Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

International Nuclear Information System (INIS)

2002-06-01

All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

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

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.)

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.)

A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

Energy Technology Data Exchange (ETDEWEB)

Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

1998-03-01

The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

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.

Organic coolants and their applications to fusion reactors

International Nuclear Information System (INIS)

Gierszewski, P.; Hollies, B.

1986-08-01

Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

International Nuclear Information System (INIS)

Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

1982-01-01

Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

Progress in design and development of series liquid lithium-lead expeirmental loops in China

International Nuclear Information System (INIS)

Wu Yican; Huang Qunying; Zhu Zhiqiang; Gao Sheng; Song Yong; Li Chunjing; Peng Lei; Liu Shaojun; Wu qingsheng; Liu Songlin; Chen Hongli; Bai Yunqing; Jin Ming; Lv Ruojun; Wang Weihua; Guo Zhihui; Chen Yaping; Ling Xinzhen; Zhang Maolian; Wang Yongliang; Wu Zhaoyang; Wang Hongyan

2009-01-01

Liquid LiPb (lithium-lead) experimental loops are the important platforms to investigate the key technologies of liquid LiPb breeder blankets for fusion reactors. Based on the development strategy for liquid LiPb breeder blankets, the technologies development of liquid LiPb experimental loops have been explored by the FDS Team for years, and a series of LiPb experimental loops named DRAGON have been designed and developed, which have independence intellectual property and multi-functional parameters. In this paper, the development route suggestion of Chinese LiPb experimental loops was elaborated, and some information for the senes experimental loops were introduced, such as the design principles, structural features, functions and related experimental researches, etc. (authors)

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.

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

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

Liquid Lithium Wall Experiments in CDX-U

International Nuclear Information System (INIS)

Doerner, R.; Kaita, R.; Majeski, R.; Luckhardt, S.

1999-01-01

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. Sputtering and erosion tests are currently underway in the PISCES device at the University of California at San Diego (UCSD). To complement this effort, plasma interaction questions in a toroidal plasma geometry will be addressed by a proposed new groundbreaking experiment in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The CDX-U plasma is intensely heated and well diagnosed, and an extensive liquid lithium plasma-facing surface will be used for the first time with a toroidal plasma. Since CDX-U is a small ST, only approximately1 liter or less of lithium is required to produce a toroidal liquid lithium limiter target, leading to a quick and cost-effective experiment

Large lithium loop experience

International Nuclear Information System (INIS)

Kolowith, R.; Owen, T.J.; Berg, J.D.; Atwood, J.M.

1981-10-01

An engineering design and operating experience of a large, isothermal, lithium-coolant test loop are presented. This liquid metal coolant loop is called the Experimental Lithium System (ELS) and has operated safely and reliably for over 6500 hours through September 1981. The loop is used for full-scale testing of components for the Fusion Materials Irradiation Test (FMIT) Facility. Main system parameters include coolant temperatures to 430 0 C and flow to 0.038 m 3 /s (600 gal/min). Performance of the main pump, vacuum system, and control system is discussed. Unique test capabilities of the ELS are also discussed

Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

International Nuclear Information System (INIS)

Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

2014-01-01

The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

Thermodynamic properties and equation of state of liquid lead and lead bismuth eutectic

Science.gov (United States)

Sobolev, V. P.; Schuurmans, P.; Benamati, G.

2008-06-01

Since the 1950s, liquid lead (Pb) and lead-bismuth eutectic (Pb-Bi) have been studied in the USA, Canada and in the former-USSR as potential coolants for nuclear installations due to their very attractive thermophysical and neutronic properties. However, experimental data on the thermal properties of these coolants in the temperature range of interest are still incomplete and often contradictory. This makes it very difficult to perform design calculations and to analyse the normal and abnormal behaviour of nuclear installations where these coolants are expected to be used. Recently, a compilation of heavy liquid metal (HLM) properties along with recommendations for its use was prepared by the OECD/NEA Working Party on Fuel Cycle (WPFC) Expert Group on Lead-Bismuth Eutectic Technology. A brief review of this compilation and some new data are presented in this article. A set of correlations for the temperature dependence of the main thermodynamic properties of Pb and Pb-Bi(e) at normal pressure, and a set of simplified thermal and caloric equations of state for the liquid phase are proposed.

Basic principles of lead and lead-bismuth eutectic application in blanket of fusion reactors

International Nuclear Information System (INIS)

Beznosov, A.V.; Pinaev, S.S.; Muraviev, E.V.; Romanov, P.V.

2005-01-01

High magnetohydrodynamic pressure drop is an important issue for liquid metal blanket concepts. To decrease magnetohydrodynamic resistance authors propose to form insulating coatings on internal surface of blanket ducts at any moment of fusion reactor exploitation. It may be achieved easily if lead or lead-bismuth eutectic is used and technology of oxidative potential handling is applied. A number of experiments carried out in NNSTU show the availability of the proposed technology. It bases on formation of the insulating coatings that consist of the oxides of components of the structural materials and of the coolant components. In-situ value of the insulating coatings characteristics ρδ is ∼ 10 -5 Ohm·m 2 for steels and 5,0x10 -6 - 5,0x10 -5 Ohm·m 2 for vanadium alloys. Thermal cycling is possible during exploitation of a blanket. The experimental research of the insulating coatings properties during thermal cycling have shown that the coatings formed into the lead and lead-bismuth coolants save there insulating properties. Experience of many years is an undoubted advantage of the lead-bismuth coolant and less of the lead coolant in comparison with lithium. Russian Federation possesses of experience of exploitation of the research and industrial facilities, of experience of creation of the pumps, steamgenerators and equipment with heavy liquid metal coolants. The unique experience of designing, assembling and exploitation of the fission reactors with lead-bismuth coolant is also available. The problem of technology of lead and lead-bismuth coolants for power high temperature radioactive facilities has been solved. Accidents, emergency situations such as leakage of steamgenerators or depressurization of gas system in facilities with lead and lead-bismuth coolants have been explored and suppressed. (author)

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

Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

Science.gov (United States)

Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

2017-11-01

The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification

Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

International Nuclear Information System (INIS)

Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

2015-01-01

At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

Corrosion by liquid lead and lead-bismuth: experimental results review and analysis

Energy Technology Data Exchange (ETDEWEB)

Zhang, Jinsuo [Los Alamos National Laboratory

2008-01-01

Liquid metal technologies for liquid lead and lead-bismuth alloy are under wide investigation and development for advanced nuclear energy systems and waste transmutation systems. Material corrosion is one of the main issues studied a lot recently in the development of the liquid metal technology. This study reviews corrosion by liquid lead and lead bismuth, including the corrosion mechanisms, corrosion inhibitor and the formation of the protective oxide layer. The available experimental data are analyzed by using a corrosion model in which the oxidation and scale removal are coupled. Based on the model, long-term behaviors of steels in liquid lead and lead-bismuth are predictable. This report provides information for the selection of structural materials for typical nuclear reactor coolant systems when selecting liquid lead or lead bismuth as heat transfer media.

Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

International Nuclear Information System (INIS)

2005-01-01

facility IFMIF with lithium target. Also, the results of research work performed with the purpose of supporting innovative technologies aimed at the use of liquid metals, e.g., heat pipe applied to nuclear power facilities, lyophobic-capillary porous systems, and others are presented here. As part of the technical program, meeting participants visited test facilities for thermal hydraulics research with liquid metal coolants (sodium, NaK, lead-bismuth, lithium) in the SSC RF IPPE

Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-07-01

facility IFMIF with lithium target. Also, the results of research work performed with the purpose of supporting innovative technologies aimed at the use of liquid metals, e.g., heat pipe applied to nuclear power facilities, lyophobic-capillary porous systems, and others are presented here. As part of the technical program, meeting participants visited test facilities for thermal hydraulics research with liquid metal coolants (sodium, NaK, lead-bismuth, lithium) in the SSC RF IPPE.

Wetting properties of liquid lithium on lithium compounds

Energy Technology Data Exchange (ETDEWEB)

Krat, S.A., E-mail: stepan.krat@gmail.com [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States); National Research Nuclear University MEPhI, Moscow (Russian Federation); Popkov, A.S. [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States); National Research Nuclear University MEPhI, Moscow (Russian Federation); Gasparyan, Yu. M.; Pisarev, A.A. [National Research Nuclear University MEPhI, Moscow (Russian Federation); Fiflis, Peter; Szott, Matthew; Christenson, Michael; Kalathiparambil, Kishor; Ruzic, David N. [Center for Plasma Material Interactions, Department of Nuclear, Plasma, and Radiological Engineering, University Illinois at Urbana-Champaign, Urbana (United States)

2017-04-15

Highlights: • Contact angles of liquid lithium and Li{sub 3}N, Li{sub 2}O, Li{sub 2}CO{sub 3} were measured. • Liquid lithium wets lithium compounds at relatively low temperatures: Li{sub 3}N at 257 °C, Li{sub 2}O at 259 °C, Li{sub 2}CO{sub 3} at 323 °C. • Li wets Li{sub 2}O and Li{sub 3}N better than previously measured fusion-relevant materials (W, Mo, Ta, TZM, stainless steel). • Li wets Li{sub 2}CO{sub 3} better than most previously measured fusion-relevant materials (W, Mo, Ta). - Abstract: Liquid metal plasma facing components (LMPFC) have shown a potential to supplant solid plasma facing components materials in the high heat flux regions of magnetic confinement fusion reactors due to the reduction or elimination of concerns over melting, wall damage, and erosion. To design a workable LMPFC, one must understand how liquid metal interacts with solid underlying structures. Wetting is an important factor in such interaction, several designs of LMPFC require liquid metal to wet the underlying solid structures. The wetting of lithium compounds (lithium nitride, oxide, and carbonate) by 200 °C liquid lithium at various surface temperature from 230 to 330 °C was studied by means of contact angle measurements. Wetting temperatures, defined as the temperature above which the contact angle is less than 90°, were measured. The wetting temperature was 257 °C for nitride, 259 °C for oxide, and 323 °C for carbonate. Surface tensions of solid lithium compounds were calculated from the contact angle measurements.

Mechanical Design of the NSTX Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

2009-02-19

The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

Mechanical Design of the NSTX Liquid Lithium Divertor

International Nuclear Information System (INIS)

Ellis, R.; Kaita, R.; Kugel, H.; Paluzzi, G.; Viola, M.; Nygren, R.

2009-01-01

The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuum compatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

Plasma interaction with liquid lithium: Measurements of retention and erosion

Energy Technology Data Exchange (ETDEWEB)

Baldwin, M.J. E-mail: mbaldwin@ferp.ucsd.edu; Doerner, R.P.; Luckhardt, S.C.; Seraydarian, R.; Whyte, D.G.; Conn, R.W

2002-11-01

This paper reports on recent studies of high flux deuterium and helium plasma interaction with liquid lithium in the Pisces-B edge plasma simulator facility. Deuterium retention is explored as a function of plasma ion fluence in the range 6x10{sup 19}-4x10{sup 22} atoms cm{sup -2} and exposure temperatures of 523-673 K. The results are consistent with full uptake of the deuterium ions incident on the liquid metal surface, independent of the temperature of the liquid lithium. Full uptake continues until the sample is volumetrically converted to lithium deuteride. Helium retention is not observed for fluences up to 5x10{sup 21} He atoms cm{sup -2}. Measurements of the erosion of lithium are found to be consistent with physical sputtering for the lithium solid phase. However, a mechanism that provides an increased evaporative-like yield and is related to ion impact events on the surface, dominates during the liquid phase leading to an enhanced loss rate for liquid lithium that is greater than the expected loss rate due to evaporation at elevated temperatures. Further, the material loss rate is found to depend linearly on the incident ion flux, even at very high temperature.

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

Energy Technology Data Exchange (ETDEWEB)

Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Capece, A.; Koel, B.; Roszell, J. [Princeton University, Princeton, New Jersey 08544 (United States); Biewer, T. M.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Kubota, S. [University of California at Los Angeles, Los Angeles, California 90095 (United States); Beiersdorfer, P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); and others

2015-05-15

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started.

High performance discharges in the Lithium Tokamak eXperiment with liquid lithium walls

International Nuclear Information System (INIS)

Schmitt, J. C.; Bell, R. E.; Boyle, D. P.; Esposti, B.; Kaita, R.; Kozub, T.; LeBlanc, B. P.; Lucia, M.; Maingi, R.; Majeski, R.; Merino, E.; Punjabi-Vinoth, S.; Tchilingurian, G.; Capece, A.; Koel, B.; Roszell, J.; Biewer, T. M.; Gray, T. K.; Kubota, S.; Beiersdorfer, P.

2015-01-01

The first-ever successful operation of a tokamak with a large area (40% of the total plasma surface area) liquid lithium wall has been achieved in the Lithium Tokamak eXperiment (LTX). These results were obtained with a new, electron beam-based lithium evaporation system, which can deposit a lithium coating on the limiting wall of LTX in a five-minute period. Preliminary analyses of diamagnetic and other data for discharges operated with a liquid lithium wall indicate that confinement times increased by 10× compared to discharges with helium-dispersed solid lithium coatings. Ohmic energy confinement times with fresh lithium walls, solid and liquid, exceed several relevant empirical scaling expressions. Spectroscopic analysis of the discharges indicates that oxygen levels in the discharges limited on liquid lithium walls were significantly reduced compared to discharges limited on solid lithium walls. Tokamak operations with a full liquid lithium wall (85% of the total plasma surface area) have recently started

Investigations on interactions between the flowing liquid lithium limiter and plasmas

International Nuclear Information System (INIS)

Ren, J.; Zuo, G.Z.; Hu, J.S.; Sun, Z.; Li, J.G.; Zakharov, L.E.; Ruzic, D.N.; Xu, W.Y.

2016-01-01

Two different designs of flowing liquid lithium limiter were first tested for power exhaust and particle removal in HT-7 in 2012 autumn campaign. During the experiments, the reliability and compatibility of the limiters within Tokamak were experimentally demonstrated, and some positive results were achieved. It was found that the flowing liquid lithium limiter was effective for suppressing H concentration and led to a low ratio of H/(H + D). O impurity was slightly decreased by using limiters as well as when using a Li coating. A significant increase of the wall retention ratio was also observed which resulted from the outstanding D particles pumping ability of flowing liquid lithium limiters. The strong interaction between plasma and lithium surface could cause lithium ejection into plasma and lead to disruptions. The stable plasmas produced by uniform Li flow were in favor of lithium control. While the limiters were applied with a uniform Li flow, the normal plasma was easy to be obtained, and the energy confinement time increased from ∼0.025 s to 0.04 s. Furthermore, it was encouraging to note that the application of flowing liquid lithium limiters could further improve the confinement of plasma by ∼10% on the basis of Li coating. These remarkable results will help for the following design of flowing liquid lithium limiter in EAST to improve the plasma operation.

Dissolved nitrogen in liquid lithium - a problem in fusion reactor chemistry

International Nuclear Information System (INIS)

Hubberstey, P.

1984-01-01

When dissolved in liquid lithium, nitrogen adopts the role filled by oxygen in liquid sodium systems, reacting readily with stainless steel containment materials to form Li 9 CrN 5 as a surface product; extended reaction leads to pronounced corrosion and embrittlement problems. It also interacts with both carbon and silicon impurities forming Li 2 NCN and Li 5 SiN 3 , respectively; it is inert, however, to oxygen impurity. Although dissolved nitrogen reacts with neither the tritium generated in the breeding process nor the lead added to act as a neutron multiplier, its presence may seriously influence tritium recovery processes since it reacts with and hence may poison the majority of the transition metals (Y,Ti,Zr) presently being considered as tritium getter materials. Its reactivity with these metals forms the basis of the hot trapping technique used to remove dissolved nitrogen from liquid lithium systems; cold trapping is ineffective because of its large solubility even at temperatures just above the melting point of pure lithium (453.6K). Whenever possible, the chemistry of nitrogen dissolved in liquid lithium is rationalised using the thermodynamic concepts and its significance to fusion reactor technology stressed. (author)

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)

Design of liquid lithium pumps for FMIT

International Nuclear Information System (INIS)

Adkins, H.E.

1983-01-01

In the Fusion Materials Irradiation Test (FMIT) facility, a jet of liquid lithium is bombarded by accelerated deuterons to generate high energy neutrons for materials testing. The lithium system will include two electromagnetic pumps, a 750 gpm main pump and a 10 gpm auxiliary pump. The larger pump was designed and built in 1982, following extensive testing of a similar pump in the Experimental Lithium System. Design of the auxiliary pump has been completed, but fabrication has not started. This paper discusses the design considerations leading to selection of the Annular Linear Induction Pump (ALIP) concept for these applications. Design parameters, fabrication procedures, and results of pump testing are also reviewed

Surface Treatment to Improve Corrosion Resistance in Lead-Alloy Coolants

International Nuclear Information System (INIS)

Todd R. Allen; Kumar Sridharan; McLean T. Machut; Lizhen Tan

2007-01-01

One of the six proposed advanced reactor designs of the Generation IV Initiative, the Lead-cooled Fast Reactor (LFR) possesses many characteristics that make it a desirable candidate for future nuclear energy production and responsible actinide management. These characteristics include favorable heat transfer, fluid dynamics, and neutronic performance compared to other candidate coolants. However, the use of a heavy liquid metal coolant presents a challenge for reactor designers in regards to reliable structural and fuel cladding materials in both a highly corrosive high temperature liquid metal and an intense radiation field. Flow corrosion studies at the University of Wisconsin have examined the corrosion performance of candidate materials for application in the LFR concept as well as the viability of various surface treatments to improve the materials compatibility. To date this research has included several focus areas, which include the formulation of an understanding of corrosion mechanisms and the examination of the effects of chemical and mechanical surface modifications on the materials performance in liquid lead-bismuth by experimental testing in Los Alamos National Laboratory's DELTA Loop, as well as comparison of experimental findings to numerical and physical models for long term corrosion prediction. This report will first review the literature and introduce the experiments and data that will be used to benchmark theoretical calculations. The experimental results will be followed by a brief review of the underlying theory and methodology for the physical and theoretical models. Finally, the results of theoretical calculations as well as experimentally obtained benchmarks and comparisons to the literature are presented

Effect of a novel amphipathic ionic liquid on lithium deposition in gel polymer electrolytes

International Nuclear Information System (INIS)

Choi, Nam-Soon; Koo, Bonjae; Yeon, Jin-Tak; Lee, Kyu Tae; Kim, Dong-Won

2011-01-01

Highlights: · Synthesis of a dimeric ionic liquid. · Gel polymer electrolytes providing uniform lithium deposit pathway. · An amphipathic ionic liquid locates at the interface between an electrolyte-rich phase and a polymer matrix in a gel polymer electrolyte. · The presence of PDMITFSI ionic liquid leads to the suppression of dendritic lithium formation on a lithium metal electrode. - Abstract: A novel dimeric ionic liquid based on imidazolium cation and bis(trifluoromethanesulfonyl) imide (TFSI) anion has been synthesized through a metathesis reaction. Its chemical shift values and thermal properties are identified via 1 H nuclear magnetic resonance (NMR) imaging and differential scanning calorimetry (DSC). The effect of the synthesized dimeric ionic liquid on the interfacial resistance of gel polymer electrolytes is described. Differences in the SEM images of lithium electrodes after lithium deposition with and without the 1,1'-pentyl-bis(2,3-dimethylimidazolium) bis(trifluoromethane-sulfonyl)imide (PDMITFSI) ionic liquid in gel polymer electrolytes are clearly discernible. This occurs because the PDMITFSI ionic liquid with hydrophobic moieties and polar groups modulates lithium deposit pathways onto the lithium metal anode. Moreover, high anodic stability for a gel polymer electrolyte with the PDMITFSI ionic liquid was clearly observed.

Interactions between drops of a molten aluminum-lithium alloy and liquid water

International Nuclear Information System (INIS)

Nelson, L.S.

1994-01-01

In certain hypothesized nuclear reactor accident scenarios, 1- to 10-g drops of molten aluminum-lithium alloys might contact liquid water. Because vigorous steam explosions have occurred when large amounts of molten aluminum-lithium alloys were released into water or other coolants, it becomes important to know whether there will be explosions if smaller amounts of these molten alloys similarly come into contact with water. Therefore, the authors released drops of molten Al-3.1 wt pct Li alloy into deionized water at room temperature. The experiments were performed at local atmospheric pressure (0.085 MPa) without pressure transient triggers applied to the water. The absence of these triggers allowed them to (a) investigate whether spontaneous initiation of steam explosions would occur with these drops and (b) study the alloy-water chemical reactions. The drop sizes and melt temperatures were chosen to simulate melt globules that might form during the hypothesized melting of the aluminum-lithium alloy components

Liquid Lithium Limiter Effects on Tokamak Plasmas and Plasma-Liquid Surface Interactions

Energy Technology Data Exchange (ETDEWEB)

R. Kaita; R. Majeski; R. Doerner; G. Antar; M. Baldwin; R. Conn; P. Efthimion; M. Finkenthal; D. Hoffman; B. Jones; S. Krashenninikov; H. Kugel; S. Luckhardt; R. Maingi; J. Menard; T. Munsat; D. Stutman; G. Taylor; J. Timberlake; V. Soukhanovskii; D. Whyte; R. Woolley; L. Zakharov

2002-10-15

We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors.

Liquid Lithium Limiter Effects on Tokamak Plasmas and Plasma-Liquid Surface Interactions

International Nuclear Information System (INIS)

Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Baldwin, M.; Conn, R.; Efthimion, P.; Finkenthal, M.; Hoffman, D.; Jones, B.; Krashenninikov, S.; Kugel, H.; Luckhardt, S.; Maingi, R.; Menard, J.; Munsat, T.; Stutman, D.; Taylor, G.; Timberlake, J.; Soukhanovskii, V.; Whyte, D.; Woolley, R.; Zakharov, L.

2002-01-01

We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors

Liquid lithium limiter effects on tokamak plasmas and plasma-liquid surface interactions

International Nuclear Information System (INIS)

Kaita, R.; Majeski, R.; Doerner, R.

