increased lithium-ion diffusion coefficient after doping. Besides, in-situ X-ray diffraction is also performed to investigate the lithium-ion insertion/extraction behaviors of SrLi_2Ti_6O_1_4, Sr_0_._5Pb_0_._5Li_2Ti_6O_1_4 and PbLi_2Ti_6O_1_4. The observed results confirm that Sr_0_._5Pb_0_._5Li_2Ti_6O_1_4 has good structural stability and reversibility for repeated lithium storage.

Analysis of actual status of works on technology of heavy liquid metal coolants

International Nuclear Information System (INIS)

Martynov, P.N.; Askhadullin, R.Sh.; Orlov, Yu.I.; Storozhenko, A.N.

2014-01-01

Principle duties in heavy liquid metal coolant technology (HLMC) are provision of the purity of coolant and surfaces of circulation loop for maintenance of design thermohydraulic characteristics, prevention of structural materials corrosion and erosion during long service life and present-day safety precautions on different stages of reactor facility operation. For this reason, current HLMC (Pb-Bi, Pb) technology must include coolant pre-operation and charging; monitoring and regulating of coolant oxygen potential; hydrogen purification of coolant and surfaces of circulation loop from lead oxides-based slags; coolant filtration; reactor cover gas purification from coolant aerosols. The current topical problem is personnel training on the questions of HLMC technology [ru

Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

International Nuclear Information System (INIS)

Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

1985-01-01

Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

11Li Breakup on 208Pb at Energies Around the Coulomb Barrier

DEFF Research Database (Denmark)

Fernández-García, J.P.; Cubero, M.; Rodríguez-Gallardo, M.

2013-01-01

The inclusive breakup for the 11Li+208Pb reaction at energies around the Coulomb barrier has been measured for the first time. A sizable yield of 9Li following the 11Li dissociation has been observed, even at energies well below the Coulomb barrier. Using the first-order semiclassical perturbation...... theory of Coulomb excitation it is shown that the breakup probability data measured at small angles can be used to extract effective breakup energy as well as the slope of B(E1) distribution close to the threshold. Four-body continuum-discretized coupled-channels calculations, including both nuclear...... and Coulomb couplings between the target and projectile to all orders, reproduce the measured inclusive breakup cross sections and support the presence of a dipole resonance in the 11Li continuum at low excitation energy....

Investigation of wetting property between liquid lead lithium alloy and several structural materials for Chinese DEMO reactor

Science.gov (United States)

Lu, Wei; Wang, Weihua; Jiang, Haiyan; Zuo, Guizhong; Pan, Baoguo; Xu, Wei; Chu, Delin; Hu, Jiansheng; Qi, Junli

2017-10-01

The dual-cooled lead lithium (PbLi) blanket is considered as one of the main options for the Chinese demonstration reactor (DEMO). Liquid PbLi alloy is used as the breeder material and coolant. Reduced activation ferritic/martensitic (RAFM) steel, stainless steel and the silicon carbide ceramic matrix composite (SiCf) are selected as the substrate materials for different use. To investigate the wetting property and inter-facial interactions of PbLi/RAFM steel, PbLi/SS316L, PbLi/SiC and PbLi/SiCf couples, in this paper, the special vacuum experimental device is built, and the 'dispensed droplet' modification for the classic sessile droplet technique is made. Contact angles are measured between the liquid PbLi and the various candidate materials at blanket working temperature from 260 to 480 °C. X-ray photoelectron spectroscopy (XPS) is used to characterize the surface components of PbLi droplets and substrate materials, in order to study the element trans-port and corrosion mechanism. Results show that SiC composite (SiCf) and SiC ceramic show poor wetting properties with the liquid PbLi alloy. Surface roughness and testing temperature only provide tiny improvements on the wetting property below 480 °C. RAFM steel performs better wetting properties and corrosion residence when contacted with molten PbLi, while SS316L shows low corrosion residence above 420 °C for the decomposition of protective surface film mainly consisted of chromic sesquioxide. The results could provide meaningful compatibility database of liquid PbLi alloy and valuable reference in engineering design of candidate structural and functional materials for future fusion blanket.

Compatibility of CLAM steel weldments with static LiPb alloy at 550 Degree-Sign C

Energy Technology Data Exchange (ETDEWEB)

Chen Xizhang, E-mail: kernel.chen@gmail.com [School of Materials Science and Engineering, Jiangsu University, ZhenJiang, Jiangsu 212013 (China); Shen Zheng; Li Peng [School of Materials Science and Engineering, Jiangsu University, ZhenJiang, Jiangsu 212013 (China); Madigan, Bruce [Montana Tech. of University of Montana, Butte, MT 59701 (United States); Huang Yuming; Lei Yucheng [School of Materials Science and Engineering, Jiangsu University, ZhenJiang, Jiangsu 212013 (China); Huang Qunying [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 031 (China); Zhou Jianzhong [School of Mechanical Engineering, Jiangsu University, ZhenJiang, Jiangsu 212013 (China)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Corrosion extent of weld zone is higher than that of HAZ. Black-Right-Pointing-Pointer Thick martensite lath and large residual stress lead to higher corrosion rate. Black-Right-Pointing-Pointer Cr on the surface of weld zone decreases by about 50%, W increases slightly. Black-Right-Pointing-Pointer After 500 h and 1000 h of corrosion, weight losses are 0.272 mg/cm{sup 2} and 0.403 mg/cm{sup 2}. Black-Right-Pointing-Pointer With the increasing of corrosion time, the corrosion rate decreases significantly. - Abstract: CLAM steel is considered as a structural material to be used in the Test Blanket Module as a barrier or blanket adjacent to liquid LiPb in fusion reactors. In this paper, CLAM steel is welded by tungsten inert gas (TIG) welding, and the compatibility of the weldment with liquid LiPb is tested. Specimens were corroded in static liquid LiPb, with corrosion times of 500 h and 1000 h, at 550 Degree-Sign C, and the corresponding weight losses are 0.272 mg/cm{sup 2} and 0.403 mg/cm{sup 2} respectively. Also the corrosion rate decreases with increased corrosion time. In the as-welded condition, corrosion resistance of the weld zone is higher than that of the HAZ (Heat Affected Zone). Likely, thick martensite lath and large residual stresses at the welding zone result in higher corrosion rates. The compatibility of CLAM steel weld joints with high temperature liquid LiPb can be improved to some extent through a post-weld tempering process. The surface of the as-welded CLAM steel is uniformly corroded and the concentration of Cr on the surface decreases by about 50% after corrosion. Penetration of LiPb into the matrix is observed for neither the as-welded nor the as-tempered conditions. Influenced by thick martensite lath and large residual stresses, the welded area, especially the weld zone, is easily corroded, therefore it is of primary importance to protect the welded area in the solid blanket of the fusion reactor.

Low-activation lead coolant for advanced small modular NPP

International Nuclear Information System (INIS)

Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

2001-01-01

The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

Study of multi nucleon transfer in "9","1"1Li + "2"0"8Pb reactions

International Nuclear Information System (INIS)

Vinodkumar, A.M.

2014-01-01

One of the most active areas of research with radioactive beams is the study of the fusion of weakly bound nuclei, such as the halo nuclei. The central issue is whether the fusion cross section will be enhanced due to the large nuclear size of the halo nucleus or whether fusion-limiting breakup of the weakly bound valence nucleons will lead to a decreased fusion cross section. The fusion of "9","1"1Li with "2"0"8Pb were reported. These measurements were carried out at TRIUMF, Canada. These measurements suggests at above barrier energies, fusion hindrance is taking place in the case of "1"1Li projectile. However, sub barrier fusion measurement need a lower energy measurement. These measurements also suggest need for further measurement of transfer and breakup channels in these reactions. So we suggest a measurement of multi nucleon transfer in the case of "9Li + "2"0"8Pb. Also, these measurement will be able to produce the same nuclei as suggested in the ISOLDE experiment by, where "2"1"2","2"1"4Pb and "2"0"8","2"1"0Hg nuclei for studying the spectroscopy of these nuclei. (author)

On blanket concepts of the Helias reactor

International Nuclear Information System (INIS)

Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

1999-07-01

The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

Energy Technology Data Exchange (ETDEWEB)

Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

2013-06-15

Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

International Nuclear Information System (INIS)

Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

2001-01-01

In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

Hydrogen transfer in Pb–Li forced convection flow with permeable wall

Energy Technology Data Exchange (ETDEWEB)

Fukada, Satoshi, E-mail: sfukada@nucl.kyushu-u.ac.jp; Muneoka, Taiki; Kinjyo, Mao; Yoshimura, Rhosuke; Katayama, Kazunari

2015-10-15

Highlights: • The paper presents experimental and analytical results of Pb–Li eutectic alloy forced convection flow. • Analytical results are in good agreement with ones of hydrogen permeation in Pb–Li forced convection flow. • The results are useful for the design of liquid blanket of fusion reactors. - Abstract: Transient- or steady-state hydrogen permeation from a primary fluid of Li{sub 17}Pb{sub 83} (Pb–Li) through a permeable tube of Inconel-625 alloy to a secondary Ar purge is investigated experimentally under a forced convection flow in a dual cylindrical tube system. Results of the overall hydrogen permeation flux are correlated in terms of diffusivity, solubility and an average axial velocity of Pb–Li and diffusivity and solubility of the solid wall. Analytical solutions under proper assumptions are derived to simulate the transient- and steady-state rates of the overall hydrogen permeation, and close agreement is obtained between experiment and analysis. Two things are clarified from the comparison: (i) how the steady-state permeation rate is affected by the mass-transfer properties and the average velocity of Pb–Li and the properties of Inconel-625, and (ii) how its transient behavior is done by the diffusivity of the two materials. The results obtained here will give important information to estimate or to analyze the tritium transfer rate in fluidized Pb–Li blankets of DEMO or the future commercial fusion reactors.

Failure Modes and Effects Analysis on ITER DFLL-TBM system

International Nuclear Information System (INIS)

Hu Liqin; Yuan Run; Chen Hongli; Bai Yunqing

2012-01-01

As required for licensing process, accident analyses of International Thermonuclear Experimental Reactor (ITER) accounting for site specifications and design changes will be updated. Chinese Dual-Functional Lithium-Lead-Test Blanket Module (DFLL-TBM) system is a key safety-related component of ITER, its detailed safety analysis, which was designated to demonstrate the integrated technologies of both Helium single coolant (SLL) blanket and Helium-LiPb dual coolant (DLL) blanket, was performed. Failure Modes and Effects Analysis (FMEA) was applied to perform the safety analysis of DFLL-TBM. This study described the process of FMEA studies on DFLL-TBM system. All safety-related Postulated Initiating Events (PIEs) was identified. And a set of PIEs recommended to be taken into account in the further deterministic transient analyses were defined for both SLL and DLL blanket concepts separately.

Numerical modeling of first experiments on PbLi MHD flows in a rectangular duct with foam-based SiC flow channel insert

Energy Technology Data Exchange (ETDEWEB)

Smolentsev, S., E-mail: sergey@fusion.ucla.edu [University of California, Los Angeles (United States); Courtessole, C.; Abdou, M.; Sharafat, S. [University of California, Los Angeles (United States); Sahu, S. [Institute of Plasma Research (India); Sketchley, T. [University of California, Los Angeles (United States)

2016-10-15

Highlights: • Numerical studies were performed as a pre-experimental analysis to the experiment on MHD PbLi flows in a rectangular duct with a flow channel insert (FCI). • Dynamic testing of foam-based SiC foam-based CVD coated FCI has been performed using MaPLE facility at UCLA. • Two physical models were proposed to explain the experimental results and 3D and 2D computations performed using COMSOL, HIMAG and UCLA codes. • The obtained results suggest that more work on FCI development, fabrication and testing has to be done to assure good hermetic properties before the implementation in a fusion device. - Abstract: A flow channel insert (FCI) is the key element of the DCLL blanket concept. The FCI serves as electrical and thermal insulator to reduce the MHD pressure drop and to decouple the temperature-limited ferritic structure from the flowing hot lead-lithium (PbLi) alloy. The main focus of the paper is on numerical computations to simulate MHD flows in the first experiments on PbLi flows in a stainless steel rectangular duct with a foam-based silicon carbide (SiC) FCI. A single uninterrupted long-term (∼6500 h) test has recently been performed on a CVD coated FCI sample in the flowing PbLi in a magnetic field up to 1.5 T at the PbLi temperature of 300 °C using the MaPLE loop at UCLA. An unexpectedly high MHD pressure drop measured in this experiment suggests that a PbLi ingress into the FCI occurred in the course of the experiment, resulting in degradation of electroinsulating FCI properties. The ingress through the protective CVD layer was further confirmed by the post-experimental microscopic analysis of the FCI. The numerical modeling included 2D and 3D computations using HIMAG, COMSOL and a UCLA research code to address important flow features associated with the FCI finite length, fringing magnetic field, rounded FCI corners and also to predict changes in the MHD pressure drop in the unwanted event of a PbLi ingress. Two physical

Breakup mechanisms for 7Li + 197Au, 204Pb systems at sub-barrier energies

Directory of Open Access Journals (Sweden)

Luong D.H.

2013-12-01

Full Text Available Coincidence measurements of breakup fragments were carried out for the 7Li + 197Au and 204Pb systems at sub-barrier energies. The mechanisms triggering breakup, and time-scales of each process, were identified through the reaction Q-values and the relative energy of the breakup fragments. Binary breakup of 7Li were found to be predominantly triggered by nucleon transfer, with p-pickup leading to 8Be → α + α decay being the preferred breakup mode. From the time-scales of each process, the coincidence yields were separated into prompt and delayed components, allowing the identification of breakup process important in the suppression of complete fusion of 7Li at above-barrier energies.

Pulse*Star Inertial Confinement Fusion Reactor: heat transfer loop and balance of plant considerations

International Nuclear Information System (INIS)

McDowell, M.W.; Murray, K.A.

1984-01-01

A conceptual heat transfer loop and balance of plant design for the Pulse*Star Inertial Confinement Fusion Reactor has been investigated and results are presented. The Pulse*Star reaction vessel, a perforated steel bell jar approximately 11 m in diameter, is immersed in Li 17 Pb 83 coolant which flows through the perforations and forms a 1.5 m thick plenum of droplets around an 8 m diameter inner chamber. The reactor and associated pumps, piping, and steam generators are contained within a 17 m diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops with flow rates of 5.5 m 3 /s each are necessary to transfer 3300 MWt of power. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. Power balance calculations based on an improved electrical conversion efficiency revealed a net electrical output of 1260 MWe to the bus bar and a resulting net efficiency of 39%. Suggested balance-of-plant layouts are also presented

Calculation and analysis of neutron and radiation characteristics of lead coolants with isotopic tailoring for future nuclear power facilities

Energy Technology Data Exchange (ETDEWEB)

Blokhin, A.I.; Ivanov, A.P.; Korobeinikov, V.V.; Lunev, V.P.; Manokhin, V.N.; Khorasanov, G.L. [SSC RF A. I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation)

2000-03-01

A new type of safe fast reactor with lead coolant was proposed in Russia. The use of coolants with low moderating properties is one of the ways to get a hard neutron spectrum and an increase in the burning of Np-237, Am-243 and other miner actinides(MA) fissionable preferentially in the fast reactor. The stable lead isotope, Pb-208, is proposed as the one of such coolants. The neutron inelastic scattering cross-section of Pb-208 is 3.0-3.5 times less than the one of other lead isotopes. Calculation of the MA transmutation rates in the standard BN-type fast reactor with different coolants is performed by Monte-Carlo method using Code MMKFK. Six various models are simulated for the fast reactor blanket with different kinds of fuel and coolant. The fast reactor with natural-lead coolant practically does not differ from the reactor with sodium coolant relative to MA incineration. The use of Pb-208 as a coolant in the fast reactor results in increasing incineration of MA from 18 to 26% in comparison with a usual fast reactor. Calculation of induced radioactivity was performed using the FISPACT-3 inventory code, also. The results include total induced radioactivity and dose rate for initial material composition and selected long-lived radionuclides. The calculations show that the coolant consisting of lead isotope, Pb-206, or Pb-207, can be considered as the low-activation one because it does not practically contain long-lived toxic radionuclides. (M. Suetake)

Fuel enrichment and temperature distribution in nuclear fuel rod in (D-T) driven hybrid reactor system

Energy Technology Data Exchange (ETDEWEB)

Osman, Ypek [Suleyman Demirel Universitesi Muhendislik-Mimarlyk Fakultesi, Isparta (Turkey)

2001-07-01

In this study, melting point of the fuel rod and temperature distribution in nuclear fuel rod are investigated for different coolants under various first wall loads (P{sub w}, =5, 6, 7, 8, 9, and 10 MWm{sup -2}) in Fusion-Fission reactor fueled with 50%LWR +50%CANDU. The fusion source of neutrons of 14.1 MeV is simulated by a movable target along the main axis of cylindrical geometry as a line source. In addition, the fusion chamber was thought as a cylindrical cavity with a diameter of 300 cm that is comparatively small value. The fissile fuel zone is considered to be cooled with four different coolants, gas, flibe (Li{sub 2}BeF{sub 4}), natural lithium (Li), and eutectic lithium (Li{sub 17}Pb{sub 83}). Investigations are observed during 4 years for discrete time intervals of{delta}t= 0.5 month and by a plant factor (PF) of 75%. Volumetric ratio of coolant-to fuel is 1:1, 45.515% coolant, 45.515% fuel, 8.971% clad, in fuel zone. (author)

Gamma ray shielding properties of PbO-Li2O-B2O3 glasses

International Nuclear Information System (INIS)

Kumar, Ashok

2017-01-01

The mass attenuation coefficients have been measured in (0.6-x) PbO-x Li 2 O-0.40 B 2 O 3 (where 0≤ x≤0.25 mol%) glasses for photon energies of 356, 662, 1173 and 1332 keV in a narrow beam geometry with an overall scatter acceptance angle of 2.31°. The experimental results are found to be within 3% of their theoretical values. These coefficients were then used to obtain the values of mean free path, effective atomic number and electron density. The shielding properties of these glasses have also been compared among themselves in terms of their mean free path and radiation protection efficiency. The shielding properties prepared glasses have also been compared with standard concretes as well as with the standard shielding glasses. It is found that the prepared glasses are the better shielding substitute to the conventional concretes as well as other standard shielding glasses. The Pb 3 B 4 O 9 has been found to be the most effective shield. - Highlights: • Shielding efficiencies of PbO-B 2 O 3 -Li 2 O glasses have been compared. • Measurements have been done for 356, 662, 1173 and 1332 keV photon energies. • Experimental values have been found to be within 3% of their theoretical ones. • Pb 3 B 4 O 9 has been found to be the most effective shield.

LiCl-LiI molten salt electrolyte with bismuth-lead positive electrode for liquid metal battery

Science.gov (United States)

Kim, Junsoo; Shin, Donghyeok; Jung, Youngjae; Hwang, Soo Min; Song, Taeseup; Kim, Youngsik; Paik, Ungyu

2018-02-01

Liquid metal batteries (LMBs) are attractive energy storage device for large-scale energy storage system (ESS) due to the simple cell configuration and their high rate capability. The high operation temperature caused by high melting temperature of both the molten salt electrolyte and metal electrodes can induce the critical issues related to the maintenance cost and degradation of electrochemical properties resulting from the thermal corrosion of materials. Here, we report a new chemistry of LiCl-LiI electrolyte and Bi-Pb positive electrode to lower the operation temperature of Li-based LMBs and achieve the long-term stability. The cell (Li|LiCl-LiI|Bi-Pb) is operated at 410 °C by employing the LiCl-LiI (LiCl:LiI = 36:64 mol %) electrolyte and Bi-Pb alloy (Bi:Pb = 55.5:44.5 mol %) positive electrode. The cell shows excellent capacity retention (86.5%) and high Coulombic efficiencies over 99.3% at a high current density of 52 mA cm-2 during 1000th cycles.

Activation analysis of tritium breeder lithium lead irradiated by fusion neutrons in FDS-II

International Nuclear Information System (INIS)

Mingliang Chen

2006-01-01

R-and-D of fusion materials, especially their activation characteristics, is one of the key issues for fusion research in the world. Research on activation characteristics for low activation materials, such as reduced activation ferritic/martensitic steels, vanadium alloys and SiCf/SiC composites, is being done throughout the world to ensure the attractiveness of fusion power regarding safety and environmental aspects. However, there is less research on the activation characteristics of the other important fusion materials, such as tritium breeder etc.. Lithium lead (Li 17 Pb 83 ) is presently considered as a primary candidate tritium breeder for fusion power reactors because of its attractive characteristics. It can serve as a tritium breeder, neutron multiplier and coolant in the blanket at the same time. The radioactivity of Li 17 Pb 83 by D-T fusion neutrons in FDS-II has been calculated and analyzed. FDS-II is a concept design of fusion power reactor, which consists of fusion core with advanced plasma parameters extrapolated from the ITER (International Thermonuclear Experimental Reactor) and two candidates of liquid lithium breeder blankets (named SLL and DLL blankets). The neutron transport and activation calculation are carried out based on the one-dimensional model for FDS-II with the home-developed multi-functional code system VisualBUS and the multi-group data library HENDL1.0/MG and European Activation File EAF-99. The effects of irradiation time on the activation characteristics of Li 17 Pb 83 were analyzed and it concludes that the irradiation time has an important effect on the activation level of Li 17 Pb 83 . Furthermore, the results were compared with the activation levels of other tritium breeders, such as Li 4 SiO 4 , Li 2 TiO 3 , Li 2 O and Li etc., under the same irradiation conditions. The dominant nuclides to dose rate and activity of Li 17 Pb 83 were analyzed as well. Tritium generated by Li has a great contribution to the afterheat and

New concepts for the recovery and isotopic separation of tritium in fusion reactors

International Nuclear Information System (INIS)

Dombra, A.H.; Holtslander, W.J.; Miller, A.I.; Canadian Fusion Fuels Technology Project, Toronto, Ontario)

1986-01-01

New concepts for the recovery of tritium from light water coolant of LiPb blankets, and high-pressure helium coolant of Li-ceramic blankets are introduced. Application of these concepts to fusion reactors is illustrated with conceptual system designs for the anticipated NET blanket requirements. (author)

Optical and electrical properties of zinc oxide thin films with low resistivity via Li-N dual-acceptor doping

Energy Technology Data Exchange (ETDEWEB)

Zhang Daoli, E-mail: zhang_daoli@mail.hust.edu.cn [Department of Electronic Science and Technology, Huazhong University of Science and Technology, No. 1037 Luoyu Road, Hongshan District, Wuhan City, Hubei Province 430074 (China); Wuhan National Laboratory for Optoelectronics, 1037 Luoyu Road, Hongshan District, Wuhan City, Hubei Province 430074 (China); Zhang Jianbing [Department of Electronic Science and Technology, Huazhong University of Science and Technology, No. 1037 Luoyu Road, Hongshan District, Wuhan City, Hubei Province 430074 (China); Wuhan National Laboratory for Optoelectronics, 1037 Luoyu Road, Hongshan District, Wuhan City, Hubei Province 430074 (China); Guo Zhe [Department of Electronic Science and Technology, Huazhong University of Science and Technology, No. 1037 Luoyu Road, Hongshan District, Wuhan City, Hubei Province 430074 (China); Miao Xiangshui [Department of Electronic Science and Technology, Huazhong University of Science and Technology, No. 1037 Luoyu Road, Hongshan District, Wuhan City, Hubei Province 430074 (China); Wuhan National Laboratory for Optoelectronics, 1037 Luoyu Road, Hongshan District, Wuhan City, Hubei Province 430074 (China)

2011-05-19

Highlights: > Zinc oxide films have been deposited on glass substrates by Li-N dual-acceptor doping method via a modified SILAR method. > The resistivity of ZnO film was found to be 1.04 {Omega} cm with a Hall mobility of 0.749 cm{sup 2} V{sup -1} s{sup -1}, carrier concentration of 8.02 x 1018 cm{sup -3}, and transmittance of about 80% in visible range showing good crystallinity with prior c-axis orientation. > A shallow acceptor level of 91 meV is identified from free-to-neutral-acceptor transitions. > Another deep level of 255 meV was ascribed to Li{sub Zn}-Li{sub i} complex. - Abstract: Zinc oxide thin films with low resistivity have been deposited on glass substrates by Li-N dual-acceptor doping method via a modified successive ionic layer adsorption and reaction process. The thin films were systematically characterized via scanning electron microscopy (SEM), atomic force microscopy (AFM), X-ray diffraction, ultraviolet-visible spectrophotometry and fluorescence spectrophotometry. The resistivity of zinc oxide film was found to be 1.04 {Omega} cm with a Hall mobility of 0.749 cm{sup 2} V{sup -1} s{sup -1} and carrier concentration of 8.02 x 10{sup 18} cm{sup -3}. The Li-N dual-acceptor doped zinc oxide films showed good crystallinity with prior c-axis orientation, and high transmittance of about 80% in visible range. Moreover, the effects of Li doping level and other parameters on crystallinity, electrical and ultraviolet emission of zinc oxide films were investigated.

Model and simulation of a vacuum sieve tray for T extraction from liquid PbLi breeding blankets

International Nuclear Information System (INIS)

Mertens, M.A.J.; Demange, D.; Frances, L.

2016-01-01

Highlights: • A simulation tool was developed to analyse, optimise and scale up VST set-ups. • This tool predicts that efficiencies higher than 90% can be reached. • Upscaling to DEMO breeding blanket flow rates results in feasibly sized designs. - Abstract: Tritium self-sufficiency within a nuclear fusion reactor is necessary to demonstrate nuclear fusion as a viable source of energy. Tritium can be produced within liquid eutectic PbLi but then has to be extracted to be refuelled to the plasma. The vacuum sieve tray (VST) method is based on the extraction of tritium from millimetre-scaled oscillating PbLi droplets falling inside a vacuum chamber. A simulation tool was developed describing the fluid dynamics occurring along the PbLi flow and was used to study the influence of the different geometrical and operational parameters on the VST performance. The simulation predicts that extraction efficiencies over 90% can be easily reached according to theory and previous experimental results. The size of the VST extraction unit for a fusion reactor is estimated based on the findings from our single-nozzle model and assuming no T reabsorption. It is found to be in the feasible range. Nevertheless, two approaches are discussed which may further reduce this size by up to 90%. The simulation tool proved to be an easy and powerful way to analyse and optimise VST set-ups at any scale.

HANARO secondary coolant management

International Nuclear Information System (INIS)

Kim, Seon Duk.

1998-02-01

In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

Radiation hazards due to activated corrosion and neutron sputtering products in fusion reactor coolant and tritium breeding fluids

International Nuclear Information System (INIS)

Klein, A.C.; Vogelsang, W.F.

1985-01-01

The accumulation of radioactive corrosion and neutron sputtering products on the surfaces of components in fusion reactor coolant and tritium breeding systems can cause significant personnel access problems. Remote maintenance techniques or special treatment may be required to limit the amount of radiation exposure to plant operational and maintenance personnel. A computer code, RAPTOR, has been developed to estimate the transport of this activated material throughout a fusion heat transfer and/or tritium breeding material loop. A method is devised which treats the components of the loop individually and determines the source rates, deposition and erosion rates, decay rates, and purification rates of these radioactive materials. RAPTOR has been applied to the MARS and Starfire conceptual reactor designs to determine the degree of the possible radiation hazard due to these products. Due to the very high corrosion release rate by HT-9 when exposed to LiPb in the MARS reactor design, the radiation fields surrounding the primary system will preclude direct contact maintenance even after shutdown. Even the removal of the radioactive LiPb from the system will not decrease the radiation fields to reasonable levels. The Starfire primary system will exhibit radiation fields similar to those found in present pressurized water reactors. (orig.)

Hydrodynamic problems of heavy liquid metal coolants technology in loop-type and mono-block-type reactor installations

International Nuclear Information System (INIS)

Orlov, Yuri I.; Efanov, Alexander D.; Martynov, Pyotr N.; Gulevsky, Valery A.; Papovyants, Albert K.; Levchenko, Yuri D.; Ulyanov, Vladimir V.

2007-01-01

In the report, the influence of hydrodynamics of the loop with heavy liquid metal coolants (Pb and Pb-Bi) on the realization methods and efficiency of the coolant technology for the reactor installations of loop, improved loop and mono-block type of design has been studied. The last two types of installations, as a rule, are characterized by the following features: availability of loop sections with low hydraulic head and low coolant velocities, large squares of coolant free surfaces; absence of stop and regulating valve, auxiliary pumps on the coolant pumping-over lines. Because of the different hydrodynamic conditions in the installation types, the tasks of the coolant technology have specific solutions. The description of the following procedures of coolant technology is given in the report: purification by hydrogen (purification using gas mixture containing hydrogen), regulation of dissolved oxygen concentration in coolant, coolant filtrating, control of dissolved oxygen concentration in coolant. It is shown that change of the loop design made with economic purpose and for improvement of the installation safety cause additional requirements to the procedures and apparatuses of the coolant technology realization

Dual-Phase CsPbBr3 -CsPb2 Br5 Perovskite Thin Films via Vapor Deposition for High-Performance Rigid and Flexible Photodetectors.

Science.gov (United States)

Tong, Guoqing; Li, Huan; Li, Danting; Zhu, Zhifeng; Xu, Enze; Li, Guopeng; Yu, Linwei; Xu, Jun; Jiang, Yang

2018-02-01

Inorganic perovskites with special semiconducting properties and structures have attracted great attention and are regarded as next generation candidates for optoelectronic devices. Herein, using a physical vapor deposition process with a controlled excess of PbBr 2 , dual-phase all-inorganic perovskite composite CsPbBr 3 -CsPb 2 Br 5 thin films are prepared as light-harvesting layers and incorporated in a photodetector (PD). The PD has a high responsivity and detectivity of 0.375 A W -1 and 10 11 Jones, respectively, and a fast response time (from 10% to 90% of the maximum photocurrent) of ≈280 µs/640 µs. The device also shows an excellent stability in air for more than 65 d without encapsulation. Tetragonal CsPb 2 Br 5 provides satisfactory passivation to reduce the recombination of the charge carriers, and with its lower free energy, it enhances the stability of the inorganic perovskite devices. Remarkably, the same inorganic perovskite photodetector is also highly flexible and exhibits an exceptional bending performance (>1000 cycles). These results highlight the great potential of dual-phase inorganic perovskite films in the development of optoelectronic devices, especially for flexible device applications. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Enhancing Hybrid Perovskite Detectability in the Deep Ultraviolet Region with Down-Conversion Dual-Phase (CsPbBr3-Cs4PbBr6) Films.

Science.gov (United States)

Tong, Guoqing; Li, Huan; Zhu, Zhifeng; Zhang, Yan; Yu, Linwei; Xu, Jun; Jiang, Yang

2018-04-05

Hybrid perovskite photodetectors (PDs) exhibit outstanding performance in the ultraviolet-visible (UV-vis) spectrum but have poor detectability in the deep ultraviolet (DUV) region (200-350 nm). In this work, a novel inorganic-hybrid architecture that incorporates a dual-phase (CsPbBr 3 -Cs 4 PbBr 6 ) inorganic perovskite material as a down-conversion window layer and a hybrid perovskite as a light capture layer was prepared to achieve faster, highly sensitive photodetection in the DUV spectrum. A dual-phase inorganic perovskite film coated on the back surface of the photodetector enables strong light absorption and tunes the incident energy into emission bands that are optimized for the perovskite photodetector. The presence of Cs 4 PbBr 6 enhances the capture and down-conversion of the incident DUV light. Due to the down-conversion and transport of the DUV photons, a self-driven perovskite photodetector with this composite structure exhibits a fast response time of 7.8/33.6 μs and a high responsivity of 49.4 mA W -1 at 254 nm without extra power supply.

Material analyses of foam-based SiC FCI after dynamic testing in PbLi in MaPLE loop at UCLA

Energy Technology Data Exchange (ETDEWEB)

Gonzalez, Maria, E-mail: maria.gonzalez@ciemat.es [LNF-CIEMAT, Avda Complutense, 40, 28040 Madrid (Spain); Rapisarda, David; Ibarra, Angel [LNF-CIEMAT, Avda Complutense, 40, 28040 Madrid (Spain); Courtessole, Cyril; Smolentsev, Sergey; Abdou, Mohamed [Fusion Science and Technology Center, UCLA (United States)

2016-11-01

Highlights: • Samples from foam-based SiC FCI were analyzed by looking at their SEM microstructure and elemental composition. • After finishing dynamic experiments in the flowing hot PbLi, the liquid metal ingress has been confirmed due to infiltration through local defects in the protective inner CVD layer. • No direct evidences of corrosion/erosion were observed; these defects could be related to the manufacturing process. - Abstract: Foam-based SiC flow channel inserts (FCIs) developed and manufactured by Ultramet, USA are currently under testing in the flowing hot lead-lithium (PbLi) alloy in the MaPLE loop at UCLA to address chemical/physical compatibility and to access the MHD pressure drop reduction. UCLA has finished the first experimental series, where a single uninterrupted long-term (∼6500 h) test was performed on a 30-cm FCI segment in a magnetic field up to 1.8 T at the temperature of 300 °C and maximum flow velocities of ∼ 15 cm/s. After finishing the experiments, the FCI sample was extracted from the host stainless steel duct and cut into slices. Few of them have been analyzed at CIEMAT as a part of the joint collaborative effort on the development of the DCLL blanket concept in the EU and the US. The initial inspection of the slices using optical microscopic analysis at UCLA showed significant PbLi ingress into the bulk FCI material that resulted in degradation of insulating properties of the FCI. Current material analyses at CIEMAT are based on advanced techniques, including characterization of FCI samples by FESEM to study PbLi ingress, imaging of cross sections, composition analysis by EDX and crack inspection. These analyses suggest that the ingress was caused by local defects in the protective inner CVD layer that might be originally present in the FCI or occurred during testing.