2003-01-01

We present results from the first experiments with a large area liquid lithium limiter in a magnetic fusion device, and its effect on improving plasma performance by reducing particle recycling. Using large area liquid metal surfaces in any major fusion device is unlikely before a test on a smaller scale. This has motivated its demonstration in the CDX-U spherical torus with a unique, fully toroidal lithium limiter. The highest current discharges were obtained with a liquid lithium limiter. There was a reduction in recycling, as indicated by a significant decrease in the deuterium-alpha emission and oxygen radiation. How these results might extrapolate to reactors is suggested in recycling/retention experiments with liquid lithium surfaces under high-flux deuterium and helium plasma bombardment in PISCES-B. Data on deuterium atoms retained in liquid lithium indicate retention of all incident ions until full volumetric conversion to lithium deuteride. The PISCES-B results also show a material loss mechanism that lowers the maximum operating temperature compared to that for the liquid surface equilibrium vapor pressure. This may restrict the lithium temperature in reactors. (author)

Chemistry of liquid metal coolants and sensors

International Nuclear Information System (INIS)

Gnanasekaran, T.

2015-01-01

Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

Mathematical Model-Based Temperature Preparation of Liquid-Propellant Components Cooled by Liquid Nitrogen in the Heat Exchanger with a Coolant

Directory of Open Access Journals (Sweden)

S. K. Pavlov

2014-01-01

Full Text Available Before fuelling the tanks of missiles, boosters, and spacecraft with liquid-propellant components (LPC their temperature preparation is needed. The missile-system ground equipment performs this operation during prelaunch processing of space-purpose missiles (SPM. Usually, the fuel cooling is necessary to increase its density and provide heat compensation during prelaunch operation of SPM. The fuel temperature control systems (FTCS using different principles of operation and types of coolants are applied for fuel cooling.To determine parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is cooled by liquid nitrogen upon contact heat exchange in the coolant reservoir, a mathematical model of this process and a design technique are necessary. Both allow us to determine design parameters of the cooling system and the required liquid nitrogen reserve to cool LPC to the appropriate temperature.The article presents an overview of foreign and domestic publications on cooling processes research and implementation using cryogenic products such as liquid nitrogen. The article draws a conclusion that it is necessary to determine the parameters of LPC cooling process through the fuel heat exchange in the heat exchanger with coolant, which is liquid nitrogen-cooled upon contact heat exchange in the coolant reservoir allowing to define rational propellant cooling conditions to the specified temperature.The mathematical model describes the set task on the assumption that a heat exchange between the LPC and the coolant in the heat exchanger and with the environment through the walls of tanks and pipelines of circulation loops is quasi-stationary.The obtained curves allow us to calculate temperature changes of LPC and coolant, cooling time and liquid nitrogen consumption, depending on the process parameters such as a flow rate of liquid nitrogen, initial coolant temperature, pump characteristics, thermal

FTU cooled liquid lithium upgrade

Energy Technology Data Exchange (ETDEWEB)

Iafrati, M., E-mail: matteo.iafrati@enea.it [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Apicella, M.L.; Boncagni, L. [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Lyublinski, I. [JSC “RED STAR”, Moscow (Russian Federation); Mazzitelli, G. [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Vertkov, A. [JSC “RED STAR”, Moscow (Russian Federation)

2017-04-15

In the framework of the liquid lithium limiter experiment in Frascati a new auxiliary system was developed in order to provide a better control of the energy fluid vector. The cooled liquid lithium system (CLL) was installed for the first time at the end of 2013, it uses overheated water to heat the lithium and to extract, at the same time, the heat from the metal surface when it gets wet by the plasma. A first version of the system, developed and presented in previous papers, has been modified to optimize the heat flux measurement on the liquid lithium surface. The changes include a new power supply logic for the heating system, new sensors and new read-out electronics compatible with the implementation of a real time control system. The prototype was updated with the aim of achieving a low cost and versatile control system.

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

Studies of corrosion resistance of Japanese steels in liquid lead-bismuth

International Nuclear Information System (INIS)

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

2003-01-01

Liquid lead-bismuth has attractive characteristics as a coolant in future fast reactors and Accelerator Driven Sub-critical Systems (ADS) applications. The corrosion behavior of structural materials in lead-bismuth eutectic is one of key problems in developing nuclear power plants and installations using lead-bismuth coolant. Our experiences with heat exchangers using liquid lead-bismuth and the results of corrosion tests of Japanese steels are reported in this paper. A series of corrosion tests was carried out in collaboration with the Institute of Physics and Power Engineering (IPPE). Test specimens of various Japanese steels were exposed in a non-isothermal forced circulation loop. The influence of maximum temperature and oxygen content in lead bismuth were chosen for study as the primary causes of corrosion in Japanese steels. After the corrosion tests, corrosion behavior was analyzed by visual inspection, measurement of weight loss and metallurgical examination of the microstructure of the corroded zone. The corrosion mechanism in liquid lead bismuth is discussed on the basis of the metallurgical examination of the corroded zone. (author)

Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor. Comparisons of the decay heat removal characteristics on lead, lead-bismuth and sodium cooled reactors

International Nuclear Information System (INIS)

Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

2000-04-01

The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. In this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube failure accidents in a steam generator. In this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in Equivalent plant' with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. In conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to confirm the heat transfer reduction by the oxidized film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance. (author)

NSTX plasma operation with a Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Ellis, R.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M.; Paul, S.F. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); and others

2012-10-15

Highlights: Black-Right-Pointing-Pointer NSTX 2010 experiments tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium molybdenum divertor surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. Black-Right-Pointing-Pointer Noteworthy improvements in plasma performance with the plasma strike point on the liquid lithium molybdenum divertor were obtained similar to those obtained previously with lithiated graphite. The role of lithium impurities in this result is discussed. Black-Right-Pointing-Pointer Inspection of the liquid lithium molybdenum divertor after the Campaign indicated mechanical damage to supports, and other hardware resulting from forces following plasma current disruptions. - Abstract: NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the

Electrochemical Model for Ionic Liquid Electrolytes in Lithium Batteries

International Nuclear Information System (INIS)

Yoo, Kisoo; Deshpande, Anirudh; Banerjee, Soumik; Dutta, Prashanta

2015-01-01

ABSTRACT: Room temperature ionic liquids are considered as potential electrolytes for high performance and safe lithium batteries due to their very low vapor pressure and relatively wide electrochemical and thermal stability windows. Unlike organic electrolytes, ionic liquid electrolytes are molten salts at room temperature with dissociated cations and anions. These dissociated ions interfere with the transport of lithium ions in lithium battery. In this study, a mathematical model is developed for transport of ionic components to study the performance of ionic liquid based lithium batteries. The mathematical model is based on a univalent ternary electrolyte frequently encountered in ionic liquid electrolytes of lithium batteries. Owing to the very high concentration of components in ionic liquid, the transport of lithium ions is described by the mutual diffusion phenomena using Maxwell-Stefan diffusivities, which are obtained from atomistic simulation. The model is employed to study a lithium-ion battery where the electrolyte comprises ionic liquid with mppy + (N-methyl-N-propyl pyrrolidinium) cation and TFSI − (bis trifluoromethanesulfonyl imide) anion. For a moderate value of reaction rate constant, the electric performance results predicted by the model are in good agreement with experimental data. We also studied the effect of porosity and thickness of separator on the performance of lithium-ion battery using this model. Numerical results indicate that low rate of lithium ion transport causes lithium depleted zone in the porous cathode regions as the porosity decreases or the length of the separator increases. The lithium depleted region is responsible for lower specific capacity in lithium-ion cells. The model presented in this study can be used for design of optimal ionic liquid electrolytes for lithium-ion and lithium-air batteries

Deuterium retention in liquid lithium

International Nuclear Information System (INIS)

Baldwin, M.J.; Doerner, R.P.; Luckhardt, S.C.; Conn, R.W.

2002-01-01

Measurements of deuterium retention in samples of lithium exposed in the liquid state to deuterium plasma are reported. Retention was measured as a function of plasma ion dose in the range 6x10 19 -4x10 22 D atoms and exposure temperature between 523 and 673 K using thermal desorption spectrometry. The results are consistent with the full uptake of all deuterium ions incident on the liquid metal surface and are found to be independent of the temperature of the liquid lithium over the range explored. Full uptake, consistent with very low recycling, continues until the sample is volumetrically converted to lithium deuteride. This occurs for exposure temperatures where the gas pressure during exposure was both below and slightly above the corresponding decomposition pressure for LiD in Li. (author)

Preliminary Design of the Liquid Lead Corrosion Test Loop

International Nuclear Information System (INIS)

Cho, Chung Ho; Cha, Jae Eun; Cho, Choon Ho; Song, Tae Yung; Kim, Hee Reyoung

2005-01-01

Recently, Lead-Bismuth Eutectic (LBE) or Lead has newly attracted considerable attraction as a coolant to get the more inherent safety. Above all, LBE is preferred as the coolant and target material for an Accelerator-Driven System (ADS) due to its high production rate of neutrons, effective heat removal, and good radiation damage properties. But, the LBE or Lead as a coolant has a challenging problem that the LBE or Lead is more corrosive to the construction materials and fuel cladding material than the sodium because the solubility of Ni, Cr and Fe is high. After all, the LBE or Lead corrosion has been considered as an important design limit factor of ADS and Liquid Metal cooled Fast Reactors (LMFR). The Korea Atomic Energy Research Institute (KAERI) has been developing an ADS called HYPER. HYPER is designed to transmute Transuranics (TRU), Tc-99 and I-129 coming from Pressurized Water Reactors (PWRs) and uses an LBE as a coolant and target material. Also, an experimental apparatuses for the compatibility of fuel cladding and structural material with the LBE or Lead are being under the construction or design. The main objective of the present paper is introduction of Lead corrosion test loop which will be built the upside of the LBE corrosion test loop by the end of October of 2005

Surface Treatment to Improve Corrosion Resistance in Lead-Alloy Coolants

Energy Technology Data Exchange (ETDEWEB)

Todd R. Allen; Kumar Sridharan; McLean T. Machut; Lizhen Tan

2007-08-29

One of the six proposed advanced reactor designs of the Generation IV Initiative, the Leadcooled Fast Reactor (LFR) possesses many characteristics that make it a desirable candidate for future nuclear energy production and responsible actinide management. These characteristics include favorable heat transfer, fluid dynamics, and neutronic performance compared to other candidate coolants. However, the use of a heavy liquid metal coolant presents a challenge for reactor designers in regards to reliable structural and fuel cladding materials in both a highly corrosive high temperature liquid metal and an intense radiation fieldi. Flow corrosion studies at the University of Wisconsin have examined the corrosion performance of candidate materials for application in the LFR concept as well as the viability of various surface treatments to improve the materials’ compatibility. To date this research has included several focus areas, which include the formulation of an understanding of corrosion mechanisms and the examination of the effects of chemical and mechanical surface modifications on the materials’ performance in liquid lead-bismuth by experimental testing in Los Alamos National Laboratory’s DELTA Loop, as well as comparison of experimental findings to numerical and physical models for long term corrosion prediction. This report will first review the literature and introduce the experiments and data that will be used to benchmark theoretical calculations. The experimental results will be followed by a brief review of the underlying theory and methodology for the physical and theoretical models. Finally, the results of theoretical calculations as well as experimentally obtained benchmarks and comparisons to the literature are presented.

Gas absorption and discharge behaviors of lead-lithium

International Nuclear Information System (INIS)

Sakabe, Toshiro; Yokomine, Takehiko; Kunugi, Tomoaki; Kawara, Zensaku; Ueki, Yoshitaka; Tanaka, Teruya

2014-01-01

Highlights: • The absorption of argon in the lead-lithium is comparable with that of helium even at the solid state. • For the molten state of lead-lithium, the absorption of argon could be larger than that of helium. • It is observed that the argon tends to desorb when the phase change of lead-lithium occurs. • It is observed from the TPD-MS analysis that the argon tends to desorb when the phase change of lead-lithium occurs. - Abstract: The absorption of rare gas in the lead-lithium has been quite low and the gas is used as a cover-gas to control the environment of experiment. In our previous thermo-fluid experiment by using lithium-lead, it was found the cover gas pressure enclosed in the very leak tight container of lithium-lead was decreased with time, that is, the gas-absorption of the solid lithium-lead occurred at room temperature under atmospheric pressure. The variation of pressure exceeded the retention of argon in lead-lithium which is expected by the published data. Therefore, we aim to confirm those phenomena under well-controlled experimental condition by using argon, nitrogen and helium. According to the results of gas exposure tests, the absorption of argon in the lead-lithium is comparable with that of helium even at the solid state. For the molten state of lead-lithium, the absorption of argon could be larger than that of helium. It is also observed from the TPD-MS analysis that the argon tends to desorb when the phase change of lead-lithium occurs. If the retention of argon in the lead-lithium cannot be ignored, the problem of Ar-41 activity should be taken into consideration as well as the problem of argon bubble in the lead-lithium

Liquid lithium limiter experiments in CDX-U

International Nuclear Information System (INIS)

Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J.; Timberlake, J.; Zakharov, L.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Soukhanovskii, V.; Maingi, R.; Finkenthal, M.; Stutman, D.; Rodgers, D.

2005-01-01

Recent experiments in the Current Drive eXperiment - Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R=34 cm, a=22 cm, B toroidal 2 kG, I P =100 kA, T e (0)∼100 eV, n e (0)∼ 5 x 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z effective <1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced. (author)

Liquid Lithium Limiter Experiments in CDX-U

International Nuclear Information System (INIS)

Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J.; Timberlake, J.; Zakharov, L.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Soukhanovskii, V.; Maingi, R.; Finkenthal, M.; Stutman, D.; Rodgers, D.

2004-01-01

Recent experiments in the Current Drive Experiment-Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, B toroidal = 2 kG, I P = 100 kA, T e (0) = 100 eV, n e (0) ∼ 5 x 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium-limited discharges are consistent with Z effective < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced

Testing of Liquid Lithium Limiters in CDX-U

International Nuclear Information System (INIS)

Majeski, R.; Kaita, R.; Boaz, M.; Efthimion, P.; Gray, T.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Maingi, R.; Maiorano, M.; Smith, S.; Rodgers, D.

2004-01-01

Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described

Testing of Liquid Lithium Limiters in CDX-U

Energy Technology Data Exchange (ETDEWEB)

R. Majeski; R. Kaita; M. Boaz; P. Efthimion; T. Gray; B. Jones; D. Hoffman; H. Kugel; J. Menard; T. Munsat; A. Post-Zwicker; V. Soukhanovskii; J. Spaleta; G. Taylor; J. Timberlake; R. Woolley; L. Zakharov; M. Finkenthal; D. Stutman; G. Antar; R. Doerner; S. Luckhardt; R. Seraydarian; R. Maingi; M. Maiorano; S. Smith; D. Rodgers

2004-07-30

Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid-metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid-lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm2, subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now be en performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid-lithium plasma-facing area was increased to 2000 cm2. Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.

Testing of liquid lithium limiters in CDX-U

International Nuclear Information System (INIS)

Majeski, R.; Kaita, R.; Boaz, M.; Efthimion, P.; Gray, T.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Maingi, R.; Maiorano, M.; Smith, S.; Rodgers, D.; Soukhanovskii, V.

2004-01-01

Part of the development of liquid metals as a first wall or divertor for reactor applications must involve the investigation of plasma-liquid metal interactions in a functioning tokamak. Most of the interest in liquid metal walls has focused on lithium. Experiments with lithium limiters have now been conducted in the Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory. Initial experiments used a liquid lithium rail limiter (L3) built by the University of California at San Diego. Spectroscopic measurements showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm 2 , subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now been performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments, the liquid lithium plasma-facing area was increased to 2000 cm 2 . Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density. The loop voltage required to sustain the plasma current was reduced from 2 V to 0.5 V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described

Preparation and transport properties of novel lithium ionic liquids

International Nuclear Information System (INIS)

Shobukawa, Hitoshi; Tokuda, Hiroyuki; Tabata, Sei-Ichiro; Watanabe, Masayoshi

2004-01-01

Novel lithium salts of borates having two electron-withdrawing groups (either 1,1,1,3,3,3-hexafluoro-2-propoxy or pentafluorophenoxy group) and two methoxy-oligo(ethylene oxide) groups (number of repeating unit: n = 3, 4, 7.2) were prepared by successive substitution-reactions from LiBH 4 . The obtained lithium salts were clear and colorless liquids at room temperature. The density, thermal property, viscosity, and ionic conductivity were measured for the lithium ionic liquids. The pulsed-gradient spin-echo NMR (PGSE-NMR) method was used to independently determine self-diffusion coefficients of the lithium cation ( 7 Li NMR) and the anion ( 19 F NMR) in the bulk. The ionic conductivity of the new lithium salts was 10 -5 to 10 -4 S cm -1 at 30 deg. C, which was lower than that of typical ionic liquids by two orders of magnitude. However, the degree of self-dissociation of the lithium ionic liquids; the ratio of the molar conductivity determined by the complex impedance method to that calculated from the self-diffusion coefficients and the Nernst-Einstein equation, ranged from 0.1 to 0.4, which are comparable values to those of a highly dissociable salt in an aprotic polar solvent and of typical ionic liquids. The main reason for the meager conductivity was high viscosities of the lithium ionic liquids. It should be noted that the lithium ionic liquids have self-dissociation ability and conduct the ions in the absence of organic solvents

Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

International Nuclear Information System (INIS)

Monta, Kazuo.

1974-01-01

Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

Liquid metal cooled blanket concept for NET

International Nuclear Information System (INIS)

Malang, S.; Casal, V.; Arheidt, K.; Fischer, U.; Link, W.; Rust, K.

1986-01-01

A blanket concept for NET using liquid lithium-lead both as breeder material and as coolant is described. The need for inboard breeding is avoided by using beryllium as neutron multiplier in the outboard blanket. Novel flow channel inserts are employed in all poloidal ducts to reduce the MHD pressure drop. The concept offers a simple mechanical design and a higher tritium breeding ratio compared to water- and gas-cooled blankets. (author)

Comparison of lithium and the eutectic lead lithium alloy, two candidate liquid metal breeder materials for self-cooled blankets

International Nuclear Information System (INIS)

Malang, S.; Mattas, R.

1994-06-01

Liquid metals are attractive candidates for both near-term and long-term fusion applications. The subjects of this comparison are the differences between the two candidate liquid metal breeder materials Li and LiPb for use in breeding blankets in the areas of neutronics, magnetohydrodynamics, tritium control, compatibility with structural materials, heat extraction system, safety, and required R ampersand D program. Both candidates appear to be promising for use in self-cooled breeding blankets which have inherent simplicity with the liquid metal serving as both breeders and coolant. The remaining feasibility question for both breeder materials is the electrical insulation between liquid metal and duct walls. Different ceramic coatings are required for the two breeders, and their crucial issues, namely self-healing of insulator cracks and radiation induced electrical degradation are not yet demonstrated. Each liquid metal breeder has advantages and concerns associated with it, and further development is needed to resolve these concerns

Experiments for liquid metal embrittlement of fusion reactor materials by liquid lithium

International Nuclear Information System (INIS)

Grundmann, M.; Borgstedt, H.U.

1984-10-01

The liquid metal embrittlement behaviour of two martensitic-ferritic steels [X22CrMoV121 (Nr. 1.4923) and X18CrMoVNb 121 (Nr. 1,4914)] and one austenite chromium-nickel-steel X5CrNi189 (Nr. 1.4301) was investigated. Tensile tests in liquid lithium at 200 and 250 0 C with two different strain rates on precorroded samples (1000 h at 550 0 C in lithium) were carried out. Reference values were gained from tensile tests in air (RT, 250 0 C). It is concluded that there is sufficient compatibility of the austenitic steel with liquid lithium. The use of the ferritic-martensitic steels in liquid lithium on the other hand, especially at temperatures of about 550 0 C, seems to be problematic. The experimental results led to a better understanding of LME, applying the theory of this material failure. (orig./IHOE) [de

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.

Control of nitrogen concentration in liquid lithium by hot trapping

International Nuclear Information System (INIS)

Sakurai, Toshiharu; Yoneoka, Toshiaki; Tanaka, Satoru; Suzuki, Akihiro; Muroga, Takeo

2002-01-01

Nitrogen concentration in liquid lithium was controlled by the method of hot trapping. V-Ti alloy and chromium were used as nitrogen gettering materials. Chromium is known to form ternary nitride with lithium. Gettering experiments were conducted at 823 K for 0.8-2.2 Ms. Under high nitrogen concentration in liquid lithium, above 10 -2 mass%, nitrogen gettering effect of chromium was found to be larger than that of V-10at.% Ti alloy. Nitrogen gettering by chromium at 823 K reached a limit at about 6.5x10 -3 mass% of nitrogen concentration in liquid lithium. Instability of ternary nitride of chromium and lithium below this nitrogen concentration in liquid lithium was considered to be the reason for this limit. The composition of the ternary nitride that was formed in this study was considered to be Li 6 Cr(III) 3 N 5 . In addition, immersion experiments of yttrium with V-10at.% Ti alloy were performed. It was found that nitriding of yttrium in liquid lithium is controlled by nitrogen gettering effect of V-10at.% Ti alloy

Diagnostics for liquid lithium experiments in CDX-U

International Nuclear Information System (INIS)

Kaita, R.; Efthimion, P.; Hoffman, D.; Jones, B.; Kugel, H.; Majeski, R.; Munsat, T.; Raftopoulos, S.; Taylor, G.; Timberlake, J.; Soukhanovskii, V.; Stutman, D.; Iovea, M.; Finkenthal, M.; Doerner, R.; Luckhardt, S.; Maingi, R.; Causey, R.

2000-01-01

A flowing liquid lithium first wall or diverter target could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls in fusion reactors. To investigate the interaction of a spherical torus plasma with liquid lithium limiters, large area diverter targets, and walls, discharges will be established in the Current Drive Experiment-Upgrade (CDX-U) where the plasma-wall interactions are dominated by liquid lithium surfaces. Among the unique CDX-U lithium diagnostics is a multi-layer mirror (MLM) array, which will monitor the 135 (angstrom) LiIII line for core lithium concentrations. Additional spectroscopic diagnostics include a grazing incidence XUV spectrometer (STRS) and a filterscope system to monitor D α and various impurity lines local to the lithium limiter. Profile data will be obtained with a multichannel tangential bolometer and a multipoint Thomson scattering system configured to give enhanced edge resolution. Coupons on th e inner wall of the CDX-U vacuum vessel will be used for surface analysis. A 10,000 frame per second fast visible camera and an IR camera will also be available

Chemical processing of liquid lithium fusion reactor blankets

International Nuclear Information System (INIS)

Weston, J.R.; Calaway, W.F.; Yonco, R.M.; Hines, J.B.; Maroni, V.A.

1979-01-01

A 50-gallon-capacity lithium loop constructed mostly from 304L stainless steel has been operated for over 6000 hours at temperatures in the range from 360 to 480 0 C. This facility, the Lithium Processing Test Loop (LPTL), is being used to develop processing and monitoring technology for liquid lithium fusion reactor blankets. Results of tests of a molten-salt extraction method for removing impurities from liquid lithium have yielded remarkably good distribution coefficients for several of the more common nonmetallic elements found in lithium systems. In particular, the equilibrium volumetric distribution coefficients, D/sub v/ (concentration per unit volume of impurity in salt/concentration per unit volume of impurity in lithium), for hydrogen, deuterium, nitrogen and carbon are approx. 3, approx. 4, > 10, approx. 2, respectively. Other studies conducted with a smaller loop system, the Lithium Mini-Test Loop (LMTL), have shown that zirconium getter-trapping can be effectively used to remove selected impurities from flowing lithium

Operational Characteristics of Liquid Lithium Divertor in NSTX

Science.gov (United States)

Kaita, R.; Kugel, H.; Abrams, T.; Bell, M. G.; Bell, R. E.; Gerhardt, S.; Jaworski, M. A.; Kallman, J.; Leblanc, B.; Mansfield, D.; Mueller, D.; Paul, S.; Roquemore, A. L.; Scotti, F.; Skinner, C. H.; Timberlake, J.; Zakharov, L.; Maingi, R.; Nygren, R.; Raman, R.; Sabbagh, S.; Soukhanovskii, V.

2010-11-01

Lithium coatings on plasma-facing components (PFC's) have resulted in improved plasma performance on NSTX in deuterium H-mode plasmas with neutral beam heating.^ Salient results included improved electron confinement and ELM suppression. In CDX-U, the use of lithium-coated PFC's and a large-area liquid lithium limiter resulted in a six-fold increase in global energy confinement time. A Liquid Lithium Divertor (LLD) has been installed in NSTX for the 2010 run campaign. The LLD PFC consists of a thin film of lithium on a temperature-controlled substrate to keep the lithium liquefied between shots, and handle heat loads during plasmas. This capability was demonstrated when the LLD withstood a strike point on its surface during discharges with up to 4 MW of neutral beam heating.

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)

Corrosion of ferrous alloys in nitrogen contaminated liquid lithium

International Nuclear Information System (INIS)

Olson, D.L.; Bradley, W.L.

1976-01-01

Liquid lithium penetration of 304L stainless steel and Armco iron grain boundaries has been studied. The penetration kinetics for the 304L stainless steel was found to be diffusion controlled. The measured temperature dependent delay time has been associated with the initial formation of the corrosion product at the grain boundary. Nitrogen in the stainless steel or the liquid lithium has been found to accelerate the rate of attack without changing the apparent activation energy. Grain boundary grooving of Armco iron in liquid lithium indicates that the controlling mass transport is also through a corrosion product present as a surface film. Stresses as small as 12 MPa have been found to give rise to a fifty-fold increase in the rate of penetration of Armco iron by liquid lithium

Coolant and ambient temperature control for chillerless liquid cooled data centers

Science.gov (United States)

Chainer, Timothy J.; David, Milnes P.; Iyengar, Madhusudan K.; Parida, Pritish R.; Simons, Robert E.

2016-02-02

Cooling control methods include measuring a temperature of air provided to a plurality of nodes by an air-to-liquid heat exchanger, measuring a temperature of at least one component of the plurality of nodes and finding a maximum component temperature across all such nodes, comparing the maximum component temperature to a first and second component threshold and comparing the air temperature to a first and second air threshold, and controlling a proportion of coolant flow and a coolant flow rate to the air-to-liquid heat exchanger and the plurality of nodes based on the comparisons.

Thawing of lithium in the SP-100 reactor core configuration

International Nuclear Information System (INIS)

Magee, P.M.; Malovrh, J.W.; REineking, W.H.

1986-01-01

The General Electric SP-100 Liquid Metal Reactor is designed to be launched with the lithium coolant in the reactor and primary loops frozen. Initial startup of the system in space, after a satisfactory orbit is achieved, will be accomplished by slowly increasing the power in the reactor core and using the heat generated to melt the lithium, first in the reactor, and then progressively down the primary loops. This technique significantly facilitates ground handling, reduces vibrational loads during vehicle launch and minimized the shuttle bay heat load. The challenge is to thaw the coolant and startup the system within an acceptable time without structural damage. The test results clearly demonstrate that thawing of the lithium in the SP-100 reactor core can be done rapidly without structural damage and, thus, support the selected concept of SP-100 launch with frozen lithium and thaw/startup in space

Liquid metal coolants for fusion-fission hybrid system: A neutronic analysis

Energy Technology Data Exchange (ETDEWEB)

Marques, Renato V.A.; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., E-mail: claubia@nuclear.ufmg.br [Universidade de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany P. [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

Based on a work already published by the UFMG Nuclear Engineering Department, it was suggested to use different coolant materials in a fusion-fission system after a fuel burnup simulation, including that one used in reference work. The goal is to compare the neutron parameters, such as the effect multiplication factor and actinide amounts in transmutation layer, for each used coolant and find the best(s) coolant material(s) to be applied in the considered system. Results indicate that the lead and lead-bismuth coolant are the most suitable choices to be applied to cool the system. (author)

Recent Liquid Lithium Limiter Experiments in CDX-U

International Nuclear Information System (INIS)

Majeski, R.; Jardin, S.; Kaita, R.; Gray, T.; Marfuta, P.; Spaleta, J.; Timberlake, J.; Zakharov, L.; Antar, G.; Doerner, R.; Luckhardt, S.; Seraydarian, R.; Soukhanovskii, V.; Maingi, R.; Finkenthal, M.; Stutman, D.; Rodgers, D.; Angelini, S.

2005-01-01

Recent experiments in the Current Drive eXperiment-Upgrade (CDX-U) provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R=34 cm, a=22 cm, B toroidal = 2 kG, I P =100 kA, T e (0) ∼ 100 eV, n e (0) ∼ 5 x 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium pool limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z effective < 1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced

Use of liquid metals in nuclear and thermonuclear engineering, and in other innovative technologies

Science.gov (United States)

Rachkov, V. I.; Arnol'dov, M. N.; Efanov, A. D.; Kalyakin, S. G.; Kozlov, F. A.; Loginov, N. I.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.

Ion transport properties of lithium ionic liquids and their ion gels

International Nuclear Information System (INIS)

Shobukawa, Hitoshi; Tokuda, Hiroyuki; Susan, Md. Abu Bin Hasan; Watanabe, Masayoshi

2005-01-01

A new series of lithium ionic liquids were prepared by introducing of two electron-withdrawing trifluoroacetyl groups in borate salts containing two methoxy-oligo(ethylene oxide) groups in the structures. Successive substitution reactions of oligo-ethylene glycol monomethyl ether and trifluroacetic acid from LiBH 4 yielded the lithium salts, which were clear and colorless liquids at room temperature. The fundamental physicochemical properties, such as density, thermal property, viscosity, ionic conductivity, self-diffusion coefficients, and electrochemical stability, were measured. The lithium ionic liquids had self-dissociation ability and conducted ions even in the absence of organic solvents. New polymer electrolytes, named 'ion gels', were prepared by radical cross-linking reactions of a poly(ethylene oxide-co-propylene oxide)tri-acrylate macromonomer in the presence the lithium ionic liquid. An increase in the glass transition temperatures (T g ) of the ion gels was very small even with increasing lithium ionic liquid concentration, and the T g 's were lower than that of the ionic liquid itself. The ionic conductivity of the ion gels surpassed that of the lithium ionic liquid in the bulk at certain compositions

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)

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

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

Factors influencing the thermodynamic isotope effect of lithium in polyetherlithium liquid-liquid extraction systems

International Nuclear Information System (INIS)

Fu Lian; Fang Shengqiang; Yao Zhongqi; Gao Zhichang; Tan Ganzhu

1989-01-01

The published data up to now concerning polyether-lithium liquid-liquid extraction systems, can be summarized by the equation, ε p = (α-1)/[1 + 0.46(1-P)], where α denotes the isotope separation factor; P - the ratio of the lithium concentration in the organic phase to the initial concentration of crown ethers; ε p -the enrichment coefficient as P = 100%. Based on the changes in ε p , P, α and D(distribution ratio), the functions of factors such as polyether's structure, polyether's side group, polyether's concentration, organic solvent, negative ion of lithium salt and lithium salt's concentration, are discussed and reported

Lithium Hideout and Return in the CANDU Heat Transport System during Shutdown and Start-up

International Nuclear Information System (INIS)

Qiu, L.; Snaglewski, A.P.

2012-09-01

Lithium hydroxide is used to control the pH a (pH apparent) of the Heat Transport System (HTS) coolant in CANDU R reactors. The recommended range of the lithium concentration in the coolant is between 0.38 ppm (5.5x10 -5 m) and 0.60 ppm (8.7x10 -5 m) to minimize carbon steel corrosion in the HTS and magnetite deposition in the core during normal operation; this corresponds to pH a values between 10.2 and 10.4. Similar pH a and lithium concentrations should be maintained during shutdown and start-up. However, maintaining the pH a of the HTS coolant within specification during shutdown and start-up has been difficult for some CANDU stations, especially when the HTS is taken to a Low Level Drain State (LLDS), because of lithium hideout and return. This paper presents the results from lithium adsorption and desorption studies on iron oxides under relevant shutdown and start-up chemistry conditions performed to elucidate the mechanisms of the observed lithium hideout and return. The results show that lithium hideout and return are driven largely by changes in the solubility of magnetite as the HTS coolant chemistry changes during shutdown; changes in lithium concentration were inversely correlated with the solubility of magnetite. When the HTS system is de-pressurized and drained to a low coolant level, the ingress of air rapidly oxidizes the dissolved Fe (II) in the coolant, 2Fe +2 + 1 / 2 O 2 + 3 H 2 = 2FEOOH + 4 H + , resulting in the formation of lepidocrocite or maghemite, which have much lower solubilities but larger surface areas than does magnetite. The large surface area of the Fe (III) oxides can adsorb significant quantities of lithium from the coolant, leading to lithium hideout and a pH a decrease. During start-up, the chemistry of the coolant changes from oxidizing to reducing, and lepidocrocite and other Fe (III) oxides are reduced to Fe (II), gradually dissolving as their solubility increases with increasing temperature. The adsorbed lithium is released

Coolant Design System for Liquid Propellant Aerospike Engines

Science.gov (United States)

McConnell, Miranda; Branam, Richard

2015-11-01

Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

A molecular dynamics study of lithium-containing aprotic heterocyclic ionic liquid electrolytes

Science.gov (United States)

Lourenço, Tuanan C.; Zhang, Yong; Costa, Luciano T.; Maginn, Edward J.

2018-05-01

Classical molecular dynamics simulations were performed on twelve different ionic liquids containing aprotic heterocyclic anions doped with Li+. These ionic liquids have been shown to be promising electrolytes for lithium ion batteries. Self-diffusivities, lithium transference numbers, densities, and free volumes were computed as a function of lithium concentration. The dynamics and free volume decreased with increasing lithium concentration, and the trends were rationalized by examining the changes to the liquid structure. Of those examined in the present work, it was found that (methyloxymethyl)triethylphosphonium triazolide ionic liquids have the overall best performance.

Trace hydrogen extraction from liquid lithium tin alloy

International Nuclear Information System (INIS)

Xie Bo; Hu Rui; Xie Shuxian; Weng Kuiping

2010-01-01

In order to finish the design of tritium extraction system (TES) of fusion fission hybrid reactor (FFHR) tritium blanket, involving the dynamic mathematical model of liquid metal in contact with a gaseous atmosphere, approximate mathematical equation of tritium in lithium tin alloy was deduced. Moreover, carrying process used for trace hydrogen extraction from liquid lithium tin alloy was investigated with hydrogen being used to simulate tritium in the study. The study results indicate that carrying process is effective way for hydrogen extraction from liquid lithium tin alloy, and the best flow velocity of carrier gas is about 4 L/min under 1 kg alloy temperatures and carrying numbers are the main influencing factors of hydrogen number. Hydrogen extraction efficiency can reach 85% while the alloy sample is treated 6 times at 823 K. (authors)

The design of a liquid lithium lens for a muon collider

International Nuclear Information System (INIS)

Balbekov, V.; Geer, S.; Hassanein, A.; Holtkamp, N.; Lebrun, P.; Neuffer, D.; Norem, J.; Palmer, R.; Reed, C.; Silvestrov, G.; Spentzouris, P.; Tollestrup, A.; Vsevolozhskaya, T. A.

1999-01-01

The last stage of ionization cooling for the muon collider requires a multistage liquid lithium lens. This system uses a large (approximately0.5 MA) pulsed current through liquid lithium to focus the beam while energy loss in the lithium removes momentum which is replaced by linacs. The beam optics are designed to maximize the 6 dimensional transmission from one lens to the next while minimizing emittance growth. The mechanical design of the lithium vessel is constrained by a pressure pulse due to the sudden ohmic heating, and the stress on the Be window. The authors describe beam optics, the liquid lithium pressure vessel, pumping, power supplies, as well as the overall optimization of the system

Liquid-metal aspects of HYLIFE

International Nuclear Information System (INIS)

Meier, W.R.; Hoffman, N.J.; McDowell, M.W.

1980-01-01

The High Yield Lithium Injection Fusion Energy (HYLIFE) converter is a reactor concept for an inertial fusion electric power plant. In this concept, flowing molten lithium protects the structures of the fusion chamber from the deleterious effects of deuterium-tritium (DT) fusion reactions and converts the pulsed fusion energy into steay thermal power. Lithium is circulated as the primary coolant to transfer heat to an intermediate sodium loop which drives a superheated steam cycle. Lithium is also the source of the tritium fuel which is recovered via a molten-salt extraction process. The liquid-metal aspects of the HYLIFE plant with particular emphasis on the lithium systems

The Liquid Lithium Limiter control system on FTU

International Nuclear Information System (INIS)

Bertocchi, A.; Panella, M.; Vitale, V.; Sinibaldi, S.

2006-01-01

In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system configuration was installed for testing on the FTU tokamak. The liquid lithium flows through capillaries from a reservoir to the side facing the plasma to form a thin liquid lithium film. The system is composed of three stainless steel sections, which contain two thermocouples each. A heating system brings the Li temperature to about 200 o C allowing the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. [M. Apicella, G. Mazzitelli et al., First experiment with Lithium Limiter on FTU, 17 o International Conference on Plasma Surface Interaction in Controlled Fusion Devices, 22 - 26 May 2006, Hefei Anhui, China]. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22 TM modules and a CORBA/PHP/MySQL software architecture [A. Bertocchi, S. Podda, V. Vitale, Fusion Eng. Des. 74 (2005) 787-791]. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLab and Java environments respectively to monitor the lithium temperature coming from thermocouples - have been also implemented. The control system allows regulating the heater temperature in each section of the LLL to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During plasma operations the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink TM tool - has been realized. (author)

High-power liquid-lithium jet target for neutron production

OpenAIRE

Halfon, S.; Arenshtam, A.; Kijel, D.; Paul, M.; Berkovits, D.; Eliyahu, I.; Feinberg, G.; Friedman, M.; Hazenshprung, N.; Mardor, I.; Nagler, A.; Shimel, G.; Tessler, M.; Silverman, I.

2013-01-01

A compact Liquid-Lithium Target (LiLiT) was built and tested with a high-power electron gun at Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the 7Li(p,n)7Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy (BNCT). The liquid-lithium jet target acts both as ...

Extraction of tritium from liquid lithium by permeation

International Nuclear Information System (INIS)

Alire, R.M.

1978-01-01

This paper assesses a method for extracting tritium from liquid lithium for specific application to the conceptual laser fusion reactor that uses a continuous lithium ''waterfall.'' The tritium diffuses through a refractory metal that contains a getter and is then stored in a hydride-forming alloy. There are various uncertainties with this method including helium-4 extraction, unknown impurities that may accumulate in liquid lithium, the effects of these impurities on tritium separation, and the maintenance of tritium-contaminated equipment. Our study indicates that major tritium losses will occur during equipment maintenance rather than as a result of permeation losses through the primary vessel

Effects of Large Area Liquid Lithium Limiters on Spherical Torus Plasmas

International Nuclear Information System (INIS)

Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Gettelfinger, G.; Gray, T.; Hoffman, D.; Jardin, S.; Kugel, H.; Marfuta, P.; Munsat, T.; Neumeyer, C.; Raftopoulos, S.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Delgado-Aparicio, L.; Seraydarian, R.P.; Antar, G.; Doerner, R.; Luckhardt, S.; Baldwin, M.; Conn, R.W.; Maingi, R.; Menon, M.; Causey, R.; Buchenauer, D.; Ulrickson, M.; Jones, B.; Rodgers, D.

2004-01-01

Use of a large-area liquid lithium surface as a first wall has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter

Effects of large area liquid lithium limiters on spherical torus plasmas

International Nuclear Information System (INIS)

Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Gettelfinger, G.; Gray, T.; Hoffman, D.; Jardin, S.; Kugel, H.; Marfuta, P.; Munsat, T.; Neumeyer, C.; Raftopoulos, S.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Delgado-Aparicio, L.; Seraydarian, R.P.; Antar, G.; Doerner, R.; Luckhardt, S.; Baldwin, M.; Conn, R.W.; Maingi, R.; Menon, M.; Causey, R.; Buchenauer, D.; Ulrickson, M.; Jones, B.; Rodgers, D.

2005-01-01

Use of a large-area liquid lithium surface as a limiter has significantly improved the plasma performance in the Current Drive Experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory. Previous CDX-U experiments with a partially-covered toroidal lithium limiter tray have shown a decrease in impurities and the recycling of hydrogenic species. Improvements in loading techniques have permitted nearly full coverage of the tray surface with liquid lithium. Under these conditions, there was a large drop in the loop voltage needed to sustain the plasma current. The data are consistent with simulations that indicate more stable plasmas having broader current profiles, higher temperatures, and lowered impurities with liquid lithium walls. As further evidence for reduced recycling with a liquid lithium limiter, the gas puffing had to be increased by up to a factor of eight for the same plasma density achieved with an empty toroidal tray limiter

Design of a permeator against vacuum for tritium extraction from eutectic lithium-lead in a DCLL DEMO

Energy Technology Data Exchange (ETDEWEB)

Garcinuño, Belit, E-mail: belit.garcinuno@ciemat.es [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Rapisarda, David [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Fernández, Iván [Fundación & Departamento de Ingeniería Energética, UNED, Madrid (Spain); CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain); Moreno, Carlos; Palermo, Iole; Ibarra, Ángel [CIEMAT-LNF (Laboratorio Nacional de Fusión), Madrid (Spain)

2017-04-15

Highlights: • A conceptual design of a Permeator Against Vacuum is presented. • The efficiency is dependent on geometry and Tritium transport. • The use of different membrane materials is discussed. • A squared PAV with alternated PbLi flowing and vacuum flat ducts is designed. • 80% efficiency of Tritium extraction is accomplished under DCLL-BB requirements. - Abstract: One of the most important issues in future fusion power plants is the extraction of tritium generated in the breeders in order to achieve self-sufficiency. When the breeder is a liquid metal one of the most promising techniques is the Permeation Against Vacuum, whose principle is based on tritium diffusion through a permeable membrane in contact with the liquid metal carrier and its further extraction by a vacuum pump. A conceptual design of permeator has been developed, taking into account the features of a DEMO reactor with a Dual Coolant Lithium Lead (DCLL) breeder blanket. The study is based on the analysis of different membranes and geometries aiming at the overall efficiency (extraction capability) of the device, as well as its compatibility with the breeder material. The permeator is based on a rectangular section multi-channel distribution where the liquid metal channels and vacuum channels are alternated in order to maximize the contact area and therefore to promote tritium transport from the bulk to the walls. The resulting permeator design has an excellent estimated extraction efficiency, of 80%, in a relatively compact device.

High-power liquid-lithium jet target for neutron production

Science.gov (United States)

Halfon, S.; Arenshtam, A.; Kijel, D.; Paul, M.; Berkovits, D.; Eliyahu, I.; Feinberg, G.; Friedman, M.; Hazenshprung, N.; Mardor, I.; Nagler, A.; Shimel, G.; Tessler, M.; Silverman, I.

2013-12-01

A compact liquid-lithium target (LiLiT) was built and tested with a high-power electron gun at the Soreq Nuclear Research Center. The lithium target, to be bombarded by the high-intensity proton beam of the Soreq Applied Research Accelerator Facility (SARAF), will constitute an intense source of neutrons produced by the 7Li(p,n)7Be reaction for nuclear astrophysics research and as a pilot setup for accelerator-based Boron Neutron Capture Therapy. The liquid-lithium jet target acts both as neutron-producing target and beam dump by removing the beam thermal power (>5 kW, >1 MW/cm3) with fast transport. The target was designed based on a thermal model, accompanied by a detailed calculation of the 7Li(p,n) neutron yield, energy distribution, and angular distribution. Liquid lithium is circulated through the target loop at ˜200 °C and generates a stable 1.5 mm-thick film flowing at a velocity up to 7 m/s onto a concave supporting wall. Electron beam irradiation demonstrated that the liquid-lithium target can dissipate electron power areal densities of >4 kW/cm2 and volume power density of ˜2 MW/cm3 at a lithium flow of ˜4 m/s while maintaining stable temperature and vacuum conditions. The LiLiT setup is presently in online commissioning stage for high-intensity proton beam irradiation (1.91-2.5 MeV, 1-2 mA) at SARAF.