Study of fission reactions induced by 4,6He and 7Li beams on 209Bi and 208Pb targets

Directory of Open Access Journals (Sweden)

Lukyanov S.M.

2013-12-01

Full Text Available Study of fission reactions induced by 4,6He and 7Li beams on 209Bi and 208Pb targets, leading to the production of 210,212A compound nuclei, was performed. It was shown that the fission excitation functions for the three reactions 4,6He + 209Bi and 7Li + 208Pb had similar behavior within the experimental error for a broad range of energy. More likely, halo structure of 6He is not reflected on the fission reaction mechanism. Otherwise, a large value of the fusion cross section was observed so far, as it could be expected in the case of weakly bound character of 6He projectile.

Dynamic Recrystallization Behavior and Corrosion Resistance of a Dual-Phase Mg-Li Alloy

Directory of Open Access Journals (Sweden)

Gang Liu

2018-03-01

Full Text Available The hot deformation and dynamic recrystallization behavior of the dual-phase Mg-9Li-3Al-2Sr-2Y alloy had been investigated using a compression test. The typical dual-phase structure was observed, and average of grain size of as-homogenized alloy is about 110 µm. It mainly contains β-Li, α-Mg, Al4Sr and Al2Y phases. The dynamic recrystallization (DRX kinetic was established based on an Avrami type equation. The onset of the DRX process occurred before the peak of the stress–strain flow curves. It shows that the DRX volume fraction increases with increasing deformation temperature or decreasing strain rate. The microstructure evolution during the hot compression at various temperatures and strain rates had been investigated. The DRX grain size became larger with the increasing testing temperature or decreasing strain rate because the higher temperature or lower strain rate can improve the migration of DRX grain boundaries. The fully recrystallized microstructure can be achieved in a small strain due to the dispersed island-shape α-Mg phases, continuous the Al4Sr phases and spheroidal Al2Y particles, which can accelerate the nucleation. The continuous Al4Sr phases along the grain boundaries are very helpful for enhancing the corrosion resistance of the duplex structured Mg-Li alloy, which can prevent the pitting corrosion and filiform corrosion.

Dynamic Recrystallization Behavior and Corrosion Resistance of a Dual-Phase Mg-Li Alloy.

Science.gov (United States)

Liu, Gang; Xie, Wen; Wei, Guobing; Yang, Yan; Liu, Junwei; Xu, Tiancai; Xie, Weidong; Peng, Xiaodong

2018-03-09

The hot deformation and dynamic recrystallization behavior of the dual-phase Mg-9Li-3Al-2Sr-2Y alloy had been investigated using a compression test. The typical dual-phase structure was observed, and average of grain size of as-homogenized alloy is about 110 µm. It mainly contains β-Li, α-Mg, Al₄Sr and Al₂Y phases. The dynamic recrystallization (DRX) kinetic was established based on an Avrami type equation. The onset of the DRX process occurred before the peak of the stress-strain flow curves. It shows that the DRX volume fraction increases with increasing deformation temperature or decreasing strain rate. The microstructure evolution during the hot compression at various temperatures and strain rates had been investigated. The DRX grain size became larger with the increasing testing temperature or decreasing strain rate because the higher temperature or lower strain rate can improve the migration of DRX grain boundaries. The fully recrystallized microstructure can be achieved in a small strain due to the dispersed island-shape α-Mg phases, continuous the Al₄Sr phases and spheroidal Al₂Y particles, which can accelerate the nucleation. The continuous Al₄Sr phases along the grain boundaries are very helpful for enhancing the corrosion resistance of the duplex structured Mg-Li alloy, which can prevent the pitting corrosion and filiform corrosion.

Neutronic optimization of a LiAlO2 solid breeder blanket

International Nuclear Information System (INIS)

Levin, P.; Ghoniem, N.M.

1986-02-01

In this report, a pressurized lobular blanket configuration is neutronically optimized. Among the features of this blanket configuration are the use of beryllium and LiAlO 2 solid breeder pins in a cross-flow configuration in a helium coolant. One-dimensional neutronic optimization calculations are performed to maximize the tritium breeding ratio (TER). The procedure involves spatial allocations of Be, LiAlO 2 , 9-C (ferritic steel), and He; in such a way as to maximize the TBR subject to several material, engineering and geometrical constraints. A TBR of 1.17 is achieved for a relatively thin blanket (approx. = 43 cm depth), and consistency with all imposed constraints

Evaluation of primary coolant pH operation methods for the domestic PWRs

International Nuclear Information System (INIS)

Paek, Seung Woo; Na, Jung Won; Kim, Yong Eak; Bae, Jae Heum

1992-01-01

Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of-core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed. (Author)

Crystal structure of a Zn-doped derivative of the Li17Ge4 compound

International Nuclear Information System (INIS)

Lacroix-Orio, L.; Tillard, M.; Belin, C.

2008-01-01

The compound Li 17-ε Zn ε Ge 4 has been obtained as a side product during the preparation of the intermetallic compound Li 8 Zn 2 Ge 3 from the elements. Its structure has been determined from single crystal X-ray diffraction intensities measured at 173 K. It crystallizes in the cubic system, F4-bar3m space group, a = 18.842(1) A, Z = 20. Its crystal structure is slightly different from those so far reported in the literature for the Zn-free phase Li 17 Ge 4 , particularly concerned are the positions and the site occupations of Li atoms. Most likely, these structural variations result from the presence of a small Zn concentration in the compound. The Zn doping atom has been found only at the specific Li 4d site (about 3 at.% Zn)

Experimental and numerical studies of pressure drop in PbLi flows in a circular duct under non-uniform transverse magnetic field

Energy Technology Data Exchange (ETDEWEB)

Li, F.-C., E-mail: lifch@hit.edu.cn; Sutevski, D.; Smolentsev, S.; Abdou, M.

2013-11-15

Highlights: • An indirect DP measurement approach for high-temperature LM MHD flow is developed. • Experiments and numerical simulations of PbLi MHD flow are performed. • Characteristics of DP in LM MHD flow under fringing magnetic field are studied. • Pressure distributions in LM MHD flow at entry and exit of magnet are different. -- Abstract: Experiments and three-dimensional (3D) numerical simulations are performed to investigate the magnetohydrodynamic (MHD) characteristics of liquid metal (LM) flows of molten lead-lithium (PbLi) eutectic alloy in an electrically conducting circular duct subjected to a transverse non-uniform (fringing) magnetic field. An indirect measurement approach for differential pressure in high temperature LM PbLi is first developed, and then detailed data on pressure drop in this PbLi MHD flow are measured. The obtained experimental results for the pressure distribution are in good agreement with numerical simulations. Using the numerical simulation results, the 3D effects caused by fringing magnetic field on the LM flow are illustrated via distributions for the axial pressure gradients and transverse pressure differences. It has been verified that a simple approach for estimation of pressure drop in LM MHD flow in a fringing magnetic field proposed by Miyazaki et al. [22] i.e., a simple integral of pressure gradient along the fringing field zone using a quasi-fully-developed flow assumption, is also applicable to the conditions of the present experiment providing the magnetic interaction parameter is large enough. Furthermore, for two different sections of the LM flow at the entry to and at the exit from the magnet, it is found that the pressure distributions in the duct cross sections in these two regions are different.

Improved Cycling Stability of Cobalt-free Li-rich Oxides with a Stable Interface by Dual Doping

International Nuclear Information System (INIS)

Xie, Dongjiu; Li, Guangshe; Li, Qi; Fu, Chaochao; Fan, Jianming; Li, Liping

2016-01-01

Highlights: • Cobalt-free Na_xLi_1_._2_-_xMn_0_._6_-_xAl_xNi_0_._2O_2 oxides are prepared by a sol-gel method. • Dual-doping strengthens the covalence of Mn-O bonds and suppresses the side reactions between cathode and electrolyte. • Doped cathode has a capacity retention over 92.2% after 100 cycles at a high temperature of 55 °C. - Abstract: Li-rich cobalt-free oxides, popularly used as a cathode with high capacity in lithium ion battery, always suffer from poor cycling stability between 2.0 and 4.8 V vs Li"+/Li, especially when cycled at high temperatures (>50 °C). To overcome this issue, Na"+ and Al"3"+ dual-doped Na_xLi_1_._2_-_xMn_0_._6_-_xAl_xNi_0_._2O_2 Li-rich cathode is prepared in this study. It is shown that the side reactions between cathode and electrolyte during cycling are suppressed. The improved cycling performance is observed for all of the doped samples, among which the sample with x = 0.03 exhibits the highest capacity retention of 86.1% after 200 cycles between 2.0 and 4.8 V at 2C (1C = 200 mA g"−"1) and shows a remarkable cycling stability, even at a high temperature of 55 °C (a capacity retention of 92.2% after 100 cycles). Moreover, the average voltage of the sample with x = 0.03 after 100 cycles at 0.5C remains at 3.11 V with a retention ratio of 86.6%. This work provides a new strategy to develop Li-rich cobalt-free cathodes with excellent cycling stability for lithium ion batteries at high temperatures.

Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

2006-07-01

The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

Assessment of corrosion phenomena in liquid lithium at T < 873 K. A Li(d,n) neutron source as case study

Energy Technology Data Exchange (ETDEWEB)

Knaster, J., E-mail: juan.knaster@ifmif.org [IFMIF/EVEDA Project Team (F4E), Rokkasho (Japan); Favuzza, P. [ENEA, Firenze (Italy)

2017-05-15

The corrosion induced by alkali metals in steels has been the subject of long decades of intense studies under both nuclear fission and fusion research programs. Li or its eutectic Pb-17Li is the liquid metal coolant choice for fusion blankets due to the tritium breeder capability of Li. Non-metal impurities enhance corrosion, but only N becomes potentially a problem given its high solubility in liquid Li and the depletion of Cr through ternary nitrides Li-Cr-N. The low solubility of C and O allow its cold trapping to values <10 wppm, however N can only be hot trapped demanding temperatures typically of 873 K. The inherent difficulties of experimentation on physicochemical kinetics related with alkali metals lead to a confusing divergence of results available in the literature; however, the understanding of the corrosion phenomena of RAFM steels exposed to flowing Li up to 873 K is mature. Next decade, 14 MeV neutrons will be available for fusion materials testing through Li(d,n) nuclear reactions. In such a facility, a concave RAFM steel backplate will be channelling 523 K flowing Li in the region where the 40 MeV deuteron beam will be impacting. If RAFM steels are considered, two main concurrent mechanisms will take place: a) mass transport of alloying elementsalong the loop and b) depletion of Cr through formation of Li{sub 9}CrN{sub 5}. Fortunately, the mass transport phenomena of Cr within the ΔT = 350 K in the loop is limited due to the poor solubility of Cr in liquid Li (0.21 wppm at 873 K). In turn, at 523 K Li the activity of N to form the ternary compound is negligible. However, the high solubility of Ni in Li (2144 wppm at 873 K), suggests the presence of mass transport phenomena of Ni from the stainless steel piping; unfortunately, the physicochemical kinetics are not fully understood. Lifus 6, in operation in Brasimone (ENEA) since the end 2015, will close in a definitive manner remaining open questions.

Conceptual design of China fusion power plant FDS-II

International Nuclear Information System (INIS)

Wu, Y.; Liu, S.; Chen, H.

2007-01-01

As one of the series of fusion system design concepts developed by the FDS Team of China, FDS-II is designated to exploit and evaluate potential attractiveness of fusion energy application for the generation of electricity on the basis of conservatively advanced plasma parameters, which can be limitedly extrapolated from the successful operation of ITER. The principle of the blanket design is established in both the feasibility and potential attractiveness of technology to meet the requirement for tritium self-sufficiency, safety margin, operation economy and environment protection etc. The plasma physics and engineering parameters of FDS-II are selected on the basis of the progress in recent experiments and associated theoretical studies of magnetic confinement fusion plasma with a fusion power of 2∝3 GW. The neutron wall load of 2∝3 MW/m 2 and the surface heat flux of 0.5∝1 MW/m 2 are considered for high effective power conversion. The ''multi-modules'' scenario is adopted in the FDS-II blanket design to reduce thermal stress and electromagnetic forces under plasma disruption, with liquid metal lithium lead (LiPb) as tritium breeder, the Reduced Activation Ferritic/Martensitic (RAFM) steel as structural material. Two options of specific liquid LiPb blanket concepts have been proposed, named the Dual-cooled Lithium Lead (DLL) breeder blanket and the Quasi-Static Lithium Lead (SLL) breeder blanket. The DLL blanket is a dual-cooled LiPb breeder system with helium gas to cool the first wall and main structure and LiPb eutectic to be self-cooled. The flow channel inserts (FCIs), e.g. SiCf/SiC composites, are designed as the thermal and electrical insulators inside the LiPb flow channels to reduce the magnetohydrodynamic (MHD) pressure drop and to allow the coolant LiPb outlet temperature up to 700 C for high thermal efficiency. The SLL blanket is another option of the FDS-II blanket with the technology developed relatively easily. To avoid or mitigate the

Electrochemical Characteristics and Li+ Ion Intercalation Kinetics of Dual-phase Li4Ti5O12/Li2TiO3 Composite in Voltage Range of 0−3 V

KAUST Repository

Bhatti, Humaira S

2016-04-20

Li4Ti5O12, Li2TiO3 and dual-phase Li4Ti5O12/Li2TiO3 composite were prepared by sol-gel method with average particle size of 1 µm, 0.3 µm and 0.4 µm, respectively. Though Li2TiO3 is electrochemically inactive, the rate capability of Li4Ti5O12/Li2TiO3 is comparable to Li4Ti5O12 at different current rates. Li4Ti5O12/Li2TiO3 also shows good rate performance of 90 mA h g-1 at high rate of 10 C in voltage range of 1−3 V, attributable to increased interfaces in the composite. While Li4Ti5O12 delivers capacity retention of 88.6 % at 0.2 C over 50 cycles, Li4Ti5O12/Li2TiO3 exhibits no capacity fading at 0.2 C (40 cycles) and capacity retention of 98.45 % at 0.5 C (50 cycles). This highly stable cycling performance is attributed to the contribution of Li2TiO3 in preventing undesirable reaction of Li4Ti5O12 with the electrolyte during cycling. CV curves of Li4Ti5O12/Li2TiO3 in 0−3 V range exhibit two anodic peaks at 1.51 V and 0.7−0.0 V, indicating two modes of lithium intercalation into the lattice sites of active material. Owing to enhanced intercalation/de-intercalation kinetics in 0−3 V, composite electrode delivers superior rate performance of 203 mAh/g at 2.85 C and 140 mAh/g at 5.7 C with good reversible capacity retention over 100 cycles.

Electrochemical Characteristics and Li+ Ion Intercalation Kinetics of Dual-phase Li4Ti5O12/Li2TiO3 Composite in Voltage Range of 0−3 V

KAUST Repository

Bhatti, Humaira S; Anjum, Dalaver H.; Ullah, Shafiq; Ahmed, Bilal; Habib, Amir; Karim, Altaf; Hasanain, Syed Khurshid

2016-01-01

Li4Ti5O12, Li2TiO3 and dual-phase Li4Ti5O12/Li2TiO3 composite were prepared by sol-gel method with average particle size of 1 µm, 0.3 µm and 0.4 µm, respectively. Though Li2TiO3 is electrochemically inactive, the rate capability of Li4Ti5O12/Li2TiO3 is comparable to Li4Ti5O12 at different current rates. Li4Ti5O12/Li2TiO3 also shows good rate performance of 90 mA h g-1 at high rate of 10 C in voltage range of 1−3 V, attributable to increased interfaces in the composite. While Li4Ti5O12 delivers capacity retention of 88.6 % at 0.2 C over 50 cycles, Li4Ti5O12/Li2TiO3 exhibits no capacity fading at 0.2 C (40 cycles) and capacity retention of 98.45 % at 0.5 C (50 cycles). This highly stable cycling performance is attributed to the contribution of Li2TiO3 in preventing undesirable reaction of Li4Ti5O12 with the electrolyte during cycling. CV curves of Li4Ti5O12/Li2TiO3 in 0−3 V range exhibit two anodic peaks at 1.51 V and 0.7−0.0 V, indicating two modes of lithium intercalation into the lattice sites of active material. Owing to enhanced intercalation/de-intercalation kinetics in 0−3 V, composite electrode delivers superior rate performance of 203 mAh/g at 2.85 C and 140 mAh/g at 5.7 C with good reversible capacity retention over 100 cycles.

Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

Science.gov (United States)

Urgorri, F. R.; Moreno, C.; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, I.; Palermo, I.; Ibarra, A.

2017-11-01

The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg h-1 while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

Rate-dependent, Li-ion insertion/deinsertion behavior of LiFePO4 cathodes in commercial 18650 LiFePO4 cells.

Science.gov (United States)

Liu, Qi; He, Hao; Li, Zhe-Fei; Liu, Yadong; Ren, Yang; Lu, Wenquan; Lu, Jun; Stach, Eric A; Xie, Jian

2014-03-12

We have performed operando synchrotron high-energy X-ray diffraction (XRD) to obtain nonintrusive, real-time monitoring of the dynamic chemical and structural changes in commercial 18650 LiFePO4/C cells under realistic cycling conditions. The results indicate a nonequilibrium lithium insertion and extraction in the LiFePO4 cathode, with neither the LiFePO4 phase nor the FePO4 phase maintaining a static composition during lithium insertion/extraction. On the basis of our observations, we propose that the LiFePO4 cathode simultaneously experiences both a two-phase reaction mechanism and a dual-phase solid-solution reaction mechanism over the entire range of the flat voltage plateau, with this dual-phase solid-solution behavior being strongly dependent on charge/discharge rates. The proposed dual-phase solid-solution mechanism may explain the remarkable rate capability of LiFePO4 in commercial cells.

The 208Pb(7Li,6He)209Bi reaction at 52 MeV

International Nuclear Information System (INIS)

Zeller, A.F.; Weisser, D.C.; Ophel, T.R.; Hebbard, D.F.

1979-11-01

Single proton transfers to low lying levels in 209 Bi from the 208 Pb(7Li, 6 He) reaction at 52 MeV have been measured and spectroscopic factors derived from an EFR-DWBA analysis. Relative spectroscopic factors are in good agreement with light ion results and previous heavy ion work. Absolute spectroscopic factors were generally too large and the peaks of the angular distributions were out of phase with the DWBA calculations by 1 0 - 4 0

Birefringence Polarimeter Using Dual LiNbO3 Electrooptic Crystal Modulators

Science.gov (United States)

Saitou, Takeshi; Nurdin Bin, Muhammad; Kowa, Hiroyuki; Umeda, Norihiro; Takizawa, Kuniharu; Kondoh, Eiichi; Jin, Lianhua

2012-08-01

A birefringence polarimeter that uses dual LiNbO3 electrooptic crystal modulators operating at a frequency ratio of 4:1 is described. The significance of this polarimeter is that the birefringent parameters of a sample are obtained only from the modulated polarization status. The measurement, therefore, avoids depolarization effects resulting from the sample itself and the rest of the optical system. The high speed and accuracy of this polarimeter are shown by measurements using a quarter-wave plate, a Babinet-Soleil compensator, and a phase modulator.

Deposition of Li{sub 4}Ti{sub 5}O{sub 12} and LiMn{sub 2}O{sub 4} films on the lithium-ion conductor of Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3} sintered pellet

Energy Technology Data Exchange (ETDEWEB)

Wu, Xian Ming, E-mail: xianmingwu@163.com [College of Chemistry and Chemical Engineering, Jishou University, Jishou Hunan 416000 (China); Xiangxi Minerals and New Materials Research and Service Center, Jishou Hunan 416000 (China); Chen, Shang [College of Chemistry and Chemical Engineering, Jishou University, Jishou Hunan 416000 (China); Xiangxi Minerals and New Materials Research and Service Center, Jishou Hunan 416000 (China); He, Ze Qiang; Chen, Shou Bin; Li, Run Xiu [College of Chemistry and Chemical Engineering, Jishou University, Jishou Hunan 416000 (China)

2015-08-31

LiMn{sub 2}O{sub 4} and Li{sub 4}Ti{sub 5}O{sub 12} films were deposited on the lithium-ion conductor of Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3} sintered pellet by spray technique. The effect of annealing temperature, annealing time, Li:Ti and Li:Mn molar ratio on the phase and crystallization of the films were investigated with X-ray diffraction. The LiMn{sub 2}O{sub 4}/Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3}/Li{sub 4}Ti{sub 5}O{sub 12} thin-film lithium-ion battery using Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3} sintered pellet as both electrolyte and substrate was also studied. The results show that the effect of annealing temperature, annealing time, Li:Ti and Li:Mn molar ratio has great effect on the phase and crystallization of Li{sub 4}Ti{sub 5}O{sub 12} and LiMn{sub 2}O{sub 4} films deposited on the Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3} sintered pellet. The optimal Li:Ti and Li:Mn molar ratio for the deposition of Li{sub 4}Ti{sub 5}O{sub 12} and LiMn{sub 2}O{sub 4} films on Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3} sintered pellet are 7.2:5 and 1.05:2, respectively. The optimal annealing temperature and time for the deposition of LiMn{sub 2}O{sub 4} film on Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3} sintered pellet are 650 °C and 10 min. While those for Li{sub 4}Ti{sub 5}O{sub 12} film are 700 °C and 10 min. The LiMn{sub 2}O{sub 4}/Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3}/Li{sub 4}Ti{sub 5}O{sub 12} thin-film battery offers a working voltage about 2.25 V and can be easily cycled. - Highlights: • LiMn{sub 2}O{sub 4} and Li{sub 4}Ti{sub 5}O{sub 12} films spray deposited on Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7}(PO{sub 4}){sub 3} sintered pellet • Film crystal phase depends on the spray solution composition and annealing conditions. • Prepared thin-film lithium-ion battery employs sintered pellet as electrolyte and substrate. • LiMn{sub 2}O{sub 4}/Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7

Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

Energy Technology Data Exchange (ETDEWEB)

Mistrangelo, C., E-mail: chiara.mistrangelo@kit.edu; Bühler, L.

2014-10-15

Highlights: • MHD flows in model geometries of DCLL blanket manifolds. • Study of velocity, pressure distributions and flow partitioning in parallel ducts. • Flow partitioning affected by 3D MHD pressure drop and velocity distribution in the expanding zone. • Reduced pressure drop in a continuous expansion compared to a sudden expansion. - Abstract: An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure. Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules. In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.

Qualification of MHD effects in dual-coolant DEMO blanket and approaches to their modelling

International Nuclear Information System (INIS)

Mas de les Valls, E.; Batet, L.; Medina, V. de; Fradera, J.; Sedano, L.A.

2011-01-01

Design refinements of vertical insulated banana-shaped liquid metal channels are being considered as a progress of conceptual design of dual-coolant liquid metal blankets (DEMO specifications). Among them: (a) optimised channel geometry and (b) improvements on flow channel inserts. Progress of channel conceptual design is conducted in parallel with underlying physics of MHD models in diverse aspects: (1) MHD models, (2) MHD turbulence, (3) LM buoyancy effects, (4) three-dimensional flows, and (5) LM/FCI/wall electrical and thermal coupling; in order to progress on common liquid metal flow characterisation, pressure drop and three-dimensional flows. The analyses are assumed as extension of those previous carried out for the DCLL blankets for new design refinements. At the present stage of the conceptual design progress, a preliminary thermofluid MHD study is of crucial interest for further design improvements and future detailed modelling. The paper overviews the ongoing modelling studies, making model refinements explicit, and anticipates some modelling results.

Study of the threshold anomaly in the scattering of polarized 7Li from 208Pb

International Nuclear Information System (INIS)

Martel, I.; Gomez-Camacho, J.; Blyth, C.O.; Davis, N.J.; Rusek, K.; Connell, K.A.; Lilley, J.S.; Bailey, M.W.

1995-01-01

Experimental data on elastic and inelastic analysing powers T 20 and inelastic cross sections for the scattering of polarized 7 Li from a 208 Pb target are presented. The experimental data are analyzed with DWBA and coupled channels calculations, which show the sensitivity of the experimental data to the real and imaginary parts of the nuclear transition form factor. This study reveals the existence of a threshold anomaly for the transition terms of the interaction. ((orig.))

Liquid metal technology in fusion

International Nuclear Information System (INIS)

Torre Cabezas, M. de la; Martin Espigares, M.; Lapena, J.

1985-01-01

Lithium (or Li-Pb) is one of the several possible coolants being considered for the blanket of magnetic toroidal fusion reactor, not only because of its good thermal and neutron properties, but also because the tritium required to fuel the reactor can be produced by neutron reactions in the lithium. In this paper two main technology tasks to be proposed in our fusion programme have been identified: 1) the development of impurity monitoring devices for use in lithium and Li-Pb environments; 2) effects of Li and Li-Pb environments on the low cycle fatigue properties of different steels. (author)

Reply to "On Vaporization of liquid Pb-Li eutectic alloy from 1000 K to 1200 K- A high temperature mass spectrometric study"

Science.gov (United States)

Jain, Uttam; Mukherjee, Abhishek

2018-03-01

This communication is in response to a letter to editor commenting on the authors' earlier paper "Vaporization of liquid Pb-Li eutectic alloy from 1000 K to 1200 K - A high temperature mass spectrometric study".

Studies on the phase diagram of Pb-Mo-O system

International Nuclear Information System (INIS)

Aiswarya, P.M.; Ganesan, Rajesh; Gnanasekaran, T.

2014-01-01

Liquid lead and Lead-Bismuth Eutectic (LBE) alloy are considered as spallation target and coolant in the accelerator driven systems and as candidate coolant in advanced nuclear reactors. Corrosion of the structural steel components in these liquid metal coolants can be minimized by the insitu formation of passive oxide layer on the steel surface under controlled oxygen concentration. A detailed knowledge of phase diagrams of Pb-M-O and Bi-M-O (M = Fe, Cr, Mo) systems and data on thermochemical properties of the ternary compounds of these systems are required for better understanding of composition and stability of these passive oxide films. In the present work, studies have been carried out to establish the ternary phase diagram of Pb-Mo-O system

Research activities for measurement of hydrogen solubility in LiPb

International Nuclear Information System (INIS)

Katayama, K.; Edao, Y.; Fukada, S.

2014-01-01

Research activities for measurement of hydrogen solubility in LiPb has been conducted under the U.S.-Japan TITAN collaboration program at STAR facility in Idaho National Laboratory. The solubility test using a quartz crucible has been tried at 300degC but the crack had been caused at about 15 hours after temperature reached 300degC. This is caused by chemical reaction of LLE and SiO_2. Hydrogen solubility and apparent diffusivity has been estimated from data obtained by solubility tests using alumina crucible by fitting method assuming one dimensional diffusion. The estimated solubility in the range from 300degC to 500degC is close to Reiter's one. However, the value of solubility increased sharply at 600degC. This seemed to be caused by chemical reaction of LLE and Al_2O_3. The estimated diffusivity was two orders of magnitude larger than literature data. This large difference may be due to natural convection of LLE in the alumina crucible. (author)

Petrogenesis of orogenic lamproites of the Bohemian Massif: Sr-Nd-Pb-Li isotope constraints for Variscan enrichment of ultra-depleted mantle domains

Czech Academy of Sciences Publication Activity Database

Krmíček, Lukáš; Romer, R. L.; Ulrych, Jaromír; Glodny, J.; Prelevič, D.

2016-01-01

Roč. 35, 1 July (2016), s. 198-216 ISSN 1342-937X Institutional support: RVO:67985831 Keywords : Silica-rich lamproites * Sr-Nb-Pb-Li isotopes * mantle metasomatism * Variscides Subject RIV: DB - Geology ; Mineralogy Impact factor: 6.959, year: 2016

Proceedings of 'workshop on Pb-alloy cooled fast reactor'

International Nuclear Information System (INIS)

Kim, Sang Ji; Kim, Yong Hee; Hong, Ser Gi

2003-06-01

The objective of 'Workshop on Pb-Alloy Cooled Fast Reactor', held in Taejeon, Korea on May 6, 2003, is to enhance the basic knowledge in this area by facilitating the exchange of information and discussions about problematic area of design aspects. There were five presentations from three different countries and about 25 participants gathered during the workshop. The topics covered in the workshop include benefits and drawbacks of Pb-alloy and Sodium coolant, two Pb-alloy cooled 900 MWt reactor designs using both B4C rods and NSTs, BREST-300 breakeven reactor and transmutation effectiveness of LLFPs in the typical thermal/fast neutron systems. The generic conclusion for the Pb-alloy cooled fast reactor from this workshop is as follows: 1) It has a potential to satisfy the goals established for the Generation-IV reactor concepts, so it has a bright future. 2) As a fast neutron system with a moderate breeding or a conversion, it is flexible in its roles and has superior safety characteristics over sodium coolant because of Pb-alloy's chemical inertness with water/air and high boiling temperature

Evaluation of Two 300 MWe Fourth Generation Pb-Bi Reactor System Concepts

International Nuclear Information System (INIS)

Miller, Laurence F.; Khuram Khan, M.; Williams, Wesley; Mynatt, F.R.

2002-01-01

This paper describes the evaluation of two 300 MWe modular Pb-Bi cooled reactor system concepts that can be field assembled from components shipped on standard rail cars or on trucks. Thus, the largest components must be smaller than 12' x 12' x 80' (3.66 m x 3.66 m x 24.4 m) and should weigh no more than 80 tons. One of these systems utilizes a cylindrical two-loop containment vessel for the core and the other is a slab design. The fuel for both designs consists of standard-sized metallic IFR fuel in 17 x 17 square array assemblies with a pitch-to-diameter ratio of 1.15. The coolant outlet temperature is limited by current material technology, which is estimated to be 550 C. The primary coolant inlet temperature is selected to be 350 C. This is well above the melting temperature of Pb-Bi, and it is expected to be sufficiently high to limit transient-induced thermal stresses to acceptable values. Coolant flow rates through the core and external piping are below 1 m/s. The results from neutronics calculations include power distributions, reactivity coefficients, and fuel depletion, and results from heat transfer calculations include temperatures and flow rates at various locations in the primary and secondary systems. The neutronic design calculations are accomplished by using a discrete ordinate transport code and a cross section processing system developed at Oak Ridge National Laboratory. Two-dimensional flux distributions are obtained with the DOORS system, and ORIGEN-S, coupled with KENO, is used for time-dependent depletion calculations. The thermal-hydraulic design of the core consists of heat transfer and fluid flow calculation for an average channel. The inlet and outlet temperatures, along with the fuel centerline temperature, are determined in conjunction with core flow rates, pumping power, and total power output. This is accomplished by using a lumped parameter steady-state model with a spreadsheet and by using a one-dimensional time-dependent model

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

NOEL: a no-leak fusion blanket concept

International Nuclear Information System (INIS)

Powell, J.R.; Yu, W.S.; Fillo, J.A.; Horn, F.L.; Makowitz, H.