Electrical detection of liquid lithium leaks from pipe joints

Energy Technology Data Exchange (ETDEWEB)

Schwartz, J. A., E-mail: jschwart@pppl.gov; Jaworski, M. A.; Mehl, J.; Kaita, R.; Mozulay, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)

2014-11-15

A test stand for flowing liquid lithium is under construction at Princeton Plasma Physics Laboratory. As liquid lithium reacts with atmospheric gases and water, an electrical interlock system for detecting leaks and safely shutting down the apparatus has been constructed. A defense in depth strategy is taken to minimize the risk and impact of potential leaks. Each demountable joint is diagnosed with a cylindrical copper shell electrically isolated from the loop. By monitoring the electrical resistance between the pipe and the copper shell, a leak of (conductive) liquid lithium can be detected. Any resistance of less than 2 kΩ trips a relay, shutting off power to the heaters and pump. The system has been successfully tested with liquid gallium as a surrogate liquid metal. The circuit features an extensible number of channels to allow for future expansion of the loop. To ease diagnosis of faults, the status of each channel is shown with an analog front panel LED, and monitored and logged digitally by LabVIEW.

Corrosion resistance investigation of vanadium alloys in liquid lithium

Energy Technology Data Exchange (ETDEWEB)

Borovitskaya, I. V., E-mail: symp@imet.ac.ru [Russian Academy of Sciences, Baikov Institute of Metallurgy and Materials Science (Russian Federation); Lyublinskiy, I. E. [JSC Red Star (Russian Federation); Bondarenko, G. G. [National Research University Higher School of Economics (Russian Federation); Paramonova, V. V. [Russian Academy of Sciences, Baikov Institute of Metallurgy and Materials Science (Russian Federation); Korshunov, S. N.; Mansurova, A. N. [National Research Center Kurchatov Institute (Russian Federation); Lyakhovitskiy, M. M. [Russian Academy of Sciences, Baikov Institute of Metallurgy and Materials Science (Russian Federation); Zharkov, M. Yu. [JSC Red Star (Russian Federation)

2016-12-15

A major concern in using vanadium alloys for first wall/blanket systems in fusion reactors is their activity with regard to nonmetallic impurities in the coolants. This paper presents the results of studying the corrosion resistance in high-purity liquid lithium (with the nitrogen and carbon content of less than 10{sup –3} wt %) of vanadium and vanadium alloys (V–1.86Ga, V–3.4Ga–0.62Si, V–4.81Ti–4.82Cr) both in the initial state and preliminarily irradiated with Ar+ ions with energy of 20 keV to a dose of 10{sup 22} m{sup –2} at an irradiation temperature of ~400°C. The degree of corrosion was estimated by measuring the changes in the weight and microhardness. Corrosion tests were carried out under static isothermal conditions at a temperature of 600°C for 400 h. The identity of corrosion mechanisms of materials both irradiated with Ar ions and not irradiated, which consisted in an insignificant penetration of nitrogen into the materials and a substantial escape of oxygen from the materials, causing the formation of a zone with a reduced microhardness near the surface, was established. The influence of the corrosive action of lithium on the surface morphology of the materials under study was found, resulting in the manifestation of grain boundaries and slip lines on the sample surface, the latter being most clearly observed in the case of preliminary irradiation with Ar ions.

Stable lithium electrodeposition in liquid and nanoporous solid electrolytes

KAUST Repository

Lu, Yingying

2014-08-10

Rechargeable lithium, sodium and aluminium metal-based batteries are among the most versatile platforms for high-energy, cost-effective electrochemical energy storage. Non-uniform metal deposition and dendrite formation on the negative electrode during repeated cycles of charge and discharge are major hurdles to commercialization of energy-storage devices based on each of these chemistries. A long-held view is that unstable electrodeposition is a consequence of inherent characteristics of these metals and their inability to form uniform electrodeposits on surfaces with inevitable defects. We report on electrodeposition of lithium in simple liquid electrolytes and in nanoporous solids infused with liquid electrolytes. We find that simple liquid electrolytes reinforced with halogenated salt blends exhibit stable long-term cycling at room temperature, often with no signs of deposition instabilities over hundreds of cycles of charge and discharge and thousands of operating hours. We rationalize these observations with the help of surface energy data for the electrolyte/lithium interface and impedance analysis of the interface during different stages of cell operation. Our findings provide support for an important recent theoretical prediction that the surface mobility of lithium is significantly enhanced in the presence of lithium halide salts. Our results also show that a high electrolyte modulus is unnecessary for stable electrodeposition of lithium.

Use of Russian technology of ship reactors with lead-bismuth coolant in nuclear power

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Chitaykin, V.I.; Gromov, B.F.; Grigoryv, O.G.; Dedoul, A.V.; Toshinsky, G.I.; Dragunov, Yu.G.; Stepanov, V.S.

2000-01-01

The experience of using lead-bismuth coolant in Russian nuclear submarine reactors has been presented. The fundamental statements of the concept of using the reactors cooled by lead-bismuth alloy in nuclear power have been substantiated. The results of developments for using lead bismuth coolant in nuclear power have been presented. (author)

The Liquid Lithium Limiter control system on FTU

Energy Technology Data Exchange (ETDEWEB)

Bertocchi, A; Panella, M; Vitale, V [Associazione EURATOM- ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati (RM) (Italy); Sinibaldi, S [Rome University ' ' Tor Vergata ' ' , Informatics, Systems and Production Dept., Via del Politecnico 1, 00133 Rome (Italy)

2006-07-01

In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system configuration was installed for testing on the FTU tokamak. The liquid lithium flows through capillaries from a reservoir to the side facing the plasma to form a thin liquid lithium film. The system is composed of three stainless steel sections, which contain two thermocouples each. A heating system brings the Li temperature to about 200 {sup o}C allowing the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. [M. Apicella, G. Mazzitelli et al., First experiment with Lithium Limiter on FTU, 17{sup o} International Conference on Plasma Surface Interaction in Controlled Fusion Devices, 22 - 26 May 2006, Hefei Anhui, China]. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22{sup TM} modules and a CORBA/PHP/MySQL software architecture [A. Bertocchi, S. Podda, V. Vitale, Fusion Eng. Des. 74 (2005) 787-791]. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLab and Java environments respectively to monitor the lithium temperature coming from thermocouples - have been also implemented. The control system allows regulating the heater temperature in each section of the LLL to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During plasma operations the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink{sup TM} tool - has been realized. (author)

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.

Results and code prediction comparisons of lithium-air reaction and aerosol behavior tests

International Nuclear Information System (INIS)

Jeppson, D.W.

1986-03-01

The Hanford Engineering Development Laboratory (HEDL) Fusion Safety Support Studies include evaluation of potential safety and environmental concerns associated with the use of liquid lithium as a breeder and coolant for fusion reactors. Potential mechanisms for volatilization and transport of radioactive metallic species associated with breeder materials are of particular interest. Liquid lithium pool-air reaction and aerosol behavior tests were conducted with lithium masses up to 100 kg within the 850-m 3 containment vessel in the Containment Systems Test Facility. Lithium-air reaction rates, aerosol generation rates, aerosol behavior and characterization, as well as containment atmosphere temperature and pressure responses were determined. Pool-air reaction and aerosol behavior test results were compared with computer code calculations for reaction rates, containment atmosphere response, and aerosol behavior. The volatility of potentially radioactive metallic species from a lithium pool-air reaction was measured. The response of various aerosol detectors to the aerosol generated was determined. Liquid lithium spray tests in air and in nitrogen atmospheres were conducted with lithium temperatures of about 427 0 and 650 0 C. Lithium reaction rates, containment atmosphere response, and aerosol generation and characterization were determined for these spray tests

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.

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

Spectroscopic measurements of lithium influx from an actively water-cooled liquid lithium limiter on FTU

Energy Technology Data Exchange (ETDEWEB)

Apruzzese, G.M., E-mail: gerarda.apruzzese@enea.it; Apicella, M.L.; Maddaluno, G.; Mazzitelli, G.; Viola, B.

2017-04-15

Since 2006, experiments using a liquid lithium limiter (LLL) were successfully performed on FTU, pointing out the problem of the quantity of lithium in the plasma, especially in conditions of strong evaporation due to the high temperature of limiter surface. In order to avoid the strong evaporation it is necessary to control the temperature by removing the heat from the limiter during the plasma exposure. To explore this issue a new actively cooled lithium limiter (CLL) has been installed and tested in FTU. Suitable monitors to detect the presence of lithium in the plasma are the spectroscopic diagnostics in the visible range that permit to measure the flux of lithium, coming from the limiter surface, through the brightness of the LiI spectral lines. For this aim an Optical Multichannel Analyser (OMA) spectrometer and a single wavelength impurities monitor have been used. The analysis of the Li influx signals has permitted to monitor the effects of interaction between the plasma and the limiter connected to the thermal load. Particular attention has been paid on the possible occurrence of sudden rise of the signals, which is an index of a strong interaction that could lead to a disruption. On the other hand, the appearance of significant signals gives useful indication if the interaction with the plasma has taken place.

Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

Science.gov (United States)

Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

2009-01-01

Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

3D flow simulation of liquid lead in the erosion test facility for ADS materials

International Nuclear Information System (INIS)

Muscher, Heinrich; Kieser, Martin; Weisenburger, Alfons; Mueller, Georg

2009-01-01

Future nuclear reactor concepts, such as GEN IV or ADS use liquid lead for neutron multiplication and coolant purposes. The design concepts assumes that the structural material is in contact with the liquid metal at temperatures up to 600 C and a flow rate of 20 m/s. Therefore a significant effect of liquid metal corrosion/erosion is expected. The paper describes the fluid dynamical simulation of the ADS erosion test facility. Earlier studies on the laminar flow modeling were completed by introduction of transient behavior and extended to 3D-models. The results for liquid lead should be transferable to LBE (lead bismuth eutectic). Further work has to include a mass transport model for modeling of the global isothermal erosion rate of the structural material dependent on time (for liquid lead and LBE).

Corrosion behavior of welds in oxygen containing liquid lead

Energy Technology Data Exchange (ETDEWEB)

Heinzel, A.; Weisenburger, A.; Mueller, G. [Karlsruhe Institute of Technology (Germany). Inst. for Pulsed Power and Microwave Technology

2012-11-01

Liquid lead (Pb) and lead-bismuth eutectic (LBE) have been considered as coolant and/or spallation target in future accelerator driven systems (ADS). Therefore, in the recent years a lot of corrosion experiments on conventional steels were carried out in these heavy liquid metals. Beside these experiments, also tests on welded joints are required. Therefore ferritic/martensitic (F/M) steels (P91, P92) as well as an ODS steel were joint with TIG (Tungsten-Inert-Gas), EB (Electron Beam) and friction stir welding. After that, specimens were exposed to 10{sup -6} and 10{sup -8}wt% oxygen containing lead at 550 C for about 2000h. Weld regions having similar chemical composition and similar structure due to a heat treatment after the welding process show a corrosion behaviour under these conditions that is similar to that of the respective bulk material. (orig.)

Compensation of equipment housing elements of reactor units with heavy liquid metal coolant vessel temperature deformations

International Nuclear Information System (INIS)

Lebedevich, V.; Ahmetshin, M.; Mendes, D.; Kaveshnikov, S.; Vinogradov, A.

2015-01-01

In Russia a lot of different versions of fast reactors (FRs) are investigated and one of these is FR cooled by liquid lead and liquid lead-bismuth alloy. In this poster we are interested by FR with concrete vessel; its components are placed in cavities inside the vessel, and connected by a channel system. During the installation the equipment components are placed in several equipment housings. Between these housings there are cavities with coolant. The alignment of the housings should be provided. It can be broken by irregular concrete vessel heating during FR starting or other transition regimes. Our goal is to suggest a list of designing steps to compensate temperature deformations of equipment housing elements. A simplified model of equipment housing was suggested. It consists of two cylinders - tunnels in the concrete vessel, separated by a cavity filled by coolant and inert gas. The bottom part was considered as heated to 420 C. degrees while in the top part temperature decreased to 45 C. degrees (on the concrete surface). According to this data, results show that temperature gradient leads to a concrete layer dislocation of about 12.5 mm, which can lead to damage and breaking alignment. We propose the following solution to compensate for temperature deformation: -) to chisel out part of the upper top of the insulating concrete; -) to install an adequate misalignment of equipment housing elements preliminary; and -) to use a torsion system like a piston-type device for providing additional strength in order to compensate deformation and vibrations

Evaluation of molten lead mixing in sodium coolant by diffusion for application to PAHR

International Nuclear Information System (INIS)

Chawla, T.C.; Pedersen, D.R.; Leaf, G.; Minkowycz, W.J.

1983-01-01

In post-accident heat removal (PAHR) applications the use of a lead slab is being considered for protecting a porous bed of steel shots in ex-vessel cavity from direct impingement of molten steel or fuel upon vessel failure following a hypothetical core dissembly accident in an LMFBR. The porous bed is provided to increase coolability of the fuel debris by the sodium coolant. The objectives of the present study are (1) to determine melting rates of lead slabs of various thicknesses in contact with sodium coolant and (2) to evaluate the extent of penetration and mixing rates of molten lead into sodium coolant by molecular diffusion alone

The liquid lithium limiter control system on FTU

Energy Technology Data Exchange (ETDEWEB)

Bertocchi, A. [EURATOM-ENEA Association, Frascati Research Center, Via E. Fermi 45, 00044 Frascati (Rome) (Italy)], E-mail: bertocchi@frascati.enea.it; Di Donna, M [Department of Informatics, Systems and Productions, University of Rome Tor Vergata, Rome (Italy); Panella, M; Vitale, V [EURATOM-ENEA Association, Frascati Research Center, Via E. Fermi 45, 00044 Frascati (Rome) (Italy)

2007-10-15

In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system (CPS) configuration was installed to test on Tokamak FTU. The liquid lithium flows through capillaries from a reservoir to the side faced to the plasma to form a thin lithium film as wall coating. The system includes three stainless steel cases, which contain two thermocouples each one. A heating system brings the Li temperature about 200 deg. C to allow the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22{sup TM} modules and a CORBA/PHP/MySQL software architecture. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLAB{sup TM} and Java environments, respectively, to monitor the lithium temperature coming from thermocouples - have been also implemented. The LLL control system allows to regulate the heater temperature in each unit to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During the plasma shot the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink{sup TM} tool - has been realized.

The liquid lithium limiter control system on FTU

International Nuclear Information System (INIS)

Bertocchi, A.; Di Donna, M.; Panella, M.; Vitale, V.

2007-01-01

In the second half of 2005, a liquid lithium limiter (LLL) with capillary porous system (CPS) configuration was installed to test on Tokamak FTU. The liquid lithium flows through capillaries from a reservoir to the side faced to the plasma to form a thin lithium film as wall coating. The system includes three stainless steel cases, which contain two thermocouples each one. A heating system brings the Li temperature about 200 deg. C to allow the liquid to flow. This temperature, monitored by thermocouples, needs to be controlled. To carry out this experimental procedure, some new features have been introduced in the existent control system based on Opto22 TM modules and a CORBA/PHP/MySQL software architecture. The historical data storage to keep the lithium temperature evolution has been added. Two graphical tools - developed in MATLAB TM and Java environments, respectively, to monitor the lithium temperature coming from thermocouples - have been also implemented. The LLL control system allows to regulate the heater temperature in each unit to reach operational conditions, where the temperature adjustment can be performed either automatically through a specific control law or manually by the operator. During the plasma shot the system switches off the limiter power supply to prevent instruments damage. Moreover, in the same experimental context, a first approach to automatically obtain executable code - starting from control laws designed by Simulink TM tool - has been realized

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

Liquid lithium blanket processing studies

International Nuclear Information System (INIS)

Talbot, J.B.; Clinton, S.D.

1979-01-01

The sorption of tritium on yttrium from flowing molten lithium and the subsequent release of tritium from yttrium for regeneration of the metal sorbent were investigated to evaluate the feasibility of such a tritium-recovery process for a fusion reactor blanket of liquid lithium. In initial experiments with the forced convection loop, yttrium samples were contacted with lithium at 300 0 C. A mass transfer coefficient of 2.5 x 10 - cm/sec, which is more than an order of magnitude less than the value measured in earlier static experiments, was determined for the flowing lithium system. Rates of tritium release from yttrium samples were measured to evaluate possible thermal regeneration of the sorbent. Values for diffusion coefficients at 505, 800, and 900 0 C were estimated to be 1.1 x 10 -13 , 4.9 x 10 -12 , and 9.3 x 10 -10 cm 2 /sec, respectively. Tritium release from yttrium was investigated at higher temperatures and with hydrogen added to the argon sweep gas to provide a reducing atmosphere

A new facility for studying plasma interacting with flowing liquid lithium surface

International Nuclear Information System (INIS)

Cao, X.; Ou, W.; Tian, S.; Wang, C.; Zhu, Z.; Wang, J.; Gou, F.; Yang, D.; Chen, S.

2014-01-01

A new facility to study plasmas interacting with flowing liquid lithium surface was designed and is constructing in Sichuan University. The integrated setup includes the liquid lithium circulating part and linear high density plasma generator. The circulating part is consisted of main loop, on-line monitor system, lithium purification system and temperature programmed desorption system. In our group a linear high density plasma generator was built in 2012. Three coils were mounted along the vessel to produce an axial magnetic field inside. The magnetic field strength is up to 0.45 T and work continuously. Experiments on plasmas interacting with free flowing liquid lithium surface will be performed

Experiments with liquid metal walls: Status of the lithium tokamak experiment

Energy Technology Data Exchange (ETDEWEB)

Kaita, Robert, E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Berzak, Laura; Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Hammett, Gregory; Jacobson, Craig M.; Jones, Andrew; Kozub, Thomas; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Lucia, Matthew; Lundberg, Daniel; Majeski, Richard; Mansfield, Dennis; Menard, Jonathan; Spaleta, Jeffrey; Strickler, Trevor; Timberlake, John [Princeton Plasma Physics Laboratory, Princeton, NJ (United States)

2010-11-15

Abstarct: Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The lithium tokamak experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the current drive experiment-upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in ohmically heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy (CHERS). Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions.

Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

International Nuclear Information System (INIS)

Kaita, Robert; Berzak, Laura; Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Hammett, Gregory; Jacobson, Craig M.; Jones, Andrew; Kozub, Thomas; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Lucia, Matthew; Lundberg, Daniel; Majeski, Richard; Mansfield, Dennis; Menard, Jonathan; Spaleta, Jeffrey; Strickler, Trevor; Timberlak, John

2010-01-01

Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy. Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions.

Lithium attenuates lead induced toxicity on mouse non-adherent bone marrow cells.

Science.gov (United States)

Banijamali, Mahsan; Rabbani-Chadegani, Azra; Shahhoseini, Maryam

2016-07-01

Lead is a poisonous heavy metal that occurs in all parts of environment and causes serious health problems in humans. The aim of the present study was to investigate the possible protective effect of lithium against lead nitrate induced toxicity in non-adherent bone marrow stem cells. Trypan blue and MTT assays represented that exposure of the cells to different concentrations of lead nitrate decreased viability in a dose dependent manner, whereas, pretreatment of the cells with lithium protected the cells against lead toxicity. Lead reduced the number and differentiation status of bone marrow-derived precursors when cultured in the presence of colony stimulating factor (CSF), while the effect was attenuated by lithium. The cells treated with lead nitrate exhibited cell shrinkage, DNA fragmentation, anion superoxide production, but lithium prevented lead action. Moreover, apoptotic indexes such as PARP cleavage and release of HMGB1 induced by lead, were protected by lithium, suggesting anti-apoptotic effect of lithium. Immunoblot analysis of histone H3K9 acetylation indicated that lithium overcame lead effect on acetylation. In conclusion, lithium efficiently reduces lead toxicity suggesting new insight into lithium action which may contribute to increased cell survival. It also provides a potentially new therapeutic strategy for lithium and a cost-effective approach to minimize destructive effects of lead on bone marrow stem cells. Copyright © 2016 Elsevier GmbH. All rights reserved.

Design Constraints for Liquid-Protected Divertors

International Nuclear Information System (INIS)

Shin, S.; Abdel-Khalik, S.I.; Yoda, M.

2005-01-01

Recent work on liquid-surface-protected plasma facing components has resulted in the establishment of operating windows for candidate liquids, as well as limits on the maximum allowable liquid surface temperature in order to limit plasma impurities from liquid evaporation. In this study, an additional constraint on the maximum allowable surface temperature gradient (i.e., heat flux gradient) has been quantified. Spatial variations in the wall and liquid surface temperatures are expected due to variations in the incident radiation and particle fluxes. Thermocapillary forces created by such temperature gradients can lead to film rupture and dry spot formation in regions of elevated local temperatures. Here, attention has been focused on ''non-flowing'' thin liquid films similar to those formed on the surface of porous wettedwall components. Future analyses will include the effects of macroscopic fluid motion, and MHD forces.A numerical model using the level contour reconstruction method was used to follow the evolution of the liquid free surface above a non-isothermal solid surface. The model was used to develop generalized charts for the maximum allowable spatial temperature gradients (i.e., the critical Marangoni number) as a function of the governing non-dimensional variables, viz. the Weber, Froude, and Prandtl numbers, and aspect ratio. Attention was focused on the asymptotic limit for thin liquid films (i.e., low aspect ratio) which provides a lower bound for the maximum allowable temperature gradients. Specific examples for lithium, Flibe, lithium-lead, tin, and gallium are presented. The generalized charts developed in this investigation will allow reactor designers to identify design windows for successful operation of liquid-protected plasma facing components for various coolants, film thicknesses, and operating conditions

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

Energy Technology Data Exchange (ETDEWEB)

Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Brooks, A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Lopes-Cardozo, N. [TU/Eindhoven, Eindhoven (Netherlands); Menard, J.; Ono, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Rindt, P. [TU/Eindhoven, Eindhoven (Netherlands); Tresemer, K. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

2016-11-15

Highlights: • An upgrade path for the NSTX-U tokamak is proposed that maintains scientific productivity while enabling exploration of novel, liquid metal PFC. • Pre-filled liquid metal divertor targets are proposed as an intermediate step that mitigates technical and scientific risks associated with liquid metal PFC. • Analysis of leading edge features show a strong link between engineering design considerations and expected performance as a PFC. • A method for optimizing porous liquid metal targets restrained by capillary forces is provided indicating pore-sizes well within current technical capabilities. - Abstract: Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. Two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

Stable lithium electrodeposition in liquid and nanoporous solid electrolytes

KAUST Repository

Lu, Yingying; Tu, Zhengyuan; Archer, Lynden A.