1980-01-01

Analysis and tests of a no-leak fusion blanket concept (NOEL-NO External Leak) are described. Coolant cannot leak into the plasma chamber even if large through-cracks develop in the first wall. Blanket modules contain a two-phase material, A, that is solid (several cm thick) on the inside of the module shell, and liquid in the interior. The solid layer is maintained by imbedded tubes carrying a coolant, B, below the freezing point of A. Most of the 14-MeV neutron energy is deposited as heat in the module interior. The thermal energy flow from the module interior to the shell keeps the interior liquid. Pressure on the liquid A interior is greater than the pressure on B, so that B cannot leak out if failures occur in coolant tubes. Liquid A cannot leak into the plasma chamber through first wall cracks because of the intervening frozen layer. The thermal hydraulics and neutronics of NOEL blankets have been investigated for various metallic (e.g., Li, Pb 2 , LiPb, Pb) and fused salt choices for material A

Electrochemical fabrication, microstructure and magnetic properties of Sm2Co17/Fe7Co3 dual phase nanocomposite

International Nuclear Information System (INIS)

Cui, Chunxiang; Chen, Fenghua; Yang, Wei; Li, Hongfang; Liu, Qiaozhi; Sun, Jibing

2015-01-01

By utilizing alternate electrochemical reaction, atomic migration and deposition of Fe, Co, Sm and other chemical substances in the electrochemical solution, a large number of Sm 2 Co 17 /Fe 7 Co 3 dual phase nanowire arrays were carried out in the anodic aluminum oxide (AAO) template with highly uniform and orderly. The Sm 2 Co 17 /Fe 7 Co 3 dual phase nanowire arrays with diameter of 50 nm and length of 12 μm have the smooth surface and uniform diameter. The morphology and microstructure of annealed Sm 2 Co 17 /Fe 7 Co 3 dual phase nanowires were observed and analyzed using SEM, TEM and HRTEM. Compared with single-phase nanowires, dual phase magnetic nanowires have higher coercivity and saturation magnetization. In this composite system, both the hard and the soft phases have a high Curie temperature, therefore, we believe that the Sm 2 Co 17 /Fe 7 Co 3 dual phase nanowire arrays is a new type of high-temperature magnetic composites. - Highlights: • Sm 2 Co 17 /Fe 7 Co 3 dual phase nanowires were prepared by electrochemical method. • The interface pinning is the main factor to improve anisotropy field of the nanowires. • The dual phase magnetic nanowires have higher coercivity and saturation magnetization

Fundamentals for the development of a low-activation lead coolant with isotopic enrichment for advanced nuclear power facilities

International Nuclear Information System (INIS)

Khorasanov, G.L.; Blokhin, A.I.

2002-01-01

The purpose of this paper is to study the prospects of new coolants for fast reactors and accelerator driven systems. The main focus is on their improvement using the isotopic tailoring technique to reduce post-irradiation activity. Calculations using the FISPACT-3 code show that irradiating natural lead (Pb-nat) for 30 years leads to the accumulation of long-lived toxic radionuclides, 207 Bi, 208 Bi and 210 Pb, which extends the cooling down period to the clearance level. This time can be shortened by using the lead isotope 206 Pb instead of Pb-nat. This substantially decreases the concentration of the most toxic polonium isotope, 210 Po. Calculations for lead activation in the hard proton-neutron ADS spectrum were performed using the CASCADE/SNT code. The time-dependent activity of the 207 Bi produced in Pb-nat and 206 Pb after irradiation for one year with a proton beam having an energy of 0.8 GeV and a current of 30 mA is given. The activity of 207 Bi is decreased by four orders of magnitude when 206 Pb is used instead of natural lead as a coolant for ADS targets. The production of such radiotoxic nuclides as 210 Po is also substantially diminished. (author)

Investigations of neutron-rich nuclei at the dripline through their analogue states : The cases of $^{10}$Li - $^{10}$Be (T=2) and $^{17}$C - $^{17}$N (T=5/2)

CERN Multimedia

2002-01-01

We propose to study the elastic resonance scattering reactions $^{9}$Li+p and $^{16}$C+p to investigate the energies, spins and parities of the lowest T=2 states in $^{10}$Be and the T=5/2 states in $^{17}$N. These are analogue states of the ground states and first excited states in $^{10}$Li and $^{17}$C.

Gamma ray shielding properties of PbO-Li2O-B2O3 glasses

Science.gov (United States)

Kumar, Ashok

2017-07-01

The mass attenuation coefficients have been measured in (0.6-x) PbO-x Li2O-0.40 B2O3 (where 0≤ x≤0.25 mol%) glasses for photon energies of 356, 662, 1173 and 1332 keV in a narrow beam geometry with an overall scatter acceptance angle of 2.31°. The experimental results are found to be within 3% of their theoretical values. These coefficients were then used to obtain the values of mean free path, effective atomic number and electron density. The shielding properties of these glasses have also been compared among themselves in terms of their mean free path and radiation protection efficiency. The shielding properties prepared glasses have also been compared with standard concretes as well as with the standard shielding glasses. It is found that the prepared glasses are the better shielding substitute to the conventional concretes as well as other standard shielding glasses. The Pb3B4O9 has been found to be the most effective shield.

Benchmark Linelists and Radiative Cooling Functions for LiH Isotopologues

Science.gov (United States)

Diniz, Leonardo G.; Alijah, Alexander; Mohallem, José R.

2018-04-01

Linelists and radiative cooling functions in the local thermodynamic equilibrium limit have been computed for the six most important isotopologues of lithium hydride, 7LiH, 6LiH, 7LiD, 6LiD, 7LiT, and 6LiT. The data are based on the most accurate dipole moment and potential energy curves presently available, the latter including adiabatic and leading relativistic corrections. Distance-dependent reduced vibrational masses are used to account for non-adiabatic corrections of the rovibrational energy levels. Even for 7LiH, for which linelists have been reported previously, the present linelist is more accurate. Among all isotopologues, 7LiH and 6LiH are the best coolants, as shown by the radiative cooling functions.

Principal Physical and Technical Advantages from the Use of Radiogenic Lead as a Coolant of Power Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

Kulikov, G.G. [International Science and Technology Center (ISTC), Krasnoproletarskaya ulitsa 32-34, Moscow, 127473 (Russian Federation); Shmelev, A.N.; Apse, V.A. [Moscow Engineering Physics Institute (State University), Kashirskoe shosse 31, Moscow, 115409 (Russian Federation); Artisyuk, V.V. [Obninsk State Technical University of Nuclear Power Engineering, Obninsk, Kaluzhskaya reg. 249040 (Russian Federation)

2009-06-15

Radiogenic lead is a final product of radioactive decay chains in uranium and thorium ores. After a number of alpha- and beta-decays, the starting isotopes {sup 232}Th, {sup 238}U and {sup 235}U are converted into stable lead isotopes: {sup 208}Pb, {sup 206}Pb and {sup 207}Pb, respectively. Radiogenic lead with a large fraction of {sup 208}Pb has unique neutron-physical properties because {sup 208}Pb is a double magic nuclide with closed proton and neutron shells in nucleus. That is why {sup 208}Pb has an extremely low cross-section of thermal neutron capture reaction ({approx}0.5 mb) in comparison with common lead ({approx}175 mb) and graphite ({approx}3.5 mb). In addition, {sup 208}Pb is a weak neutron moderator through inelastic scattering of fast neutrons owing to the higher first energy excitation level of nucleus ({approx}2.7 MeV for {sup 208}Pb as against {approx}0.8 MeV for common lead) and through elastic scattering owing to a high atomic number. So, high {sup 208}Pb content in lead coolant of fast reactor allows a decrease in the unfavorable spectral component in a coolant temperature reactivity coefficient [1]. In spite of {sup 208}Pb content being as high as 52% in common lead, the remaining lead fraction (mainly {sup 207}Pb and {sup 204}Pb isotopes) is characterized both by a large neutron capture cross-section and essential inelastic scattering. Radiogenic lead from thorium and uranium-thorium ores has a very low fraction of these unfavorable isotopes. The use of radiogenic lead as a coolant and graphite as a structural material creates favorable conditions for development of high-temperature and high-flux reactors. Such a high-temperature reactor differs profitably from He-cooled HTGR by low pressure and natural circulation of coolant, while such a high-flux reactor makes it possible to transmute radioactive isotopes with extremely low neutron capture cross-sections, like {sup 90}Sr and {sup 137}Cs. Plutonium in ({sup 238}U-Pu-Th-{sup 233}U

European research and development programme for water-cooled lithium-lead blankets: present status and future work

International Nuclear Information System (INIS)

Giancarli, L.; Leroy, P.; Proust, E.; Raepsaet, X.

1992-01-01

The European R and D programme in support of the development of water-cooled Pb-17Li blankets for DEMO aims at improving the data base concerning tritium behaviour and compatibility between blanket materials. The four main areas of the experimental programme are structural material corrosion by Pb-17Li, tritium extraction and permeation control.=, Pb-17Li physico-chemistry, and water/Pb-17Li interaction. This paper describes the most significant results obtained to date in the various experiments performed in Europe and the future programme required to complete the data base by 1994. 28 refs

Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

Energy Technology Data Exchange (ETDEWEB)

Kramer, Kevin James [Univ. of California, Berkeley, CA (United States)

2010-04-08

This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 μm of tungsten to mitigate x-ray damage. The first wall is cooled by Li₁₇Pb₈₃ eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li₁₇Pb₈₃ flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li₁₇Pb₈₃, separated from the Li₁₇Pb₈₃ by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF₂), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles

Implementation of the chemical PbLi/water reaction in the SIMMER code

Energy Technology Data Exchange (ETDEWEB)

Eboli, Marica, E-mail: marica.eboli@for.unipi.it [DICI—University of Pisa, Largo Lucio Lazzarino 2, 56122 Pisa (Italy); Forgione, Nicola [DICI—University of Pisa, Largo Lucio Lazzarino 2, 56122 Pisa (Italy); Del Nevo, Alessandro [ENEA FSN-ING-PAN, CR Brasimone, 40032 Camugnano, BO (Italy)

2016-11-01

Highlights: • Updated predictive capabilities of SIMMER-III code. • Verification of the implemented PbLi/Water chemical reactions. • Identification of code capabilities in modelling phenomena relevant to safety. • Validation against BLAST Test No. 5 experimental data successfully completed. • Need for new experimental campaign in support of code validation on LIFUS5/Mod3. - Abstract: The availability of a qualified system code for the deterministic safety analysis of the in-box LOCA postulated accident is of primary importance. Considering the renewed interest for the WCLL breeding blanket, such code shall be multi-phase, shall manage the thermodynamic interaction among the fluids, and shall include the exothermic chemical reaction between lithium-lead and water, generating oxides and hydrogen. The paper presents the implementation of the chemical correlations in SIMMER-III code, the verification of the code model in simple geometries and the first validation activity based on BLAST Test N°5 experimental data.

Assessment of the integration of a He-cooled divertor system in the power conversion system for the dual-coolant blanket concept (TW2-TRP-PPCS12D8)

International Nuclear Information System (INIS)

Norajitra, P.; Kruessmann, R.; Malang, S.; Reimann, G.

2002-12-01

Application of a helium-cooled divertor together with the dual-coolant blanket concept is considered favourable for achieving a high thermal efficiency of the power plant due to its relatively high coolant outlet temperature. A new FZK He-cooled modular divertor concept with integrated pin arrays (HEMP) is introduced. Its main features and function are described in detail. The result of the thermalhydraulic analysis shows that the HEMP divertor concept has the potential of resisting, a heat flow density of at least 10-15 MW/m 2 at a reachable heat transfer coefficient of approx. 60 kW/m 2 K and a reasonable pumping power. Integration of this divertor concept into the power conversion system using a closed Brayton gas turbine system with three-stage compression leads to a net efficiency of the blanket/divertor cycle of about 43%. (orig.)

The astrophysical reaction 8Li(n,gamma)9Li from measurements by reverse kinematics

OpenAIRE

Bertulani, Carlos A.

1998-01-01

We study the breakup of 9Li projectiles in high energy (28.5 MeV/u) collisions with heavy nuclear targets (208Pb). The wave functions are calculated using a single-particle model for 9Li, and a simple optical potential model for the scattering part. A good agreement with measured data is obtained with insignificant E2 contribution.

Pulse Star Inertial Confinement Fusion Reactor: Heat transfer loop and balance-of-plant considerations

International Nuclear Information System (INIS)

McDowell, M.W.; Blink, J.A.; Curlander, K.A.

1983-01-01

A conceptual heat transfer loop and balance-of-plant design for the Pulse Star Inertial Confinement Fusion Reactor has been investigated and the results are presented. The Pulse Star reaction vessel, a perforated steel bell jar about11 m in diameter, is immersed in Li 17 Pb 83 coolant, which flows through the perforations and forms a 1.5-m-thick plenum of droplets around a 8-m-diameter inner chamber. The bell jar and associated pumps, piping, and steam generators are contained within a 17-m-diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops, each with a flow rate of 5.5 m 3 /s, are necessary to transfer 3300 MWt of power. Liquid metal is pumped to the top of the pool, where it flows downward through eight vertical steam generators. Double-walled tubes are used in the steam generators to assure tritium containment without intermediate heat transfer loops. Each pump is a mixed flow type and has a required NPSH of 3.4 m, a speed of 278 rpm, and an impeller diameter of 1.2 m. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. The design minimizes the total cost (heat exchanger area plus pumping) for the plant lifetime. The power required for the pumps is 36 MWe. Each resulting steam generator is 12 m high and 1.6 m in diameter, with 2360 tubes. The steam generators and pumps fit easily in the pool between the reactor chamber and the pool wall

PbBr3 Perovskite Crystals

KAUST Repository

Wei, Tzu-Chiao; Mokkapati, Sudha; Li, Ting-You; Lin, Chun-Ho; Lin, Gong-Ru; Jagadish, Chennupati; He, Jr-Hau

2018-01-01

, such as lithium niobate (LiNbO3), LiTaO3, KTiOPO4, and KH2PO4. Such a strong two-photon absorption effect in CH3NH3PbBr3 can be used to modulate the spectral and spatial profiles of laser pulses, as well as to reduce noise, and can be used to strongly control

Comparison of thermohydraulic characteristics in the use of various coolants

International Nuclear Information System (INIS)

Muramatsu, Toshiharu; Suda, Kazunori; Yamaguchi, Akira

2000-11-01

Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiated based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1) There is no remarkable difference between liquid sodium and liquid Pb-Bi in characteristics of internal flows and free surface characteristics based on Fr number. (2) The AQUA-VOF code has a potential enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [Thermal Stratification Phenomena] (1) On-set position of thermal entrainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. On the other hand, the position in the case of CO 2 gas was shifted to upstream side with decreasing of Ri number. (2) Destruction speed of the thermal stratification interface was dependent on thermal diffusivity as fluid properties. Therefore it was concluded that an elimination method is necessary for the interface generated in CO 2 gas. [Thermal Striping Phenomena] (1) Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO 2 gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2) To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it is necessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phynomlena] (1) Fundamental behavior of the natural convection in various coolant follows buoyant jet

Pressure effect on the transport properties of superconducting Li0.9MobO17bronze

International Nuclear Information System (INIS)

Filippini, C.E.; Boujida, M.; Marcus, J.; Schlenker, C.; Beille, J.

1989-01-01

The electrical resistivity of Li 0.9 Mo 6 O 17 single crystal has been studied between 1.5 K and 300 K under hydrostatic pressures up to 20 k bar. A large increase of the superconducting transition temperature, from 1.7 K to 2.5 K, is associated to a sharp decease of the temperature T m of the electronic, probably CDW instability

Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

International Nuclear Information System (INIS)

Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

2009-01-01

The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

The effect of dual complexing agents of lactic and citric acids on the formation of sol-gel derived Ag–PbTiO3 percolative thin film

International Nuclear Information System (INIS)

Su, Yanbo; Hu, Tao; Tang, Liwen; Weng, Wenjian; Han, Gaorong; Ma, Ning; Du, Piyi

2014-01-01

Controlling the formation of conductive particles to be nano-scale is important for achieving percolation effect in metal dispersed thin film composite to contribute extraordinary dielectric properties required for miniaturization of electronic devices. In this paper, lactic acid (LA) and citric acid (CA) were used as dual complexing agents to prepare a typical Ag nanoparticle dispersed PbTiO 3 (PTO) composite thin film by using a sol-gel method. The phase structure of the thin film and the coordination effect between complexing agent and metallic ions were investigated. It revealed that LA coordinated with Ti 4+ and Pb 2+ and CA coordinated with Ag + . Lead was fixed inside the gel network by LA and restricted to evaporate during heat treatment thus the pyrochlore phase was prevented from forming in the thin film. Ag + was coordinated by CA and the diffusion and thus aggregation of silver during gelation and annealing process were weakened. Silver nanoparticles dispersed in the PTO matrix formed with dual complexing agents of LA and CA introduced during the preparation process. The composite thin film of perfect perovskite phase with silver nanoparticles embedded was obtained at the molar ratio of LA/lead = 0.5 and CA/lead = 0.5. The dielectric constant of the thin film with silver nanoparticles is 5 times higher than that without silver nanoparticles. - Highlights: • Ag nanoparticle–PbTiO 3 percolative film with high dielectric property is prepared. • Evaporation of lead was prevented by coordinating Pb with lactic acid agent. • Dual complexing agents contribute block and pinning effects to form Ag nanoparticles

Pb3O4 type antimony oxides MSb2O4 (M = Co, Ni) as anode for Li-ion batteries

International Nuclear Information System (INIS)

Jibin, A.K.; Reddy, M.V.; Subba Rao, G.V.; Varadaraju, U.V.; Chowdari, B.V.R.

2012-01-01

Graphical abstract: Isostructural Pb 3 O 4 type MSb 2 O 4 (M = Co, Ni) compounds were investigated as possible anodes for lithium ion batteries. The reversible capacity is due to electrochemically active Sb and the transition metal and Li 2 O form an inactive matrix which buffers volume variations associated with alloying-de-alloying of antimony. Highlights: ► Isostructural MSb 2 O 4 (M = Co, Ni) were studied as anode for LIBs for first time. ► Li/MSb 2 O 4 (M = Co, Ni) cells displayed reversibility due to electrochemically active Sb. ► CoSb 2 O 4 showed good reversibility compared to NiSb 2 O 4 . - Abstract: Polycrystalline samples of isostructural MSb 2 O 4 (M = Co, Ni) have been prepared by solid state synthesis and lithium-storage is investigated as possible anode materials for lithium-ion batteries. The reaction mechanism of lithium with MSb 2 O 4 (M = Co, Ni) is explored by galvanostatic cycling, cyclic voltammogram and ex situ studies. Both CoSb 2 O 4 and NiSb 2 O 4 exhibit similar electrochemical behavior and show reversible capacity of 490 and 412 mAh g −1 respectively in the first cycle. Reversible alloying de-alloying of Li x Sb takes place in an amorphous matrix of M (Co, Ni) and Li 2 O during electrochemical cycling.

Neutronics Evaluation of Lithium-Based Ternary Alloys in IFE Blankets

Energy Technology Data Exchange (ETDEWEB)

Jolodosky, A. [Univ. of California, Berkeley, CA (United States); Fratoni, M. [Univ. of California, Berkeley, CA (United States)

2015-09-22

Lithium is often the preferred choice as breeder and coolant in fusion blankets as it offers excellent heat transfer and corrosion properties, and most importantly, it has a very high tritium solubility and results in very low levels of tritium permeation throughout the facility infrastructure. However, lithium metal vigorously reacts with air and water and exacerbates plant safety concerns. For this reason, over the years numerous blanket concepts have been proposed with the scope of reducing concerns associated with lithium. The European helium cooled pebble bed breeding blanket (HCPB) physically confines lithium within ceramic pebbles. The pebbles reside within a low activation martensitic ferritic steel structure and are cooled by helium. The blanket is composed of the tritium breeding lithium ceramic pebbles and neutron multiplying beryllium pebbles. Other blanket designs utilize lead to lower chemical reactivity; LiPb alone can serve as a breeder, coolant, neutron multiplier, and tritium carrier. Blankets employing LiPb coolants alongside silicon carbide structural components can achieve high plant efficiency, low afterheat, and low operation pressures. This alloy can also be used alongside of helium such as in the dual-coolant lead-lithium concept (DCLL); helium is utilized to cool the first wall and structural components made up of low-activation ferritic steel, whereas lithium-lead (LiPb) acts as a self-cooled breeder in the inner channels of the blanket. The helium-cooled steel and lead-lithium alloy are separated by flow channel inserts (usually made out of silicon carbide) which thermally insulate the self-cooled breeder region from the helium cooled steel walls. This creates a LiPb breeder with a much higher exit temperature than the steel which increases the power cycle efficiency and also lowers the magnetohydrodynamic (MHD) pressure drop [6]. Molten salt blankets with a mixture of lithium, beryllium, and fluorides (FLiBe) offer good tritium breeding

Theoretical basis of oxygen pressure control in liquid Pb-Bi using YSZ

International Nuclear Information System (INIS)

Jung, S. H.; Hwang, I. S.; Park, B. K.

2002-01-01

To develop a liquid Pb-Bi cooled reactor, it is necessary to solve the structural material corrosion problem caused by Pb-Bi. This experiment examine the fundamental behaviors to practically test the oxide film formation on the surface of structural material known as solution of corrosion inhibition in liquid Pb-Bi. The corrosion inhibition through oxide film formation is to prevent metals from dissolving into liquid Pb-Bi though not forming coolants slug resulted from oxidation. In this paper, we examined the oxygen pressure controllability using YSZ in cover gas, and theoretically derived the relationship between oxygen cover gas pressure and dissolved oxygen in liquid Pb-Bi

Coolant clean-up system in the primary coolant circuit for nuclear reactor

International Nuclear Information System (INIS)

Saito, Michio.

1981-01-01

Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

Corrosion of austenitic and martensitic stainless steels in flowing 17Li83Pb alloy

Science.gov (United States)

Broc, M.; Flament, T.; Fauvet, P.; Sannier, J.

1988-07-01

With regard to the behaviour of 316 L stainless steel at 400°C in flowing anisothermal 17Li83Pb the mass transfer suffered by this steel appears to be quite important without noticeable influence of constant or cyclic stress. Evaluation made from solution-annealed specimens leads to a corrosion rate of approximately 30 μm yr -1 at steady state to which a depth of 25 μm has to be added to take into account the initial period phenomena. On the other hand, with semi-stagnant 17Li83Pb at 400° C, the mass transfer of 316 L steel appears to be lower and more acceptable after a 3000-h exposure; but long-time kinetics data have to be achieved in order to see if that better behaviour is persistent and does not correspond to a longer incubation period. As for the martensitic steels their corrosion rate at 450°C in the thermal convection loop TULIP is constant up to 3000 h and five times lower than that observed for 316 L steel in the same conditions.

Graphite|LiFePO4 lithium-ion battery working at the heat engine coolant temperature

Science.gov (United States)

Lewandowski, Andrzej; Kurc, Beata; Swiderska-Mocek, Agnieszka; Kusa, Natalia

2014-11-01

Electrochemical properties of the graphite anode and the LiFePO4 cathode, working together with the 1 M LiPF6 in TMS (sulpholane) at 90 °C have been studied. The general aim of the investigation was to demonstrate a potential application for a Li-ion cell working in the cooling system of a car heat engine (90 °C). Electrodes were characterized with the use of electrochemical impedance spectroscopy (EIS), scanning electron microscopy (SEM) as well as galvanostatic charging/discharging tests. SEM images of both electrodes after charging/discharging processes were covered with a film (electrochemical SEI formation). The charge transfer resistance at 90 °C, Rct, of the C6Li|Li+ anode and the LiFePO4 cathode was 24 Ω and 110 Ω, respectively. Reversible capacity of the LiC6 anode after 10-20 cycles, at a low current rate was close to the theoretical value of 370 mAh g-1 however an increasing current rate decreased to ca. 200 mAh g-1 (for 1C). The reversibility of the process was close to 95%. The capacity of the LiFePO4 cathode was ca. 150 mAh g-1, almost independent of the current rate and close to the theoretical value of 170 mAh g-1.

Electrochemical fabrication, microstructure and magnetic properties of Sm{sub 2}Co{sub 17}/Fe{sub 7}Co{sub 3} dual phase nanocomposite

Energy Technology Data Exchange (ETDEWEB)

Cui, Chunxiang, E-mail: hutcui@hebut.edu.cn [Key Lab. for New Type of Functional Materials in Hebei Province, Hebei University of Technology, No.8, Road No.1, Dingzigu, Hongqiao District, Tianjin 300130 (China); Chen, Fenghua [Tianjin Sanhuan Lucky New Materials Inc., Tianjin Economical-Technological Development Area (TEDA), Tianjin 300457 (China); Yang, Wei; Li, Hongfang; Liu, Qiaozhi; Sun, Jibing [Key Lab. for New Type of Functional Materials in Hebei Province, Hebei University of Technology, No.8, Road No.1, Dingzigu, Hongqiao District, Tianjin 300130 (China)

2015-06-15

By utilizing alternate electrochemical reaction, atomic migration and deposition of Fe, Co, Sm and other chemical substances in the electrochemical solution, a large number of Sm{sub 2}Co{sub 17}/Fe{sub 7}Co{sub 3} dual phase nanowire arrays were carried out in the anodic aluminum oxide (AAO) template with highly uniform and orderly. The Sm{sub 2}Co{sub 17}/Fe{sub 7}Co{sub 3} dual phase nanowire arrays with diameter of 50 nm and length of 12 μm have the smooth surface and uniform diameter. The morphology and microstructure of annealed Sm{sub 2}Co{sub 17}/Fe{sub 7}Co{sub 3} dual phase nanowires were observed and analyzed using SEM, TEM and HRTEM. Compared with single-phase nanowires, dual phase magnetic nanowires have higher coercivity and saturation magnetization. In this composite system, both the hard and the soft phases have a high Curie temperature, therefore, we believe that the Sm{sub 2}Co{sub 17}/Fe{sub 7}Co{sub 3} dual phase nanowire arrays is a new type of high-temperature magnetic composites. - Highlights: • Sm{sub 2}Co{sub 17}/Fe{sub 7}Co{sub 3} dual phase nanowires were prepared by electrochemical method. • The interface pinning is the main factor to improve anisotropy field of the nanowires. • The dual phase magnetic nanowires have higher coercivity and saturation magnetization.

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

Science.gov (United States)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

Multicomponent diffusion in molten salt LiF-BeF{sub 2}: Dynamical correlations and Maxwell–Stefan diffusivities

Energy Technology Data Exchange (ETDEWEB)

Chakraborty, Brahmananda, E-mail: brahma@barc.gov.in; Ramaniah, Lavanya M. [High Pressure & Synchrotron Radiation Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai-400085 (India)

2015-06-24

Applying Green–Kubo formalism and equilibrium molecular dynamics (MD) simulations, we have studied the dynamic correlation, Onsager coeeficients and Maxwell–Stefan (MS) Diffusivities of molten salt LiF-BeF{sub 2}, which is used as coolant in high temperature reactor. All the diffusive flux correlations show back-scattering or cage dynamics which becomes pronouced at higher temperature. Although the MS diffusivities are expected to depend very lightly on the composition due to decoupling of thermodynamic factor, the diffusivity Đ{sub Li-F} and Đ{sub Be-F} decreases sharply for higher concentration of LiF and BeF{sub 2} respectively. Interestingly, all three MS diffusivities have highest magnitude for eutectic mixture at 1000K (except Đ{sub Be-F} at lower LiF mole fraction) which is desirable from coolant point of view. Although the diffusivity for positive-positive ion pair is negative it is not in violation of the second law of thermodynamics as it satisfies the non-negative entropic constraints.

Synthesis and Optical Characterization of Nd3+ doped TeO2-PbO-Li2O

Directory of Open Access Journals (Sweden)

M. Rahim Sahar

2012-02-01

Full Text Available Glass based on Nd3+-doped TeO2-PbO-Li2O has successfully been made by melt quenching technique and their thermal parameters have been determined using Differential Thermal Analyzer (DTA. The glass is then nucleated and/or growth by controlled heat treatment at slightly below the crystallization temperature. The X-ray diffraction (XRD technique is used to estimate the nano-crystallite size. Meanwhile, the optical characterization has been determined using the Photoluminescence Spectroscopy. It is found out that the crystallite size is about 20 nm and very much depending on the heat-treatment time. Meanwhile, the intensity of the luminescence spectra is very much depending on the concentration of the dopant.   Keyword: tellurium glasses, melt quenching technique, optical characterization

Five-nucleon simultaneous and sequential transfer in the 12C(11B,6Li)17O and 12C(d,7Li)7Be reactions

International Nuclear Information System (INIS)

Jarczyk, L.; Kamys, B.; Kistryn, M.; Magiera, A.; Rudy, Z.; Strzal/kowski, A.; Barna, R.; DAmico, V.; De Pasquale, D.; Italiano, A.; Licandro, M.

1996-01-01

Measurements of the angular distributions of the 12 C( 11 B, 6 Li) 17 O reaction were performed at three energies of a 11 B beam: 28, 35, and 40 MeV. The results were analyzed in the frame of the exact finite range distorted wave Born approximation of the first and the second order assuming the simultaneous and sequential transfer of the neutron and the α particle. Such an analysis was also performed for previously measured angular distributions of the 12 C(d, 7 Li) 7 Be reaction at E lab = 78 MeV. In both reactions under investigation dominance was found of the simultaneous transfer of the α particle and the nucleon correlated to the 5 He ( 5 Li) cluster in the ground or the first excited state. copyright 1996 The American Physical Society

The EC conceptual design proposal of a water-cooled convertible blanket for ITER

International Nuclear Information System (INIS)

Giancarli, L.; Proust, E.; Baraer, L.; Bielak, B.; Raepsaet, X.; Salavy, J.F.; Sedano, L.; Szczepanski, J.; Quintric-Bossy, J.; Severi, Y.

1993-01-01

For several years the EC laboratories have developed breeding blankets for DEMO. From this experience, it has been derived a proposal of tritium breeding blanket for the Extended Performance Phase (EPP) of ITER. The general basic ideas are the following: (i) the switch from the shielding blanket used during the BPP to the breeding blanket for the EPP should not require segments replacement ('convertible' blanket): (ii) its use should not have significant impact on the Basic Performance Phase (BPP); (iii) design and used materials should assure good safety standards and acceptable public perception; (iv) the blanket coolant should be compatible with the coolant required in the high heat-flux components (e.g. divertor, etc.; (v) the required R and D should fit with the ITER time schedule; (vi) the blanket should be able to withstand large power excursions and to accept long downtimes. The proposed design consists of a water-cooled liquid metal blanket, using the eutectic Pb-17Li during the EPP and a non-breeding Pb-alloy (Pb-18Mg or Pb-50Bi) during the BPP. Each segment is basically formed by a box containing the alloy, cooled by an array of poloidal hairpin-type cooling tubes and reinforced by toroidal and radial stiffeners. The coolant tubes are double-walled tubes allowing leak detections. The selected First Wall (FW) is a toroidally-drilled steel plate with brazed water-cooling U-tube. The structural material is austenitic stainless steel (316L(N)) which limits the maximum acceptable neutron fluence to about 1 MWa/m 2 . The advantages of using other structural materials requiring longer leadtimes, such as ferritic/martensitic steels, are also briefly discussed

Technologies for hydrogen production based on direct contact of gaseous hydrocarbons and evaporated water with Molten Pb or Pb-Bi

International Nuclear Information System (INIS)

Gulevich, A. V.; Martynov, P. N.; Gulevsky, V. A.; Ulyanov, V. V.

2007-01-01

Results of studies intended for the substantiation of a new energy-saving and safe technology for low cost hydrogen production have been presented. The technology's basis is direct mixing of water and (or) gaseous hydrocarbons with heavy liquid metal coolants (HLMC) Pb or Pb-Bi. Preliminary research has been done on thermal dynamics and kinetics of the processes taking place in the interaction of HLMC with hydrocarbon-containing gases. It has been shown as a result that water and gaseous hydrocarbons interact with molten Pb and Pb-Bi relatively quietly in chemical aspect (without ignition and explosions). Therefore, (and taking into account the thermal physics, physical and chemical properties of HLMC such as low pressure of saturated vapor of Pb and Pb- Bi in enhanced temperatures, their good heat conductivity and heat capacity, low viscosity, etc.) heat transfer is possible from the molten metal to water and hydrocarbons without heat transferring partitions (that is, by direct contact of the working media). Devices to implement this method of heating liquid and gaseous media provide essential advantages: - A simple design; - None heat-transferring surfaces subject to corrosion, contamination, thermal fatigue, vibration impacts; - A high effectiveness owing to a larger heat exchanging surface per volume unit; - A small hydraulic resistance. The possibility and effectiveness of heating various gaseous and liquid media in their direct contact with molten Pb and Pb-Bi has been substantiated convincingly by experimental results at IPPE. Besides, the following processes of hydrogen-containing media conversion have been proved feasible thereby. 1. Water decomposition into hydrogen and oxygen. The process can develop at temperatures of 400-1000 degree C. It is necessary to provide constant removal of oxygen from the reaction zone and maintain a minimum possible content of chemically active oxygen in the melt. 2. Pyrolytic decomposition of hydrocarbons into carbon and

Conceptual design tool development for a Pb-Bi cooled reactor

International Nuclear Information System (INIS)

Lee, K. G.; Chang, S. H.; No, H. C.; Chunm, M. H.