2014-01-01

of these metals and their inability to form uniform electrodeposits on surfaces with inevitable defects. We report on electrodeposition of lithium in simple liquid electrolytes and in nanoporous solids infused with liquid electrolytes. We find that simple liquid

Neutron Physics aspects of using lead as a coolant in Fast Reactors

International Nuclear Information System (INIS)

Kiefhaber, E.

1991-02-01

The use of lead as a coolant for fast reactors is being considered as an attractive alternative in the USSR, especially with respect to its inherent safety features. In order to come to an own assessment at KfK, some investigations have been performed concerning a comparison of the nuclear characteristics of fast reactors with lead and sodium cooling. The studies have shown, that the nuclear and thermal hydraulic design calculations do not face special problems and that the nuclear characteristics of both types of cores do not differ essentially, except for the coolant density or void effect, which is more favourable for smaller sized lead cooled cores. A proper safety assessment of lead cooled cores will however require more detailed safety studies. Crucial points of lead cooling are the strong corrosion of austenitic steels in lead and the unknown behavior of ferritic steels in lead and under irradiation

High-flux neutron source based on a liquid-lithium target

Science.gov (United States)

Halfon, S.; Feinberg, G.; Paul, M.; Arenshtam, A.; Berkovits, D.; Kijel, D.; Nagler, A.; Eliyahu, I.; Silverman, I.

2013-04-01

A prototype compact Liquid Lithium Target (LiLiT), able to constitute an accelerator-based intense neutron source, was built. The neutron source is intended for nuclear astrophysical research, boron neutron capture therapy (BNCT) in hospitals and material studies for fusion reactors. The LiLiT setup is presently being commissioned at Soreq Nuclear research Center (SNRC). The lithium target will produce neutrons through the 7Li(p,n)7Be reaction and it will overcome the major problem of removing the thermal power generated by a high-intensity proton beam, necessary for intense neutron flux for the above applications. The liquid-lithium loop of LiLiT is designed to generate a stable lithium jet at high velocity on a concave supporting wall with free surface toward the incident proton beam (up to 10 kW). During off-line tests, liquid lithium was flown through the loop and generated a stable jet at velocity higher than 5 m/s on the concave supporting wall. The target is now under extensive test program using a high-power electron-gun. Up to 2 kW electron beam was applied on the lithium flow at velocity of 4 m/s without any flow instabilities or excessive evaporation. High-intensity proton beam irradiation will take place at SARAF (Soreq Applied Research Accelerator Facility) superconducting linear accelerator currently in commissioning at SNRC.

Optimization of the first wall for the DEMO water cooled lithium lead blanket

Energy Technology Data Exchange (ETDEWEB)

Aubert, Julien, E-mail: julien.aubert@cea.fr [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Aiello, Giacomo [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Bachmann, Christian [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Di Maio, Pietro Alessandro [Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, Rosario [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy); Li Puma, Antonella; Morin, Alexandre [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Tincani, Amelia [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy)

2015-10-15

Highlights: • This paper presents the optimization of the first wall of the water cooled lithium lead DEMO blanket with pressurized water reactor condition and circular channels in order to find the best geometry that can allow the maximum heat flux considering design criteria since an estimate of the engineering limit of the first wall heat load capacity is an essential input for the decision to implement limiters in DEMO. • An optimization study was carried out for the flat first wall design of the DEMO Water-Cooled Lithium Lead considering thermal and mechanical constraint functions, assuming T{sub inlet}/T{sub outlet} equal to 285 °C/325 °C, based on geometric design parameters. • It became clear that through the optimization the advantages of a waved First Wall are diminished. • The analysis shows that the maximum heat load could achieve 2.53 MW m{sup −2}, but considering assumptions such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total first wall thickness ≤22 mm, heat flux is limited to 1.57 MW m{sup −2}. - Abstract: The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m{sup −2} with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet

Evaluation of compatibility of flowing liquid lithium curtain for blanket with core plasma in fusion reactors

International Nuclear Information System (INIS)

Deng Baiquan; Huang Jinhua; Peng Lilin; Yan Jiancheng

2003-01-01

A global model analysis of the compatibility of flowing liquid lithium curtain for blanket with core plasma has been performed. The relationships between the surface temperature of lithium curtain and mean effective plasma charges, fuel dilution and produced fusion power have been obtained. Results show that under normal circumstances, the evaporation of liquid lithium does not affect Z eff seriously, but affects fuel dilution and fusion power sensitively. The authors have investigated the relationships between the flow velocity of liquid lithium and its surface temperature rise based on the conditions of the option II of the fusion experimental breeder (FEB-E) design with reversed shear configuration and fairly high power density. The authors concluded that the effects of evaporation from liquid lithium curtain for FEB-E on plasma are negligible even if the flow velocity of liquid lithium is as low as 0.5 m·s -1 . Finally, the sputtering yield of liquid lithium saturated by hydrogen isotopes is briefly discussed

Ionic Liquid-Nanoparticle Hybrid Electrolytes and their Application in Secondary Lithium-Metal Batteries

KAUST Repository

Lu, Yingying; Das, Shyamal K.; Moganty, Surya S.; Archer, Lynden A.

2012-01-01

Ionic liquid-tethered nanoparticle hybrid electrolytes comprised of silica nanoparticles densely grafted with imidazolium-based ionic liquid chains are shown to retard lithium dendrite growth in rechargeable batteries with metallic lithium anodes

Core performance of equilibrium fast reactors for different coolant materials and fuel types

International Nuclear Information System (INIS)

Mizutani, Akihiko; Sekimoto, Hiroshi

1998-01-01

Parametric studies with several coolant and fuel materials in the equilibrium state are performed for fast reactors in which natural uranium is fed and all of the actinides are confined. Sodium, sodium-potassium, lead, lead-bismuth and helium coolant materials, and oxide, nitride and metal fuels are employed to compare the neutronic characteristics in the equilibrium state. As to the criticality performance, sodium-potassium shows the best performance among the liquid metal coolants and the metallic fuel indicates the best performance

Conceptual design study of the hylife lithium waterfall laser fusion chamber. FY 1978 annual report to Lawrence Livermore Laboratory

International Nuclear Information System (INIS)

1978-01-01

Conceptual design studies of the target chamber defined the general configuration and dimensions of the chamber and the inlet plenum, orifice plate, and nozzle plate concepts required to generate the desired lithium jet fall. Preliminary studies were performed of the target chamber interfaces with the liquid lithium supply system, the laser system, the pellet injection system, and the target chamber mounting and support system. Target chamber environmental effects resulting from typical thermonuclear burns were evaluated. The outlet region of the target chamber was outlined conceptually, and preliminary design considerations were given to the annular graphite reflector regions of the target chamber and the associated liquid lithium coolant passages

Hydrogen and helium recycling from stirred liquid lithium under steady state plasma bombardment

Energy Technology Data Exchange (ETDEWEB)

Hirooka, Yoshi, E-mail: hirooka.yoshihiko@nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); The Graduate School for Advanced Studies, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Zhou, Haishan [The Graduate School for Advanced Studies, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Ono, Masa [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States)

2014-12-15

For improved core performance via edge plasma-wall boundary control, solid and liquid lithium has been used as a plasma-facing material in a number of confinement experiments over the past several decades. Unfortunately, it is unavoidable that lithium is saturated in the surface region with implanted hydrogenic species as well as oxygen-containing impurities. For steady state operation, a flowing liquid lithium divertor with forced convection would probably be required. In the present work, the effects of liquid stirring to simulate forced convection have been investigated on the behavior of hydrogen and helium recycling from molten lithium at temperatures up to ∼350 °C. Data indicate that liquid stirring reactivates hydrogen pumping via surface de-saturation and/or uncovering impurity films, but can also induce helium release via surface temperature change.

Properties of lithium and its handling

International Nuclear Information System (INIS)

Asada, Takashi; Kano, Shigeki; Tachi, Toshiaki; Kawai, Masataka

2000-09-01

Lithium is one of good coolants because of high boiling point (1317degC), small specific gravity (0.47 at 600degC) and large specific heat (1 cal/g/degC). Therefore if lithium will be used in fast reactor for coolant, the heat efficiency of reactor will largely increase. Here the fundamental properties of lithium and the results of examination on chemical reaction, combustion and extinction are shown. These examinations were also carried out on sodium to compare with lithium. The differences between both are that lithium reacts more moderately with water, not explosive, and is not combustible but after ignition burns at higher temperature and longer. (author)

Development of a large lithium coolant system for operation under vacuum

International Nuclear Information System (INIS)

Kolowith, R.; Schwartz, K.E.; Meadows, G.E.; Berg, J.D.

1983-11-01

Argon and vacuum systems for the Experimental Lithium System (ELS) were tested to demonstrate vacuum-break capability, vacuum pumping performance, and vacuum sensor compatibility with a hostile liquid metal vapor/aerosol environment. Mechanical, diffusion and cryogenic vacuum pumps were evaluated. High-vacuum levels in the 10 -3 Pa range were achieved over a 270 0 C flowing lithium system. Ionization, thermal conductivity, capacitance manometer, and compound-type pressure sensors were evaluated to determine the effects of this potentially deleterious environment. Screening elbows were evaluated as pressure sensor protective devices. A dual-purpose vacuum-level/nitrogen partial-pressure sensor was evaluated as a means of detecting air in-leakage. Several types of static mechanical vacuum seals were also evaluated. Measurements of the vapor/aerosol generation were made at several system locations and operating conditions

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

Effect of elevated lithium on the waterside corrosion of zircaloy-4: Experimental and predictive studies

International Nuclear Information System (INIS)

Pecheur, D.; Giordano, A.; Picard, E.; Billot, P.; Thomazet, J.

1997-01-01

Lithium and boron content in the coolant are known to influence the oxidation behaviour of the fuel cladding. Since new PWR operating conditions could consist in an increase of the lithium and the boron concentration in the coolant early in the cycle, a specific study has been conducted to analyze and to predict the effect of such new water chemistry conditions on the oxidation kinetics of the Zircaloy-4 material. Experimental studies have been performed in out-of-pile loop tests, under one and two phase flow heat transfer in various water chemistry conditions (0≤Li≤350 ppm, 0≤B≤1000 ppm, 0≤K≤56 ppm). A simulation of the effect of elevated lithium on the corrosion has been made using the semi-empirical COCHISE corrosion code. Under one phase flow heat transfer conditions, the addition of lithium hydroxide in the coolant increases the oxidation rate, essentially in the post-transition regime for low lithium levels (≤ 75 ppm) and immediately in the pre-transition phase for very high lithium level (350 ppm). Under two phase flow heat transfer, an enhancement of the corrosion is observed in the area of the rod submitted to boiling. Based on the out-of-pile loop test performed in presence of KOH instead of LiOH, such an enhancement of the corrosion appears to be due to a lithium enrichment in the oxide layer induced by boiling and not to a pH effect. The simulation of the increase of lithium content in the coolant from 2.2 to 3.5 ppm leads to an enhancement in corrosion rates which becomes only significant at high burn up. This predictive result of elevated lithium effect on corrosion is then compared with oxidation data derived from reactors operating under an elevated lithium regime. (author). 14 refs, 9 figs, 3 tabs

Effect of elevated lithium on the waterside corrosion of zircaloy-4: Experimental and predictive studies

Energy Technology Data Exchange (ETDEWEB)

Pecheur, D; Giordano, A; Picard, E; Billot, P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France); Thomazet, J [FRAMATOME, Nuclear Fuel Div., Lyon (France)

1997-02-01

Lithium and boron content in the coolant are known to influence the oxidation behaviour of the fuel cladding. Since new PWR operating conditions could consist in an increase of the lithium and the boron concentration in the coolant early in the cycle, a specific study has been conducted to analyze and to predict the effect of such new water chemistry conditions on the oxidation kinetics of the Zircaloy-4 material. Experimental studies have been performed in out-of-pile loop tests, under one and two phase flow heat transfer in various water chemistry conditions (0{<=}Li{<=}350 ppm, 0{<=}B{<=}1000 ppm, 0{<=}K{<=}56 ppm). A simulation of the effect of elevated lithium on the corrosion has been made using the semi-empirical COCHISE corrosion code. Under one phase flow heat transfer conditions, the addition of lithium hydroxide in the coolant increases the oxidation rate, essentially in the post-transition regime for low lithium levels ({<=} 75 ppm) and immediately in the pre-transition phase for very high lithium level (350 ppm). Under two phase flow heat transfer, an enhancement of the corrosion is observed in the area of the rod submitted to boiling. Based on the out-of-pile loop test performed in presence of KOH instead of LiOH, such an enhancement of the corrosion appears to be due to a lithium enrichment in the oxide layer induced by boiling and not to a pH effect. The simulation of the increase of lithium content in the coolant from 2.2 to 3.5 ppm leads to an enhancement in corrosion rates which becomes only significant at high burn up. This predictive result of elevated lithium effect on corrosion is then compared with oxidation data derived from reactors operating under an elevated lithium regime. (author). 14 refs, 9 figs, 3 tabs.

Liquid-liquid extraction to lithium isotope separation based on room-temperature ionic liquids containing 2,2'-binaphthyldiyl-17-crown-5

International Nuclear Information System (INIS)

Sun Xiaoli; Zhou Wen; Gu Lin; Qiu Dan; Ren Donghong; Gu Zhiguo; Li Zaijun

2015-01-01

A novel liquid-liquid extraction system was investigated for the selective separation of lithium isotopes using ionic liquids (ILs = C 8 mim + PF 6 - , C 8 mim + BF 4 - , and C 8 mim + NTf 2 - ) as extraction solvent and 2,2'-binaphthyldiyl-17-crown-5 (BN-17-5) as extractant. The effects of the concentration of lithium salt, counter anion of lithium salt, initial pH of aqueous phase, extraction temperature, and time on the lithium isotopes separation were discussed. Under optimized conditions, the maximum single-stage separation factor α of 6 Li/ 7 Li obtained in the present study was 1.046 ± 0.002, indicating the lighter isotope 6 Li was enriched in IL phase while the heavier isotope 7 Li was concentrated in the solution phase. The formation of 1:1 complex Li(BN-17-5) + in the IL phase was determined on the basis of slope analysis method. The large value of the free energy change (-ΔG° = 92.89 J mol -1 ) indicated the high separation capability of the Li isotopes by BN-17-5/IL system. Lithium in Li(BN-17-5) + complex was stripped by 1 mol L -1 HCl solution. The extraction system offers high efficiency, simplicity, and green application prospect to lithium isotope separation. (author)

VUV/XUV measurements of impurity emission in plasmas with liquid lithium surfaces on LTX

International Nuclear Information System (INIS)

Tritz, Kevin; Finkenthal, Michael; Stutman, Dan; Bell, Ronald E; Boyle, Dennis; Kaita, Robert; Kozub, Tom; Lucia, Matthew; Majeski, Richard; Merino, Enrique; Schmitt, John; Beiersdorfer, Peter; Clementson, Joel; Kubota, Shigeyuki

2014-01-01

The VUV/XUV spectrum has been measured on the Lithium Tokamak eXperiment (LTX) using a transmission grating imaging spectrometer (TGIS) coupled to a direct-detection x-ray charge-coupled device camera. TGIS data show significant changes in the ratios between the lithium and oxygen impurity line emission during discharges with varying lithium wall conditions. Lithium coatings that have been passivated by lengthy exposure to significant levels of impurities contribute to a large O/Li ratio measured during LTX plasma discharges. Furthermore, previous results have indicated that a passivated lithium film on the plasma facing components will function as a stronger impurity source when in the form of a hot liquid layer compared to a solid lithium layer. However, recent TGIS measurements of plasma discharges in LTX with hot stainless steel boundary shells and a fresh liquid lithium coating show lower O/Li impurity line ratios when compared to discharges with a solid lithium film on cool shells. These new measurements help elucidate the somewhat contradictory results of the effects of solid and liquid lithium on plasma confinement observed in previous experiments. (paper)

Extraction of lithium from salt lake brine using room temperature ionic liquid in tributyl phosphate

International Nuclear Information System (INIS)

Shi, Chenglong; Jia, Yongzhong; Zhang, Chao; Liu, Hong; Jing, Yan

2015-01-01

Highlights: • We proposed a new system for Li recovery from salt lake brine by extraction using an ionic liquid. • Cation exchange was proposed to be the mechanism of extraction followed in ionic liquid. • This ionic liquid system shown considerable extraction ability for lithium and the single extraction efficiency of lithium reached 87.28% under the optimal conditions. - Abstract: Lithium is known as the energy metal and it is a key raw material for preparing lithium isotopes which have important applications in nuclear energy source. In this work, a typical room temperature ionic liquid (RTILs), 1-butyl-3-methyl-imidazolium hexafluorophosphate ([C 4 mim][PF 6 ]), was used as an alternative solvent to study liquid/liquid extraction of lithium from salt lake brine. In this system, the ionic liquid, NaClO 4 and tributyl phosphate (TBP) were used as extraction medium, co-extraction reagent and extractant respectively. The effects of solution pH value, phase ratio, ClO 4 − amount and other factors on lithium extraction efficiency had been investigated. Optimal extraction conditions of this system include the ratio of TBP/IL at 4/1 (v/v), O/A at 2:1, n(ClO 4 − )/n(Li + ) at 2:1, the equilibration time of 10 min and unadjusted pH. Under the optimal conditions, the single extraction efficiency of lithium was 87.28% which was much higher than the conventional extraction system. Total extraction efficiency of 99.12% was obtained by triple-stage countercurrent extraction. Study on the mechanism revealed that the use of ionic liquid increased the extraction yield of lithium through cation exchange in this system. Preliminary results indicated that the use of [C 4 mim][PF 6 ] as an alternate solvent to replace traditional organic solvents (VOCs) in liquid/liquid extraction was very promising

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.

A low-risk aqueous lithium salt blanket for engineering test reactors

International Nuclear Information System (INIS)

Gierszewski, P.

1986-09-01

A simple blanket concept is proposed based on 1-3 wt.% lithium dissolved as a salt in low temperature (80 degrees C) and low pressure (0.1 MPa) water. This concept can provide, for example, a 0.5 tritium breeding ratio with 60% steel structure and 70% coverage. The use of neutron multipliers, other structural materials (especially zirconium alloys), higher coverage and higher lithium salt concentrations allows tritium breeding ratios over unity if necessary. Other advantages of this concept include the simple shield-like geometry, substantial structural volume for mechanical strength, excellent heat transfer ability of water coolant, efficient neutron and gamma shielding through the combination of high-Z structure and low-Z water, and conventional tritium recovery and control technology. This concept could initially provide the shielding needs for an engineering test reactor and later, by the addition of lithium salt and tritium recovery systems, also provide tritium breeding. This staged operation and liquid breeder/coolant allows control over the tritium inventory in the device without machine disassembly. 14 refs

Correlation of cylinder-head temperatures and coolant heat rejections of a multicylinder, liquid-cooled engine of 1710-cubic-inch displacement

Science.gov (United States)

Lundin, Bruce T; Povolny, John H; Chelko, Louis J

1949-01-01

Data obtained from an extensive investigation of the cooling characteristics of four multicylinder, liquid-cooled engines have been analyzed and a correlation of both the cylinder-head temperatures and the coolant heat rejections with the primary engine and coolant variables was obtained. The method of correlation was previously developed by the NACA from an analysis of the cooling processes involved in a liquid-cooled-engine cylinder and is based on the theory of nonboiling, forced-convection heat transfer. The data correlated included engine power outputs from 275 to 1860 brake horsepower; coolant flows from 50 to 320 gallons per minute; coolants varying in composition from 100 percent water to 97 percent ethylene glycol and 3 percent water; and ranges of engine speed, manifold pressure, carburetor-air temperature, fuel-air ratio, exhaust-gas pressure, ignition timing, and coolant temperature. The effect on engine cooling of scale formation on the coolant passages of the engine and of boiling of the coolant under various operating conditions is also discussed.

Liquid Crystals of Lithium Dodecylbenzenesulfonate for Electric Double Layer Capacitors

International Nuclear Information System (INIS)

Kuzmin, Andrey Vasil’evich; Yurtov, Evgeny V.

2016-01-01

Ionic lyotropic liquid crystals based on lithium dodecylbenzenesulfonate were used as electrolytes for electric double layer capacitors with carbon fibrous electrodes. The capacitors were tasted by cyclic voltammetry, galvanostatic charge and discharge, and impedance spectroscopy. The highest specific capacitance was achieved for electrical double layer capacitor equipped with ionic lyotropic liquid crystal of lithium dodecylbenzenesulfonate 35 wt% in water. The specific capacitance of capacitor was calculated from galvanostatic discharge curves – 15 F/g of carbon fibrous material

Lithium uptake and the corrosion of zirconium alloys in aqueous lithium hydroxide solutions

International Nuclear Information System (INIS)

Ramasubramanian, N.