2000-01-01

Conceptual design is generally ill-structured and mysterious problem solving. This leads the experienced experts to be still responsible for the most of synthesis and analysis task, which are not amenable to logical formulations in design problems. Especially because a novel reactor such as a Pb-Bi cooled reactor is going on a conceptual design stage, it will be very meaningful to develop the conceptual design tool. This tool consists of system design module with artificial intelligence, scaling module, and validation module. System design decides the optimal structure and the layout of a Pb-Bi cooled reactor, using design synthesis part and design analysis part. The designed system is scaled to be optimal with desired power level, and then the design basis accidents (Dbase) are analyzed in validation module. Design synthesis part contains the specific data for reactor components and the general data for a Pb-Bi cooled reactor. Design analysis part contains several design constraints for formulation and solution of a design problem. In addition, designer's intention may be externalized through emphasis on design requirements. For the purpose of demonstration, the conceptual design tool is applied to a Pb-Bi cooled reactor with 125 M Wth of power level. The Pb-Bi cooled reactor is a novel reactor concept in which the fission-generated heat is transferred from the primary coolant to the secondary coolant through a reactor vessel wall of a novel design. The Pb-Bi cooled reactor is to deliver 125 M Wth per module for 15 effective full power years without any on-site fuel handling. The conceptual design tool investigated the feasibility of a Pb-Bi cooled reactor. Application of the conceptual design tool will be, in detail, presented in the full paper. (author)

Coolant leakage detection device

International Nuclear Information System (INIS)

Ito, Takao.

1983-01-01

Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

Neutron flux measurement with 6Li and 7Li dual glass scintillators by γ compensation method

International Nuclear Information System (INIS)

Ji Changsong; Zhang Shulan; Zhang Shuheng

1996-01-01

Based on the characteristics of 6 Li glass scintillator which is sensitive to both neutron and gamma rays, and 7 Li glass scintillator which is sensitive to gamma rays only, a new method of detecting weak neutron flux under interference of strong gamma radiation has been investigated by means of 6 Li- 7 Li pair glass scintillator gamma compensation method. The result of neutron flux measurement by above-mentioned method with an error of about 1% when the gamma ray interference is up to 18.7% has been obtained

Determination of factor of hydrogen permeation reduction (PRF) for different protective coatings over vanadium

International Nuclear Information System (INIS)

Afanasyev, S.; Kulsartov, T.; Shestakov, V.; Chikhray, Y.; Smith, D.

2002-01-01

Selection of structural materials for liquid-metal system as well as for another system and constructions of nuclear energy plants must be carried out and based on specified demands depending on conditions of these materials functioning. Specific demand is its compatibility with liquid metals. Design of reactors with liquid-metal coolant (Li, PbLi 17 ) which reproduces tritium arise additional demand to structural materials. This demand is a creation of structural material or protective barrier with minimum acceptable value of tritium permeation through itself or with maximum permeation reduction factor (PRF). Vanadium and vanadium alloys are supposed to be use as a blanket structural material in such nuclear energy plants. Worked out at first stage of studies vanadium coatings should have stability of its characteristics at temperature 800 deg. C under influence of hydrogen. Given work shows the experimental results on testing of protective coatings over vanadium: glass-ceramic coating and CaO-base coating. PRF for every coating and its changes depending on thermo-capacity of vanadium sample with coating was determined by method of hydrogen permeation. The results of experiments would be used at the development of cooling loops of reactor core protection with liquid-metal coolant

Minimizing core deposits radiation fields in PWRs by coordinated Li/B chemistry

International Nuclear Information System (INIS)

Roesmer, J.

1983-01-01

The effect of coolant chemistry on the buildup and composition of core deposits and on out-of-core radiation fields was investigated in the Beaver Valley and Trojan plants. Coordinated Li/B coolant chemistry led to an appreciable reduction of the surface concentration of core deposits, decreased greatly the formation of crud films on fresh fuel, and resulted in a reduction in the rate and level of radiation field buildup in the out-of-core regions of the primary circuits. (author)

A novel dual-salts of LiTFSI and LiODFB in LiFePO4-based batteries for suppressing aluminum corrosion and improving cycling stability

Science.gov (United States)

Li, Faqiang; Gong, Yan; Jia, Guofeng; Wang, Qinglei; Peng, Zhengjun; Fan, Wei; Bai, Bing

2015-11-01

The strong corrosion behavior at the Al current collector restricts the application range of lithium bis (trifluoromethanesulfonylimide) (LiTFSI), despite its high stability against water and thermal. SEM, LSV and Tafel curves proved that adding LiODFB into LiTFSI-based electrolytes could suppress aluminum corrosion caused by LiTFSI-based electrolytes. The cycling stability and rate capability of LiFePO4-based batteries using LiTFSI0.6-LiODFB0.4-based electrolytes is excellent as compared to LiFePO4-based batteries using LiPF6-based electrolytes.

Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

International Nuclear Information System (INIS)

Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

2008-01-01

In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

Coupling of electromagnetics and structural/fluid dynamics - application to the dual coolant blanket subjected to plasma disruptions

International Nuclear Information System (INIS)

Jordan, T.

1996-01-01

Some aspects concerning the coupling of quasi-stationary electromagnetics and the dynamics of structure and fluid are investigated. The necessary equations are given in a dimensionless form. The dimensionless parameters in these equations are used to evaluate the importance of the different coupling effects. A finite element formulation of the eddy-current damping in solid structures is developed. With this formulation, an existing finite element method (FEM) structural dynamics code is extended and coupled to an FEM eddy-current code. With this program system, the influence of the eddy-current damping on the dynamic loading of the dual coolant blanket during a centered plasma disruption is determined. The analysis proves that only in loosely fixed or soft structures will eddy-current damping considerably reduce the resulting stresses. Additionally, the dynamic behavior of the liquid metal in the blankets' poloidal channels is described with a simple two-dimensional magnetohydrodynamic approach. The analysis of the dimensionless parameters shows that for small-scale experiments, which are designed to model the coupled electromagnetic and structural/fluid dynamic effects in such a blanket, the same magnetic fields must be applied as in the real fusion device. This will be the easiest way to design experiments that produce transferable results. 10 refs., 7 figs

Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

International Nuclear Information System (INIS)

Hunsbedt, A.; Boardman, C.E.

1993-01-01

A dual passive cooling system for liquid metal cooled nuclear fission reactors is described, comprising the combination of: a reactor vessel for containing a pool of liquid metal coolant with a core of heat generating fissionable fuel substantially submerged therein, a side wall of the reactor vessel forming an innermost first partition; a containment vessel substantially surrounding the reactor vessel in spaced apart relation having a side wall forming a second partition; a first baffle cylinder substantially encircling the containment vessel in spaced apart relation having an encircling wall forming a third partition; a guard vessel substantially surrounding the containment vessel and first baffle cylinder in spaced apart relation having a side wall forming a forth partition; a sliding seal at the top of the guard vessel edge to isolate the dual cooling system air streams; a second baffle cylinder substantially encircling the guard vessel in spaced part relationship having an encircling wan forming a fifth partition; a concrete silo substantially surrounding the guard vessel and the second baffle cylinder in spaced apart relation providing a sixth partition; a first fluid coolant circulating flow course open to the ambient atmosphere for circulating air coolant comprising at lent one down comer duct having an opening to the atmosphere in an upper area thereof and making fluid communication with the space between the guard vessel and the first baffle cylinder and at least one riser duct having an opening to the atmosphere in the upper area thereof and making fluid communication with the space between the first baffle cylinder and the containment vessel whereby cooling fluid air can flow from the atmosphere down through the down comer duct and space between the forth and third partitions and up through the space between the third and second partition and the riser duct then out into the atmosphere; and a second fluid coolant circulating flow

Li3V2(PO4)3-coated Li1.17Ni0.2Co0.05Mn0.58O2 as the cathode materials with high rate capability for Lithium ion batteries

International Nuclear Information System (INIS)

Liu, Yi; Huang, Xiao; Qiao, Qiqi; Wang, Yonglong; Ye, Shihai; Gao, Xueping

2014-01-01

In this work, Lithium rich layered oxide Li 1.17 Ni 0.2 Co 0.05 Mn 0.58 O 2 (LNCMO) is prepared and coated with Li 3 V 2 (PO 4 ) 3 (LVP) by a chemical deposition method. The surface modification with LVP is introduced into Li-rich layered oxides LNCMO for the first time. After 100 cycles of charging and discharging at various rates, the Li 3 V 2 (PO 4 ) 3 -coated Li 1.17 Ni 0.2 Co 0.05 Mn 0.58 O 2 (LVP-coated LNCMO) (5 wt%) still provides a large capacity of 261.4 mAh g -1 , much higher than the pristine LNCMO (211.5 mAh g -1 ). At 5 C rate, the LVP-coated LNCMO exhibits a stable cyclic capacity of 153.4 mAh g -1 , higher than 114.1 mAh g -1 of the pristine LNCMO. The electrochemical impedance spectroscopy (EIS) analysis demonstrates the LVP coating layer can suppress interaction between the cathode surface and the electrolyte and enhance the kinetics of lithium-ion diffusion, contributing to the stable cyclic performance with more cyclic capacity as well as at the high current density

LIBRA-LiTE: A commercial size light ion fusion power plant

International Nuclear Information System (INIS)

Badger, B.; Choi, B.; Engelstad, R.L.; Kulcinski, G.L.; Lovell, E.G.; MacFarlane, J.J.; Mogehed, E.A.; Moses, G.A.; Peterson, R.R.; Rutledge, S.; Sawan, M.E.; Sviatoslavsky, G.; Sviatoslavsky, I.N.; Wittenberg, L.J.

1992-05-01

LIBRA-LiTE is a concept study for future 1000 MWe nuclear fusion reactors operating on the principle of inertial confinement. Light ions, e.g. lithium ions, are given an energy of 25-35 MeV in an accelerator and focused symmetrically onto a target (deuterium-tritium filled sphere of 7 mm diameter) in a reactor chamber. The fusion reaction is ignited by shock wave induced compression of the target. The radiation (photons, neutrons, ions) is absorbed in a blanket where the thermal power is removed by a coolant and tritium is rebred. The LIBRA-LiTE concept study is the continuation of the earlier LIBRA study (330 MWe) with a modified concept of light ion beam focusing. Starting from an ion source (diode), the lithium ion beams are focused ballistically onto the target. For this to be achieved, lithium must be used as the coolant in the reactor chamber and the blanket concept must be slightly modified by providing steel tubes (HT-9) as guiding tubes for the coolant flow. A particular engineering problem to be solved are the ion beam focusing magnets, which have to extend rather closely up to the center of the reactor chamber. (orig.) [de

PbBr3 Perovskite Crystals

KAUST Repository

Wei, Tzu-Chiao

2018-01-31

Researchers have recently revealed that hybrid lead halide perovskites exhibit ferroelectricity, which is often associated with other physical characteristics, such as a large nonlinear optical response. In this work, the nonlinear optical properties of single crystal inorganic–organic hybrid perovskite CH3NH3PbBr3 are studied. By exciting the material with a 1044 nm laser, strong two-photon absorption-induced photoluminescence in the green spectral region is observed. Using the transmission open-aperture Z-scan technique, the values of the two-photon absorption coefficient are observed to be 8.5 cm GW−1, which is much higher than that of standard two-photon absorbing materials that are industrially used in nonlinear optical applications, such as lithium niobate (LiNbO3), LiTaO3, KTiOPO4, and KH2PO4. Such a strong two-photon absorption effect in CH3NH3PbBr3 can be used to modulate the spectral and spatial profiles of laser pulses, as well as to reduce noise, and can be used to strongly control the intensity of incident light. In this study, the superior optical limiting, pulse reshaping, and stabilization properties of CH3NH3PbBr3 are demonstrated, opening new applications for perovskites in nonlinear optics.

Evaluation of thermal conductivity for liquid lead lithium alloys at various Li concentrations based on measurement and evaluation of density, thermal diffusivity and specific heat of alloys

Energy Technology Data Exchange (ETDEWEB)

Kondo, Masatoshi, E-mail: kondo.masatoshi@nr.titech.ac.jp [Tokyo Institute of Technology, 2-12-1, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Nakajima, Yuu; Tsuji, Mitsuyo [Tokai University, 4-1-1 Kitakaname, Hiratsuka-shi, Kanagawa 259-1292 (Japan); Nozawa, Takashi [Japan Atomic Energy Agency, Rokkasyo-mura, Kamikita-gun, Aomori 039-3212 (Japan)

2016-11-01

Graphical abstract: Thermal diffusivities and thermal conductivities of liquid Pb–Li alloys (Pb–5Li, Pb–11Li and Pb–17Li). - Highlights: • The densities and specific heats of liquid Pb–Li alloys are evaluated based on the previous studies, and mathematically expressed in the equations with the functions of temperature and Li concentration. • The thermal diffusivities of liquid Pb–Li alloys (i.e., Pb–5Li, Pb–11Li and Pb–17Li) are obtained by laser flash method, and mathematically expressed in the equations with the functions of temperature and Li concentration. • The thermal conductivities of liquid Pb–Li alloys were evaluated and mathematically expressed in the equations with the functions of temperature and Li concentration. - Abstract: The thermophysical properties of lead lithium alloy (Pb–Li) are essential for the design of liquid Pb–Li blanket system. The purpose of the present study is to make clear the density, the thermal diffusivity and the heat conductivity of the alloys as functions of temperature and Li concentration. The densities of the solid alloys were measured by means of the Archimedean method. The densities of the alloys at 300 K as a function of Li concentration (0 at% < χ{sub Li} < 28 at%) were obtained in the equation as ρ{sub (300} {sub K)} [g/cm{sup 3}] = −6.02 × 10{sup −2} × χ{sub Li} + 11.3. The density of the liquid alloys was formulated as functions of temperature and Li concentration (0 at% < χ{sub Li} < 30 at%), and expressed in the equation as ρ [g/cm{sup 3}] = (9.00 × 10{sup −6} × T − 7.01 × 10{sup −2}) × χ{sub Li} + 11.4 − 1.19 × 10{sup −3}T. The thermal diffusivity of Pb, Pb–5Li, Pb–11Li and Pb–17Li were measured by means of laser flash method. The thermal diffusivity of Pb–17Li was obtained in the equation as α{sub Pb–17Li} [cm{sup 2}/s] = 3.46 × 10{sup −4}T + 1.05 × 10{sup −1} for the temperature range between 573 K and 773 K. The thermal conductivity of

Transfer of 6Li break-up fragments at 6Li projectile energies far above the coulomb barrier

International Nuclear Information System (INIS)

Neumann, B.; Buschmann, J.; Rebel, H.; Gils, H.J.; Klewe-Nebenius, H.

1979-05-01

Transfer of beam-velocity fragments has been experimentally investigated in 6 Li induced reactions on 208 Pb and 209 Bi in the energy range Esub(Li) = 60-156 MeV. The experimental techniques involve the observation of the target residues and measurements of the recoil ranges of heavy residual nuclei produced by charged particle bombardment. The determination of the recoil energy enables the discrimination of different reaction paths leading to the same residual nuclei. ( 6 Li, xn+p) excitation functions prove to be very similar to (α,(x-1)n) reactions at Esub(α) approximately 2/3 x Esub(Li). The results present experimental evidence for a particular reaction type indicated in previous experiments: Dissociation of the 6 Li projectile with capture of the beam-velocity alpha particle indicating an (α,xn) reaction ('internal break-up'). (orig.) [de

Microstructure evolution and texture development in thermomechanically processed Mg-Li-Al based alloys

Energy Technology Data Exchange (ETDEWEB)

Kumar, Vinod [Department of Materials Science and Engineering, IIT Kanpur (India); Govind [Vikram Sarabhai Space Center, Trivandrum (India); Shekhar, Rajiv; Balasubramaniam, R. [Department of Materials Science and Engineering, IIT Kanpur (India); Balani, Kantesh, E-mail: kbalani@iitk.ac.in [Department of Materials Science and Engineering, IIT Kanpur (India)

2012-06-15

Highlights: Black-Right-Pointing-Pointer Thermomechanical processing of novel LAT 971 and LATZ 9531 Mg-Al-Li based alloys. Black-Right-Pointing-Pointer Microstructural deviation from the equilibrium phase diagram. Black-Right-Pointing-Pointer Disparity in texture of these alloys after hot-rolling (recrystallization and grain growth). Black-Right-Pointing-Pointer Role of alloying and phase distribution in affecting the texture/interplaner spacing. - Abstract: In the present study, the influence of alloying and thermomechanical processing on the microstructure and texture evolution on the two Mg-Li-Al based alloys, namely Mg-9 wt% Li-7 wt% Al-1 wt% Sn (LAT971) and Mg-9 wt% Li-5 wt% Al-3 wt% Sn-1 wt% Zn (LATZ9531) has been elicited. Novel Mg-Li-Al based alloys were cast (induction melting under protective atmosphere) followed by hot rolling at {approx}573 K with a cumulative reduction of five. A contrary dual phase dendritic microstructure rich in {alpha}-Mg, instead of {beta}-Li phase predicted by equilibrium phase diagram of Mg-Li binary alloy was observed. Preferential presence of Mg-Li-Sn primary precipitates (size 4-10 {mu}m) within {alpha}-Mg phase and Mg-Li-Al secondary precipitates (<3 {mu}m) interspersed in {beta}-Li indicated their degree of dissolution during hot-rolling and homogenization in the dual phase matrix. Presence of Al, Sn and Zn alloying elements in the Mg-Li based alloy has resulted an unusual dual-phase microstructure, change in the lattice parameter, and intriguing texture evolution after hot-rolling of cast LAT 971 and LATZ9531 alloy. Strong texture was absent in the as-cast samples whereas texture development after hot-rolling revealed an increased activity of the non-basal (101{sup Macron }0) slip planes. The quantification of the grain average misorientation (less than 2 Degree-Sign ) using electron backscattered diffraction confirmed the presence of strain free grains in majority of the grains (fraction >0.75) after hot-rolling of Mg-Li

Engineering design and development of lead lithium loop for thermo-fluid MHD studies

International Nuclear Information System (INIS)

Kumar, M.; Patel, Anita; Jaiswal, A.; Ranjan, A.; Mohanta, D.; Sahu, S.; Saraswat, A.; Rao, T.S.; Mehta, V.; Bhattacharyay, R.; Rajendra Kumar, E.

2017-01-01

In the frame of the design and development of LLCB TBM, number of R and D activities is in progress in the area of Pb-Li technology development. Molten Pb-Li is used as a tritium breeder and also as a coolant for the internals of the TBM structure. In presence of strong plasma confining toroidal magnetic field, motion of electrically conducting Pb-Li leads to Magneto Hydro Dynamic (MHD) phenomena, as a consequence of which the flow profile of Pb-Li is significantly modified inside the Pb-Li channels of TBM. This causes additional pressure drop inside TBM and affects the heat transfer from internal structure. The detail studies of these MHD effects are of prime importance for successful design of LLCB TBM and its performance evaluation. Although, various numerical MHD codes have been developed, validated in simple flow configuration and are being used to study MHD phenomena in LLCB TBM, experimental validation of these codes in TBM relevant complex flow geometry is yet to be performed. A Pb-Li MHD experimental loop is, therefore, being developed at IPR to perform thermo-fluid MHD experiments in various LLCB TBM relevant flow configuration. MHD experiments are planned with different test sections instrumented with potential pins, thermo couples, etc. under a uniform magnetic field of ∼1.4 T. The obtained experimental data will be analyzed to understand the MHD phenomena in TBM like flow configuration and also for validation of MHD codes. This paper describes the detailed process as well as engineering design of the Pb-Li MHD loop and its major components along with the plan of MHD experiments in various test mock ups. (author)

The heat capacity and entropy of the lithium silicides Li17Si4 and Li16.42Si4 in the temperature range from (2 to 873) K

International Nuclear Information System (INIS)

Thomas, Daniel; Zeilinger, Michael; Gruner, Daniel; Hüttl, Regina; Seidel, Jürgen; Wolter, Anja U.B.; Fässler, Thomas F.; Mertens, Florian

2015-01-01

Highlights: • High quality experimental heat capacities of the new lithium rich silicides Li 17 Si 4 and Li 16.42 Si 4 are reported. • Two different calorimeters have been used to cover the broad temperature range from (2 to 873) K. • Samples were prepared and characterized (XRD) by the original authors who firstly described these new silicide phases in 2013. • Supply of polynomial heat capacity functions for four temperature intervals. • Calculation of standard entropies and entropies of formation of the lithium silicides. - Abstract: This work presents the heat capacities and standard entropies of the recently described lithium rich silicide phases Li 17 Si 4 and Li 16.42 Si 4 as a function of temperature in the range from (2 to 873) K. The measurements were carried out using two different calorimeters. The heat capacities were determined in the range from T = (2 to 300) K by a relaxation technique using a Physical Properties Measurement System (PPMS) from Quantum Design, and in the range from T = (283 to 873) K by means of a Sensys DSC from Setaram applying the C p -by-step method. The experimental data are given with an accuracy of (1 to 2)% above T = 20 K and the error increases up to 7% below T = 20 K. The results of the measurements at low temperatures permit the calculation of additional thermodynamic parameters such as the standard entropy as well as the temperature coefficients of electronic and lattice contributions to the heat capacity. Additionally, differential scanning calorimetric (DSC) measurements were carried out to verify the phase transition temperatures of the studied lithium silicide phases. The results represent a significant contribution to the data basis for thermodynamic calculations (e.g. CALPHAD) and to the understanding of the phase equilibria in the (Li + Si) system, especially in the lithium rich region

Corrosion of ferrous alloys in eutectic lead-lithium environments

International Nuclear Information System (INIS)

Chopra, O.K.; Smith, D.L.

1983-09-01

Corrosion data have been obtained on austenitic prime candidate alloy (PCA) and Type 316 stainless steel and ferritic HT-9 and Fe-9Cr-1Mo steels in a flowing Pb-17 at. % Li environment at 727 and 700 K (454 and 427 0 C). The results indicate that the dissolution rates for both austenitic and ferritic steels in Pb-17Li are an order of magnitude greater than in flowing lithium. The influence of time, temperature, and alloy composition on the corrosion behavior in Pb-17Li is similar to that in lithium. The weight losses for the austenitic steels are an order of magnitude greater than for the ferritic steels. The rate of weight loss for the ferritic steels is constant, whereas the dissolution rates for the austenitic steels decrease with time. After exposure to Pb-17Li, the austenitic steels develop a very weak and porous ferrite layer which easily spalls from the specimen surface

Lead coolant test facility systems design, thermal hydraulic analysis and cost estimate

Energy Technology Data Exchange (ETDEWEB)

Khericha, Soli, E-mail: slk2@inel.gov [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Harvego, Edwin; Svoboda, John; Evans, Robert [Battelle Energy Alliance, LLC, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Dalling, Ryan [ExxonMobil Gas and Power Marketing, Houston, TX 77069 (United States)

2012-01-15

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed below: Bullet Develop and demonstrate feasibility of submerged heat exchanger. Bullet Develop and demonstrate open-lattice flow in electrically heated core. Bullet Develop and demonstrate chemistry control. Bullet Demonstrate safe operation. Bullet Provision for future testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimated. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 Degree-Sign C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.

Isospin character of low-lying pygmy dipole states in 208Pb via inelastic scattering of 17O ions.

Science.gov (United States)

Crespi, F C L; Bracco, A; Nicolini, R; Mengoni, D; Pellegri, L; Lanza, E G; Leoni, S; Maj, A; Kmiecik, M; Avigo, R; Benzoni, G; Blasi, N; Boiano, C; Bottoni, S; Brambilla, S; Camera, F; Ceruti, S; Giaz, A; Million, B; Morales, A I; Vandone, V; Wieland, O; Bednarczyk, P; Ciemała, M; Grebosz, J; Krzysiek, M; Mazurek, K; Zieblinski, M; Bazzacco, D; Bellato, M; Birkenbach, B; Bortolato, D; Calore, E; Cederwall, B; Charles, L; de Angelis, G; Désesquelles, P; Eberth, J; Farnea, E; Gadea, A; Görgen, A; Gottardo, A; Isocrate, R; Jolie, J; Jungclaus, A; Karkour, N; Korten, W; Menegazzo, R; Michelagnoli, C; Molini, P; Napoli, D R; Pullia, A; Recchia, F; Reiter, P; Rosso, D; Sahin, E; Salsac, M D; Siebeck, B; Siem, S; Simpson, J; Söderström, P-A; Stezowski, O; Theisen, Ch; Ur, C; Valiente-Dobón, J J

2014-07-04

The properties of pygmy dipole states in 208Pb were investigated using the 208Pb(17O, 17O'γ) reaction at 340 MeV and measuring the γ decay with high resolution with the AGATA demonstrator array. Cross sections and angular distributions of the emitted γ rays and of the scattered particles were measured. The results are compared with (γ, γ') and (p, p') data. The data analysis with the distorted wave Born approximation approach gives a good description of the elastic scattering and of the inelastic excitation of the 2+ and 3- states. For the dipole transitions a form factor obtained by folding a microscopically calculated transition density was used for the first time. This has allowed us to extract the isoscalar component of the 1- excited states from 4 to 8 MeV.

Development of nuclear transmutation technology - A study on the thermal-hydraulic characteristics of Pb-Bi coolant material

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Yang, Hui Chang; Huh, Byung Gil [Seoul National University, Seoul (Korea)

2000-03-01

The objective of this study is to provide the direction of HYPER design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of lead-bismuth material as a HYPER coolant and of proton accelerator target system. In this study, in order to evaluate the thermal-hydraulic characteristics of HYPER system, the FLUENT calculation is performed with liquid metal lead-bismuth(43%) and the turbulent Prandtl number model is developed. Also, the heat transfer analyses including temperature rising are performed for accelerator beam window, solid tungsten target and liquid target which is composed of liquid lead and lead-bismuth, respectively and the thermal stress analyses are performed for accelerator beam window. Through this study, the BASECASE whose parameter is HYPER system design specification is calculated by FLUENT. It is shown that the coolant velocity must exceeds 1.6 m/s for supporting the core coolant temperature in operating temperature range. The suggested turbulent Prandtl number model is applicable to liquid metal. And in order to maintain the integrity of proton beam target system, it is necessary to investigate the target structure associated with smoothing the flow path and beam window cooling. 43 refs., 67 figs., 27 tabs. (Author)

Design analysis of a lead–lithium/supercritical CO2 Printed Circuit Heat Exchanger for primary power recovery

International Nuclear Information System (INIS)

Fernández, Iván; Sedano, Luis

2013-01-01

Highlights: • A design for a PbLi/CO 2 (SC) Printed Circuit Heat Exchanger which optimizes the pressure drop performance is proposed. • Numerical analyses have been performed to optimize the airfoil fins shape and arrangement. • SiC is proposed as structural material and tritium permeation barrier for the PCHE. • The integrated flux is larger than expected and allows reducing the CO 2 mass flow in this sector of the power cycle. • A transport model has been developed to evaluate the permeation of tritium from the liquid metal to the secondary CO 2 . -- Abstract: One of the key issues for fusion power plant technology is the efficient, reliable and safe recovery of the power extracted by the primary coolants. An interesting design option for power conversion cycles based on Dual Coolant Breeding Blankets (DCBB) is a Printed Circuit Heat Exchanger, which is supported by the advantages of its compactness, thermal effectiveness, high temperature and pressure capability and corrosion resistance. This work presents a design analysis of a silicon carbide Printed Circuit Heat Exchanger for lead–lithium/supercritical CO 2 at DEMO ranges (4× segmentation)

In-situ electrochemical study of Zr1nb alloy corrosion in high temperature Li{sup +} containing water

Energy Technology Data Exchange (ETDEWEB)

Krausová, Aneta [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Macák, Jan, E-mail: macakj@vscht.cz [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Sajdl, Petr [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Novotný, Radek [JRC-IET, Westerduinveg 3, 1755 LE Petten (Netherlands); Renčiuková, Veronika [University of Chemistry and Technology, Technická 3, 166 28 Prague 6 (Czech Republic); Vrtílková, Věra [ÚJP a.s., Nad Kamínkou 1345, 156 10 Prague 5 (Czech Republic)

2015-12-15

Long-term in-situ corrosion tests were performed in order to evaluate the influence of lithium ions on the corrosion of zirconium alloy. Experiments were carried out in a high-pressure high-temperature loop (280 °C, 8 MPa) in a high concentration water solution of LiOH (70 and 200 ppm Li{sup +}) and in a simulated WWER primary coolant environment. The kinetic parameters characterising the oxidation process have been explored using in-situ electrochemical impedance spectroscopy and slow potentiodynamic polarization. Also, a suitable equivalent circuit was suggested, which would approximate the impedance characteristics of the corrosion of Zr–1Nb alloy. The Mott–Schottky approach was used to determine the semiconducting character of the passive film. - Highlights: • Zr1Nb alloy was tested in WWER coolant and in LiOH solutions at 280 °C. • Corrosion rates were estimated in-situ from electrochemical data. • Electrochemical data agreed well with weight gains and metallography data. • Increase of corrosion rate in LiOH appeared after short exposure (300–500 h). • Very high donor densities (1.1–1.2 × 10{sup 20} cm{sup −3}) of Zr oxide grown in LiOH were found.

Investigation of tritium inventory and permeation behaviour in the liquid breeder blanket concept of Demo as a function of design and material parameters

International Nuclear Information System (INIS)

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

1991-01-01

A numerical code has been used to estimate the time dependence of tritium inventory and of tritium transport into the coolant, into the first wall boxes and through the liquid breeder in the Pb-17Li blanket concept of DEMO. Several issues in both design and material parameters have been considered and the effect on inventory and permeation of coatings with low surface recombination coefficient and/or low diffusivity at various surfaces of the structural material has been studied. TiC has been chosen as reference material for these calculations and a general database on coating efficiency as a function of its properties has also been produced on the basis of TiC data

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

Energy dependence of particle ratio fluctuations in central Pb+Pb collisions from $\\sqrt{s_{_{NN}}} =$~6.3 to 17.3 GeV

CERN Document Server

Alt, C; Baatar, B; Barna, D; Bartke, J; Betev, L; Bialkowska, H; Blume, C; Boimska, B; Botje, M; Bracinik, J; Bramm, R; Bunccic, P; Cerny, V; Christakoglou, P; Chvala, O; Cramer, J G; Csató, P; Dinkelaker, P; Eckardt, V; Flierl, D; Fodor, Z; Foka, P; Friese, V; Gál, J; Gazdzicki, M; Genchev, V; Georgopoulos, G; Gladysz, E; Grebieszkow, K; Hegyi, S; Höhne, C; Kadija, K; Karev, A; Kliemant, M; Kniege, S; Kolesnikov, V I; Kornas, E; Korus, R; Kowalski, M; Kraus, I; Kreps, M; Kresan, D; Van Leeuwen, M; Lévai, P; Litov, L; Lungwitz, B; Makariev, M; Malakhov, A I; Mateev, M; Melkumov, G L; Mischke, A; Mitrovski, M; Molnár, J; Mrówczynski, St; Nicolic, V; Pálla, G; Panagiotou, A D; Panayotov, D; Petridis, A; Pikna, M; Prindle, D; Pühlhofer, F; Renfordt, R; Roland, C; Roland, G; Rybczynski, M; Rybicki, A; Sandoval, A; Schmitz, N; Schuster, T; Seyboth, P; Siklér, F; Sitár, B; Skrzypczak, E; Stefanek, G; Stock, R; Ströbele, H; Susa, T; Szentpétery, I; Sziklai, J; Szymanski, P; Trubnikov, V; Varga, D; Vassiliou, M; Veres, G I; Vesztergombi, G; Vranic, D; Wetzler, A; Wlodarczyk, Z; Yoo, I K

2009-01-01

We present recent measurements of the energy dependence of event-by-event fluctuations in the K/pi and (p + \\bar{p})/pi multiplicity ratios in heavy ion collisions at the CERN SPS. The particle ratio fluctuations were obtained for central Pb+Pb collisions at five collision energies, \\sqrt{s_{_{NN}}}, between 6.3 and 17.3 GeV. After accounting for the effects of finite-number statistics and detector resolution, we extract the strength of non-statistical fluctuations at each energy. For the K/pi ratio, larger fluctuations than expected for independent particle production are found at all collision energies. The fluctuations in the (p + \\bar{p})/pi ratio are smaller than expectations from independent particle production, indicating correlated pion and proton production from resonance decays. For both ratios, the deviation from purely statistical fluctuations shows an increase towards lower collision energies. The results are compared to transport model calculations, which fail to describe the energy dependence o...

The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

International Nuclear Information System (INIS)

Urquidi-Macdonald, Mirna; Pitt, Jonathan; Macdonald, Digby D.

2007-01-01

A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH(T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields

Compatibility of 316L stainless steel with tritium breeders for fusion reactors

International Nuclear Information System (INIS)

Broc, M.; Fauvet, P.; Flament, T.; Sannier, J.

1986-06-01

Compatibility problems with structural materials are a concern for the choice of the tritium breeder for fusion reactors. In the frame of the European Programme on Fusion Technology, two types of blankets are considered: liquid (eutectic lithium-lead alloy at 0.68 wt % Li: 17Li83Pb) and solid (lithium aluminate or silicate) breeders. This paper is devoted to compatibility studies of 316L stainless steel with 17Li83Pb alloy and γ-LiA10 2 ceramic

Compartmentalized safety coolant injection system

International Nuclear Information System (INIS)

Johnson, F.T.