1991-01-01

This paper reports on corrosion films on zirconium alloys that were analyzed for lithium by Atomic Absorption Spectroscopy (AAS), Secondary Ion Mass Spectrometry (SIMS), and Infrared Reflection Absorption Spectroscopy (IRAS). The oxides grown in reactor in dilute lithium hydroxide solution, specimens cut from Zircaloy, and Zr-2.5Nb alloy pressure tubes removed from CANDU (Canada Deuterium Uranium, Registered Trademark) reactors showed low concentrations of lithium (4 to 50 ppm). The lithium was not leachable in a warm dilute acid. 6 Li undergoes transmutation by the 6 Li(n,t) 4 He reaction. However, SIMS profiles for d 7 Li were identical through the bulk oxide and the isotopic ratio was close to the natural abundance value. The lithium in the oxide, existing as adsorbed lithium on the surface, has been in dynamic equilibrium with lithium in the coolant, and, in spite of many Effective Full Power Years (EFPY) of operation, lithium added to the CANDU coolant at ∼2.5 ppm is not concentrating in the oxides. On the other hand, corrosion films grown in the laboratory in concentrated lithium hydroxide solutions were very porous and contained hundreds of ppm of lithium in the oxide

Ionic Liquid-Doped Gel Polymer Electrolyte for Flexible Lithium-Ion Polymer Batteries

Science.gov (United States)

Zhang, Ruisi; Chen, Yuanfen; Montazami, Reza

2015-01-01

Application of gel polymer electrolytes (GPE) in lithium-ion polymer batteries can address many shortcomings associated with liquid electrolyte lithium-ion batteries. Due to their physical structure, GPEs exhibit lower ion conductivity compared to their liquid counterparts. In this work, we have investigated and report improved ion conductivity in GPEs doped with ionic liquid. Samples containing ionic liquid at a variety of volume percentages (vol %) were characterized for their electrochemical and ionic properties. It is concluded that excess ionic liquid can damage internal structure of the batteries and result in unwanted electrochemical reactions; however, samples containing 40–50 vol % ionic liquid exhibit superior ionic properties and lower internal resistance compared to those containing less or more ionic liquids.

Extraction of lithium from salt lake brine using room temperature ionic liquid in tributyl phosphate

Energy Technology Data Exchange (ETDEWEB)

Shi, Chenglong [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China); University of Chinese Academy of Sciences, 100049 Beijing (China); Jia, Yongzhong [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China); Zhang, Chao [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China); University of Chinese Academy of Sciences, 100049 Beijing (China); Liu, Hong [Qinghai Salt Chemical Products Supervision and Inspection Center, 816000 Golmud (China); Jing, Yan, E-mail: 1580707906@qq.com [Key Laboratory of Salt Lake Resources and Chemistry, Qinghai Institute of Salt Lakes, Chinese Academy of Sciences, 810008 Xining (China)

2015-01-15

Highlights: • We proposed a new system for Li recovery from salt lake brine by extraction using an ionic liquid. • Cation exchange was proposed to be the mechanism of extraction followed in ionic liquid. • This ionic liquid system shown considerable extraction ability for lithium and the single extraction efficiency of lithium reached 87.28% under the optimal conditions. - Abstract: Lithium is known as the energy metal and it is a key raw material for preparing lithium isotopes which have important applications in nuclear energy source. In this work, a typical room temperature ionic liquid (RTILs), 1-butyl-3-methyl-imidazolium hexafluorophosphate ([C{sub 4}mim][PF{sub 6}]), was used as an alternative solvent to study liquid/liquid extraction of lithium from salt lake brine. In this system, the ionic liquid, NaClO{sub 4} and tributyl phosphate (TBP) were used as extraction medium, co-extraction reagent and extractant respectively. The effects of solution pH value, phase ratio, ClO{sub 4}{sup −} amount and other factors on lithium extraction efficiency had been investigated. Optimal extraction conditions of this system include the ratio of TBP/IL at 4/1 (v/v), O/A at 2:1, n(ClO{sub 4}{sup −})/n(Li{sup +}) at 2:1, the equilibration time of 10 min and unadjusted pH. Under the optimal conditions, the single extraction efficiency of lithium was 87.28% which was much higher than the conventional extraction system. Total extraction efficiency of 99.12% was obtained by triple-stage countercurrent extraction. Study on the mechanism revealed that the use of ionic liquid increased the extraction yield of lithium through cation exchange in this system. Preliminary results indicated that the use of [C{sub 4}mim][PF{sub 6}] as an alternate solvent to replace traditional organic solvents (VOCs) in liquid/liquid extraction was very promising.

Extraction of lithium ion from alkaline aqueous media by a liquid surfactant membrane

International Nuclear Information System (INIS)

Kinugasa, Takumi; Ono, Yuri; Kawamura, Yuko; Watanabe, Kunio; Takeuchi, Hiroshi.

1995-01-01

Extraction of lithium ion from aqueous alkaline media by a liquid surfactant membrane was performed using a mixture of LIX54 and TOPO as the extractant. Stripping of lithium from the kerosene solution to the acid solution was suppressed with increasing content of polyamine (ECA) surfactant. The extraction rate of lithium by the liquid membrane could be interpreted taking account of an interfacial resistance due to ECA. It was confirmed that swelling of the (W/O) emulsion drops by water permeation through the liquid membrane is evaluated in terms of a change in osmotic pressure gradient between the external and internal aqueous phases during the lithium extraction. In the present operation, the extraction ratio of Li + from the external feed and the uptake into the internal phase reached as high as 95%. (author)

Detection of coolant void in lead-cooled fast reactors

International Nuclear Information System (INIS)

Wolniewicz, Peter; Håkansson, Ane; Jansson, Peter

2015-01-01

Highlights: • We model the ALFRED LFR using different Monte-Carlo codes. • We study the impact on coolant void on the fission cross section in fission chambers. • We develop a methodology to detect coolant void. • We study the impact of detector fissile coating burn-up. • We conclude that the developed methodology may be an attractive complement to LFR monitoring. - Abstract: Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242 Pu and 235 U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor. In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology

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)

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

High-efficiency technology for lithium isotope separation using an ionic-liquid impregnated organic membrane

International Nuclear Information System (INIS)

Hoshino, Tsuyoshi; Terai, Takayuki

2011-01-01

The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 ( 6 Li) in tritium breeding materials. New lithium isotope separation technique using ionic-liquid impregnated organic membranes (Ionic-Liquid-i-OMs) have been developed. Lithium ions are able to move by electrodialysis through certain Ionic-Liquid-i-OMs between the cathode and the anode in lithium solutions. In this report, the effects of protection cover and membrane thickness on the durability of membrane and the efficiency of isotope separation were evaluated. In order to improve the durability of the Ionic-Liquid-i-OM, we developed highly-durable Ionic-Liquid-i-OM. Both surfaces of the Ionic-Liquid-i-OM were covered by a nafion 324 overcoat or a cation exchange membrane (SELEMION TM CMD) to prevent the outflow of the ionic liquid. It was observed that the durability of the Ionic-Liquid-i-OM was improved by a nafion 324 overcoat. On the other hand, the organic membrane selected was 1, 2 or 3 mm highly-porous Teflon film, in order to efficiently impregnate the ionic liquid. The 6 Li isotope separation factor by electrodialysis using highly-porous Teflon film of 3 mm thickness was larger than using that of 1 or 2 mm thickness.

Ionic Liquid-Doped Gel Polymer Electrolyte for Flexible Lithium-Ion Polymer Batteries

Directory of Open Access Journals (Sweden)

Ruisi Zhang

2015-05-01

Full Text Available Application of gel polymer electrolytes (GPE in lithium-ion polymer batteries can address many shortcomings associated with liquid electrolyte lithium-ion batteries. Due to their physical structure, GPEs exhibit lower ion conductivity compared to their liquid counterparts. In this work, we have investigated and report improved ion conductivity in GPEs doped with ionic liquid. Samples containing ionic liquid at a variety of volume percentages (vol % were characterized for their electrochemical and ionic properties. It is concluded that excess ionic liquid can damage internal structure of the batteries and result in unwanted electrochemical reactions; however, samples containing 40–50 vol % ionic liquid exhibit superior ionic properties and lower internal resistance compared to those containing less or more ionic liquids.

Spherical Torus Plasma Interactions with Large-area Liquid Lithium Surfaces in CDX-U

International Nuclear Information System (INIS)

Kaita, R.; Majeski, R.; Boaz, M.; Efthimion, P.; Jones, B.; Hoffman, D.; Kugel, H.; Menard, J.; Munsat, T.; Post-Zwicker, A.; Soukhanovskii, V.; Spaleta, J.; Taylor, G.; Timberlake, J.; Woolley, R.; Zakharov, L.; Finkenthal, M.; Stutman, D.; Antar, G.; Doerner, R.; Luckhardt, S.; Maingi, R.; Maiorano, M.; Smith, S.

2002-01-01

The Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego. Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance

Spherical Torus Plasma Interactions with Large-area Liquid Lithium Surfaces in CDX-U

Energy Technology Data Exchange (ETDEWEB)

R. Kaita; R. Majeski; M. Boaz; P. Efthimion; B. Jones; D. Hoffman; H. Kugel; J. Menard; T. Munsat; A. Post-Zwicker; V. Soukhanovskii; J. Spaleta; G. Taylor; J. Timberlake; R. Woolley; L. Zakharov; M. Finkenthal; D. Stutman; G. Antar; R. Doerner; S. Luckhardt; R. Maingi; M. Maiorano; S. Smith

2002-01-18

The Current Drive Experiment-Upgrade (CDX-U) device at the Princeton Plasma Physics Laboratory (PPPL) is a spherical torus (ST) dedicated to the exploration of liquid lithium as a potential solution to reactor first-wall problems such as heat load and erosion, neutron damage and activation, and tritium inventory and breeding. Initial lithium limiter experiments were conducted with a toroidally-local liquid lithium rail limiter (L3) from the University of California at San Diego. Spectroscopic measurements showed a clear reduction of impurities in plasmas with the L3, compared to discharges with a boron carbide limiter. The evidence for a reduction in recycling was less apparent, however. This may be attributable to the relatively small area in contact with the plasma, and the presence of high-recycling surfaces elsewhere in the vacuum chamber. This conclusion was tested in subsequent experiments with a fully toroidal lithium limiter that was installed above the floor of the vacuum vessel. The new limiter covered over ten times the area of the L3 facing the plasma. Experiments with the toroidal lithium limiter have recently begun. This paper describes the conditioning required to prepare the lithium surface for plasma operations, and effect of the toroidal liquid lithium limiter on discharge performance.

Real-time algorithm for the measurement of liquid metal coolant flow velocity with correlated thermal signals

International Nuclear Information System (INIS)

Moazzeni, Taleb; Jiang, Yingtao; Ma, Jian; Li, Ning

2009-01-01

One flow meter was developed especially for the environment of high irradiation, pressure, and temperature. The transit time of natural random temperature fluctuation in process, for example nuclear reactor, can be obtained based on the cross-correlation method, which has already been shown that it is capable in situations where no other flow meter can be used. Thereby, the flow rate can be derived in pipe flow if the area of cross-section is known. In practice, the evaluation of the integrals over the measurement time in cross-correlation method will lead errors caused by peak detection from flat cross correlation coefficient distribution or additional peaks. One Auto-Adaptive Impulse Response Function estimation is introduced and significantly narrower peak will be obtained. Fiber optic sensors are advantageous for temperature measurements in the reactor pressure vessels. However, the corrosive coolant (as liquid lead/lead alloy or molten salt coolant) is a barrier of the optic sensor in such application. Thermocouple with grounded stainless steel shielding material would have same life time with structure material in reactor, although thermocouple has relatively slow response. The degradation due to corrosion/erosion will not introduce measurement error or necessary calibration, because only the correlation between signals is taken into consideration during measurements. Experiments conducted in a testing hydraulic facility approved the considerable improvement of accuracy by this new algorithm using thermocouple temperature sensors. (author)

Transient heat transfer phenomena of the liquid metal layer cooled by overlying R113 coolant

International Nuclear Information System (INIS)

Cho, J. S.; Seo, K. R.; Jung, C. H.; Park, R. J.; Kim, S. B.

1999-01-01

To understand the fundamental relationship of the natural convection heat transfer in the molten metal pool and the boiling mechanism of the overlying coolant, experiments were performed for the transient heat transfer of the liquid metal pool with overlying R113 coolant with boiling. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. The metal pool is heated from the bottom surface and the coolant is injected onto the molten metal pool. Tests were conducted by changing the bottom surface boundary condition. The bottom heating condition was varied from 8kW to 14kW. As a result the boiling mechanism of the R113 coolant is changed from the nuclear boiling to film boiling. The Nusselt number and the Rayleigh number in the molten metal pool region obtained as functions of time. Analysis was made for the relationship between the heat flux and the temperature difference of the metal layer surface temperature and the boiling coolant bulk temperature

Interactions of liquid lithium with various atmospheres, concretes, and insulating materials; and filtration of lithium aerosols

International Nuclear Information System (INIS)

Jeppson, D.W.

1979-06-01

This report describes the facilities and experiments and presents test results of a program being conducted at the hanford Engineering Development Laboratory (HEDL) in support of the fusion reactor development effort. This experimental program is designed to characterize the interaction of liquid lithium with various atmospheres, concretes, and insulating materials. Lithium-atmosphere reaction tests were conducted in normal humidity air, pure nitrogen, and carbon dioxide. These tests are described and their results, such as maximum temperatures, aerosol generated, and reaction rates measured, are reported. Initial lithium temperatures for these tests ranged between 224 0 C and 843 0 C. A lithium-concrete reaction test, using 10 kg of lithium at 327 0 C, and lithium-insulating materials reaction tests, using a few grams of lithium at 350 0 C and 600 0 C, are also described and results are presented. In addition, a lithium-aerosol filter loading test was conducted to determine the mass loading capacity of a commercial high efficiency particulate air (HEPA) filter. The aerosol was characterized, and the loading-capacity-versus-pressure-buildup across the filter is reported

Research and development of lithium isotope separation using an ionic-liquid impregnated organic membrane

International Nuclear Information System (INIS)

Hoshino, Tsuyoshi

2013-01-01

The tritium needed as a fuel for fusion reactors is produced by the neutron capture reaction of lithium-6 ( 6 Li) in tritium breeding materials. However, natural Li contains only about 7.6 at.% 6 Li. In Japan, new lithium isotope separation technique using ionic-liquid impregnated organic membranes have been developed. The improvement in the durability of the ionic-liquid impregnated organic membrane is one of the main issues for stable, long-term operation of electrodialysis cells while maintaining good performance. Therefore, we developed highly-durable ionic-liquid impregnated organic membrane. Both ends of the ionic-liquid impregnated organic membrane were covered by a nafion 324 overcoat to prevent the outflow of the ionic liquid. The transmission of Lithium aqueous solution after 10 hours under the highly-durable ionic-liquid impregnated organic membrane is almost 13%. So this highly-durable ionic-liquid impregnated organic membrane for long operating of electrodialysis cells has been developed through successful prevention of ion liquid dissolution. (J.P.N.)

Solubility of lithium deuteride in liquid lithium

International Nuclear Information System (INIS)

Veleckis, E.; Yonco, R.M.; Maroni, V.A.

1977-01-01

The solubility of LiD in liquid lithium between the eutectic and monotectic temperatures was measured using a direct sampling method. Solubilities were found to range from 0.0154 mol.% LiD at 199 0 C to 3.32 mol.% LiD at 498 0 C. The data were used in the derivation of an expression for the activity coefficient of LiD as a function of temperature and composition and an equation relating deuteride solubility and temperature, thus defining the liquidus curve. Similar equations were also derived for the Li-LiH system using the existing solubility data. Extrapolation of the liquidus curves yielded the eutectic concentrations (0.040 mol.% LiH and 0.035 mol.% LiD) and the freezing point depressions (0.23 0 C for Li-LiH and 0.20 0 C for Li-LiD) at the eutectic point. The results are compared with the literature data for hydrogen and deuterium. The implications of the relatively high solubility of hydrogen isotopes in lithium just above the melting point are discussed with respect to the cold trapping of tritium in fusion reactor blankets. (Auth.)

An investigation of core liquid level depression in small break loss-of-coolant accidents

International Nuclear Information System (INIS)

Schultz, R.R.; Watkins, J.C.; Motley, F.E.; Stumpf, H.; Chen, Y.S.

1991-08-01

Core liquid level depression can result in partial core dryout and heatup early in a small break loss-of-coolant accident (SBLOCA) transient. Such behavior occurs when steam, trapped in the upper regions of the reactor primary system (between the loop seal and the core inventory), moves coolant out of the core region and uncovers the rod upper elevations. The net result is core liquid level depression. Core liquid level depression and subsequent core heatups are investigated using subscale data from the ROSA-IV Program's 1/48-scale Large Scale Test Facility (LSTF) and the 1/1705-scale Semiscale facility. Both facilities are Westinghouse-type, four-loop, pressurized water reactor simulators. The depression phenomena and factors which influence the minimum core level are described and illustrated using examples from the data. Analyses of the subject experiments, conducted using the TRAC-PF1/MOD1 (Version 12.7) thermal-hydraulic code, are also described and summarized. Finally, the response of a typical Westinghouse four-loop plant (RESAR-3S) was calculated to qualitatively study coal liquid level depression in a full-scale system. 31 refs., 37 figs., 6 tabs

Interactions of solid and liquid lithium with steady state hydrogen and helium plasmas

International Nuclear Information System (INIS)

Hirooka, Y.; Nishikawa, M.; Ohgaki, H.; Ohtsuka, Y.

2005-01-01

A variety of innovative Plasma-Facing Component (PFC) concepts, employing moving solid or liquid surfaces, have recently been proposed in order to resolve technical issues, associated with the applications of currently used PFCs in future steady state fusion devices. As the first step to evaluate the concept using flowing-liquids for PFCs, steady state hydrogen and helium plasma interactions with solid and standing liquid lithium have been investigated in the present work, using the H α and He-I spectroscopy at the ion bombarding energies up to 150eV and at the lithium temperatures between room temperature and 480 deg C. Data indicate that hydrogen recycling over liquid lithium is clearly reduced, relative to that over solid lithium, whereas helium recycling does not show the same trend. From the kinetic analysis of these recycling time constant data, the activation energies for the overall recycling processes have been evaluated to be 0.02±0.01eV, both for hydrogen and helium plasmas. Also, it has been found that the activation energy is nearly independent of ion bombarding energy. (author)

Hydrophobic ionic liquids based on the 1-butyl-3-methylimidazolium cation for lithium/seawater batteries

Science.gov (United States)

Zhang, Yancheng; Urquidi-Macdonald, Mirna

Two hydrophobic ionic liquids (room temperature molten salts) based on 1-butyl-3-methylimidazolium cation (BMI +), BMI +PF 6- and BMI +Tf 2N -, were used in developing a highly efficient lithium anode system for lithium/seawater batteries. The lithium anode system was composed of lithium metal/ionic liquid/Celgard membrane. Both BMI +PF 6-and BMI +Tf 2N - maintained high apparent anodic efficiency (up to 100%) under potentiostatic polarization (at +0.5 V versus open-circuit potential (OCP)) in a 3% NaCl solution. Eventually, traces of water contaminated the ionic liquid and a bilayer film (LiH and LiOH) on the lithium surface was formed, decreasing the rate of lithium anodic reaction and hence the discharge current density. BMI +Tf 2N - prevented traces of water from reaching the lithium metal surface longer than BMI +PF 6- (60 h versus 7 h). However, BMI +PF 6- was better than BMI +Tf 2N - in keeping a constant current density (˜0.2 mA cm -2) before the traces of water contaminated the lithium surface due to the non-reactivity of BMI +PF 6- with the lithium metal that kept the bare lithium surface. During the discharge process, BMI +PF 6- and BMI +Tf 2N - acted as ion transport media of Li +, Cl -, OH - and H 2O, but did not react with them because of the excellent chemical stability, high conductivity, and high hydrophobicity of these two ionic liquids. Both BMI +PF 6- and BMI +Tf 2N - gels were tentative approaches used to delay the traces of water coming in contact with the lithium surface.

Conference on heat mass transfer and properties of liquid metals TF-2002

International Nuclear Information System (INIS)

Efanov, A.D.; Kozlov, F.A.

2003-01-01

Results of the conference TF-2002 devoted to the combined approach to problems of harnessing liquid metals as coolants for NPU are presented. The conference takes place in Obninsk, 29 - 31 October, 2002. Papers of the conference involve items on thermal hydraulics, mass transfer and safety of NPU with liquid metal coolants, structure, physical and chemical properties of liquid metal and liquid metal solutions, decommissioning of units and ecology, application of liquid metals divorced with NPU. Most of the papers of the conference are devoted to the investigation into lead and lead-bismuth coolants [ru

Control of nitrogen concentration in liquid lithium by iron-titanium alloy

International Nuclear Information System (INIS)

Hirakane, Shinji; Yoneoka, Toshiaki; Tanaka, Satoru

2006-01-01

Reducing the nitrogen concentration in liquid lithium is one of the most important steps in creating a liquid lithium blanket system. In this study, in order to verify the nitrogen gettering performance of Fe-Ti alloy, the variation in the nitrogen concentration in liquid lithium, into which Fe-10 at.% Ti or Fe-5 at.% Ti getter was immersed, was examined. The results confirmed a gettering performance of Fe-Ti alloy comparable to that of V-Ti alloy, although the effects were not durable in either the Fe-Ti or the V-Ti alloy. After the immersion test, the existing states of nitrogen absorbed in the gettering material were analyzed by means of XRD, XMA and XPS. TiN and some nitrogen dissolved in α-Fe without forming TiN were observed. It was indicated that nitrogen gettering is prevented not only by the surface nitrides, but also by the internal diffusion barriers originating from the absorbed nitrogen

Chemical sensors for monitoring non-metallic impurities in liquid sodium coolant

International Nuclear Information System (INIS)

Ganesan, Rajesh; Jayaraman, V.; Rajan Babu, S.; Sridharan, R.; Gnanasekaran, T.