1983-01-01

A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

‘Green’-synthesized near-infrared PbS quantum dots with silica-PEG dual-layer coating: ultrastable and biocompatible optical probes for in vivo animal imaging

Science.gov (United States)

Wang, D.; Qian, J.; Cai, F.; He, S.; Han, S.; Mu, Y.

2012-06-01

In this paper, PbS semiconductor quantum dots (QDs) with near-infrared (NIR) photoluminescence were synthesized in oleic acid and paraffin liquid mixture by using an easily handled and ‘green’ approach. Surface functionalization of the QDs was accomplished with a silica and polyethylene glycol (PEG) phospholipid dual-layer coating and the excellent chemical stability of the nanoparticles is demonstrated. We then successfully applied the ultrastable PbS QDs to in vivo sentinel lymph node (SLN) mapping of mice. Histological analyses were also carried out to ensure that the intravenously injected nanoparticles did not produce any toxicity to the organism of mice. These experimental results suggested that our ultrastable NIR PbS QDs can serve as biocompatible and efficient probes for in vivo optical bioimaging and has great potentials for disease diagnosis and clinical therapies in the future.

‘Green’-synthesized near-infrared PbS quantum dots with silica–PEG dual-layer coating: ultrastable and biocompatible optical probes for in vivo animal imaging

International Nuclear Information System (INIS)

Wang, D; Qian, J; Cai, F; He, S; Han, S; Mu, Y

2012-01-01

In this paper, PbS semiconductor quantum dots (QDs) with near-infrared (NIR) photoluminescence were synthesized in oleic acid and paraffin liquid mixture by using an easily handled and ‘green’ approach. Surface functionalization of the QDs was accomplished with a silica and polyethylene glycol (PEG) phospholipid dual-layer coating and the excellent chemical stability of the nanoparticles is demonstrated. We then successfully applied the ultrastable PbS QDs to in vivo sentinel lymph node (SLN) mapping of mice. Histological analyses were also carried out to ensure that the intravenously injected nanoparticles did not produce any toxicity to the organism of mice. These experimental results suggested that our ultrastable NIR PbS QDs can serve as biocompatible and efficient probes for in vivo optical bioimaging and has great potentials for disease diagnosis and clinical therapies in the future. (paper)

Research on coolant radiochemistry

International Nuclear Information System (INIS)

Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

2004-04-01

The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

Central as well as peripheral attentional bottlenecks in dual-task performance activate lateral prefrontal cortices

Directory of Open Access Journals (Sweden)

Andre J Szameitat

2016-03-01

Full Text Available Human information processing suffers from severe limitations in parallel processing. In particular, when required to respond to two stimuli in rapid succession, processing bottlenecks may appear at central and peripheral stages of task processing. Importantly, it has been suggested that executive functions are needed to resolve the interference arising at such bottlenecks. The aims of the present study were to test whether central attentional limitations (i.e., bottleneck at the decisional response selection stage as well as peripheral limitations (i.e., bottleneck at response initiation both demand executive functions located in the lateral prefrontal cortex. For this, we re-analysed two previous studies, in which a total of 33 participants performed a dual-task according to the paradigm of the psychological refractory period (PRP during fMRI. In one study (N=17, the PRP task consisted of two two-choice response tasks known to suffer from a central bottleneck (CB group. In the other study (N=16, the PRP task consisted of two simple-response tasks known to suffer from a peripheral bottleneck (PB group. Both groups showed considerable dual-task costs in form of slowing of the second response in the dual-task (PRP effect. Imaging results are based on the subtraction of both single-tasks from the dual-task within each group. In the CB group, the bilateral middle frontal gyri and inferior frontal gyri were activated. Higher activation in these areas was associated with lower dual-task costs. In the PB group, the right middle frontal and inferior frontal gyrus were activated. Here, higher activation was associated with higher dual-task costs. In conclusion we suggest that central and peripheral bottlenecks both demand executive functions located in lateral prefrontal cortices. Differences between the CB and PB groups with respect to the exact prefrontal areas activated and the correlational patterns suggest that the executive functions resolving

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

17 CFR 41.27 - Prohibition of dual trading in security futures products by floor brokers.

Science.gov (United States)

2010-04-01

... trading in a security futures product on a designated contract market or registered derivatives...) Registered derivatives transaction execution facilities. Prior to listing a security futures product for... 17 Commodity and Securities Exchanges 1 2010-04-01 2010-04-01 false Prohibition of dual trading in...

A new carbon additive compounded Li3V1.97Zn0.05(PO4)3/C cathode for plug-in hybrid electric vehicles

International Nuclear Information System (INIS)

Wang, Wenhui; Zhang, Jiaolong; Lin, Yue; Ding, Fei; Chen, Zhenyu; Dai, Changsong

2015-01-01

The application of lithium ion batteries in plug-in hybrid electric vehicles (PHEVs) requires safety, high energy density, high power density, excellent cyclability and good low temperature performance. On the basis of thermally stable Li 3 V 2 (PO 4 ) 3 /C and cost-effective performance carbon additives, we designed a Li 3 V 1.97 Zn 0.05 (PO 4 ) 3 /(C+10PB) (PB stands for performance carbon additives PBX101) cathode that meets the above requirements for PHEVs battery. Firstly, its Ragone plot presents an excellent energy density retention at high power rates; secondly, the excellent capacity retention and high Coulombic efficiency of Li 3 V 1.97 Zn 0.05 (PO 4 ) 3 /(C+10PB)-Li half-cell clearly indicates a potential good cyclability of full cells based on Li 3 V 1.97 Zn 0.05 (PO 4 ) 3 /(C+10PB) cathode. Finally, we believe the good low temperature performance of Li 3 V 1.97 Zn 0.05 (PO 4 ) 3 /(C+10PB) (i.e. retains 91.6% and 76.3% of its capacity at ∼25 °C, when cycled at 0 and -15 °C) is also beneficial to its application in PHEVs

Technology for cleaning of Pb-Bi adhering to steel (1). Basic tests

International Nuclear Information System (INIS)

Saito, Shigeru; Sasa, Toshinobu; Umeno, Makoto; Kurata, Yuji; Kikuchi, Kenji; Futakawa, Masatoshi

2004-12-01

The accelerator driven system (ADS) is proposed to transmute minor actinides (MA) in high-level waste from spent fuels of nuclear power reactors. Liquid Pb-Bi alloy is a candidate material for spallation target and coolant of ADS. Pb-Bi cleaning technology is required to reduce radiation exposure during maintenance service and to decontaminate replaced components. In this study, three cleaning methods were tested; silicon oil cleaning at 170degC, mixture of acetic acid and nitric acid cleaning. Specimens were prepared by immersion in melted Pb-Bi. After silicon oil tests, most of Pb-Bi remained on the surface of the specimens. It was found that blushing was needed to remove Pb-Bi effectively. On the other hands, Pb-Bi was easily dissolved and almost removed in the mixed acid and nitric acid. Silicon oil cleaning did not affect on base metals. The surface of base metals was slightly blacked after mixed acid cleaning. F82H base metals were corroded by nitric acid. (author)

Coolant system decontamination

International Nuclear Information System (INIS)

Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

1981-01-01

An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

11Li structural information from inclusive break-up measurements

Directory of Open Access Journals (Sweden)

Fernández-García J. P.

2015-01-01

Full Text Available Structure information of 11Li halo nucleus has been obtained from the inclusive break-up measurements of the 11Li+208Pb reactions at energies around the Coulomb barrier (Elab = 24.3 and 29.8 MeV. The effective break-up energy and the slope of B(E1 distribution close to the threshold have been extracted from the experimental data.

Occupational Radiation Exposure Analysis of US ITER DCLL TBM

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad J; Cadwallader, Lee C; Dagher, Mohamad

2007-08-01

This report documents an Occupational Radiation Exposure (ORE) analysis that was performed for the US International Thermonuclear Experimental Reactor (ITER) Dual Coolant Lead Lithium (DCLL) Test Blanket Module (TBM). This analysis was performed with the QADMOD dose code for anticipated maintenance activities for this TBM concept and its ancillary systems. The QADMOD code was used to model the PbLi cooling loop of this TBM concept by specifying gamma ray source terms that simulated radioactive material within the piping, valves, heat exchanger, permeator, pump, drain tank, and cold trap of this cooling system. Estimates of the maintenance tasks that will have to be performed and the time required to perform these tasks where developed based on either expert opinion or on industrial maintenance experience for similar technologies. This report details the modeling activity and the calculated doses for the maintenance activities envisioned for the US DCLL TBM.

Dual-chamber/dual-anode proportional counter incorporating an intervening thin-foil solid neutron converter

International Nuclear Information System (INIS)

Boatner, Lynn A.; Neal, John S.; Blackston, Matthew A.; Kolopus, James A.; Ramey, Joanne O.

2012-01-01

A dual-chamber/dual-anode gas proportional counter utilizing thin solid 6 LiF or 10 B neutron converters coated on a 2-micon-thick Mylar film that is positioned between the two counter chambers and anodes has been designed, fabricated, and tested using a variety of fill gases—including naturally abundant helium. In this device, neutron conversion products emitted from both sides of the coated converter foil are detected—rather than having half of the products absorbed in the wall of a conventional tube-type counter where the solid neutron converter is deposited on the tube wall. Geant4-based radiation transport calculations were used to determine the optimum neutron converter coating thickness for both isotopes. Solution methods for applying these optimized-thickness coatings on a Mylar film were developed that were carried out at room temperature without any specialized equipment and that can be adapted to standard coating methods such as silk screen or ink jet printing. The performance characteristics of the dual-chamber/dual-anode neutron detector were determined for both types of isotopically enriched converters. The experimental performance of the 6 LiF-converter-based detector was described well by modeling results from Geant4. Additional modeling studies of multiple-foil/multiple-chamber/anode configurations addressed the basic issue of the relatively longer absorption range of neutrons versus the shorter range of the conversion products for 6 LiF and 10 B. Combined with the experimental results, these simulations indicate that a high-performance neutron detector can be realized in a single device through the application of these multiple-foil/solid converter, multiple-chamber detector concepts.

Nuclear reactor coolant and cover gas system

International Nuclear Information System (INIS)

George, J.A.; Redding, A.H.; Tower, S.N.

1976-01-01

A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

Coolant clean-up and recycle systems

International Nuclear Information System (INIS)

Ito, Takao.

1979-01-01

Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

Self-cooled liquid-metal blanket concept

International Nuclear Information System (INIS)

Malang, S.; Arheidt, K.; Barleon, L.

1988-01-01

A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

Energy dependence of kaon-to-proton ratio fluctuations in central Pb+Pb collisions from $\\sqrt{s_{NN}}$ = 6.3 to 17.3 GeV

CERN Document Server

Anticic, T.; Barna, D.; Bartke, J.; Beck, H.; Betev, L.; Bialkowska, H.; Blume, C.; Bogusz, M.; Boimska, B.; Book, J.; Botje, M.; Buncic, P.; Cetner, T.; Christakoglou, P.; Chung, P.; Chvala, O.; Cramer, J.G.; Eckardt, V.; Fodor, Z.; Foka, P.; Friese, V.; Gazdzicki, M.; Grebieszkow, K.; Hohne, C.; Kadija, K.; Karev, A.; Kolesnikov, V.I.; Kollegger, T.; Kowalski, M.; Kresan, D.; Laszlo, A.; Lacey, R.; van Leeuwen, M.; Mackowiak, M.; Makariev, M.; Malakhov, A.I.; Mateev, M.; Melkumov, G.L.; Mitrovski, M.; Mrowczynski, St.; Nicolic, V.; Palla, G.; Panagiotou, A.D.; Peryt, W.; Pluta, J.; Prindle, D.; Puhlhofer, F.; Renfordt, R.; Roland, C.; Roland, G.; Rybczynski, M.; Rybicki, A.; Sandoval, A.; Schmitz, N.; Schuster, T.; Seyboth, P.; Sikler, F.; Skrzypczak, E.; Slodkowski, M.; Stefanek, G.; Stock, R.; Strobele, H.; Susa, T.; Szuba, M.; Utvic, M.; Varga, D.; Vassiliou, M.; Veres, G.I.; Vesztergombi, G.; Vranic, D.; Wlodarczyk, Z.; Wojtaszek-Szwarc, A.

2011-01-01

Kaons and protons carry large parts of two conserved quantities, strangeness and baryon number. It is argued that their correlation and thus also fluctuations are sensitive to conditions prevailing at the anticipated parton-hadron phase boundary. Fluctuations of the $(\\mathrm{K}^+ + \\mathrm{K}^-)/(\\mathrm{p}+\\bar{\\mathrm{p}})$ and $\\mathrm{K}^+/\\mathrm{p}$ ratios have been measured for the first time by NA49 in central Pb+Pb collisions at 5 SPS energies between $\\sqrt{s_{NN}}$= 6.3~GeV and 17.3~GeV. Both ratios exhibit a change of sign in $\\sigma_{\\mathrm{dyn}}$, a measure of non-statistical fluctuations, around $\\sqrt{s_{NN}}$ = 8~GeV. Below this energy, $\\sigma_{\\mathrm{dyn}}$ is positive, indicating higher fluctuation compared to a mixed event background sample, while for higher energies, $\\sigma_{\\mathrm{dyn}}$ is negative, indicating correlated emission of kaons and protons. The results are compared to UrQMD calculations which which give a good description at the higher SPS energies, but fail to reproduc...

Heat-pulse flowmeter for a liquid breeder blanket

International Nuclear Information System (INIS)

Kondo, Masatoshi; Shibano, Kyohei; Tanaka, Teruya; Muroga, Takeo

2013-01-01

Liquid metals Li, Pb-17Li and Sn-20Li are candidate liquid breeders in fusion reactors. The development of a flowmeter that can be applied to high-temperature liquid metals is an important issue. A heat-pulse flowmeter is proposed in the present study. Its basic performance was investigated by means of a loop experiment with Pb-17Li and a numerical simulation. The temperature distribution in flowing Pb-17Li was obtained by local transient heating of the outer surface of a loop tube. The temperature distribution gradually changed and resembled the movement of a hot spot, which had a higher temperature than its surroundings. This hot spot moved along the flow and passed through the tips of the thermocouples. The change in temperature distribution with the movement of the hot spot was monitored by three thermocouples exposed to the Pb-17Li flow. The results of the loop experiments were numerically simulated by assuming a certain flow rate, and the temperature profile obtained in the loop experiment was in agreement with the simulation results. The time taken by the hot spot to pass through the tips of the thermocouples was measured and simulated, and the correlation between this time and the average flow velocity was evaluated. The results indicated the average flow velocity can be obtained using the heat-pulse flowmeter proposed in this study. (author)

Coolant leakage detecting device

International Nuclear Information System (INIS)

Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

1995-01-01

The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

Measurement of the mass attenuation coefficients and electron densities for BiPbSrCaCuO superconductor at different energies

Science.gov (United States)

Çevik, U.; Baltaş, H.

2007-03-01

The mass attenuation coefficients for Bi, Pb, Sr, Ca, Cu metals, Bi2O3, PbO, SrCO3, CaO, CuO compounds and solid-state forms of Bi1.7Pb0.3Sr2Ca2Cu3O10 superconductor were determined at 57.5, 65.2, 77.1, 87.3, 94.6, 122 and 136 keV energies. The samples were irradiated using a 57Co point source emitted 122 and 136 keV γ-ray energies. The X-ray energies were obtained using secondary targets such as Ta, Bi2O3 and (CH3COO)2UO22H2O. The γ- and X-rays were counted by a Si(Li) detector with a resolution of 0.16 keV at 5.9 keV. The effect of absorption edges on electron density, effective atomic numbers and their variation with photon energy in composite superconductor samples was discussed. Obtained values were compared with theoretical values.

Coolant inlet device for nuclear reactors

International Nuclear Information System (INIS)

Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

1969-01-01

Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

Design and fabrication of magnetic coolant filter

Science.gov (United States)

Prashanth, B. N.

2017-07-01

Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.E.; Cheng, E.T.

1985-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li/sub 17/Pb/sub 83/ and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the TBR to group structure and weighting spectrum increases and Li enrichment decrease with up to 20% discrepancies for thin natural Li/sub 17/Pb/sub 83/ blankets

Numerical simulation of magnetohydrodynamic (MHD) flow with internal heat generation

International Nuclear Information System (INIS)

Bokade, Vipin; Bhandarkar, U.V.; Bodi, Kowsik

2016-01-01

A strong magnetic field is used to confine the plasma in a fusion reactor. This magnetic field also affects the flow of Lead-Lithium (breeder/coolant) in the breeding blanket. So it is important to study MHD flow of Lead-Lithium (Pb-Li). Open-source toolbox, OpenFOAM, is used to study single phase behaviour of Pb-Li. As the induced magnetic field is very small, Ni et al. electric potential algorithm is employed in OpenFOAM and validated with analytical results. This solver can also solve the temperature field with heat source term. Simulations are carried out in 2D straight channel for various values of Hartmann Number ranging from 100 to 5000 and velocity profile, temperature, current density and pressure drop are studied. (author)

Organic coolant for ARIES-III

International Nuclear Information System (INIS)

Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

1991-04-01

ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

Excitation energy transfer to luminescence centers in M{sup II}MoO{sub 4} (M{sup II}=Ca, Sr, Zn, Pb) and Li{sub 2}MoO{sub 4}

Energy Technology Data Exchange (ETDEWEB)

Spassky, D.A., E-mail: deris2002@mail.ru [Skobeltsyn Institute of Nuclear Physics, M.V. Lomonosov Moscow State University, Leninskie Gory 1, bld.2, 119991 Moscow (Russian Federation); National University of Science and Technology (MISiS), Leninsky Prospekt 4, 119049 Moscow (Russian Federation); Kozlova, N.S. [National University of Science and Technology (MISiS), Leninsky Prospekt 4, 119049 Moscow (Russian Federation); Nagirnyi, V. [Institute of Physics, University of Tartu, W. Ostwaldi 1, 50411 Tartu (Estonia); Savon, A.E. [Skobeltsyn Institute of Nuclear Physics, M.V. Lomonosov Moscow State University, Leninskie Gory 1, bld.2, 119991 Moscow (Russian Federation); Hizhnyi, Yu.A.; Nedilko, S.G. [Taras Shevchenko National University of Kyiv, Volodymyrska str. 64/13, 01601 Kyiv (Ukraine)

2017-06-15

Based on the results of spectroscopy studies and electronic band structure calculations, the analysis of excitation energy transformation into luminescence is performed for a set of molybdates M{sup II}MoO{sub 4} (M{sup II}=Ca, Sr, Zn, Pb) and Li{sub 2}MoO{sub 4}. The bandgap energies were determined from comparison of experimental and calculated reflectivity spectra as 3.3 eV for PbMoO{sub 4}, 4.3 eV for ZnMoO{sub 4}, 4.4 eV for CaMoO{sub 4}, 4.7 eV for SrMoO{sub 4}, and 4.9 eV for Li{sub 2}MoO{sub 4}. It is shown that photoluminescence excitation spectra of these materials reveal the specific features of their conduction bands. The threshold of separated charge carriers’ creation is shown to be by 1.3–1.9 eV higher than the bandgap energy in CaMoO{sub 4}, SrMoO{sub 4} and ZnMoO{sub 4}. The effect is explained by the peculiarities of conduction band structure, namely to the presence of gap between the subbands of the conduction band and to the low mobility of electrons in the lower sub-band of the conduction band.

Blanket comparison and selection study. Final report. Volume 1

International Nuclear Information System (INIS)

1984-09-01

The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li 2 O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N 2 ) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

Blanket comparison and selection study. Final report. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1984-09-01

The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li/sub 2/O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N/sub 2/) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li/sub 2/O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue.

Blanket comparison and selection study. Final report. Volume 3

International Nuclear Information System (INIS)

1984-09-01

The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li 2 O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N 2 ) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concept are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concepts are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

Blanket comparison and selection study. Final report. Volume 2

International Nuclear Information System (INIS)

1984-09-01

The study focused on: (1) Development of reference design guidelines, evaluation criteria, and a methodology for evaluating and ranking candidate blanket concepts. (2) Compilation of the required data base and development of a uniform systems analysis for comparison. (3) Development of conceptual designs for the comparative evaluation. (4) Evaluation of leading concepts for engineering feasibility, economic performance, and safety. (5) Identification and prioritization of R and D requirements for the leading blanket concepts. Sixteen concepts (nine TMR and seven tokamak) which were identified as leading candidates in the early phases of the study, were evaluated in detail. The overall evaluation concluded that the following concepts should provide the focus for the blanket R and D program: (Breeder/Coolant/Structure), Lithium/Lithium/Vanadium Alloy, Li 2 O/Helium/Ferritic Steel, LiPb Alloy/LiPb Alloy/Vanadium Alloy, and Lithium/Helium/Ferritic Steel. The primary R and D issues for the Li/Li/V concept are the development of an advanced structural alloy, resolution of MHD and corrosion problems, provision for an inert atmosphere (e.g., N 2 ) in the reactor building, and the development of non-water cooled near-plasma components, particularly for the tokamak. The main issues for the LiPb/LiPb/V concepts are similar to the Li/Li/V blanket with the addition of resolving the tritium recovery issue. The R and D issues for Li 2 O/He/FS concept include resolution of the tritium recovery/containment issue, achieving adequate tritium breeding and resolving other solid breeder issues such as swelling and fabrication concerns. Major concerns for the Li/He/FS concept are related to its rather poor economic performance. Improvement of its economic performance will be somewhat concept-dependent and will be more of a systems engineering issue

Fusion reactor blanket-main design aspects

International Nuclear Information System (INIS)

Strebkov, Yu.; Sidorov, A.; Danilov, I.

1994-01-01

The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

The sodium coolant

International Nuclear Information System (INIS)

Rodriguez, G.

2004-01-01

The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

Study of core characteristics on fuel and coolant type. Results of F/S phase-I

International Nuclear Information System (INIS)

Ikegami, Tetsuo; Hayashi, Hideyuki; Sasaki, Makoto; Mizuno, Tomoyasu; Yamadate, Megumi; Takaki, Naoyuki; Kurosawa, Norifumi; Sakashita, Yoshiaki; Naganuma, Masayuki

2001-03-01

The phase-I of the Feasibility Study of Commercialized Fast Reactor Cycle Systems (F/S) were started from July, 1999 and terminated at the end of FY2000 in order to executed examination about technology alternatives of various commercialized fast reactor (FR) recycle concepts, in response to the JNC middle long term enterprise plan. In the phase-I of this F/S, a number of conceptual candidates have been selected from the following 5 viewpoints: a) ensuring safety, b) economic competitiveness to future LWRs, c) efficient utilization of resources, d) reduction of environmental burden, e) enhancement of nuclear non-proliferation. As for this study from the above viewpoints, core characteristics of many kinds of reactors have been investigated, analyzed and examined a core / a fuel characteristic in the combinations of fuel and coolant types and power output scales. Based on these results, R and D plans of the phase-II to be performed have been proposed, and a database to select candidate reactor concepts has been prepared. The conclusions have been obtained in the phase-I are as follows: (1) Evaluation of a fuel form in every each coolant was compared. A promising fuel form was extracted as follows: an oxide and a metal fuel for sodium coolant cores, a metal and a nitride fuel for heavy metal coolant cores, an oxide and a nitride fuel for carbon dioxide coolant cores and a nitride fuel for He gas coolant cores. (2) As the general idea that performance of a core nucleus can be compatible with re-criticality evasion in sodium coolant large-sized oxide fuel cores, a axial blanket particle elimination radial heterogeneous core is one influential candidate. (3) In case of Pb-Bi coolant nature circulation medium size core with an oxide fuel, it is difficult to simultaneously achieve higher discharged burn-up and higher breeding ratio according to the viewpoints of the phase-I. (4) Core characteristics of a carbon dioxide coolant core shows to be almost equivalent to that of

Out-of-pile chemical compatibility of Pb-Bi eutectic alloy with graphite

Energy Technology Data Exchange (ETDEWEB)

Sengupta, A.K.; Bhagat, R.K.; Jarvis, T.; Majumdar, S. [Radiometallurgy Div., Bhabha Atomic Research Centre, Mumbai (India); Laik, A.; Kale, G.B. [Material Science Div., Bhabha Atomic Research Centre, Mumbai (India); Kamath, H.S. [Nuclear Fuels Group, Bhabha Atomic Research Centre, Mumbai (India)

2006-06-15

Lead Bismuth eutectic alloy (Pb: 55.5 wt.%, Bi: 44.5 wt.%) is a potential candidate coolant material for high-temperature reactors because of its low melting point (124 C), high thermal conductivity, heat capacity, and better neutronic properties. Out-of-pile chemical compatibility studies of this coolant with graphite (coolant channel) have been carried out by isothermal annealing of the liquid alloy in a graphite crucible at 800, 900, 1000, and 1100 C for times ranging from 100 h to 1000 h. Formation of a reaction layer is observed. The growth rate of the reaction layer follows a parabolic law. Reaction layer thicknesses of 61.3 {mu}m and 121 {mu}m are estimated from the growth rate vs. time relation after 1 year and 5 years respectively. The growth of the reaction layer is diffusion-controlled and the activation energy of the reaction is estimated to be 100 KJ/mol. (orig.)

Out-of-pile chemical compatibility of Pb-Bi eutectic alloy with graphite

International Nuclear Information System (INIS)

Sengupta, A.K.; Bhagat, R.K.; Jarvis, T.; Majumdar, S.; Laik, A.; Kale, G.B.; Kamath, H.S.

2006-01-01

Lead Bismuth eutectic alloy (Pb: 55.5 wt.%, Bi: 44.5 wt.%) is a potential candidate coolant material for high-temperature reactors because of its low melting point (124 C), high thermal conductivity, heat capacity, and better neutronic properties. Out-of-pile chemical compatibility studies of this coolant with graphite (coolant channel) have been carried out by isothermal annealing of the liquid alloy in a graphite crucible at 800, 900, 1000, and 1100 C for times ranging from 100 h to 1000 h. Formation of a reaction layer is observed. The growth rate of the reaction layer follows a parabolic law. Reaction layer thicknesses of 61.3 μm and 121 μm are estimated from the growth rate vs. time relation after 1 year and 5 years respectively. The growth of the reaction layer is diffusion-controlled and the activation energy of the reaction is estimated to be 100 KJ/mol. (orig.)

Benchmark calculations for fusion blanket development

International Nuclear Information System (INIS)

Sawan, M.L.; Cheng, E.T.

1986-01-01

Benchmark problems representing the leading fusion blanket concepts are presented. Benchmark calculations for self-cooled Li 17 Pb 83 and helium-cooled blankets were performed. Multigroup data libraries generated from ENDF/B-IV and V files using the NJOY and AMPX processing codes with different weighting functions were used. The sensitivity of the tritium breeding ratio to group structure and weighting spectrum increases as the thickness and Li enrichment decrease with up to 20% discrepancies for thin natural Li 17 Pb 83 blankets. (author)

A Li-particulate blanket concept for ITER

International Nuclear Information System (INIS)

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

1989-01-01

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

A Strangelet and Particle Search in Pb-Pb Collisions

CERN Multimedia

Lohmann, K-D; Linden, T

2002-01-01

%NA52 %title\\\\ \\\\The NA52 experiment aims to detect strangelets, \\textit{i.e.} small drops of strange quark matter, which might result from the extreme energy and baryon densities attained in Pb+Pb collisions at a beam momentum of 158~A GeV/c. The experiment uses the H6 beam line as a spectrometer equipped with wire chambers, time of flight measurements over a path of 524~m and a hadronic calorimeter which is placed at the end of the setup.\\\\ \\\\During the 17 day run in fall of 1994 we accumulated data of 1.8~\\cdot~10$^{12}$~Pb ions on our Pb targets. The average beam intensity was 2~\\cdot~10$^{7}$~ions per spill for the NA52 experiment. We were running mainly with a 40~mm target at spectrometer rigidities of $\\pm$100 and $-$200~GeV/c and with a 16~mm target at $+$200~GeV/c. Per setting 10$^{11}$ Pb+Pb collisions were recorded. During the Pb-ion run in 1995 the statistics for the strangelet search at a rigidity of $-$200~GeV/c has been improved by about one order of magnitude. This was mainly due to a factor o...

Minimum thickness blanket-shield for fusion reactors

International Nuclear Information System (INIS)

Karni, Y.; Greenspan, E.

1989-01-01

A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

Coolant monitoring systems for PWR reactors

International Nuclear Information System (INIS)

Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

1987-01-01

The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

Thermohydraulics design and thermomechanics analysis of two European breeder blanket concepts for DEMO. Pt. 1 and Pt. 2. Pt. 1: BOT helium cooled solid breeding blanket. Pt. 2: Dual coolant self-cooled liquid metal blanket

International Nuclear Information System (INIS)

Norajitra, P.

1995-06-01

Two different breeding blanket concepts are being elaborated at Forschungszentrum Karlsruhe within the framework of the DEMO breeding blanket development, the concept of a helium cooled solid breeding blanket and the concept of a self-cooled liquid metal blanket. The breeder material used in the first concept is Li 4 SiO 4 as a pebble bed arranged separate from the beryllium pebble bed, which serves as multiplier. The breeder material zone is cooled by several toroidally-radially configurated helium cooling plates which, at the same time, act as reinforcements of the blanket structures. In the liquid metal blanket concept lead-lithium is used both as the breeder material and the coolant. It flows at low velocity in poloidal direction downwards and back in the blanket front zone. In both concepts the First Wall is cooled by helium gas. This report deals with the thermohydraulics design and thermomechanics analysis of the two blanket concepts. The performance data derived from the Monte-Carlo computations serve as a basis for the design calculations. The coolant inlet and outlet temperatures are chosen with the design criteria and the economics aspects taken into account. Uniform temperature distribution in the blanket structures can be achieved by suitable branching and routing of the coolant flows which contributes to reducing decisively the thermal stress. The computations were made using the ABAQUS computer code. The results obtained of the stresses have been evaluated using the ASME code. It can be demonstrated that all maximum values of temperature and stress are below the admissible limit. (orig.) [de

Coupled electrochemical thermal modelling of a novel Li-ion battery pack thermal management system

International Nuclear Information System (INIS)

Basu, Suman; Hariharan, Krishnan S.; Kolake, Subramanya Mayya; Song, Taewon; Sohn, Dong Kee; Yeo, Taejung

2016-01-01

Highlights: • Three-dimensional electrochemical thermal model of Li-ion battery pack using computational fluid dynamics (CFD). • Novel pack design for compact liquid cooling based thermal management system. • Simple temperature estimation algorithm for the cells in the pack using the results from the model. • Sensitivity of the thermal performance to contact resistance has been investigated. - Abstract: Thermal management system is of critical importance for a Li-ion battery pack, as high performance and long battery pack life can be simultaneously achieved when operated within a narrow range of temperature around the room temperature. An efficient thermal management system is required to keep the battery temperature in this range, despite widely varying operating conditions. A novel liquid coolant based thermal management system, for 18,650 battery pack has been introduced herein. This system is designed to be compact and economical without compromising safety. A coupled three-dimensional (3D) electrochemical thermal model is constructed for the proposed Li-ion battery pack. The model is used to evaluate the effects of different operating conditions like coolant flow-rate and discharge current on the pack temperature. Contact resistance is found to have the strongest impact on the thermal performance of the pack. From the numerical solution, a simple and novel temperature correlation of predicting the temperatures of all the individual cells given the temperature measurement of one cell is devised and validated with experimental results. Such coefficients have great potential of reducing the sensor requirement and complexity in a large Li-ion battery pack, typical of an electric vehicle.

Organic coolant in Winnipeg riverbed sediments

International Nuclear Information System (INIS)

Guthrie, J.E.; Acres, O.E.

1979-03-01

Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

Influence of SiO{sub 2} on conduction and relaxation mechanism of Li{sup +} ions in binary network former lead silicate glasses

Energy Technology Data Exchange (ETDEWEB)

Ahlawat, Navneet [Department of Physics, Chaudhary Devi Lal University, Sirsa 125055, Haryana (India); Ahlawat, Neetu, E-mail: neetugju@yahoo.co.in [Department of Applied Physics, Guru Jambheshwar University of Science and Technology, Hisar 125001, Haryana (India); Aghamkar, Praveen [Department of Physics, Chaudhary Devi Lal University, Sirsa 125055, Haryana (India); Agarwal, Ashish; Sanghi, Sujata; Sindhu, Monica [Department of Applied Physics, Guru Jambheshwar University of Science and Technology, Hisar 125001, Haryana (India)

2013-04-01

Ion conducting glasses having composition 30Li{sub 2}O·(70−x)PbO·xSiO{sub 2} were prepared by the normal melt quench technique. The compositional variations in density, molar volume and glass transition temperature confirm the dual role of PbO acting as a network modifying oxide as well as a network forming oxide. Conduction and relaxation mechanisms in these glasses were studied using impedance spectroscopy in the frequency range from 1 Hz to 7 MHz and in a temperature range below glass transition temperature. The ac and dc conductivities, activation energy of the dc conductivity and relaxation frequency were extracted from the impedance spectra. Similar values of activation energy for dc conduction and for conductivity relaxation time indicate that the ions have to overcome the same energy barrier while conducting and relaxing. The increase in dc conductivity for silica rich compositions is attributed to the presence of mixed former effect in the studied glasses. The study of conductivity spectra reveals a transition from non-random to random hopping motion of lithium ions on successive replacement of PbO by SiO{sub 2} in glass matrix. The conduction and relaxation mechanism in the studied glasses are well explained with the concept of mismatch and relaxation (CMR) model.

Evaluation of steam as a potential coolant for nonbreeding blanket designs

International Nuclear Information System (INIS)

Stevens, H.C.; Misra, B.; Youngdahl, C.K.

1978-01-01

A steam-cooled nonbreeding blanket design has been developed as an evolution of the Argonne Experimental Power Reactor (EPR) studies. This blanket concept complete with maintenance considerations is to function at temperatures up to 650 0 C utilizing nickel-based alloys such as Inconel 625. Thermo-mechanical analyses were carried out in conjunction with thermal hydraulic analysis to determine coolant chennel arrangements that permit delivery of superheated steam at 500 0 C directly to a modern fossil plant-type turbine. A dual-cycle system combining a pressurized water circuit coupled with a superheated steam circuit can produce turbine plant conversion efficiencies approaching 41.5%

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

Energy Technology Data Exchange (ETDEWEB)

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki [Nuclear Research Group, FMIPA, Bandung Institute of Technology Jl. Ganesha 10, Bandung 40132 (Indonesia); Miura, Ryosuke; Takaki, Naoyuki [Department of Nuclear Safety Engineering, Tokyo City University 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan); Sekimoto, H. [Emerritus Prof. of Research Laboratory for Nuclear Reactors, Tokyo Inst. of Technology (Japan)

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

Sedimentary input into the source of Martinique lavas: a Li perspective

Science.gov (United States)

Tang, M.; Chauvel, C.; Rudnick, R. L.

2013-12-01

The Lesser Antilles arc is known for the prominent continental crustal signatures in its lavas. It thus provides an ideal target for studying crustal recycling in subduction zones. Martinique Island, located in the middle of the Lesser Antilles arc, has been well characterized for its elemental and radiogenic isotope geochemistry (Labanieh et al., 2012). We measured Li isotopes in the Martinique lavas as well as sediments cored at the southern (Site 144) and northern part (Site 543) of the subducting slab. The sediments show a large isotopic variation (δ7Li ~ -4.2‰ to +3.2‰) but the average δ7Li of -1.1 × 2.4‰ (1 σ, n = 15) is significantly lower than that of N-MORB (δ7Li = + 3.4 × 0.7‰, 1 σ, Tomascak et al., 2008), reflecting the influence of chemical weathering in the continental provenance. Although the subducting sediments display marked mineralogical and chemical shifts from south to north due to different deposition distances to the continental platform (Carpentier et al., 2009), their average Li isotopic compositions are indiscernible from each other. With a few exceptions, the Li isotopic compositions of the Martinique lavas are systematically lighter than MORB, giving an average δ7Li of 1.6 × 1.4‰ (1 σ, n = 25, 4 exceptions excluded). The δ7Li values show no correlation with any radiogenic isotope ratios (206Pb/204Pb, 87Sr/86Sr, 143Nd/144Nd and 176Hf/177Hf), Li/Y ratio, La/Sm ratio and SiO2 content. Therefore, the light Li isotopic composition likely reflects the source characteristics rather than contamination within the arc crust. Incorporation of the isotopically light sediments from Site 144 and 543 in the source may explain the depletion of 7Li in the Martinique lavas. A two-end-member mixing model requires 2-5% addition of the sediments into the depleted mantle source, compared with 1-10% sediments constrained by radiogenic isotopes (Carpentier et al., 2008). References Carpentier, M., Chauvel, C., & Mattielli, N., 2008. Pb

Commercial tandem mirror reactor design with thermal barriers: WITAMIR-I

International Nuclear Information System (INIS)

Kulcinski, G.L.; Emmert, G.A.; Maynard, C.W.

1980-10-01

A conceptual design of a near term commercial tandem mirror power reactor is presented. The basic configuration utilizes yin-yang minimum-B plugs with inboard thermal barriers. The maximum magnetic fields are 6.1 T, 8.1 T, and 15 T in the central cell, yin-yang, and thermal barrier magnets, respectively. The blanket utilizes Pb 83 Li 17 as the coolant and HT-9 as the structural material. This yields a high energy multiplication (1.37), a sufficient tritium breeding ratio (1.07) and has a major advantage with respect to maintenance. The plasma Q is 28 at a fusion power level of 3000 MW(t); the net electrical output is 1530 MW(e); and the overall efficiency is 39%. Cost estimates indicate that WITAMIR-I is competitive with recent tokamak power reactor designs

Effects of Imide–Orthoborate Dual-Salt Mixtures in Organic Carbonate Electrolytes on the Stability of Lithium Metal Batteries

Energy Technology Data Exchange (ETDEWEB)

Li, Xing [Energy and Environment; School of Materials Science and Engineering, Southwest Petroleum University, Chengdu, Sichuan 610500, China; Zheng, Jianming [Energy and Environment; Engelhard, Mark H. [Environmental Molecular; Mei, Donghai [Physical and Computational Sciences Directorate, Pacific Northwest National Laboratory, Richland, Washington 99354, United States; Li, Qiuyan [Energy and Environment; Jiao, Shuhong [Energy and Environment; Liu, Ning [Physical and Computational Sciences Directorate, Pacific Northwest National Laboratory, Richland, Washington 99354, United States; State Key Laboratory of Chemical Resource Engineering, Beijing University of Chemical Technology, Beijing, 100029, China; Zhao, Wengao [Energy and Environment; School of Energy Research, Xiamen University, Xiamen, Fujian 361102, China; Zhang, Ji-Guang [Energy and Environment; Xu, Wu [Energy and Environment

2018-01-09

The effects of lithium imide and lithium orthoborate dual-salt electrolytes of different salt chemistries in carbonate solvents on the cycling stability of Li metal batteries were systematically and comparatively investigated. Two imide salts (LiTFSI and LiFSI) and two orthoborate salts (LiBOB and LiDFOB) were chosen for this study and compared with the conventional LiPF6 salt. The cycling stability of the Li metal cells with the electrolytes follows the order from good to poor as LiTFSI-LiBOB > LiTFSI-LiDFOB > LiPF6 > LiFSI-LiBOB > LiFSI-LiDFOB, indicating that LiTFSI behaves better than LiFSI and LiBOB over LiDFOB in these four dual-salt mixtures. The LiTFSI-LiBOB can effectively protect the Al substrate and form a more robust surface film on Li metal anode, while the LiFSI-LiBOB results in serious corrosion to the stainless steel cell case and a thicker and looser surface film on Li anode. Computational calculations indicate that the chemical and electrochemical stabilities also follow the order of LiTFSI-LiBOB > LiTFSI-LiDFOB > LiFSI-LiBOB > LiFSI-LiDFOB. The key findings of this work emphasize that the salt chemistry is critically important for enhancing the interfacial stability of Li metal anode and should be carefully manipulated in the development of high performance Li metal batteries.

The impact of tritium solubility and diffusivity on inventory and permeation in liquid breeder blankets

International Nuclear Information System (INIS)

Caorlin, M.; Gervasini, G.; Reiter, F.

1988-01-01

The authors reviewed hydrogen solubility and diffusivity data for liquid lithium-based compounds which are potential breeding blanket materials in NET-type fusion devices. These data have been used to assess tritium permeation and inventory in separately cooled NET blankets and in self cooled blankets with a vanadium first wall. The results for the separately cooled NET-liquid breeder show that tritium permeation is negligible for lithium, a serious problem for Pb-17Li and a critical one for Flibe. The total tritium inventory is lowest in lithium, high in Pb-17Li and very high in Flibe. The high tritium partial pressure for Flibe or Pb-17Li can be reduced in a self cooled blanket with a vanadium first wall. Permeation into the plasma reduces the blanket tritium inventory and permeation. Tritium recovery can be combined with the plasma exhaust

CANDU with supercritical water coolant: conceptual design features

International Nuclear Information System (INIS)

Spinks, N.

1997-01-01

An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

Calorimetric investigation of (Pb0.45Bi0.55)-U system

International Nuclear Information System (INIS)

Agarwal, Renu; Samui, Pradeep; Mukerjee, S.K.; Ramakumar, K.L.

2016-01-01

Lead-bismuth eutectic (LBE) is being considered as a coolant of future high temperature reactors. As lead and bismuth are good spallation target material, they are planned to be used in accelerator driven reactor systems (ADS). Under the clad breach conditions these elements may come in direct contact with uranium of metallic fuel. In our labs, we had earlier investigated binary interactions of U-Pb and U-Bi. To assess interaction behaviour of 'U' with the eutectic melt, it was planned to measure enthalpy of mixing of LBE-U and compares it with the binary mixing. SEM-EDS studies of the product formed after mixing of LBE and 'U' were carried out to establish coexisting phases and their compositions. UPb 3 is Pb-rich compound of U-Pb and UBi 2 is Bi-rich compound of U-Bi. So addition of 'U' in (Pb 0.45 Bi 0.55 ) will result in formation of the more stable compound among UPb 3 and UBi 2

The ARIES-AT advanced tokamak, Advanced technology fusion power plant

International Nuclear Information System (INIS)

Najmabadi, Farrokh; Abdou, A.; Bromberg, L.

2006-01-01

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R and D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ x = 2.2) which is the result of a 'thinner' blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β N . ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb-17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 deg. C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb-17Li to about 1000 deg. C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES

Addition of soluble and insoluble neutron absorbers to the reactor coolant system of TMI-2

International Nuclear Information System (INIS)

Hansen, R.F.; Silverman, J.; Queen, S.P.; Ryan, R.F.; Austin, W.E.

1984-07-01

The physical and chemical properties of six elements were studied and combined with cost estimates to determine the feasibility of adding them to the TMI-2 reactor coolant to depress k/sub eff/ to less than or equal to 0.95. Both soluble and insoluble forms of the elements B, Cd, Gd, Li, Sm, and Eu were examined. Criticality calculations were performed by Oak Ridge National Laboratory to determine the absorber concentration required to meet the 0.95 k/sub eff/ criterion. The conclusion reached is that all elements with the exception of boron have overriding disadvantages which preclude their use in this reactor. Solubility experiments in the reactor coolant show that boron solubility is the same as that of boron in pure aqueous solutions of sodium hydroxide and boric acid; consequently, solubility is not a limiting factor in reaching the k/sub eff/ criterion. Examination of the effect of pH on sodium requirements and costs for processing to remove radionuclides revealed a sharp dependence; small decreases in pH lead to a large decrease in both sodium requirements and processing costs. Boron addition to meet any contemplated reactor safety requirements can be accomplished with existing equipment; however, this addition must be made with the reactor coolant system filled and pressurized to ensure uniform boron concentration

Role of dual-laser ablation in controlling the Pb depletion in epitaxial growth of Pb(Zr0.52Ti0.48)O3 thin films with enhanced surface quality and ferroelectric properties

Science.gov (United States)

Mukherjee, Devajyoti; Hyde, Robert; Mukherjee, Pritish; Srikanth, Hariharan; Witanachchi, Sarath

2012-03-01

Pb depletion in Pb(Zr0.52Ti0.48)O3 (PZT) thin films has remained as a major setback in the growth of defect-free PZT thin films by pulsed laser ablation techniques. At low excimer (KrF) laser fluences, the high volatility of Pb in PZT leads to non-congruent target ablation and, consequently, non-stoichiometric films, whereas, at high laser fluences, the inherent ejection of molten droplets from the target leads to particulate laden films, which is undesirable in heterostructure growth. To overcome these issues, a dual-laser ablation (PLDDL) process that combines an excimer (KrF) laser and CO2 laser pulses was used to grow epitaxial PZT films on SrTiO3 (100) and MgO (100) substrates. Intensified-charge-coupled-detector (ICCD) images and optical emission spectroscopy of the laser-ablated plumes in PLDDL revealed a broader angular expansion and enhanced excitation of the ablated species as compared to those for single-laser ablation (PLDSL). This led to the growth of particulate-free PZT films with higher Pb content, better crystallinity, and lower surface roughness as compared to those deposited using PLDSL. For FE measurements, PZT capacitors were fabricated in situ using the latticed-matched metallic oxide, La0.7Sr0.3MnO3, as the top and bottom electrodes. PZT films deposited using PLDDL exhibited enhanced polarization for all driving voltages as compared to those deposited using PLDSL. A highest remanent polarization (Pr) of ˜91 μC/cm2 and low coercive field of ˜40 kV/cm was recorded at 9 V driving voltage. Fatigue characterization revealed that PZT films deposited using PLDDL showed unchanging polarization, even after 109 switching cycles.

17. aprill : rahvusvaheline tuutorprojekt

Index Scriptorium Estoniae

2001-01-01

17. apr. 2001 toimus esimene rahvusvaheline workshop "International Tutoring", eesmärk : koostöö arendamine välisülikoolidega tuutoritegevuse edendamisel ning kogemuste vahetamisel. Osalesid Eesti ülikoolide üliõpilased ja 45 Vantaa-Espoo polütehnikumi üliõpilast Soomest. Eestvedajateks TPÜ üliõpilasesinduse liikmed Katriina Vasarik, Kadri Kiigema ja Sulev Senkel

THE NEW DETECTIONS OF 7Li/6Li ISOTOPIC RATIO IN THE INTERSTELLAR MEDIA

International Nuclear Information System (INIS)

Kawanomoto, S.; Kajino, T.; Aoki, W.; Ando, H.; Noguchi, K.; Tanaka, W.; Bessell, M.; Suzuki, T. K.; Honda, S.; Izumiura, H.; Kambe, E.; Okita, K.; Watanabe, E.; Yoshida, M.; Sadakane, K.; Sato, B.; Tajitsu, A.; Takada-Hidai, M.

2009-01-01

We have determined the isotopic abundance ratio of 7 Li/ 6 Li in the interstellar media (ISMs) along lines of sight to HD169454 and HD250290 using the High-Dispersion Spectrograph on the Subaru Telescope. We also observed ζ Oph for comparison with previous data. The observed abundance ratios were 7 Li/ 6 Li = 8.1 +3.6 -1.8 and 6.3 +3.0 -1.7 for HD169454 and HD250290, respectively. These values are in reasonable agreement with those observed previously in the solar neighborhood ISMs within ±2σ error bars and are also consistent with our measurement of 7 Li/ 6 Li = 7.1 +2.9 -1.6 for a cloud along the line of sight to ζ Oph. This is good evidence for homogeneous mixing and instantaneous recycling of the gas component in the Galactic disk. We also discuss several source compositions of 7 Li, Galactic cosmic-ray interactions, stellar nucleosynthesis, and big bang nucleosynthesis.

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

The effect of disorder on electronic and magnetic properties of quaternary Heusler alloy CoFeMnSi with LiMgPbSb-type structure

International Nuclear Information System (INIS)

Feng, Yu; Chen, Hong; Yuan, Hongkuan; Zhou, Ying; Chen, Xiaorui

2015-01-01

Thin films based on Heusler alloy often lost their theoretical predicted ultra-high spin polarization owing to the appearance of disorder. Using the first-principles calculations within density functional theory (DFT), we investigate the effect of disorder including antisite and swap on electronic and magnetic properties of quaternary Heusler alloy CoFeMnSi with LiMgPbSb-type structure. Twelve kinds of antisites and six kinds of swap disorders are proposed and studied comprehensively. In our calculations, Co(Fe)-, Mn(Fe)-, Si(Mn)-antisite and Co–Fe swap disorders are most favorable due to their lowest formation energies. Moreover, the positive binding energies of Co–Fe, Co–Si, Fe–Si and Mn–Si swap disorders with respect to their corresponding antisite disorders indicate that these complex swap disorders are more stable compared with their corresponding isolated antisite disorders. The investigations on density of states (DOS) show that the spin down energy gap of disordered structures suffers contraction and their DOS entirely move towards lower zone. Besides, the 100% spin polarization is maintained in all structures with antisite and swap disorders except for those with Co(Mn)-, Co(Si)-antisite and Co–Mn, Co–Si swap disorders. Therefore, the half-metallicity of quaternary Heusler alloy CoFeMnSi is quite robust against interfering effects such as Si(Mn), Co(Fe) and Co–Fe disorders most possibly formed in the growth. - Highlights: • CoFeMnSi with LiMgPbSb-type structure is found to be a half-metallic ferromagnet. • Si(Mn), Co(Fe), Mn(Fe) antisites and Co–Fe swap disorders are most likely to form. • The half-metallicity of CoFeMnSi is robust against the most possible disorders. • The magnetic moments of the most possible disorders follow the Pauli-Slater rule

Deformation mechanism in LiF single crystals at 1.7 to 330 K

International Nuclear Information System (INIS)

Niaz, S.; Butt, M.Z.

1999-01-01

The experimental data appertaining to the influence of temperature on the critical resolved shear stress (CRSS) of LiF ionic single crystals containing 10/sup -3/ wt% of divalent metal impurities in the range 1.7 to 330 K have been analyzed within the framework of the kink-pair nucleation (KPN) model of plastic flow in crystalline materials. The CRSS-T data when plotted in log-linear coordinates exhibit three distinct regions represented by straight lines of different slopes. In the temperature range 1.7 to 90 K, the CRSS 6 determined primarily by the stress-assisted thermally-activated escape of screw dislocations trapped in the Peierls troughs. At temperatures between 90 and 260 K, the rate process of plastic deformation is unpinning of edge-dislocation segments from short was rows of randomly dispersed point defects, e.g. residual metal impurities atoms, divalent metal ion-vacancy dipoles, induced defects formed during the pre-yield stage etc. 4. However, at higher temperatures up to 330 K, the CRSS decreases rapidly with rise in temperature, probably due to the mobility of the point defects referred to, and the KPN model becomes inapplicable. (author)

Lithium1.3Aluminum0.3Titanium1.7Phosphate as a solid state Li-ion conductor: Issues with microcracking and stability in aqueous solutions

Science.gov (United States)

Jackman, Spencer D.

Lithium aluminum titanium phosphate (LATP) with formula Li1.3Al0.3Ti1.7(PO4)3 was analyzed and tested to better understand its applicability as a solid state ion conducting ceramic material for electrochemical applications. Sintered samples were obtained from Ceramatec, Inc. in Salt Lake City and characterized in terms of density, phase-purity, fracture toughness, Young's modulus, thermal expansion behavior, mechanical strength, a.c. and d.c. ionic conductivity, and susceptibility to static and electrochemical corrosion in aqueous Li salt solutions. It was shown that LATP is prone to microcrack generation because of high thermal expansion anisotropy. A.c. impedance spectra of high-purity LATP of varying grain sizes showed that microcracking had a negative impact on the ionic conduction of Li along grain boundaries, with fine-grained (1.7±0.7 µm) LATP having twice the ionic conductivity of the same purity of coarse-grained (4.8±1.9 µm) LATP at 50°C. LATP with detectible secondary phases had lower ionic conductivity for similar grain sizes, as would be expected. The Young's modulus of fine-grained LATP was measured to be 115 GPa, and the highest biaxial strength was 191±11 MPa when tested in mineral oil, 144±13 MPa as measured in air, and 26±7 MPa after exposure to deionized water, suggesting that LATP undergoes stress-corrosion cracking. After exposure to LiOH, the strength was 76±19 MPa. This decrease in strength was observed despite there being no measureable change in a.c. impedance spectra, X-ray diffraction, or sample mass, suggesting phosphate glasses at grain boundaries. The chemical and electrochemical stability of high-purity LATP in aqueous electrochemical cells was evaluated using LiOH, LiCl, LiNO3, and LiCOOCH3 salts as the Li source. LATP was found to be most stable between pH 8-9, with the longest cell operating continuously at 25 mA cm-2 for 625 hours at 40°C in LiCOOCH3. At pH values outside of the 7-10 range, eventual membrane degradation

Flow boiling test of GDP replacement coolants

International Nuclear Information System (INIS)

Park, S.H.

1995-01-01

The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

Cleaning of aluminum after machining with coolants

International Nuclear Information System (INIS)

Roop, B.

1992-01-01

An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

Bi-Abundance Ionisation Structure of the Wolf-Rayet Planetary Nebula PB 8

Science.gov (United States)

Danehkar, A.

2018-01-01

The planetary nebula PB 8 around a [WN/WC]-hybrid central star is one of planetary nebulae with moderate abundance discrepancy factors (ADFs 2-3), which could be an indication of a tiny fraction of metal-rich inclusions embedded in the nebula (bi-abundance). In this work, we have constructed photoionisation models to reproduce the optical and infrared observations of the planetary nebula PB 8 using a non-LTE stellar model atmosphere ionising source. A chemically homogeneous model initially used cannot predict the optical recombination lines. However, a bi-abundance model provides a better fit to most of the observed optical recombination lines from N and O ions. The metal-rich inclusions in the bi-abundance model occupy 5.6% of the total volume of the nebula, and are roughly 1.7 times cooler and denser than the mean values of the surrounding nebula. The N/H and O/H abundance ratios in the metal-rich inclusions are 1.0 and 1.7 dex larger than the diffuse warm nebula, respectively. To reproduce the Spitzer spectral energy distribution of PB 8, dust grains with a dust-to-gas ratio of 0.01 (by mass) were also included. It is found that the presence of metal-rich inclusions can explain the heavy element optical recombination lines, while a dual-dust chemistry with different grain species and discrete grain sizes likely produces the infrared continuum of this planetary nebula. This study demonstrates that the bi-abundance hypothesis, which was examined in a few planetary nebulae with large abundance discrepancies (ADFs > 10), could also be applied to those typical planetary nebulae with moderate abundance discrepancies.

Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

International Nuclear Information System (INIS)

Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

2015-01-01

In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

Coolant make-up device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

Research on Coolant Radiochemistry

International Nuclear Information System (INIS)

Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

2007-06-01

The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

Electrophysical properties of PMN-PT-PS-PFN:Li ceramics

Directory of Open Access Journals (Sweden)

R. Skulski

2013-01-01

Full Text Available We present the technology of obtaining and the electrophysical properties of a multicomponent material 0.61PMN-0.20PT-0.09PS-0.1PFN:Li (PMN-PT-PS-PFN:Li. The addition of PFN into PMN-PT decreases the temperature of final sintering which is very important during technological process (addition of Li decreases electric conductivity of PFN. Addition of PS i.e., PbSnO3 (which is unstable in ceramic form permits to shift the temperature of the maximum of dielectric permittivity. One-step method of obtaining ceramic samples from oxides and carbonates has been used. XRD, microstructure, scanning calorimetry measurements and the main dielectric, ferroelectric and electromechanical properties have been investigated for the obtained samples.

Development of Liquid Type Breeder Technology for ITER-TBM

Energy Technology Data Exchange (ETDEWEB)

Jung, Ki Sok; Hong, Bong Geun; Lee, Dong Won

2008-07-15

In relation to liquid type TBM technology development, various works are performed. We established a test loop concept to test the MHD effects and materials compatibility for the Pb-17Li breeder material. For the loop construction, electromagnetic pump and storage tank for the Pb-17Li loop was manufactured and some technical requirements are summarised. As a reference, technical literatures relevant to the liquid type TBM materials and the tritium extraction from breeder materials are also surveyed.

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

Reactor coolant cleanup device

International Nuclear Information System (INIS)

Igarashi, Noboru.

1986-01-01

Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

Geochemical and Sr-Nd-Pb-Li isotopic characteristics of volcanic rocks from the Okinawa Trough: Implications for the influence of subduction components and the contamination of crustal materials

Science.gov (United States)

Guo, Kun; Zhai, Shikui; Yu, Zenghui; Wang, Shujie; Zhang, Xia; Wang, Xiaoyuan

2018-04-01

The Okinawa Trough is an infant back-arc basin developed along the Ryukyu arc. This paper provides new major and trace element and Sr-Nd-Pb-Li isotope data of volcanic rocks in the Okinawa Trough and combines the published geochemical data to discuss the composition of magma source, the influence of subduction component, and the contamination of crustal materials, and calculate the contribution between subduction sediment and altered oceanic crust in the subduction component. The results showed that there are 97% DM and 3% EMI component in the mantle source in middle trough (MS), which have been influenced by subduction sediment. The Li-Nd isotopes indicate that the contribution of subduction sediment and altered oceanic crust in subduction component are 4 and 96%, respectively. The intermediate-acidic rocks suffer from contamination of continental crust material in shallow magma chamber during fractional crystallization. The acidic rocks in south trough have experienced more contamination of crustal material than those from the middle and north trough segments.

Coolant cleanup method in a nuclear reactor

International Nuclear Information System (INIS)

Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

1983-01-01

Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

Lead Coolant Test Facility Technical and Functional Requirements, Conceptual Design, Cost and Construction Schedule

International Nuclear Information System (INIS)

Soli T. Khericha

2006-01-01

This report presents preliminary technical and functional requirements (T and FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic. Based on review of current world lead or lead-bismuth test facilities and research need listed in the Generation IV Roadmap, five broad areas of requirements of basis are identified: Develop and Demonstrate Prototype Lead/Lead-Bismuth Liquid Metal Flow Loop Develop and Demonstrate Feasibility of Submerged Heat Exchanger Develop and Demonstrate Open-lattice Flow in Electrically Heated Core Develop and Demonstrate Chemistry Control Demonstrate Safe Operation and Provision for Future Testing. These five broad areas are divided into twenty-one (21) specific requirements ranging from coolant temperature to design lifetime. An overview of project engineering requirements, design requirements, QA and environmental requirements are also presented. The purpose of this T and FRs is to focus the lead fast reactor community domestically on the requirements for the next unique state of the art test facility. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 420 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M. It is also estimated that the facility will require two years to be constructed and ready for operation

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

Investigation of corrosion resistance of 18Cr-14NNi-1.5Si austenitic steel in molten PbBi eutectic

International Nuclear Information System (INIS)

Rivai, A.K.; Heinzel, H.; Effendi, N.

2013-01-01

Full-text: The development of high corrosion resistant materials for the fuel cladding and structural materials in liquid lead-bismuth (Pb-Bi) eutectic environment especially at high temperature is a critical issue for the deployment of LFR (Lead alloy-cooled fast reactor) and ADS (Accelerator driven Transmutation System). Pb-Bi eutectic is a coolant for LFR which is one of the future nuclear reactors in the world (Generation IV reactors), and also a spallation target material and a coolant for ADS. In this study, corrosion test of an austenitic steel was done in COSTA Pb-Bi eutectic corrosion test facility at Pulsed Power and Microwave Technology, Karlsruhe Institute of Technology, Germany. The sample was an 18Cr-14Ni-1.5Si austenitic steel which has been developed in Center For Technology of Nuclear Industry Materials, Indonesian National Nuclear Energy Agency. The test was done in stagnant molten Pb-Bi eutectic at 550 degree Celsius of temperature for about 300 hours with an oxygen concentration of 1 x 10 -6 wt %. The characterization was carried out using OM (Optical Microscope), SEM-EDS (Scanning Electron Microscope and Energy Dispersive X-ray Spectroscope) and AFM (Atomic Force Microscope). The corrosion test result showed the formation of a duplex oxide layer for example an outer iron oxide layer with about 3-3.4 μm in thickness. Furthermore, there was no penetration of Pb-Bi into the bulk of the specimen because of the protection from the protective oxide layer. (author)

Multirods burst tests under loss-of-coolant conditions

International Nuclear Information System (INIS)

Kawasaki, S.; Uetsuka, H.; Furuta, T.

1983-01-01

In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

Study of interactions between liquid lead-lithium alloy and austenitic and martensitic steels

International Nuclear Information System (INIS)

Simon, N.

1992-06-01

In the framework of Fusion Technology, the behaviour of structural materials in presence of liquid alloy Pb17Li is investigated. First, the diffusion coefficients of Fe and Cr have been determined at 500 deg C. Then mass transfer experiments in Pb17Li have been conducted in an anisothermal container with pure metals (Fe, Cr, Ni), Fe-Cr steels and austenitic steels. These experiments showed a very high loss of Nickel, which is an accordance with its high solubility, and Cr showed mass-losses one order of magnitude higher than for pure iron, as the diffusion coefficient of Cr is three orders of magnitude higher than for pure Fe. The corrosion rate of binary Fe-Cr and pure Fe are identical. In austenitic steels, the gamma lattice allows a higher mass-transfer of Cr than the alpha lattice, the presence of Cr slows downs the dissolution of Ni, and the porosity of corrosion layers results of losses of Cr and Ni. Finally, a review of our results and those of other laboratories allowed an identification of the corrosion limiting step. In the case of 1.4914 martensitic steel it is the diffusion of Fe in Pb17Li, while in the case of 316L austenitic steel it is the diffusion of Cr in Pb17Li

Extended Life Coolant Testing

Science.gov (United States)

2016-06-06

number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

Inorganic alkali lead iodide semiconducting APbI3 (A = Li, Na, K, Cs and NH4PbI3 films prepared from solution: Structure, morphology, and electronic structure

Directory of Open Access Journals (Sweden)

Eric Mankel

2016-06-01

Full Text Available APbI3 alkali lead iodides were prepared from aqueous (A= Na, Cs, ammonium NH4+, and methyl­ammonium CH3NH3+ and acetone (A= Li, K solutions by a self-organization low temperature process. Diffraction analysis revealed that the methylammonium-containing system (MAPbI3 crystallizes into a tetragonal perovskite structure, whereas the alkali and NH4+ systems adopt orthorhombic structures. Morphological inspection confirmed the influence of the cation on the growth mechanism: for A = Cs and NH4+, needle-like crystallites with lengths up to 3–4 mm; for A = K, thin stripes with lengths up to 5–6 mm; and for A = MA+, dodecahedral crystallites were observed. For A = Li and Na, the APbI3 systems typically resulted in polycrystalline aggregates. Optical absorption measurements demonstrated large energy band gaps for the alkali and ammonium systems with values between 2.19 and 2.40 eV. For electronic and chemical characterization by photoelectron spectroscopy, the as-prepared powders were dissolved in di-methylformamide and re-crystallized as thin films on F:SnO2 substrates by spin-coating. The binding energy differences between Pb4f and I3d core levels are highly similar in the investigated systems and close to the value measured for PbI2, indicating similar relative partial charges and formal oxidation states. The binding energies of the alkali ions are in accordance with oxidation state +1. The X-ray excited valence band spectra of the investigated APbI3 systems exhibited similar line shapes in the region between the valence band maximum and 4.5 eV higher binding energy due to common PbI6 octahedra which dominate the electronic structure. While the ionization energy values are quite similar (6.15 ± 0.25 eV, the Fermi-level positions of the unintentionally doped materials vary for different cations and different batches of the same material, which indicates that the position of the Fermi level can be influenced by changing the process parameters.

11Li Breakup on 208 at energies around the Coulomb barrier.

Science.gov (United States)

Fernández-García, J P; Cubero, M; Rodríguez-Gallardo, M; Acosta, L; Alcorta, M; Alvarez, M A G; Borge, M J G; Buchmann, L; Diget, C A; Falou, H A; Fulton, B R; Fynbo, H O U; Galaviz, D; Gómez-Camacho, J; Kanungo, R; Lay, J A; Madurga, M; Martel, I; Moro, A M; Mukha, I; Nilsson, T; Sánchez-Benítez, A M; Shotter, A; Tengblad, O; Walden, P

2013-04-05

The inclusive breakup for the (11)Li + (208)Pb reaction at energies around the Coulomb barrier has been measured for the first time. A sizable yield of (9)Li following the (11)Li dissociation has been observed, even at energies well below the Coulomb barrier. Using the first-order semiclassical perturbation theory of Coulomb excitation it is shown that the breakup probability data measured at small angles can be used to extract effective breakup energy as well as the slope of B(E1) distribution close to the threshold. Four-body continuum-discretized coupled-channels calculations, including both nuclear and Coulomb couplings between the target and projectile to all orders, reproduce the measured inclusive breakup cross sections and support the presence of a dipole resonance in the (11)Li continuum at low excitation energy.

Limits to fuel/coolant mixing

International Nuclear Information System (INIS)

Corradini, M.L.; Moses, G.A.