2011-01-01

Liquid sodium is the coolant of choice for fast breeder reactors. Liquid sodium is highly compatible with structural steels when the concentration of dissolved non-metallic impurities such as oxygen and carbon are low. However, when their concentrations are above certain threshold limits, enhanced corrosion and mass transfer and carburization of the steels would occur. The threshold concentration levels of oxygen in sodium are determined by thermochemical aspects of various ternary oxides of Na-M-O systems (M alloying elements in steels) which take part in corrosion and mass transfer. Dissolved carbon also influences these threshold levels by establishing relevant carbide equilibria. An event of steam leak into sodium at the steam generator, if undetected at its inception itself, can lead to extensive wastage of the tubes of the steam generator and prolonged shutdown. Air ingress into the argon cover gas and leak of hydrocarbon oil used as cooling fluids of the shafts of the centrifugal pumps of sodium are the sources of oxygen and carbon impurities in sodium. Continuous monitoring of the concentration of dissolved hydrogen, carbon and oxygen in sodium coolant will help identifying their ingress at inception itself. An electrochemical hydrogen sensor based on CaHBr-CaBr 2 hydride ion conducting solid electrolyte has been developed for detecting the steam leak during normal operating conditions of the reactor. A nickel diffuser based sensor system using thermal conductivity detector (TCD) and Pd-doped tin oxide thin film sensor has been developed for use during low power operations of the reactor or during its start up. For monitoring carbon in sodium, an electrochemical sensor with molten Na 2 CO 3 -LiCO 3 as the electrolyte and pure graphite as reference electrode has been developed. Yttria Doped Thoria (YDT) electrolyte based oxygen sensor is under development for monitoring dissolved oxygen levels in sodium. Fabrication, assembly, testing and performance of

Results of neutron irradiation of liquid lithium saturated with deuterium

International Nuclear Information System (INIS)

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

2017-01-01

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

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)

Nucleation and growth of lead oxide particles in liquid lead-bismuth eutectic.

Science.gov (United States)

Gladinez, Kristof; Rosseel, Kris; Lim, Jun; Marino, Alessandro; Heynderickx, Geraldine; Aerts, Alexander

2017-10-18

Liquid lead-bismuth eutectic (LBE) is an important candidate to become the primary coolant of future, generation IV, nuclear fast reactors and Accelerator Driven System (ADS) concepts. One of the main challenges with the use of LBE as a coolant is to avoid its oxidation which results in solid lead oxide (PbO) precipitation. The chemical equilibria governing PbO formation are well understood. However, insufficient kinetic information is currently available for the development of LBE-based nuclear technology. Here, we report the results of experiments in which the nucleation, growth and dissolution of PbO in LBE during temperature cycling are measured by monitoring dissolved oxygen using potentiometric oxygen sensors. The metastable region, above which PbO nucleation can occur, has been determined under conditions relevant for the operation of LBE cooled nuclear systems and was found to be independent of setup geometry and thus thought to be widely applicable. A kinetic model to describe formation and dissolution of PbO particles in LBE is proposed, based on Classical Nucleation Theory (CNT) combined with mass transfer limited growth and dissolution. This model can accurately predict the experimentally observed changes in oxygen concentration due to nucleation, growth and dissolution of PbO, using the effective interfacial energy of a PbO nucleus in LBE as a fitting parameter. The results are invaluable to evaluate the consequences of oxygen ingress in LBE cooled nuclear systems under normal operating and accidental conditions and form the basis for the development of cold trap technology to avoid PbO formation in the primary reactor circuit.

Thermal property of holmium doped lithium lead borate glasses

Science.gov (United States)

Usharani, V. L.; Eraiah, B.

2018-04-01

The new glass system of holmium doped lithium lead borate glasses were prepared by conventional melt quenching technique. The thermal stability of the different compositions of Ho3+ ions doped lithium lead borate glasses were studied by using TG-DTA. The Tg values are ranging from 439 to 444 °C with respect to the holmium concentration. Physical parameters like polaron radius(rp), inter-nuclear distance (ri), field strength (F) and polarizability (αm) of oxide ions were calculated using appropriate formulae.

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)

Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

Energy Technology Data Exchange (ETDEWEB)

Sakai, Takaaki; Enuma, Yasuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwasaki, Takashi [Nuclear Energy System Inc., Tokyo (Japan); Ohyama, Kazuhiro [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

2001-05-01

The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

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

Results of neutron irradiation of liquid lithium saturated with deuterium

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-15

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

Development Plan and R and D Status of China Lead-based Reactor

Energy Technology Data Exchange (ETDEWEB)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS [Institute of Nuclear Energy Safety Technology, Beijing (Switzerland)

2013-07-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

Development Plan and R and D Status of China Lead-based Reactor

International Nuclear Information System (INIS)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS

2013-01-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

Ionic Liquid-Nanoparticle Hybrid Electrolytes and their Application in Secondary Lithium-Metal Batteries

KAUST Repository

Lu, Yingying

2012-07-12

Ionic liquid-tethered nanoparticle hybrid electrolytes comprised of silica nanoparticles densely grafted with imidazolium-based ionic liquid chains are shown to retard lithium dendrite growth in rechargeable batteries with metallic lithium anodes. The electrolytes are demonstrated in full cell studies using both high-energy Li/MoS2 and high-power Li/TiO2 secondary batteries. Copyright © 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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.)

Kinetics of liquid lithium reaction with oxygen-nitrogen mixtures

International Nuclear Information System (INIS)

Gil, T.K.; Kazimi, M.S.

1986-01-01

A series of experiments have been conducted in order to characterize the kinetics of lithium chemical reaction with a mixture of oxygen and nitrogen. Three mixed gas compositions were used; 80% N 2 and 20% O 2 , 90% N 2 and 10% O 2 , and 95% N 2 and 5% O 2 . The reaction rate was obtained as a function of lithium temperature and the oxygen fraction. Liquid lithium temperature varied from 400 to 1100 0 C. By varying the composition, the degree of inhibition of the lithium-nitrogen reaction rate due to the presence of oxygen was observed. The results indicate that the lithium-nitrogen reaction rate depended on both the fraction of oxygen present and lithium temperature. The lithium nitride layer formed from the reaction also had a significant inhibition effect on the lithium-nitrogen reaction rate while the lithium-oxygen reaction rate was not as greatly hindered. LITFIRE, a computer code which simulates temperature and pressure history in a containment building following lithium spills, was modified by including (1) an improved model for the lithium-nitrogen reaction rate and (2) a model for the lithium-CO 2 reaction. LITFIRE was used to simulate HEDL's LC-2 and LA-5 experiments, and the predicted temperatures and pressures were in a reasonable agreement. Furthermore, LITFIRE was applied to a prototypical fusion reactor containment in order to simulate the consequences of a lithium spill accident. The result indicated that if nitrogen was used as containment building gas during the accident, the consequences of the accident would be less severe than those with air. The pressure rise in the building was found to be reduced by 50% and the maximum temperature of the combustion zone was limited to 900 0 C instead of 1200 0 C in the case of air

Measurements of time-dependent liquid-metal magnetohydrodynamic flows in a flat rectangular duct

International Nuclear Information System (INIS)

Buehler, L.; Horanyi, S.

2009-01-01

In the helium-cooled lead lithium (HCLL) blanket, which has been chosen as a reference concept for a liquid-metal breeding blanket to be tested in ITER, the heat is removed by helium cooled plates aligned with the strong toroidal magnetic field that confines the fusion plasma. The liquid breeder lead lithium circulates through gaps of rectangular cross-section between the cooling plates to transport the generated tritium towards external extraction facilities. Under the action of the strong magnetic field, liquid metal flows in conducting rectangular ducts exhibit jet-like velocity profiles in the thin boundary layers near the side walls, which are parallel to the magnetic field like the cooling plates in HCLL blankets. The velocity in these side layers may exceed several times the mean velocity in the duct and it is known that these layers become unstable for sufficiently high Reynolds numbers. The present paper summarizes experimental results for such unstable time-dependent flows in strong magnetic fields, which have been obtained in the MEKKA liquid metal laboratory of the Forschungszentrum Karlsruhe. In particular, spatial and temporal scales of perturbation patterns are identified. The results suggest that the flow between cooling plates in a HCLL blanket is laminar and stable. The observed time-dependent flow behavior appears at larger velocities so that the present results are more relevant for applications in dual coolant concepts where high-velocity jets have been predicted along side walls.

Some safety considerations of liquid lithium as a fusion breeder material

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.

1986-01-01

Test results and conclusions are presented for the reaction of steam with a high temperature lithium pool and for the reaction of high temperature lithium spray with a nitrogen atmosphere. The reactions are characterized and evaluated in regard to the potential for mobilization of radioactive species associated with the liquid breeder under postulated fusion reactor accident conditions. These evaluations include measured lithium temperature responses, atmosphere temperature and pressure responses, gas consumption and generation, aerosol quantities and particle size characterization, and potentially radioactive species releases. Conclusions are made as to the consequences of these safety considerations for the use of lithium as a fusion reactor breeder material

Verification of the hydraulic design of the FMIT liquid lithium target

International Nuclear Information System (INIS)

Miles, R.R.; Annese, C.E.; Ingham, J.G.

1983-01-01

A liquid lithium target is being developed to generate a neutron flux for material testing in a fusion-like environment. The target consists of a thin, high speed, curved wall jet of lithium which is formed by an asymmetric nozzle. A prototype target was designed using potential flow analysis and was tested in water. Measurements of jet thickness and velocity in water and thickness in lithium were compared with isothermal design predictions and were shown to match within 1% for thickness and 5% for jet velocity

Development of aluminide coatings on vanadium-base alloys in liquid lithium

International Nuclear Information System (INIS)

Park, J.H.; Dragel, D.

1993-01-01

Aluminide coatings were produced on vanadium and vanadium-base alloys by exposure of the materials to liquid lithium that contained 3/5 at.% dissolved aluminum in sealed V and V-20 wt.% Ti capsules at temperatures between 775 and 880 degrees C. After each test, the capsules were opened and the samples were examined by optical microscopy and scanning electron microscopy (SEM), and analyzed by electron-energy-dispersive spectroscopy (EDS) and X-ray diffraction. Hardness of the coating layers and bulk alloys was determined by microidentation techniques. The nature of the coatings, i.e., surface coverage, thickness, and composition, varied with exposure time and temperature, solute concentration in lithium, and alloy composition. Solute elements that yielded adherent coatings on various substrates can provide a means of developing in-situ electrical insulator coatings by reaction of the reactive layers with dissolved nitrogen in liquid lithium

CDX-U Operation with a Large Area Liquid Lithium Limiter

International Nuclear Information System (INIS)

R. Majeski; M. Boaz; D. Hoffman; B. Jones; R. Kaita; H. Kugel; T. Munsat; J. Spaleta; V. Soukhanovskii; J. Timberlake; L. Zakharov; G. Antar; R. Doerner; S. Luckhardt; R.W. Conn; M. Finkenthal; D. Stutman; R. Maingi; M. Ulrickson

2002-01-01

The Current Drive experiment-Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has begun experiments with a fully toroidal liquid lithium limiter. CDX-U is a compact [R = 34 cm, a = 22 cm, B(subscript)toroidal = 2 kG, I(subscript)P = 100 kA, T(subscript)e(0) ∼ 100 eV, n(subscript)e(0) ∼ 5 x 10 19 m -3 ] short-pulse (<25 msec) spherical torus with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, can be filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium prior to the experiment. Surface coatings are evident on part of the lithium. Despite the surface coatings, tokamak discharges operated in contact with the lithium-filled tray show evidence of reduced impurities and recycling. The reduction in recycling is largest when the lithium is liquefied by heating to 250 degrees Celsius

Retention/Diffusivity Studies in Free-Surface Flowing Liquid Lithium

International Nuclear Information System (INIS)

R.A. Stubbers; G.H. Miley; M. Nieto; W. Olczak; D.N. Ruzic; A. Hassanein

2004-01-01

FLIRE was designed to measure the hydrogen and helium retention and diffusivity in a flowing stream of liquid lithium, and it has accomplished these goals. Retention coefficients for helium in the flowing liquid stream were 0.1-2% for flow speeds of 44 cm/s and implantation energies between 500 and 2000 eV. The energy dependence of retention is linear for the energy range considered, as expected, and the dependence of retention on flow velocity fits the expected square-root of flow speed dependence. Estimates of the helium diffusion coefficient in the flowing lithium stream were ∼ 4 x 10 -7 cm 2 /s, and are independent of implantation energy. This value is much lower than expected, which could be due to several factors, such as mixing, bubble formation or surface film formation. In the case of hydrogen, long term retention and release mechanisms are of greatest importance, since this relates to tritium inventory in flowing lithium PFCs for fusion applications. The amount of hydride formation was measured for flowing lithium exposed to neutral deuterium gas. Thermal desorption spectroscopy (TDS) measurements indicate that the hydride concentration was between 0.1 and 0.2% over a wide range of pressures (6.5 x 10 -5 to 1 Torr). This result implies that the deuterium absorption rate is limited by the surface dissociation rate, since deuterium (hydrogen/tritium) is absorbed in its atomic form, not its molecular form

Retention/Diffusivity Studies in Free-Surface Flowing Liquid Lithium

Energy Technology Data Exchange (ETDEWEB)

R.A. Stubbers; G.H. Miley; M. Nieto; W. Olczak; D.N. Ruzic; A. Hassanein

2004-12-14

FLIRE was designed to measure the hydrogen and helium retention and diffusivity in a flowing stream of liquid lithium, and it has accomplished these goals. Retention coefficients for helium in the flowing liquid stream were 0.1-2% for flow speeds of 44 cm/s and implantation energies between 500 and 2000 eV. The energy dependence of retention is linear for the energy range considered, as expected, and the dependence of retention on flow velocity fits the expected square-root of flow speed dependence. Estimates of the helium diffusion coefficient in the flowing lithium stream were {approx} 4 x 10{sup -7} cm{sup 2}/s, and are independent of implantation energy. This value is much lower than expected, which could be due to several factors, such as mixing, bubble formation or surface film formation. In the case of hydrogen, long term retention and release mechanisms are of greatest importance, since this relates to tritium inventory in flowing lithium PFCs for fusion applications. The amount of hydride formation was measured for flowing lithium exposed to neutral deuterium gas. Thermal desorption spectroscopy (TDS) measurements indicate that the hydride concentration was between 0.1 and 0.2% over a wide range of pressures (6.5 x 10{sup -5} to 1 Torr). This result implies that the deuterium absorption rate is limited by the surface dissociation rate, since deuterium (hydrogen/tritium) is absorbed in its atomic form, not its molecular form.

Coolant Chemistry Control: Oxygen Mass Transport in Lead Bismuth Eutectic

International Nuclear Information System (INIS)

Weisenburger, A.; Mueller, G.; Bruzzese, C.; Glass, A.

2015-01-01

In lead-bismuth cooled transmutation systems, oxygen, dissolved in the coolant at defined quantities, is required for stable long-term operation by assuring the formation of protective oxide scales on structural steel surfaces. Extracted oxygen must be permanently delivered to the system and distributed in the entire core. Therefore, coolant chemistry control involves detailed knowledge on oxygen mass transport. Beside the different flow regimes a core might have stagnant areas at which oxygen delivery can only be realised by diffusion. The difference between oxygen transport in flow paths and in stagnant zones is one of the targets of such experiments. To investigate oxygen mass transport in flowing and stagnant conditions, a dedicated facility was designed based on computational fluid dynamics (CFD). CFD also was applied to define the position of oxygen sensors and ultrasonic Doppler velocimetry transducers for flow measurements. This contribution will present the test facility, design relevant CFD calculations and results of first tests performed. (authors)

A study on the corrosion characteristics of lead-bismuth liquid metal

International Nuclear Information System (INIS)

Tak, Nam Il; Park, Won S.; Han, Seok Jung; Jeong, Won Seok

1999-03-01

Pb-Bi eutectic has been adopted as a coolant and spallation target material of HYPER (Hybrid Power Extraction Reactor), an accelerator driven subcritical transmutation system. The contents and scope of the present study are to implement systematic survey and analyses of available results on the corrosion characteristics of Pb-Bi liquid metal which are considered to be the most important among Pb-Bi coolant technologies and to provide fundamental bases for future research efforts. Major parameters affecting the corrosion of structural materials in liquid metals are temperature, flow velocity, contents of impurities in coolant, compositions of structural materials, and so forth. It was already known that for traditional commercial austenitic steels of 18Cr-10Ni-Ti type and 12%Cr ferritic steels, the operating temperatures of Pb-Bi coolant cannot be raised above 400 dg C and 450 dg C, respectively. However, extensive researches have been performed to protect structural materials under higher operating temperature such as the development of various kinds of coating methods for steels and the investigations of coolant inhibition by different chemical elements. The available experimental results show that the effective methods to improve the performance of structural materials in Pb-Bi coolant are the development of suitable steel alloys, the creation of oxide type coatings, and the control of oxygen inhibition. According to the recently presented research results of URRS, utilization of these methods makes it possible to raise the operating temperature limit to 620-650 dg C. It provides the possibility of usage of Pb-Bi coolant for the transmutation system, HYPER some day. (Author). 27 refs., 6 tabs., 15 figs

A study on the corrosion characteristics of lead-bismuth liquid metal

Energy Technology Data Exchange (ETDEWEB)

Tak, Nam Il; Park, Won S.; Han, Seok Jung; Jeong, Won Seok

1999-03-01

Pb-Bi eutectic has been adopted as a coolant and spallation target material of HYPER (Hybrid Power Extraction Reactor), an accelerator driven subcritical transmutation system. The contents and scope of the present study are to implement systematic survey and analyses of available results on the corrosion characteristics of Pb-Bi liquid metal which are considered to be the most important among Pb-Bi coolant technologies and to provide fundamental bases for future research efforts. Major parameters affecting the corrosion of structural materials in liquid metals are temperature, flow velocity, contents of impurities in coolant, compositions of structural materials, and so forth. It was already known that for traditional commercial austenitic steels of 18Cr-10Ni-Ti type and 12%Cr ferritic steels, the operating temperatures of Pb-Bi coolant cannot be raised above 400 dg C and 450 dg C, respectively. However, extensive researches have been performed to protect structural materials under higher operating temperature such as the development of various kinds of coating methods for steels and the investigations of coolant inhibition by different chemical elements. The available experimental results show that the effective methods to improve the performance of structural materials in Pb-Bi coolant are the development of suitable steel alloys, the creation of oxide type coatings, and the control of oxygen inhibition. According to the recently presented research results of URRS, utilization of these methods makes it possible to raise the operating temperature limit to 620-650 dg C. It provides the possibility of usage of Pb-Bi coolant for the transmutation system, HYPER some day. (Author). 27 refs., 6 tabs., 15 figs.

Study of the corrosion behaviors of 304 austenite stainless steel specimens exposed to static liquid lithium at 600 K

Energy Technology Data Exchange (ETDEWEB)

Meng, Xiancai [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zuo, Guizhong; Ren, Jun; Xu, Wei; Sun, Zhen; Huang, Ming [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Wangyu [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China); Hu, Jiansheng, E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Deng, Huiqiu, E-mail: hqdeng@hnu.edu.cn [Department of Applied Physics, School of Physics and Electronics, Hunan University, Changsha 410082 (China)

2016-11-15

Investigation of corrosion behavior of stainless steel served as one kind of structure materials exposed to liquid lithium (Li) is one of the keys to apply liquid Li as potential plasma facing materials (PFM) or blanket coolant in the fusion device. Corrosion experiments of 304 austenite stainless steel (304 SS) were carried out in static liquid Li at 600 K and up to1584 h at high vacuum with pressure less than 4 × 10{sup −4} Pa. After exposure to liquid Li, it was found that the weight of 304 SS slightly decreased with weight loss rate of 5.7 × 10{sup −4} g/m{sup 2}/h and surface hardness increased by about 50 HV. Lots of spinel-like grains and holes were observed on the surface of specimens measured by SEM. By further EDS, XRD and metallographic analyzing, it was confirmed that the main compositions of spinel-like grains were M{sub 23}C{sub 6} carbides, and 304 SS produced a non-uniform corrosion behavior by preferential grain boundary attack, possibly due to the easy formation of M{sub 23}C{sub 6} carbides and/or formation of Li compound at grain boundaries.

Thermal Aging of Anions in Ionic Liquids containing Lithium Salts by IC/ESI-MS

International Nuclear Information System (INIS)

Pyschik, Marcelina; Kraft, Vadim; Passerini, Stefano; Winter, Martin; Nowak, Sascha

2014-01-01

Highlights: • Thermal aging investigation of TFSI- and FSI- based ionic liquids and their mixtures with Li salts. • PYR 13 FSI shows thermal decomposition when mixed with LiPF 6 and LiClO 4 . • PYR 13 TFSI does not show any decomposition products with the electrolyte salts. • LiPF 6 dissolved in ionic liquids suffers of thermal aging as in conventional Li-ion battery electrolytes. - Abstract: The stability of 1-methyl-1-propylpyrrolidinium bis(trifluoromethanesulfonyl)imide (PYR 13 TFSI) and 1-methyl-1-propylpyrrolidinium bis(fluorosulfonyl)imide (PYR 13 FSI) ionic liquids at elevated temperatures (60 °C) is investigated by ion chromatography. Additionally, the influence of the electrolyte salts, lithium hexafluorophosphate (LiPF 6 ), lithium bis(trifluoromethanesulfonyl)imide (LiTFSI) and lithium perchlorate (LiClO 4 ), on the decomposition of both the ionic liquids was analysed over a long term stability study. It has been found out that TFSI has a much higher thermal stability than FSI. The addition of LiTFSI did not show any effect on the aging of both ionic liquid anions. However, PYR 13 FSI degraded when mixed with the electrolyte salts LiPF 6 and LiClO 4 , while PYR 13 TFSI did not. Finally, LiPF 6 forms the same hydrolysis products in the investigated ionic liquids as in the commonly used electrolytes based on organic solvents in lithium-ion batteries

The solubility of carbon in low-nitrogen liquid lithium

International Nuclear Information System (INIS)

Yonco, R.M.; Homa, M.I.