1985-01-01

The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

Li+ transport properties of W substituted Li7La3Zr2O12 cubic lithium garnets

Directory of Open Access Journals (Sweden)

L. Dhivya

2013-08-01

Full Text Available Lithium garnet Li7La3Zr2O12 (LLZ sintered at 1230 °C has received considerable importance in recent times as result of its high total (bulk + grain boundary ionic conductivity of 5 × 10−4 S cm−1 at room temperature. In this work we report Li+ transport process of Li7−2xLa3Zr2−xWxO12 (x = 0.3, 0.5 cubic lithium garnets. Among the investigated compounds, Li6.4La3Zr1.7W0.3O12 sintered relatively at lower temperature 1100 °C exhibits highest room temperature (30 °C total (bulk + grain boundary ionic conductivity of 7.89 × 10−4 S cm−1. The temperature dependencies of the bulk conductivity and relaxation frequency in the bulk are governed by the same activation energy. Scaling the conductivity spectra for both Li6.4La3Zr1.7W0.3O12 and Li6La3Zr1.5W0.5O12 sample at different temperatures merges on a single curve, which implies that the relaxation dynamics of charge carriers is independent of temperature. The shape of the imaginary part of the modulus spectra suggests that the relaxation processes are non-Debye in nature. The present studies supports the prediction of optimum Li+ concentration required for the highest room temperature Li+ conductivity in LixLa3M2O12 is around x = 6.4 ± 0.1.

Conceptual design and testing strategy of a dual functional lithium-lead test blanket module in ITER and EAST

International Nuclear Information System (INIS)

Wu, Y.

2007-01-01

A dual functional lithium-lead (DFLL) test blanket module (TBM) concept has been proposed for testing in the International Thermonuclear Experimental Reactor (ITER) and the Experimental Advanced Superconducting Tokamak (EAST) in China to demonstrate the technologies of the liquid lithium-lead breeder blankets with emphasis on the balance between the risks and the potential attractiveness of blanket technology development. The design of DFLL-TBM concept has the flexibility of testing both the helium-cooled quasi-static lithium-lead (SLL) blanket concept and the He/PbLi dual-cooled lithium-lead (DLL) blanket concept. This paper presents an effective testing strategy proposed to achieve the testing target of SLL and DLL DEMO blankets relevant conditions, which includes three parts: materials R and D and small-scale out-of-pile mockups testing in loops, middle-scale TBMs pre-testing in EAST and full-scale consecutive TBMs testing corresponding to different operation phases of ITER during the first 10 years. The design of the DFLL-TBM concept and the testing strategy ability to test TBMs for both blanket concepts in sequence and or in parallel for both ITER and EAST are discussed

Reverse osmosis and its use at the nuclear power plants. Purification of primary circuit coolant by the means of reverse osmosis

International Nuclear Information System (INIS)

Kus, Pavel; Vonkova, Katerina; Kunesova, Katerina; Bartova, Sarka; Skala, Martin; Moucha, Tomáš

2014-01-01

This contribution is focused on the use of membrane technologies (e.g. reverse osmosis) for the primary coolant purification at the nuclear power plants. Currently, boric acid present in the primary coolant is preconcentrated at the evaporators, but their operation is very inefficient and expensive. Therefore, reverse osmosis was proposed as one of promising methods possibly replacing evaporators. The aim of the purification process is to achieve boric acid solution of a defined concentration (40 g/l) in the retentate stream in order to recycle it and reuse it in the primary circuit. Additionally, permeate flow should consist solely of pure water. To study the efficiency of several reverse osmosis modulus in the boric acid removal form the water solutions, experimental apparatus was constructed in our laboratory. It consists of the solution reservoir, pump and reverse osmosis modulus. The arrangement of experiments was batch and the retentate flow was refluxed to the feed solution. Several modulus of commercial reverse osmosis membranes were tested. The feed solution contained various concentrations of H 3 BO 3 , KOH, LiOH and NH 3 in order to simulate real primary coolant composition. Based on the experimental results, mathematical model was developed in order to optimize experimental conditions for the best results in primary coolant purification and boric acid preconcentration. (author)

Sorption of radionuclides from Pb-Bi melt. Report 1

International Nuclear Information System (INIS)

Konovalov, Eh.E.; Il'icheva, N.S.; Trifonova, O.E.

2015-01-01

Results of laboratory investigations of sorption and interfacial distribution of 54 Mn, 59 Fe, 60 Co, 106 Ru, 125 Sb, 137 Cs, 144 Ce, 154,155 Eu and 235,238 U radionuclides in the system Pb-Bi melt - steel surface are analyzed. It is shown that 106 Ru and 125 Sb are concentrated in Pb-Bi melt and other radionuclides with higher oxygen affinity are sorbed on oxide deposits on structural materials. Temperature dependences of sorption efficiency of radionuclides are studied. It is shown that there is sharp increase of this value for all radionuclides near the temperature range 350-400 deg C. Recommendations are given on the use of 106 Ru and 125 Sb as a reference for fuel element rupture detection system with radiometric monitoring of coolant melt samples and 137 Cs, 134 Cs, 134m Cs with radiometric monitoring of sorbing samples [ru

Reactor coolant pump transportation incident

International Nuclear Information System (INIS)

Noce, D.

1992-01-01

This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

Energy Technology Data Exchange (ETDEWEB)

Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-10-01

This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

Primary coolant circuits in FBR type reactors

International Nuclear Information System (INIS)

Kutani, Masushiro.

1985-01-01

Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

Thermophysical properties of potential breeder materials for fusion technology

International Nuclear Information System (INIS)

Schulz, B.

1987-01-01

The paper presents the results of the experimental determination of the thermophysical properties of liquid Li(17)Pb(83). The eutectic is characterized by metallography, thermal-, differential thermal and chemical analysis. Based on assumptions of the chemical bonding in Li-Pb-intermetallics, physical properties of these compounds in solid state are given. For Li 2 SiO 3 and Y-LiAlO 2 the thermal diffusivity and conductivity were determined as well as specific heat and thermal expansion. In general the important role of characterization in discussing thermophysical properties is pointed out. (author)

Effect of doping of trivalent cations Ga{sup 3+}, Sc{sup 3+}, Y{sup 3+} in Li{sub 1.3}Al{sub 0.3}Ti{sub 1.7} (PO{sub 4}){sub 3} (LATP) system on Li{sup +} ion conductivity

Energy Technology Data Exchange (ETDEWEB)

Kothari, Dharmesh H.; Kanchan, D.K., E-mail: dkkanchan.ssi@gmail.com

2016-11-15

We report the effect of trivalent cations dopants in the Li{sub 1.3}Al{sub 0.3−x}R{sub x}Ti{sub 1.7}(PO{sub 4}){sub 3} (R=Ga{sup 3+}, Sc{sup 3+}, Y{sup 3+}) NASICON ceramic system in the concentration range x=0.01,0.03,0.05,0.07, on the Li{sup +} ion conducting properties using impedance spectroscopy. The samples were prepared by solid state reaction method and characterized by X-Ray Diffraction and density measurements. The electrical properties were studied using impedance spectroscopy in frequency range 10 Hz to 20 MHz and temperature range 303 K to 423 K. Although the porosity of the material decreased with doping, the overall Li{sup +} ion conductivity of the system did not improve with doping. Ionic radii of the dopant cations was found to be an important factor in formation of impurity phases and low Li{sup +} ion conductivity. Gallium doped samples exhibited a higher Li{sup +} ion conductivity compared to its scandium and yttrium doped counterparts.

Upper internals of PWR with coolant flow separator

International Nuclear Information System (INIS)

Chevereau, G.; Heuze, A.

1989-01-01

The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

Coolant processing device for nuclear reactor

International Nuclear Information System (INIS)

Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

1981-01-01

Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

Fatigue management considering LWR coolant environments

International Nuclear Information System (INIS)

Park, Heung Bae; Jin, Tae eun

2000-01-01

Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

Development of lead-bismuth coolant technology for nuclear device

International Nuclear Information System (INIS)

Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

2004-01-01

Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

Device for preventing coolant outflow in a reactor

International Nuclear Information System (INIS)

Nemoto, Kiyomitsu; Mochizuki, Keiichi.

1975-01-01

Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

Method of charging instruments into liquid metal coolant

International Nuclear Information System (INIS)

Yamazaki, Hiroshi

1980-01-01

Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

The Philosophy and Feasibility of Dual Readout Calorimetry

International Nuclear Information System (INIS)

Hauptman, John

2006-01-01

I will discuss the general physical ideas behind dual-readout calorimetry, their implementation in DREAM (Dual REAdout Module) with exact separation of scintillation and Cerenkov light, implementation with mixed light in DREAM fibers, anticipated implementation in PbWO4 crystals with applications to the 4th Concept detector and to CMS, use in high energy gamma-ray and cosmic ray astrophysics with Cerenkov and N2 fluorescent light, and implementation in the 4th Concept detector for muon identification

210Pb geochronology and trace metal fluxes (Cd, Cu and Pb) in the Gulf of Tehuantepec, South Pacific of Mexico

International Nuclear Information System (INIS)

Ruiz-Fernandez, Ana Carolina; Paez-Osuna, Federico; Machain-Castillo, Maria Luisa; Arellano-Torres, Elsa

2004-01-01

Distributions of Al, Cd, Cu, Fe, Li, Mn and Pb were analyzed in a sediment core collected in the Gulf of Tehuantepec, an important fisheries region located in the South Pacific of Mexico, where data on metal accumulation and accretion rates were previously almost nonexistent. Depth profiles of metal concentrations were converted to time-based profiles by using a 210 Pb-derived vertical accretion rate, estimated to be 0.05 cm year -1 on the average. Sediments were dated up to 8 cm depth, corresponding to a layer of ca. 140 years old. The historical changes of metal accumulation along the sediment core have shown a moderate enrichment of Cd, Cu and Pb concentrations at present, of about threefold the corresponding background concentrations. Chronological trace metal records showed that metal fluxes have increased over the last 20 years, reaching the maximum values at present of 2.5, 22.5 and 45.8 (μg cm -2 year -1 ) for Cd, Pb and Cu, respectively. These increments in metal fluxes are likely influenced by the development of anthropogenic land-based activities since over this period of time oil production activities in the region have had a significant development

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

Coulomb excitation of {sup 8}Li

Energy Technology Data Exchange (ETDEWEB)

Assuncao, Marlete; Britos, Tatiane Nassar [Universidade Federal de Sao Paulo (UNIFESP), SP (Brazil). Dept. de Ciencias Exatas e da Terra; Descouvemont, Pierre [Universite Libre de Bruxelles (ULB), Brussels (Belgium). Physique Nucleaire Theorique et Physique Mathematique; Lepine-Szily, Alinka; Lichtenthaler Filho, Rubens; Barioni, Adriana; Silva, Diego Medeiros da; Pereira, Dirceu; Mendes Junior, Djalma Rosa; Pires, Kelly Cristina Cezaretto; Gasques, Leandro Romero; Morais, Maria Carmen; Added, Nemitala; Neto Faria, Pedro; Rec, Rafael [Universidade de Sao Paulo (IF/USP), SP (Brazil). Inst. de Fisica. Dept. de Fisica Nuclear

2012-07-01

Full text: This work shows the Coulomb Excitation of {sup 8}Li on targets that have effectively behavior of Rutherford in angles and energies of interest for determining the value of the B(E2) electromagnetic transition. Theoretical aspects involved in this type of measure, known as COULEX [1], and some results in the literature [2-3] will be presented. Some problems with the targets and measurement system while performing an experiment on Coulomb Excitation of {sup 8}Li will be discussed: the energy resolution, background, possible contributions of the primary beam and also the excited states of the target near the region of elastic and inelastic peaks. They will be illustrated by measurements of the Coulomb Excitation of {sup 8}Li on targets of {sup 197}Au and {sup 208}Pb using the system RIBRAS(Brazilian Radioactive Ion Beam). In this case, the {sup 8}Li beam(T{sub 1/2} = 838 ms)is produced by {sup 9}Be({sup 7}Li;{sup 8} Li){sup 8}Be reaction from RIBRAS system which is installed at Instituto de Fisica of the Universidade de Sao Paulo. The primary {sup 7L}i beam is provided by Pelletron Accelerator. [1] K. Alder and A. Winther, Electromagnetic Excitation, North-Holland, New York, 1975; [2] P. Descouvemont and D. Baye, Phys. Letts. B 292, 235-238, 1992; [3] J. A. Brown, F. D. Becchetti, J. W. Jaenecke, K, Ashktorab, and D. A. Roberts, J. J. Kolata, R. J. Smith, and K. Lamkin, R. E. Warner, Phys. Rev. Letts., 66, 19, 1991; [4] R. J. Smith, J. J Kolata, K. Lamkin and A. Morsard, F. D. Becchetti, J. A. Brown, W. Z. Liu, J. W. Jaenecke, and D. A. Roberts, R. E. Warner, Phys. Rev. C, 43, 5, 1991. (author)

Continuous surveillance of reactor coolant circuit integrity

International Nuclear Information System (INIS)

1986-01-01

Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

Dual functions of zirconium modification on improving the electrochemical performance of Ni-rich LiNi0.8Co0.1Mn0.1O2

Energy Technology Data Exchange (ETDEWEB)

Li, Xing; Zhang, Kangjia; Wang, Mingshan; Liu, Yang; Qu, Meizhen; Zhao, Wengao; Zheng, Jianming

2018-02-28

Trace amount of Zirconium (Zr) has been adopted to modify the crystal structure and surface of the Ni-rich LiNi0.8Co0.1Mn0.1O2 (NCM811) cathode material. During cycling at 1.0C, the Zr-modified NCM811 shows an improved capacity retention of 92% after 100 cycles, higher than 75% for pristine NMC811. In addition, the Zr-modified NCM811 is capable of delivering a discharge capacity of 107 mAh g-1 at 10.0C rate, much higher than 28 mAh g-1 delivered by pristine material. These improved electrochemical performances are ascribed to the dual functions of Zr modification. On one hand, part of the Zr enters the crystal lattice, which is beneficial for reducing the Li/Ni cation mixing and enhancing the crystal stability of the cathode. On the other hand, the rest of the Zr forms a 1~2 nm thick coating layer on the surface of the NCM811 cathode, which effectively prevents the direct contact between NCM and the electrolyte, thus suppressing the detrimental interfacial reactions. Therefore, the Zr-modified LiNi0.8Co0.1Mn0.1O2 exhibited significantly enhanced cycling stability and charging/discharging rate capability in comparison with the untreated counterpart.

Contribution to the optimization of the chemical and radiochemical purification of pressurized water nuclear power plants primary coolant

International Nuclear Information System (INIS)

Elain, L.

2004-12-01

The primary coolant of pressurised water reactors is permanently purified thanks to a device, composed of filters and the demineralizers furnished with ion exchange resins (IER), located in the chemical and volume control system (CVCS). The study of the retention mechanisms of the radio-contaminants by the IER implies, initially, to know the speciation of the primary coolant percolant through the demineralizers. Calculations of theoretical speciation of the primary coolant were carried out on the basis of known composition of the primary coolant and thanks to the use of an adapted chemical speciation code. A complementary study, dedicated to silver behaviour, considered badly extracted, suggests metallic aggregates existence generated by the radiolytic reduction of the Ag + ions. An analysis of the purification curves of the elements Ni, Fe, Co, Cr, Mn, Sb and their principal radionuclides, relating to the cold shutdown of Fessenheim 1-cycle 20 and Tricastin 2-cycle 21, was carried out, in the light of a model based on the concept of a coupling well term - source term. Then, a thermodynamic modelling of ion exchange phenomena in column was established. The formation of the permutation front and the enrichment zones planned was validated by frontal analysis experiments of synthetic fluids (mixtures of Ni(B(OH) 4 ) 2 , LiB(OH) 4 and AgB(OH) 4 in medium B(OH) 3 )), and of real fluid during the putting into service of the device mini-CVCS at the time of Tricastin 2 cold shutdown. New tools are thus proposed, opening the way with an optimised management of demineralizers and a more complete interpretation of the available experience feedback. (author)

Li diffusion and the effect of local structure on Li mobility in Li2O-SiO2 glasses.

Science.gov (United States)

Bauer, Ute; Welsch, Anna-Maria; Behrens, Harald; Rahn, Johanna; Schmidt, Harald; Horn, Ingo

2013-12-05

Aimed to improve the understanding of lithium migration mechanisms in ion conductors, this study focuses on Li dynamics in binary Li silicate glasses. Isotope exchange experiments and conductivity measurements were carried out to determine self-diffusion coefficients and activation energies for Li migration in Li2Si3O7 and Li2Si6O13 glasses. Samples of identical composition but different isotope content were combined for diffusion experiments in couples or triples. Diffusion profiles developed between 511 and 664 K were analyzed by femtosecond laser ablation combined with multiple collector inductively coupled plasma mass spectrometry (fs LA-MC-ICP-MS) and secondary ion mass spectrometry (SIMS). Analyses of diffusion profiles and comparison of diffusion data reveal that the isotope effect of lithium diffusion in silicate glasses is rather small, consistent with classical diffusion behavior. Ionic conductivity of glasses was measured between 312 and 675 K. The experimentally obtained self-diffusion coefficient, D(IE), and ionic diffusion coefficient, D(σ), derived from specific DC conductivity provided information about correlation effects during Li diffusion. The D(IE)/D(σ) is higher for the trisilicate (0.27 ± 0.05) than that for the hexasilicate (0.17 ± 0.02), implying that increasing silica content reduces the efficiency of Li jumps in terms of long-range movement. This trend can be rationalized by structural concepts based on nuclear magnetic resonance (NMR) and Raman spectroscopy as well as molecular dynamic simulations, that is, lithium is percolating in low-dimensional, alkali-rich regions separated by a silica-rich matrix.

Evaluation of compatibility between different types of adhesives and dual-cured resin cement.

Science.gov (United States)

Franco, Eduardo B; Lopes, Lawrence G; D'alpino, Paulo H P; Pereira, José C; Mondelli, Rafael F L; Navarro, Maria F L

2002-01-01

The objective of this in vitro study was to evaluate the bonding compatibility between different adhesives and a dual-cured resin cement, using a conventional tensile bond test. The adhesives used were: Prime & Bond (PB) (Dentsply) (PB), Scotchbond Multi Purpose (SB) (3M), and the activator Self Cure (SC) (Dentsply). The dual-curing resin cement used was Enforce (EF) (Dentsply). Six groups with five specimens in each were tested: G1: EF/PB/EF (light cured); G2: EF/SB/EF (light cured); G3: EF/PB+SC/EF (light cured); G4: EF/PB+SC/EF (only chemically cured); G5: EF/EF (light cured); G6: EF/EF (only chemically cured). The resin cement was applied in two stainless steel molds with a cone-shaped perforation measuring 4 mm in diameter and 1 mm in thickness, and the adhesive was applied between them. Ten minutes after specimens were cured, the tensile strength was measured in a universal testing machine at a crosshead speed of 0.5 mm/min. The mean values (MPa) +/- SD obtained in each experimental group were: G1: 1.4 +/- 0.2; G2: 1.3 +/- 0.2; G3: 1.2 +/- 0.4; G4: 0.8 +/- 0.2; G5: 1.2 +/- 0.1; G6: 0.7 +/- 0.1. The results were statistically evaluated using nonparametric Kruskal-Wallis and Dunn tests (p adhesives used with dual-cured resin cement. The lowest tensile bond strength values occurred in the absence of photoactivation.

Decontamination of main coolant pumps

International Nuclear Information System (INIS)

Roofthooft, R.

1988-01-01

Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

Highly sensitive analysis of boron and lithium in aqueous solution using dual-pulse laser-induced breakdown spectroscopy.

Science.gov (United States)

Lee, Dong-Hyoung; Han, Sol-Chan; Kim, Tae-Hyeong; Yun, Jong-Il

2011-12-15

We have applied a dual-pulse laser-induced breakdown spectroscopy (DP-LIBS) to sensitively detect concentrations of boron and lithium in aqueous solution. Sequential laser pulses from two separate Q-switched Nd:YAG lasers at 532 nm wavelength have been employed to generate laser-induced plasma on a water jet. For achieving sensitive elemental detection, the optimal timing between two laser pulses was investigated. The optimum time delay between two laser pulses for the B atomic emission lines was found to be less than 3 μs and approximately 10 μs for the Li atomic emission line. Under these optimized conditions, the detection limit was attained in the range of 0.8 ppm for boron and 0.8 ppb for lithium. In particular, the sensitivity for detecting boron by excitation of laminar liquid jet was found to be excellent by nearly 2 orders of magnitude compared with 80 ppm reported in the literature. These sensitivities of laser-induced breakdown spectroscopy are very practical for the online elemental analysis of boric acid and lithium hydroxide serving as neutron absorber and pH controller in the primary coolant water of pressurized water reactors, respectively.

Hydrogen diffusion in Pb β''-alumina

International Nuclear Information System (INIS)

Bates, J.B.; Dudney, N.J.; Wang, J.C.

1985-01-01

The mobile Na + ions in Na β''-alumina can be completely exchanged with Pb 2+ ions by treatment in molten PbCl 2 . When this exchange was carried out in the presence of air, protons in the form of OH - were introduced into the conduction layers along with lead ions. Although the concentration of OH - was low, on the order of 5 x 10 -3 per formula unit of Pb/sub 0.84/Mg/sub 0.67/Al/sub 10.33/O_1_7, the distribution of OH - after ion exchange indicated that the proton mobility in Pb β''-alumina is high. The potential use of Pb β''-alumina as a fast proton conductor that is stable at 400 0 C motivated further studies of hydrogen diffusion. In this report, the results of tracer diffusion measurements by isotope exchange will be presented

Triboengineering problems of lead coolant in innovative fast reactors

International Nuclear Information System (INIS)

Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

2013-01-01

Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

Liquid metal coolant disposal from UKAEA reactors at Dounreay

International Nuclear Information System (INIS)

Adam, E.R.

1997-01-01

As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

Study of the γ decay of high-lying states in 208Pb via inelastic scattering of 17O ions

Directory of Open Access Journals (Sweden)

Crespi F.C.L.

2014-03-01

Full Text Available A measurement of the high-lying states in 208Pb has been made using 17O beams at 20 MeV/u. The gamma decay following inelastic excitation was measured with the detector system AGATA Demonstrator based on segmented HPGe detectors, coupled to an array of large volume LaBr3:Ce scintillators and to an array of Si detectors. Preliminary results in comparison with (γ,γ’ data, for states in the 5-8 MeV energy interval, are presented.

Biological conversion of anglesite (PbSO(4)) and lead waste from spent car batteries to galena (PbS).

Science.gov (United States)

Weijma, Jan; De Hoop, Klaas; Bosma, Wobby; Dijkman, Henk

2002-01-01

Lead paste, a solid mixture containing PbSO(4), PbO(2), PbO/Pb(OH)(2) precipitate, and elemental Pb, is one of the main waste fractions from spent car batteries. Biological sulfidation represents a new process for recovery of lead from this waste. In this process the lead salts in lead paste are converted to galena (PbS) by sulfate-reducing bacteria. This paper investigates a continuous process for sulfidation of anglesite (PbSO(4)), the main constituent of lead paste, and lead paste, consisting of a laboratory-scale gas-lift bioreactor to which a slurry of anglesite or lead paste was supplied. Sulfate or elemental sulfur was added as an additional sulfur source. Hydrogen gas served as an electron donor for the biological reduction of sulfate and elemental sulfur to sulfide by sulfate- and sulfur-reducing bacteria. Anglesite was almost completely converted to galena at a loading rate of 19 kg of PbSO(4) m(-)(3) day(-)(1), producing a sludge of which the crystalline lead phases consisted of >98% PbS (galena) and 1-2% elemental Pb. With lead paste, stable sulfidation rates of up to 17 kg of lead paste m(-)(3) day(-)(1) were demonstrated, producing a sludge of which the crystalline lead phases consisted of an estimated >96% PbS, 1-2% elemental Pb, and 1-2% PbO(2).

A helium-cooled blanket design of the low aspect ratio reactor

International Nuclear Information System (INIS)

Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.

1998-03-01

An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh

Evaluation of lithium alloy anode materials for Li-TiS2 cells

Science.gov (United States)

Huang, C.-K.; Subbarao, S.; Shen, D. H.; Deligiannis, F.; Attia, A.; Halpert, G.

1991-01-01

A study was performed to select candidate lithium alloy anode materials and establish selection criteria. Some of the selected alloy materials were evaluated for their electrochemical properties and performance. This paper describes the criteria for the selection of alloys and the findings of the studies. Li-Si and Li-Cd alloys have been found to be unstable in the EC+2-MeTHF-based electrolyte. The Li-Al alloy system was found to be promising among the alloy systems studied in view of its stability and reversibility. Unfortunately, the large volume changes of LiAl alloys during charge/discharge cycling cause considerable 'exfoliation' of its active mass. This paper also describes ways how to address this problem. The rate of disintegration of this anode would probably be surpressed by the presence of an inert solid solution or a uniform distribution of precipitates within the grains of the active mass. It was discovered that the addition of a small quantity of Mn may improve the mechanical properties of LiAl. In an attempt to reduce the Li-Al alloy vs. Li voltage, it was observed that LiAlPb(0.1)Cd(0.3) material can be cycled at 1.5 mA/sq cm without exfoliation of the active mass.

Microwave-assisted reactive sintering and lithium ion conductivity of Li1.3Al0.3Ti1.7(PO4)3 solid electrolyte

Science.gov (United States)

Hallopeau, Leopold; Bregiroux, Damien; Rousse, Gwenaëlle; Portehault, David; Stevens, Philippe; Toussaint, Gwenaëlle; Laberty-Robert, Christel

2018-02-01

Li1.3Al0.3Ti1.7(PO4)3 (LATP) materials are made of a three-dimensional framework of TiO6 octahedra and PO4 tetrahedra, which provides several positions for Li+ ions. The resulting high ionic conductivity is promising to yield electrolytes for all-solid-state Li-ion batteries. In order to elaborate dense ceramics, conventional sintering methods often use high temperature (≥1000 °C) with long dwelling times (several hours) to achieve high relative density (∼90%). In this work, an innovative synthesis and processing approach is proposed. A fast and easy processing technique called microwave-assisted reactive sintering is used to both synthesize and sinter LATP ceramics with suitable properties in one single step. Pure and crystalline LATP ceramics can be achieved in only 10 min at 890 °C starting from amorphous, compacted LATP's precursors powders. Despite a relative density of 88%, the ionic conductivity measured at ambient temperature (3.15 × 10-4 S cm-1) is among the best reported so far. The study of the activation energy for Li+ conduction confirms the high quality of the ceramic (purity and crystallinity) achieved by using this new approach, thus emphasizing its interest for making ion-conducting ceramics in a simple and fast way.

Condition monitoring of main coolant pumps, Dhruva

International Nuclear Information System (INIS)

Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

2002-01-01

Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

Supercritical CO2 Brayton power cycles for DEMO (demonstration power plant) fusion reactor based on dual coolant lithium lead blanket

International Nuclear Information System (INIS)

Linares, José Ignacio; Cantizano, Alexis; Moratilla, Beatriz Yolanda; Martín-Palacios, Víctor; Batet, Lluis

2016-01-01

This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles as alternative energy conversion systems for a future fusion reactor based on a DCLL (dual coolant lithium-lead) blanket, as prescribed by EUROfusion. The main issue dealt is the optimization of the integration of the different thermal sources with the power cycle in order to achieve the highest electricity production. The analysis includes the assessment of the pumping consumption in the heating and cooling loops, taking into account additional considerations as control issues and integration of thermal energy storage systems. An exergy analysis has been performed in order to understand the behavior of each layout. Up to ten scenarios have been analyzed assessing different locations for thermal sources heat exchangers. Neglecting the worst four scenarios, it is observed less than 2% of variation among the other six ones. One of the best six scenarios clearly stands out over the others due to the location of the thermal sources in a unique island, being this scenario compatible with the control criteria. In this proposal 34.6% of electric efficiency (before the self-consumptions of the reactor but including pumping consumptions and generator efficiency) is achieved. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of DCLL fusion reactor. • Integration of different available thermal sources has been analyzed considering ten scenarios. • Neglecting the four worst scenarios the electricity production varies less than 2%. • Control and energy storage integration issues have been considered in the analysis. • Discarding the vacuum vessel and joining the other sources in an island is proposed.

The effect of coolant quantity on local fuel–coolant interactions in a molten pool

International Nuclear Information System (INIS)

Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

2015-01-01

Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

Quaternary system LiF-LiCl-LiVO3-Li2MoO4

International Nuclear Information System (INIS)

Anipchenko, B.V.; Garkushin, I.K.

2000-01-01

Interactions in the LiF-LiCl-LiVO 3 -Li 2 MoO 4 system are studied by differential thermal analysis. Rate of heating/cooling of the samples comprised 15 Grad/min, mass of sample composed 0.2 g. The system was investigated in the 300-650 Deg C range. X-ray diffraction method was used for determination of purity of the reagents. Composition and temperature of quaternary component eutectics are determined: 16.5 mol. % of LiF, 47.0 mol. % of LiCl, 28.8 mol. % of LiVO 3 , 7.6 mol. % of Li 2 MoO 4 ; 387 Deg C. Mean value of melting enthalpy of quaternary eutectics mixture in the LiF-LiCl-LiVO 3 -Li 2 MoO 4 system on the results of the tests was in the range of 222 kJ/kg [ru

On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

Energy Technology Data Exchange (ETDEWEB)

Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

1994-12-31

Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

International Nuclear Information System (INIS)

Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

1994-01-01

Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

LWR primary coolant pipe rupture test rig

International Nuclear Information System (INIS)

Yoshitoshi, Shyoji

1978-01-01

The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

Science.gov (United States)

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Terahertz-wave differential detection based on simultaneous dual-wavelength up-conversion

Directory of Open Access Journals (Sweden)

Yuma Takida

2017-03-01

Full Text Available We report a terahertz (THz-wave differential detection based on simultaneous dual-wavelength up-conversion in a nonlinear optical MgO:LiNbO3 crystal with optical and electronic THz-wave sources. The broadband parametric gain and noncollinear phase-matching of MgO:LiNbO3 provide efficient conversion from superposed THz waves to spatially distributed near-infrared (NIR beams to function as a dispersive THz-wave spectrometer without any additional dispersive element. We show that the μW-level THz waves from two independent sources, a 0.78-THz injection-seeded THz-wave parametric generator (is-TPG and a 1.14-THz resonant tunneling diode (RTD, are simultaneously up-converted to two NIR waves and then detected with two NIR photodetectors. By applying a balanced detection scheme to this dual-frequency detection, we demonstrate THz-wave differential imaging of maltose and polyethylene pellets in the transmission geometry. This dual-wavelength detection is applicable to more than three frequencies and broadband THz-wave radiation for real-time THz-wave spectroscopic detection and imaging.

The tin-rich copper lithium stannides: Li3Cu6Sn4 and Li2CuSn2

International Nuclear Information System (INIS)

Fuertauer, Siegfried; Flandorfer, Hans; Effenberger, Herta S.

2015-01-01

The Sn rich ternary intermetallic compounds Li 3 Cu 6 Sn 4 (CSD-427097) and Li 2 CuSn 2 (CSD-427098) were synthesized from the pure elements by induction melting and annealing at 400 C. Structural investigations were performed by powder- and single-crystal XRD. Li 3 Cu 6 Sn 4 crystallizes in space group P6/mmm; it is structurally related to but not isotypic with MgFe 6 Ge 6 (a = 5.095(2) Aa, c = 9.524(3) Aa; wR 2 = 0.059; 239 unique F 2 -values, 17 free variables). Li 3 Cu 6 Sn 4 is characterized by two sites with a mixed Cu:Sn occupation. In contrast to all other Cu-Li-Sn compounds known so far, any mixed occupation was found for Cu-Li pairs only. In addition, one Li site is only half occupied. The second Sn rich phase is Li 2 CuSn 2 (space group I4 1 /amd, a = 4.4281(15) Aa, c = 19.416(4) Aa; wR 2 = 0.033; 213 unique F 2 -values, 12 atom free variables); it is the only phase in the Cu-Li-Sn system which is noted for full ordering. Both crystal structures exhibit 3D-networks which host Li atoms in channels. They are important for understanding the lithiation mechanism in Cu-Sn electrodes for Li-ion batteries.

Coolant cleanup system for BWR type reactor

International Nuclear Information System (INIS)

Kinoshita, Shoichiro; Araki, Hidefumi.

1993-01-01

The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Nuclear Fusion Project. Semi-annual report of the Association KfK/EURATOM

International Nuclear Information System (INIS)

Kast, G.

1989-04-01

Report on technology tasks: Blanket design studies; development of computational tools for neutronics; corrosion and fatigue of structural materials in flowing Pb-17Li; tritium extraction from liquid Pb-17Li by the use of solid getters; ceramic breeder materials; high field composite conductors; superconducting poloidal field coil development; design and construction of a poloidal field coil for TORE SUPRA as NET-prototype coil; structural materials fatigue characterization at 4 K; low electrical conductivity structural development; MANET 1; pre- and post-irradiation fatigue properties of 1.4914 martensitic steel; in-pile creep-fatigue testing of type 316 and 1.4914 steel; ceramics for first-wall protection and for RF windows; fatigue under dual beam irradiation; low activation ferritic-martensitic steels; plasma facing components; procedures and tools for structural design evaluation; remote maintenance; mechanical component assembly; handling equipment for in-vessel components; safety aspects of the cryosystem and of superconducting magnets; safety guidelines for the design of NET; generic environmental impact assessment for a fusion facility; large components for plasma exhaust pumping; optimization of cryogenic vacuum pumping of He; solid particle separators for plasma exhaust; plasma exhaust purification; adsorption of DT on heated metal beds other than U; catalyst development for the exhaust purification process; ECRH power sources; vacuum and exhaust performance of NET; NET TF pancake tests; CAD DATA exchange between NET and KfK; catalytic plasma exhaust gas clean-up facility for NET; NET blanket handling device; electrical connectors for remote handling. (orig./HP)

Primary coolant recycling device for FBR type reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru; Tokiwai, Moriyasu

1998-01-01

A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

Fuel coolant interaction experiment by direct electrical heating method

International Nuclear Information System (INIS)

Takeda, Tsuneo; Hirano, Kenmei

1979-01-01

In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

Nitride and carbide thin films as hydrogen permeation barrier on Manet steel

International Nuclear Information System (INIS)

Benamati, G.; Checchetto, R.; Bonelli, M.; Gratton, L.M.; Guzman, L.; Miotello, A.; Terlain, A.

1995-01-01

TiC / TiN bilayers, ∼ 1.2 μm thick, were deposited on Manet II steel by the ion beam assisted deposition technique to investigate the possible use of this ceramic coating as hydrogen barrier. Hydrogen permeation experiments in the temperature range 470-570 K showed indeed that this coating is a very efficient barrier to the hydrogen permeation being able to reduce the hydrogen flux up to two order of magnitude with respect to the uncoated steel. Preliminary compatibility tests between coated Manet II and Pb-17Li showed no attack of Pb-17Li to the steel. (orig.)

Maxwell–Stefan diffusion and dynamical correlation in molten LiF-KF: A molecular dynamics study

Energy Technology Data Exchange (ETDEWEB)

Jain, Richa Naja, E-mail: ltprichanaja@gmail.com; Chakraborty, Brahmananda; Ramaniah, Lavanya M. [High Pressure & Synchrotron Radiation Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai-85 (India)

2016-05-23

In this work our main objective is to compute Dynamical correlations, Onsager coefficients and Maxwell-Stefan (MS) diffusivities for molten salt LiF-KF mixture at various thermodynamic states through Green–Kubo formalism for the first time. The equilibrium molecular dynamics (MD) simulations were performed using BHM potential for LiF–KF mixture. The velocity autocorrelations functions involving Li ions reflect the endurance of cage dynamics or backscattering with temperature. The magnitude of Onsager coefficients for all pairs increases with increase in temperature. Interestingly most of the Onsager coefficients has almost maximum magnitude at the eutectic composition indicating the most dynamic character of the eutectic mixture. MS diffusivity hence diffusion for all ion pairs increases in the system with increasing temperature. Smooth variation of the diffusivity values denies any network formation in the mixture. Also, the striking feature is the noticeable concentration dependence of MS diffusivity between cation-cation pair, Đ{sub Li-K} which remains negative for most of the concentration range but changes sign to become positive for higher LiF concentration. The negative MS diffusivity is acceptable as it satisfies the non-negative entropy constraint governed by 2{sup nd} law of thermodynamics. This high diffusivity also vouches the candidature of molten salt as a coolant.

Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

2016-11-01

Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Hobbins, R.R.

1981-01-01

The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

International Nuclear Information System (INIS)

Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

1976-01-01

In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

Decay analysis of compound nuclei formed in reactions with exotic neutron-rich 9Li projectile and the synthesis of 217At* within the dynamical cluster-decay model

Science.gov (United States)

Kaur, Arshdeep; Kaushal, Pooja; Hemdeep; Gupta, Raj K.

2018-01-01

The decay of various compound nuclei formed via exotic neutron-rich 9Li projectile is studied within the dynamical cluster-decay model (DCM). Following the earlier work of one of us (RKG) and collaborators (M. Kaur et al. (2015) [1]), for an empirically fixed neck-length parameter ΔRemp, the only parameter in the DCM, at a given incident laboratory energy ELab, we are able to fit almost exactly the (total) fusion cross section σfus =∑x=16σxn for 9Li projectile on 208Pb and other targets, with σfus depending strongly on the target mass of the most abundant isotope and its (magic) shell structure. This result shows the predictable nature of the DCM. The neck-length parameter ΔRemp is fixed empirically for the decay of 217At* formed in 9Li + 208Pb reaction at a fixed laboratory energy ELab, and then the total fusion cross section σfus calculated for all other reactions using 9Li as a projectile on different targets. Apparently, this procedure could be used to predict σfus for 9Li-induced reactions where experimental data are not available. Furthermore, optimum choice of "cold" target-projectile combinations, forming "hot" compact configurations, are predicted for the synthesis of compound nucleus 217At* with 8Li + 209Pb as one of the target-projectile combination, or another (t , p) combination 48Ca + 169Tb, with a doubly magic 48Ca, as the best possibility.

Ultrasonic-energy enhance the ionic liquid-based dual microextraction to preconcentrate the lead in ground and stored rain water samples as compared to conventional shaking method.

Science.gov (United States)

Nizamani, Sooraj; Kazi, Tasneem G; Afridi, Hassan I

2018-01-01

An efficient preconcentration technique based on ultrasonic-assisted ionic liquid-based dual microextraction (UA-ILDµE) method has been developed to preconcentrate the lead (Pb +2 ) in ground and stored rain water. In the current proposed method, Pb +2 was complexed with a chelating agent (dithizone), whereas an ionic liquid (1-butyl-3-methylimidazolium hexafluorophosphate) was used for extraction purpose. The ultrasonic irradiation and electrical shaking system were applied to enhance the dispersion and extraction of Pb +2 complex in aqueous samples. For second phase, dual microextraction (DµE phase), the enriched Pb +2 complex in ionic liquid, extracted back into the acidic aqueous solution and finally determined by flame atomic absorption spectrometry. Some major analytical parameters that influenced the extraction efficiency of developed method, such as pH, concentration of ligand, volume of ionic liquid and samples, time of shaking in thermostatic electrical shaker and ultrasonic bath, effect of back extracting HNO 3 volume, matrix effect, centrifugation time and rate were optimized. At the sample volume of 25mL, the calculated preconcentration factor was 62.2. The limit of detection of proposed procedure for Pb +2 ions was found to be 0.54μgL -1 . The validation of developed method was performed by the analysis of certified sample of water SRM 1643e and standard addition method in a real water sample. The extraction recovery of Pb +2 was enhanced≥2% with shaking time of 80s in ultrasonic bath as compared to used thermostatic electrical shaker, where for optimum recovery up to 10min was required. The developed procedure was successfully used for the enrichment of Pb +2 in ground and stored rain water (surface water) samples of an endemic region of Pakistan. The resulted data indicated that the ground water samples were highly contaminated with Pb +2 , while some of the surface water samples were also have higher values of Pb +2 than permissible limit of

Bioaccessibility of U, Th and Pb in particulate matter from an abandoned uranium mine

Science.gov (United States)

Millward, Geoffrey; Foulkes, Michael; Henderson, Sam; Blake, William

2016-04-01

Currently, there are approximately 150 uranium mines in Europe at various stages of either operation, development, decommissioning, restoration or abandonment (wise-uranium.com). The particulate matter comprising the mounds of waste rock and mill tailings poses a risk to human health through the inadvertent ingestion of particles contaminated with uranium and thorium, and their decay products, which exposes recipients to the dual toxicity of heavy elements and their radioactive emissions. We investigated the bioaccessibility of 238U, 232Th and 206,214,210Pb in particulate samples taken from a contaminated, abandoned uranium mine in South West England. Sampling included a mine shaft, dressing floor and waste heap, as well as soils from a field used for grazing. The contaminants were extracted using the in-vitro Unified Bioaccessibility Research Group of Europe Method (UBM) in order to mimic the digestion processes in the human stomach (STOM) and the combined stomach and gastrointestinal tract (STOM+INT). Analyses of concentrations of U, Th and Pb in the extracts were by ICP-MS and the activity concentrations of radionuclides were determined on the same particles, before and after extraction, using gamma spectroscopy. 'Total' concentrations of U, Th and Pb for all samples were in the range 57 to 16,200, 0.28 to 3.8 and 69 to 4750 mg kg-1, respectively. For U and Pb the concentrations in the STOM fraction were lower than the total and STOM+INT fractions were even lower. However, for Th the STOM+INT fractions were higher than the STOM due to the presence of Th carbonate species within the gastrointestinal fluid. Activity concentrations for 214Pb and 210Pb, including total, STOM and STOM+INT, were in the range 180 to samples were 39% and 8% in the STOM and STOM+INT, respectively, whereas the respective BAFs for 232Th were 3% and 9%. For stable 206Pb the STOM and STOM+INT BAFs were 16% and 3% for the most contaminated samples, whereas those from the field had 44% in the

Zero waste machine coolant management strategy at Los Alamos National Laboratory

International Nuclear Information System (INIS)

Carlson, B.; Algarra, F.; Wilburn, D.

1998-01-01

Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extended coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98

Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

International Nuclear Information System (INIS)

Asher, R.C.; Davies, D.; Jones, P.G.

1976-12-01

A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

Energy Technology Data Exchange (ETDEWEB)

Carlsson, Johan

2003-03-01

in the risers of the reactor vessel. For a pool type design, this approach has advantages in the case of heat-exchanger tube failures, particularly if water is used as the secondary fluid. This is because a leakage of water from the secondary circuit into the Pb/Bi-cooled primary circuit leads to upward sweeping of steam bubbles, which would collect in the gas plenum. In the case of heat exchangers in the downcomer steam bubbles may be dragged into the ADS core and add reactivity. Bypass routes are employed to increase the flow speed in loss-of-flow events for this design. It is shown that the 200 MW{sub th} accelerator-driven system with heat-exchangers in the riser copes reasonably well with both a loss-of-flow accident with the beam on and an unprotected loss-of-heat-sink accident. For a total-loss-of-power (station blackout) and an immediate beam-stop the core outlet temperature peaks at 680 K. After a combined loss-of-flow and loss-of-heat-sink accident the beam should be shut off within 4 minutes to avoid exceeding the ASME level D of 977 K, and within 8 minutes to avoid fast creep. Assuming the same core inlet temperature, both the reactor design with heat-exchanger in the risers and the downcomers have similar temperature evolutions after a total-loss-of power accident. A large accelerator-driven system of 800 MW{sub th} with a 17 m tall vessel may eventually become a standard size. For this higher power ADS, the location of the heat-exchangers has greater impact on the natural convection capability. This is due to that larger heat exchangers have more influence on the distance between the thermal centers during a loss of- flow accident. The design with heat-exchangers in the downcomers, the long-term vessel temperature peaks at 996 K during a loss-of-flow accident with the beam on. This does not pose a threat of creep rupture for the vessel. However, the location of the heat-exchangers in the downcomers will probably require secondary coolant other than

Inherent Safety Features and Passive Prevention Approaches for Pb/Bi-cooled Accelerator-Driven Systems

International Nuclear Information System (INIS)

Carlsson, Johan

2003-03-01

risers of the reactor vessel. For a pool type design, this approach has advantages in the case of heat-exchanger tube failures, particularly if water is used as the secondary fluid. This is because a leakage of water from the secondary circuit into the Pb/Bi-cooled primary circuit leads to upward sweeping of steam bubbles, which would collect in the gas plenum. In the case of heat exchangers in the downcomer steam bubbles may be dragged into the ADS core and add reactivity. Bypass routes are employed to increase the flow speed in loss-of-flow events for this design. It is shown that the 200 MW th accelerator-driven system with heat-exchangers in the riser copes reasonably well with both a loss-of-flow accident with the beam on and an unprotected loss-of-heat-sink accident. For a total-loss-of-power (station blackout) and an immediate beam-stop the core outlet temperature peaks at 680 K. After a combined loss-of-flow and loss-of-heat-sink accident the beam should be shut off within 4 minutes to avoid exceeding the ASME level D of 977 K, and within 8 minutes to avoid fast creep. Assuming the same core inlet temperature, both the reactor design with heat-exchanger in the risers and the downcomers have similar temperature evolutions after a total-loss-of power accident. A large accelerator-driven system of 800 MW th with a 17 m tall vessel may eventually become a standard size. For this higher power ADS, the location of the heat-exchangers has greater impact on the natural convection capability. This is due to that larger heat exchangers have more influence on the distance between the thermal centers during a loss of- flow accident. The design with heat-exchangers in the downcomers, the long-term vessel temperature peaks at 996 K during a loss-of-flow accident with the beam on. This does not pose a threat of creep rupture for the vessel. However, the location of the heat-exchangers in the downcomers will probably require secondary coolant other than water, like for example

ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

Science.gov (United States)

Morrison, Russell H.; Holt, Mike

2005-01-01

The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

Full reactor coolant system chemical decontamination qualification programs

Energy Technology Data Exchange (ETDEWEB)

Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-03-01

Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

Nuclear reactor of pressurized liquid coolant type

International Nuclear Information System (INIS)

Costes, D.

1976-01-01

The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

On the use of tin-lithium alloys as breeder material for blankets of fusion power plants

International Nuclear Information System (INIS)

Fuetterer, M.A.; Aiello, G.; Barbier, F.; Giancarli, L.; Poitevin, Y.; Sardain, P.; Szczepanski, J.; Li Puma, A.; Ruvutuso, G.; Vella, G.

2000-01-01

Tin-lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead-lithium (Pb-17Li) by a suitable tin-lithium alloy: (i) for the European water-cooled Pb-17Li (WCLL) blanket concept with reduced activation ferritic-martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiC f /SiC as the structural material. It was found that in none of these blankets Sn-Li alloys would lead to significant advantages, in particular due to the low tritium breeding capability. Only in forced convection cooled divertors with W-alloy structure, Sn-Li alloys would be slightly more favorable. It is concluded that Sn-Li alloys are only advantageous in free surface cooled reactor internals, as this would make maximum use of the principal advantage of Sn-Li, i.e., the low vapor pressure

Steam as turbine blade coolant: Experimental data generation

Energy Technology Data Exchange (ETDEWEB)

Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

1995-10-01

Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

Stress Analysis of Fuel Rod under Axial Coolant Flow

Energy Technology Data Exchange (ETDEWEB)

Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

2010-05-15

A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

Stress Analysis of Fuel Rod under Axial Coolant Flow

International Nuclear Information System (INIS)

Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

2010-01-01

A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

Dual functional MoS2/graphene interlayer as an efficient polysulfide barrier for advanced lithium-sulfur batteries

International Nuclear Information System (INIS)

Guo, Pengqian; Liu, Dequan; Liu, Zhengjiao; Shang, Xiaonan; Liu, Qiming; He, Deyan

2017-01-01

Highlights: •Dual functional MoS 2 /graphene interlayer was first used as an efficient polysulfide-trapping shield for lithium-sulfur batteries. •MoS 2 /graphene interlayer shows strong chemical interactions with LiPSs. •MoS 2 /graphene interlayer forms a 3D network to facilitate electron and ion transfer during the discharge-charge processes. •The resultant lithium-sulfur batteries exhibit a superior rate capacity and improved cycling capacity. -- Abstract: A dual functional interlayer consisted of composited two-dimensional MoS 2 and graphene has been developed as an efficient polysulfide barrier for lithium-sulfur batteries (LSBs). With such a configuration, LSBs show a superior rate capacity and improved cycling capacity. The excellent electrochemical performance can be attributed to the strong bonding interactions between the MoS 2 /graphene interlayer and the formed lithium polysulfides (LiPSs) as well as the good electrical conductivity of the MoS 2 /graphene composite. The MoS 2 /graphene interlayer can physically block LiPSs by the graphene nanosheets and chemically suppress the dissolution of LiPSs by the polar MoS 2 nanoflowers. Such a dual functional interlayer further provides a good contact with the surface of the sulfur cathode, acts as an upper current collector and greatly improves the sulfur utilization and the rate capability of LSBs.

Device for preventing coolant in a reactor from being lost

International Nuclear Information System (INIS)

Maruyama, Hiromi; Matsumoto, Tomoyuki.

1975-01-01

Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

2009-01-01

A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

Modelling the power conversion unit of a generic nuclear fusion plant, with a dual coolant blanket and a supercritical CO2 power cycle, by means of RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Batet, L.

2015-07-01

In the framework of the Spanish fusion program TECNO-FUS, a dual coolant blanket design was proposed for DEMO. A generic power conversion system (supercritical recompression CO2 cycle) based on this proposal has been simulated using RELAP5-3D, a multipurpose system thermal-hydraulic code developed by the Idaho National Laboratory (USA). The code allows the dynamic simulation of thermal-hydraulic systems, including the control features. A model has been set up by assembling the available RELAP5-3D components: pipe, branch, pump, compressor, turbine, etc. Thermal fluxes between fluids in heat exchangers are simulated by means of heat structures, which are used as well to simulate the heating from plasma. A number of control features have been designed for the simulated plant, and their parameters have been adjusted. The code is then able to simulate robustly the dynamics of the system with a few boundary conditions. This paper exemplifies the usefulness of the code and model to understand the behavior of the plant and to perform sensitivity analyses of the control parameters or other design features. (Author)

Simulation of steam explosion in stratified melt-coolant configuration

International Nuclear Information System (INIS)

Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

2016-01-01

Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

Mathematical model of the reactor coolant pump

International Nuclear Information System (INIS)

Kozuh, M.

1989-01-01

The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

The role of lead oxide on structural and physical properties of lithium diborate glasses

International Nuclear Information System (INIS)

Kashif, I.; Abd El-Maboud, A.; El-said, R.; Sakr, E.M.; Soliman, A.A.

2012-01-01

Highlights: ► We prepare Li 2 B 4 O 7 –Pb 3 O 4 glass samples by the quenched method as bulk. ► The effects of substitution Li 2 B 4 O 7 with Pb 3 O 4 in glass composition are studied. ► The structure, density, Vickers hardness, glass transition temperature and electrical properties have been influenced by these substitution. - Abstract: Pseudo-binary (100 − x)Li 2 B 4 O 7 –xPb 3 O 4 , with x = 0–70 mol% PbO have been prepared and their properties investigated. The glass transition temperature, density and molar volume have been determined as a function of composition. The values of T g and the molar volume decrease non-linearly while the density increases as the Pb 3 O 4 content is raised. Infrared spectra of the glasses reveal that a strong network consisting of diborate units breaks up by the addition of Pb 3 O 4 . The absorption bands below 620 cm −1 show that PbO is one of the network formers of the glasses 70 ⩾ Pb 3 O 4 ⩾ 10; as they can be associated with vibrations of (PbO 4 ) 2− grouping. PbO plays a dual role in the glass network. The calculated values of N 4 [the fraction of borons which are tetrahedral] slightly decrease with PbO content up to 30 mol% and then increase with Pb 3 O 4 content up to 50 mol%, then followed by a decrease as the Pb 3 O 4 content rises further. The Vickers hardness of the glasses varies as a function of the PbO content in the same manner as the variation of N 4 . The dc conductivity decreases with the Pb 3 O 4 concentration up to about 30 mol% and then increases thereafter.

On-Line Coolant Chemistry Analysis

International Nuclear Information System (INIS)

LM Bachman

2006-01-01

Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level

WITAMIR-I: A tandem mirror power reactor

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Beyer, J.B.

1983-01-01

A conceptual design of a near term commercial tandem mirror power reactor will be presented. The basic configuration utilizes Yin-yang minimum B end plugs with inboard thermal barriers, which are pumped by neutral beam injection. The maximum magnetic fields are 6.1 T, 8.1 T and 15 T in the central cell, Yin-yang, and thermal barrier magnets, respectively. The blanket utilizes Pb 83 Li 17 as the coolant and breeder, and HT-9 as the structural material. This configuration yields a high energy multiplication (1.37), a sufficient tritium breeding ratio (1.07) and has a major advantage with respect to maintenance. A single stage direct convertor is used at one end and an electron thermal dump at the other end. The plasma Q is 28 at a fusion power level of 3000 MWsub(th); the net electrical output is 1530 MWe and the overall efficiency is 39%. Cost estimates indicate that WITAMIR-I is competitive with recent tokamak power reactor designs. (author)

Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor

Directory of Open Access Journals (Sweden)

Bong Guen Hong

2018-02-01

Full Text Available A configuration of a fusion-driven transmutation reactor with a low aspect ratio tokamak-type neutron source was determined in a self-consistent manner by using coupled analysis of tokamak systems and neutron transport. We investigated the impact of blanket configuration on the characteristics of a fusion-driven transmutation reactor. It was shown that by merging the TRU burning blanket and tritium breeding blanket, which uses PbLi as the tritium breeding material and as coolant, effective transmutation is possible. The TRU transmutation capability can be improved with a reduced blanket thickness, and fast fluence at the first wall can be reduced.  Article History: Received: July 10th 2017; Received: Dec 17th 2017; Accepted: February 2nd 2018; Available online How to Cite This Article: Hong, B.G. (2018 Impact of Blanket Configuration on the Design of a Fusion-Driven Transmutation Reactor. International Journal of Renewable Energy Development, 7(1, 65-70. https://doi.org/10.14710/ijred.7.1.65-70

Water-cooled lithium-lead box-shaped blanket concept for Demo: thermo-mechanical optimization and manufacturing sequence proposal

International Nuclear Information System (INIS)

Baraer, L.; Dinot, N.; Giancarli, L.; Proust, E.; Salavy, J.F.; Severi, Y.; Quintric-Bossy, J.

1992-01-01

The development of the water-cooled lithium-lead box-shaped blanket concept for DEMO has now reached the stage of thermo-mechanical optimization. In the previous design phases the preliminary dimensioning of the cooling circuit has permitted to define the water proportions required in the breeder region and to demonstrate, after a minimization of steel proportion and thicknesses, that this concept could reach tritium breeding self-sufficiency. In the present analysis the location of the coolant pipes has been optimized for the whole equatorial plane cross-section of both inboard and outboard segments in order to maintain the maximum Pb-17Li/steel interface temperature below 480 deg C and to minimize the thermal gradients along the steel structures. The consequent thermo-mechanical analysis has shown that the thermal stresses always remain below the allowable limits. Segment fabricability and removal are the next design issues to be analyzed. Within this strategy, a first manufactury sequence for the outboard segment is proposed

Nitride and carbide thin films as hydrogen permeation barrier on MANET steel

International Nuclear Information System (INIS)

Benamati, G.

1994-01-01

TiC/TiN bilayers, - 1.2 μm thick, were deposited on Manet II steel by the ion beam assisted deposition technique to investigate the possible use of this ceramic coating as hydrogen barrier. Hydrogen permeation experiments in the temperature range 470-570 K showed indeed that this coating is a very efficient barrier to the hydrogen permeation being able to reduce the hydrogen flux up to two order of magnitude with respect to the uncoated steel. Preliminary compatibility tests between coated Manet II and Pb-17Li showed no attack of Pb-17Li to the steel. (author) 8 refs.; 2 figs.; 1 tab

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

International Nuclear Information System (INIS)

Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

2000-01-01

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

Electrochemical Behavior of LiBr, LiI, and Li2Se in LiCl Molten Salt

International Nuclear Information System (INIS)

Choi, In Kyu; Do, Jae Bum; Hong, Sun Seok; Seo, Chung Seok

2006-03-01

The effect of fission products on the electrolytic reduction of uranium oxide has been studied. It has been reported that volatile fission products, such as Br, I, and Se, react with Li metal which is a reductant in the process to give LiBr, LiI, and Li 2 Se. These compounds are dissociated as corresponding anions and cations in the LiCl molten salt at 650 .deg. C. In this experiment, oxidation and reduction reaction of 3wt% of each compound in LiCl molten salt were investigated by cyclic voltammetry. For LiBr, redox reactions of cation and anion were reversible, while redox reactions of Li + and I - were irreversible. For Li 2 Se, about half of the produced Li metal was disappeared at the cathode and two anodic current curves were appeared. After the cyclic voltammetric measurements for each compound, chronopotentiometric experiment was carried out for one hour with 100 - 400 mA. After the electrolysis, no compounds gave Li metal in the porous MgO filter in which Li metal was produced at the cathode. However, LiCl salt was covered with Br 2 for LiBr electrolysis. Dark red color of Br 2 was easily removed by water. For LiI electrolysis, salt gave black color and I 2 was deposited on the Pt anode. For Li 2 Se electrolysis, black fine powders were precipitated in the salt. After the separation and dryness of the precipitates, it was analyzed with XRD and it turned out PtSe 2 . From the electrochemical experimental results, it was concluded that these compounds may affect the electrolytic reduction process of uranium oxide in the spent fuel

Fuel-Coolant Interactions: Visualization and Mixing Measurements

International Nuclear Information System (INIS)

Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

2002-01-01

Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

The installation welding of pressure water reactor coolant piping

International Nuclear Information System (INIS)

Deng Feng

2010-01-01

Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

International Nuclear Information System (INIS)

Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

2003-01-01

This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

International Nuclear Information System (INIS)

Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

2010-01-01

Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

All-inorganic CsPbBr3 perovskite quantum dots embedded in dual-mesoporous silica with moisture resistance for two-photon-pumped plasmonic nanoLasers.

Science.gov (United States)

Chen, Yu; Yu, Minghuai; Ye, Shuai; Song, Jun; Qu, Junle

2018-04-05

Lead halide perovskite nanocrystals with efficient two-photon absorption and ease of achieving population inversion have been recognized as good candidates to achieve frequency up-conversion for biophotonics applications, but suffer from the limitation of the miniaturization of the device and its corresponding poor stability when exposed to atmospheric moisture. Here we demonstrate the miniaturization of plasmonic nanolasers via embedding perovskite quantum dots (QDs) in rationally designed dual-mesoporous silica with gold nanocore. The nanocomposite supports resonant surface plasmon-polaritons (SPPs), which overlap both spatially and spectrally with the CsPbBr3 QDs. The outcoupling between surface plasmon oscillations and photonics modes within a wavelength range completely overcomes the loss of localized surface plasmons, and finally contributes to a novel application of two-photon-pumped nanolasers. Large optical gain under two-photon excitation was observed as a result of resonant energy transfer from excited perovskite QDs to surface plasmon oscillations and stimulated emission of surface plasmons in a luminous mode. The outmost organic-inorganic hybrid shells of the dual-mesoporous silica nanocomposites act as a protective layer of the perovskite QDs against water and endow the nanocomposites with superhydrophobicity. This work provides an alternative inspiration for the design of new two-photon pumped nanolasers.

Investigation of tritium and 233U breeding in a fission-fusion hybrid reactor fuelling with ThO2

International Nuclear Information System (INIS)

Yildiz, K.; Sahin, S.; Sahin, H. M.; Acir, A.; Yalcin, S.; Altinok, T.; Bayrak, M.; Alkan, M.; Durukan, O.

2007-01-01

In the world, thorium reserves are three times more than natural Uranium reserves. It is planned in the near future that nuclear reactors will use thorium as a fuel. Thorium is not a fissile isotope because it doesn't make fission with thermal neutrons so it could be converted to 2 33U isotope which has very high quality fission cross-section with thermal neutrons. 2 33U isotope can be used in present LWRs as an enrichment fuel. In the fusion reactors, tritium is the most important fossil fuel. Because tritium is not natural isotope, it has to be produced in the reactor. The purpose of this work is to investigate the tritium and 2 33U breeding in a fission-fusion hybrid reactor fuelling with ThO 2 for Δt=10 days during a reactor operation period in five years. The neutronic analysis is performed on an experimental hybrid blanket geometry. In the center of the hybrid blanket, there is a line neutron source in a cylindrical cavity, which simulates the fusion plasma chamber where high energy neutrons (14.1 MeV) are produced. The conventional fusion reaction delivers the external neutron source for blankets following, 2 D + 3 T →? 4 He (3.5 MeV) + n (14.1 MeV). (1) The fuel zone made up of natural-ThO 2 fuel and it is cooled with different coolants. In this work, five different moderator materials, which are Li 2 BeF 4 , LiF-NaF-BeF 2 , Li 2 0Sn 8 0, natural Lithium and Li 1 7Pb 8 3, are used as coolants. The radial reflector, called tritium breeding zones, is made of different Lithium compounds and graphite in sandwich structure. In the work, eight different Lithium compounds were used as tritium breeders in the tritium breeding zones, which are Li 3 N, Li 2 O, Li 2 O 2 , Li 2 TiO 3 , Li 4 SiO 3 , Li 2 ZrO 3 , LiBr and LiH. Neutron transport calculations are conducted in spherical geometry with the help of SCALE4.4A SYSTEM by solving the Boltzmann transport equation with code CSAS and XSDRNPM, under consideration of unresolved and resolved resonances, in S 8 -P 3

Electroweak bosons in Pb+Pb and $p$+Pb collisions

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00356981; The ATLAS collaboration

2016-01-01

Electroweak boson ( W , Z , γ ) measurements in Pb+Pb collisions at sNN=2.76 TeV and in p +Pb collisions at sNN=5.02 TeV are presented with the ATLAS detector at the LHC. In Pb+Pb, electroweak boson yields are shown to be independent of centrality. Differential measurements in absolute pseudorapidity are used to investigate nuclear effects to the free-proton parton distribution function (PDF). The distributions lack the experimental precision to unambiguously identify the presence of nuclear modifications. In p +Pb, the Z boson cross section is measured as a function of center-of-mass rapidity yZ⁎ and the momentum fraction of the lead-going parton (Bjorken xPb ). The distributions are asymmetric and model predictions underestimate the data at large xPb . The overall shape is best described by including nuclear effects. The differential cross section is also measured in different centrality classes and shows evidence of spatially-dependent nuclear PDFs. The Z boson production yields are measured as a functi...

Petrogenesis of Variscan lamproites of the Bohemian Massif

Czech Academy of Sciences Publication Activity Database

Krmíček, Lukáš; Romer, R. L.; Glodny, J.

2015-01-01

Roč. 17, - (2015) ISSN 1607-7962. [European Geosciences Union General Assembly. 12.04.2015-17.04.2015, Vienna] Institutional support: RVO:67985831 Keywords : orogenic lamproites * mineralogy * geochemistry * Sr-Nd-Pb-Li isotopes Subject RIV: DD - Geochemistry

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Qualitative analysis of As, Ba, Cd, Cr, Zn, Fe, Mn, K, Hg, Pb y Cu, as constituents of Amatitlan Lake sediment by XRF

International Nuclear Information System (INIS)

Beltran, P.A.E.; Morales, E.A.

1987-10-01

Samples of fifteen sampling points were analyzed. Molybdenum X-ray tube with secondary excitation assembly, SiLi detector and deconvolution software AXIL were employed; self-standardization method based upon incoherent ratio was used for quantitative analysis of some elements. Ca, P, S, Ti, Mn, Fe, Cr, Zn, Cu, Ni, Ga, As, Pb, Ge, Sr and Pb, were found. As, Pb and Cu concentrations lower than 109 mg/lt, 119 mg/lt, and 500mg/lt, respectively, were measured. Hg was not detected. (author)

Revision of the Li13Si4 structure.

Science.gov (United States)

Zeilinger, Michael; Fässler, Thomas F

2013-11-06

Besides Li17Si4, Li16.42Si4, and Li15Si4, another lithium-rich representative in the Li-Si system is the phase Li13Si4 (trideca-lithium tetra-silicide), the structure of which has been determined previously [Frank et al. (1975 ▶). Z. Naturforsch. Teil B, 30, 10-13]. A careful analysis of X-ray diffraction patterns of Li13Si4 revealed discrepancies between experimentally observed and calculated Bragg positions. Therefore, we redetermined the structure of Li13Si4 on the basis of single-crystal X-ray diffraction data. Compared to the previous structure report, decisive differences are (i) the introduction of a split position for one Li site [occupancy ratio 0.838 (7):0.162 (7)], (ii) the anisotropic refinement of atomic displacement parameters for all atoms, and (iii) a high accuracy of atom positions and unit-cell parameters. The asymmetric unit of Li13Si4 contains two Si and seven Li atoms. Except for one Li atom situated on a site with symmetry 2/m, all other atoms are on mirror planes. The structure consists of isolated Si atoms as well as Si-Si dumbbells surrounded by Li atoms. Each Si atom is either 12- or 13-coordinated. The isolated Si atoms are situated in the ab plane at z = 0 and are strictly separated from the Si-Si dumbbells at z = 0.5.

Optical model studies of 6Li elastic scattering at 156 MeV

International Nuclear Information System (INIS)

Cook, J.; Gils, H.J.; Rebel, H.; Klewe-Nebenius, H.

1981-11-01

Differential cross sections for 6 Li elastic scattering at 156 MeV from 12 C, 40 Ca 90 Zr and 208 Pb are presented. The sensitivity to various potential forms is established by using Saxon Woods, Saxon-Woods-squred, density independent and density dependent folded potentials. The extent to which the experimental data determine the potentials and related quantities is discussed. (orig.) [de

Prussian Blue Mg-Li Hybrid Batteries.

Science.gov (United States)

Sun, Xiaoqi; Duffort, Victor; Nazar, Linda F

2016-08-01