1986-01-01

The solubility of carbon in liquid lithium containing 0 C and compared with the solubility in lithium containing proportional 2600 wppm nitrogen in that same temperature range. A direct sampling method was employed in which filtered samples of the saturated solution were taken at randomly selected temperatures. The entire sample was analyzed for carbon by the acetylene evolution method. The analytical method was examined critically and it was found that (1) all of the carbon in solution, including carbon introduced as lithium cyanamide is detected and (2) ethylene and ethane must also be measured and included with the acetylene to get complete recovery of the carbon content of the sample. The solubility of carbon in low-nitrogen lithium can be expressed by the equations ln S=6.731-8617T -1 and log Ssup(*)=7.459-3740T -1 , where S is the mole percent Li 2 C 2 and Ssup(*) is in weight parts per million carbon. The presence of proportional 2600 wppm nitrogen does not affect the solubility of carbon in lithium at temperatures above proportional 350 0 C, but at lower temperatures it increased the solubility by as much as an order of magnitude compared to the solubility in low-nitrogen lithium. (orig.)

High-power liquid-lithium target prototype for accelerator-based boron neutron capture therapy.

Science.gov (United States)

Halfon, S; Paul, M; Arenshtam, A; Berkovits, D; Bisyakoev, M; Eliyahu, I; Feinberg, G; Hazenshprung, N; Kijel, D; Nagler, A; Silverman, I

2011-12-01

A prototype of a compact Liquid-Lithium Target (LiLiT), which will possibly constitute an accelerator-based intense neutron source for Boron Neutron Capture Therapy (BNCT) in hospitals, was built. The LiLiT setup is presently being commissioned at Soreq Nuclear Research Center (SNRC). The liquid-lithium target will produce neutrons through the (7)Li(p,n)(7)Be reaction and it will overcome the major problem of removing the thermal power generated using a high-intensity proton beam (>10 kW), necessary for sufficient neutron flux. In off-line circulation tests, the liquid-lithium loop generated a stable lithium jet at high velocity, on a concave supporting wall; the concept will first be tested using a high-power electron beam impinging on the lithium jet. High intensity proton beam irradiation (1.91-2.5 MeV, 2-4 mA) will take place at Soreq Applied Research Accelerator Facility (SARAF) superconducting linear accelerator currently in construction at SNRC. Radiological risks due to the (7)Be produced in the reaction were studied and will be handled through a proper design, including a cold trap and appropriate shielding. A moderator/reflector assembly is planned according to a Monte Carlo simulation, to create a neutron spectrum and intensity maximally effective to the treatment and to reduce prompt gamma radiation dose risks. Copyright © 2011 Elsevier Ltd. All rights reserved.

Response of NSTX liquid lithium divertor to high heat loads

Energy Technology Data Exchange (ETDEWEB)

Abrams, T., E-mail: tabrams@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaworski, M.A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Kallman, J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Foley, E.L. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); Gray, T.K. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kugel, H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Levinton, F. [Nova Photonics, Inc., Princeton, NJ 08543 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

2013-07-15

Samples of the NSTX Liquid Lithium Divertor (LLD) with and without an evaporative Li coating were directly exposed to a neutral beam ex-situ at a power of ∼1.5 MW/m{sup 2} for 1–3 s. Measurements of front face and bulk sample temperature were obtained. Predictions of temperature evolution were derived from a 1D heat flux model. No macroscopic damage occurred when the “bare” sample was exposed to the beam but microscopic changes to the surface were observed. The Li-coated sample developed a lithium hydroxide (LiOH) coating, which did not change even when the front face temperature exceeded the pure Li melting point. These results are consistent with the lack of damage to the LLD surface and imply that heating alone may not expose pure liquid Li if the melting point of surface impurities is not exceeded. This suggests that flow and heat are needed for future PFCs requiring a liquid Li surface.

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

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

Science.gov (United States)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

Environmental considerations for alternative fusion reactor blankets

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Young, J.R.

1975-01-01

Comparisons of alternative fusion reactor blanket/coolant systems suggest that environmental considerations will enter strongly into selection of design and materials. Liquid blankets and coolants tend to maximize transport of radioactive corrosion products. Liquid lithium interacts strongly with tritium, minimizing permeation and escape of gaseous tritium in accidents. However, liquid lithium coolants tend to create large tritium inventories and have a large fire potential compared to flibe and solid blankets. Helium coolants minimize radiation transport, but do not have ability to bind the tritium in case of accidental releases. (auth)

Advantages and Challenges of Radiative Liquid Lithium Divertor

Science.gov (United States)

Ono, Masayuki

2017-10-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

Development of liquid-lithium film jet-flow for the target of (7)Li(p,n)(7)Be reactions for BNCT.

Science.gov (United States)

Kobayashi, Tooru; Miura, Kuniaki; Hayashizaki, Noriyosu; Aritomi, Masanori

2014-06-01

A feasibility study on liquid lithium target in the form of a flowing film was performed to evaluate its potential use as a neutron generation target of (7)Li(p,n)(7)Be reaction in BNCT. The target is a windowless-type flowing film on a concave wall. Its configuration was adapted for a proton beam which is 30mm in diameter and with energy and current of up to 3MeV and 20mA, respectively. The flowing film of liquid lithium was 0.6mm in thickness, 50mm in width and 50mm in length. The shapes of the nozzle and concave back wall, which create a stable flowing film jet, were decided based on water experiments. A lithium hydrodynamic experiment was performed to observe the stability of liquid lithium flow behavior. The flowing film of liquid lithium was found to be feasible at temperatures below the liquid lithium boiling saturation of 342°C at the surface pressure of 1×10(-3)Pa. Using a proto-type liquid lithium-circulating loop for BNCT, the stability of the film flow was confirmed for velocities up to 30m/s at 220°C and 250°C in vacuum at a pressure lower than 10(-3) Pa. It is expected that for practical use, a flowing liquid lithium target of a windowless type can solve the problem of radiation damage and target cooling. Copyright © 2013 Elsevier Ltd. All rights reserved.

Ferrous alloy metallurgy, liquid lithium corrosion and welding. Final report, April 1, 1973-March 31, 1984

International Nuclear Information System (INIS)

Olson, D.L.; Matlock, D.K.

1984-01-01

This research program consists of two parts: an evaluation of the corrosion behavior of ferrous alloys in liquid lithium, and a study of microstructure development and properties of dissimilar metal weldments. A ten-year overview of the research accomplishments made is presented. The effects of liquid lithium on both uniform corrosion and grain boundary penetration in ferrous alloys has been investigated as a function of time, temperature, base metal alloy content, and liquid lithium nitrogen content. Kinetic equations for the various corrosion processes have been developed and analyzed with respect to models for corrosion and corrosion product development. The effects of liquid lithium on mechanical properties, particularly fatigue, have been studied. Results have shown that in both austenitic stainless steels and ferritic steels, liquid lithium significantly reduces the mechanical integrity of all materials by inducing liquid metal embrittlement. A model for liquid metal embrittlement induced damage during fatigue was developed and shown to correlate with the experimental results. Microstructural development in austenitic weld metal, with particular emphasis on new grades with reduced chromium contents, has been investigated. The microstructures have been correlated with alloy content and the basics of a thermodynamic model for predicting weld metal microstructure has been developed. The high temperature mechanical behavior of dissimilar metal weldments (austenitic stainless steel to ferritic steel) has been investigated with the impression-creep test technique. Observed microstructural changes with position across the weldment are shown to correlate directly with creep behavior. A model based on deformation of composite materials was developed

Compatibility of yttria (Y{sub 2}O{sub 3}) with liquid lithium

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

Compatibility of Y{sub 2}O{sub 3} sintered specimens with liquid lithium was tested at 773K. No configuration change was observed with a slight increase of thickness for 1419 hr. Lithium-yttrium complex oxide (LiYO{sub 2}) was formed on the surface, and the inner part changed to gray or black nonstoichiometric Y{sub 2}O{sub 3-X} with lower electrical resistibility. It is concluded that Y{sub 2}O{sub 3} has a possibility as a ceramic coating material for liquid blankets if it can be made into a dense coating on the surface of piping materials. (author)

Formation of electrically insulating coatings on aluminided vanadium-base alloys in liquid lithium

International Nuclear Information System (INIS)

Park, J.H.; Dragel, G.

1993-01-01

Aluminide coatings were produced on vanadium and vanadium-base alloys by exposure of the materials to liquid lithium that contained 3-5 at.% dissolved aluminum in sealed capsules at temperatures between 775 and 880 degrees C. Reaction of the aluminide layer with dissolved nitrogen in liquid lithium provides a means of developing an in-situ electrical insulator coating on the surface of the alloys. The electrical resistivity of A1N coatings on aluminided V and V-20 wt.% Ti was determined in-situ

Hydrodynamics of heavy liquid metal coolant processes and filtering apparatus

International Nuclear Information System (INIS)

Albert K Papovyants; Yuri I Orlov; Pyotr N Martynov; Yuri D Boltoev

2005-01-01

Full text of publication follows: To optimize the design of filters for cleaning heavy liquid metal coolant (HLMC) from suspended impurities and choose appropriate filter material, the contribution is considered of different mechanisms of delivery and retention of these impurities from the coolant flow, which is governed by its specificity as a thermodynamically instable disperse system to a large extent. It is shown that the buildup of deposits in the filter is favored by the hydrodynamic regime with minimum filtration rates being due to the predominance in the suspension of the fine-dispersed solid phase (oxides Fe 3 O 4 , Cr 2 O 3 and so on). With concentrating the last mentioned phase in filter material pores or stagnant zones, coagulation structuration is possible, which is accompanied by sharp local increase in the viscosity and strength of the solid phase medium being built from liquid metal, i.e. slag sedimentary deposits. In rather extended pores, disintegration of such structures is possible, which is accompanied by sedimentation of large particles produced due to sticking together at coagulation. The analytical solution of the problem of particle sedimentation due to diffusion indicated that in the case under consideration, this mechanism takes place for particles less than ∼ 0,05 μm in size, which is specified by the fact that the time of their delivery to the filter material surface is longer than that of the coolant being in the filter. The London-Van-der-Waals molecular forces play a crucial role in the stage of retention of a separate particle. The constant of the molecular interaction between a spherical particle and the flat surface has been estimated for the chosen value of the gap between the contacting bodies, being dependent on the wetting angle. The sufficient condition for d p -diameter particle capture by the adhesion force field (with a gap of H ≅ 30 nm) is that it be brought by the appropriate forces at a distance from the wall equal

Fuel-coolant interaction in a shock tube with initially-established film boiling

International Nuclear Information System (INIS)

Sharon, A.; Bankoff, S.G.

1979-01-01

A new mode of thermal interaction has been employed, in which liquid metal is melted in a crucible within a shock tube; the coolant level is raised to overflow the crucible and establish subcooled film boiling with known bulk metal temperature; and a pressure shock is then initiated. With water and lead-tin alloy an initial splash of metal may be obtained after the vapor film has collapsed, due primarily to thermal interaction, followed by a successive cycle of bubble growth and collapse. To obtain large interactions, the interfacial contact temperature must exceed the spontaneous nucleation temperature of the coolant. Other cutoff behavior is observed with respect to the initial system pressure and temperatures and with the shock pressure and rise time. Experiments with butanol and lead-tin alloy show only relatively mild interactions. Qualitative explanations are proposed for the different behaviors of the two liquids

Deuterium trapping in liquid lithium irradiated by deuterium plasma

International Nuclear Information System (INIS)

Pisarev, A.; Moshkunov, K.; Vizgalov, I.; Gasparyan, Yu.

2013-01-01

Liquid lithium was irradiated by deuterium plasma to a low fluence of 10 22 –10 23 D/m 2 , cooled down to room temperature, and then slowly heated. The temperature and release rate were measured during heating. Two plateaus on the temperature–time dependence were observed at 180 °C and 660 °C. The first one corresponds to melting of Li and the second one – either to melting or to decomposition of solid LiD. Features of deuterium release in TDS were interpreted in terms of decomposition of lithium deuterides formed during plasma irradiation

Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies - 2015 Edition

International Nuclear Information System (INIS)

Fazio, Concetta; Sobolev, V.P.; Aerts, A.; Gavrilov, S.; Lambrinou, K.; Schuurmans, P.; Gessi, A.; Agostini, P.; Ciampichetti, A.; Martinelli, L.; Gosse, S.; Balbaud-Celerier, F.; Courouau, J.L.; Terlain, A.; Li, N.; Glasbrenner, H.; Neuhausen, J.; Heinitz, S.; Zanini, L.; Dai, Y.; Jolkkonen, M.; Kurata, Y.; Obara, T.; Thiolliere, N.; Martin-Munoz, F.J.; Heinzel, A.; Weisenburger, A.; Mueller, G.; Schumacher, G.; Jianu, A.; Pacio, J.; Marocco, L.; Stieglitz, R.; Wetzel, T.; Daubner, M.; Litfin, K.; Vogt, J.B.; Proriol-Serre, I.; Gorse, D.; Eckert, S.; Stefani, F.; Buchenau, D.; Wondrak, T.; Hwang, I.S.

2015-01-01

Heavy liquid metals such as lead or lead-bismuth have been proposed and investigated as coolants for fast reactors since the 1950's. More recently, there has been renewed interest worldwide in the use of these materials to support the development of systems for the transmutation of radioactive waste. Heavy liquid metals are also under evaluation as a reactor core coolant and accelerator-driven system neutron spallation source. Several national and international R and D programmes are ongoing for the development of liquid lead-alloy technology and the design of liquid lead-alloy-cooled reactor systems. In 2007, a first edition of the handbook was published to provide deeper insight into the properties and experimental results in relation to lead and lead-bismuth eutectic technology and to establish a common database. This handbook remains a reference in the field and is a valuable tool for designers and researchers with an interest in heavy liquid metals. The 2015 edition includes updated data resulting from various national and international R and D programmes and contains new experimental data to help understand some important phenomena such as liquid metal embrittlement and turbulent heat transfer in a fuel bundle. The handbook provides an overview of liquid lead and lead-bismuth eutectic properties, materials compatibility and testing issues, key aspects of thermal-hydraulics and existing facilities, as well as perspectives for future R and D. (authors)

Cycling performance of lithium polymer cells assembled by in situ polymerization of a non-flammable ionic liquid monomer

International Nuclear Information System (INIS)

Lee, Yoon-Sung; Kim, Dong-Won

2013-01-01

Highlights: • Gel polymer electrolytes were synthesized by in situ polymerization of ionic liquid in the lithium polymer cells. • Flammability of the electrolyte was significantly reduced by polymerizing electrolyte containing a non-flammable ionic liquid monomer. • The cells assembled with polymeric ionic liquid-based electrolytes exhibited reversible cycling behavior with good capacity retention. -- Abstract: Lithium polymer cells composed of a lithium negative electrode and a LiCoO 2 positive electrode were assembled with a gel polymer electrolyte obtained by in situ polymerization of an electrolyte solution containing an ionic liquid monomer with vinyl groups. The polymerization of the electrolyte solution containing the non-flammable ionic liquid monomer resulted in a significant reduction of the flammability of the gel polymer electrolytes. The lithium polymer cell assembled with the stable gel polymer electrolyte delivered a discharge capacity of 134.3 mAh g −1 at ambient temperature and exhibited good capacity retention

Parameters promoting liquid metal embrittlement of the T91 steel in lead-bismuth eutectic alloy

International Nuclear Information System (INIS)

Proriol Serre, I.; Ye, C.; Vogt, J.B.

2015-01-01

The use of liquid lead-bismuth eutectic (LBE) as a spallation target and a coolant in accelerator-driven systems raises the question of the reliability of structural materials, such as T91 martensitic steel in terms of liquid metal assisted damage and corrosion. In this study, the mechanical behaviour of the T91 martensitic steel was examined in liquid lead-bismuth eutectic (LBE) and in inert atmosphere. Several conditions showed the most sensitive embrittlement factor. The Small Punch Test technique was employed using smooth specimens. In this standard heat treatment, T91 appeared in general as a ductile material, and became brittle in the considered conditions if the test was performed in LBE. It turns out that the loading rate appeared as a critical parameter for the occurrence of liquid metal embrittlement (LME) of the T91 steel in LBE. Loading the T91 very slowly instead of rapidly in oxygen saturated LBE resulted in brittle fracture. Furthermore, low-oxygen content in LBE and an increase in temperature promote LME. (authors)

Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

Energy Technology Data Exchange (ETDEWEB)

Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

2016-11-01

Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

The 1994 loss of coolant incident at Pickering NGS

Energy Technology Data Exchange (ETDEWEB)

Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

A Technique for Dynamic Corrosion Testing in Liquid Lead Alloys

Energy Technology Data Exchange (ETDEWEB)

Loewen, Eric Paul; Davis, Cliff Bybee; Mac Donald, Philip Elsworth

2001-04-01

An experimental apparatus for the investigation of the flow-assisted dissolution and precipitation (corrosion) of potential fuel cladding and structural materials to be used in liquid lead alloy cooled reactors has been designed. This experimental project is part of a larger research effort between Idaho National Engineering and Environmental Laboratory (INEEL) and Massachusetts Institute of Technology to investigate the suitability of lead, lead-bismuth, and other lead alloys for cooling fast reactors designed to produce low-cost electricity as well as for actinide burning. The INEEL forced convection corrosion cell consists of a small heated vessel with a shroud and gas flow system. The gas flow rates, heat input, and shroud and vessel dimensions have been adjusted so that a controlled coolant flow rate, temperature, and oxygen potential are created within the downcomer located between the shroud and vessel wall. The ATHENA computer code was used to design the experimental apparatus and estimate the fluid conditions. The corrosion cell will test steel that is commercially available in the U. S. to temperatures above 650oC.

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

Effect of Energetic Plasma Flux on Flowing Liquid Lithium Surfaces

Science.gov (United States)

Kalathiparambil, Kishor; Jung, Soonwook; Christenson, Michael; Fiflis, Peter; Xu, Wenyu; Szott, Mathew; Ruzic, David

2014-10-01

An operational liquid lithium system with steady state flow driven by thermo-electric magneto-hydrodynamic force and capable of constantly refreshing the plasma exposed surface have been demonstrated at U of I. To evaluate the system performance in reactor relevant conditions, specifically to understand the effect of disruptive plasma events on the performance of the liquid metal PFCs, the setup was integrated to a pulsed plasma generator. A coaxial plasma generator drives the plasma towards a theta pinch which preferentially heats the ions, simulating ELM like flux, and the plasma is further guided towards the target chamber which houses the flowing lithium system. The effect of the incident flux is examined using diagnostic tools including triple Langmuir probe, calorimeter, rogowski coils, Ion energy analyzers, and fast frame spectral image acquisition with specific optical filters. The plasma have been well characterized and a density of ~1021 m-3, with electron temperature ~10 - 20 eV is measured, and final plasma velocities of 34 - 74 kms-1 have been observed. Calorimetric measurements using planar molybdenum targets indicate a maximum plasma energy (with 6 kV plasma gun and 20 kV theta pinch) of 0.08 MJm-2 with plasma divergence effects resulting in marginal reduction of 40 +/- 23 J in plasma energy. Further results from the other diagnostic tools, using the flowing lithium targets and the planar targets coated with lithium will be presented. DOE DE-SC0008587.

Fabrication of lithium/C-103 alloy heat pipes for sharp leading edge cooling

Science.gov (United States)

Ai, Bangcheng; Chen, Siyuan; Yu, Jijun; Lu, Qin; Han, Hantao; Hu, Longfei

2018-05-01

In this study, lithium/C-103 alloys heat pipes are proposed for sharp leading edge cooling. Three models of lithium/C-103 alloy heat pipes were fabricated. And their startup properties were tested by radiant heat tests and aerothermal tests. It is found that the startup temperature of lithium heat pipe was about 860 °C. At 1000 °C radiant heat tests, the operating temperature of lithium/C-103 alloy heat pipe is lower than 860 °C. Thus, startup failure occurs. At 1100 °C radiant heat tests and aerothermal tests, the operating temperature of lithium/C-103 alloy heat pipe is higher than 860 °C, and the heat pipe starts up successfully. The startup of lithium/C-103 alloy heat pipe decreases the leading edge temperature effectively, which endows itself good ablation resistance. After radiant heat tests and aerothermal tests, all the heat pipe models are severely oxidized because of the C-103 poor oxidation resistance. Therefore, protective coatings are required for further applications of lithium/C-103 alloy heat pipes.

Solutions of group IV elements in liquid lithium

International Nuclear Information System (INIS)

Dadd, A.T.; Hubberstey, P.; Roberts, P.G.

1982-01-01

The solubilities of tin (0.00 = 22 Sn 5 . A simple thermochemical cycle is used to demonstrate that, whereas carbon dissolves endothermically in both liquid lithium and liquid sodium, the heavier Group IV elements dissolve exothermically. A similar cycle is used to derive solvation enthalpies (for the neutral gaseous species) for all Group IV elements in the two solvents. The trend in solvation enthalpy: C > Si > Ge > Sn > Pb is indicative of a diminishing affinity of solvent for solute and is attributed to the increasing metallic character of the solute as the Group is descended. (author)

Homogeneous lithium electrodeposition with pyrrolidinium-based ionic liquid electrolytes.

Science.gov (United States)

Grande, Lorenzo; von Zamory, Jan; Koch, Stephan L; Kalhoff, Julian; Paillard, Elie; Passerini, Stefano

2015-03-18

In this study, we report on the electroplating and stripping of lithium in two ionic liquid (IL) based electrolytes, namely N-butyl-N-methylpyrrolidinium bis(fluorosulfonyl) imide (Pyr14FSI) and N-butyl-N-methylpyrrolidinium bis(trifluoromethanesulfonyl)imide (Pyr14TFSI), and mixtures thereof, both on nickel and lithium electrodes. An improved method to evaluate the Li cycling efficiency confirmed that homogeneous electroplating (and stripping) of Li is possible with TFSI-based ILs. Moreover, the presence of native surface features on lithium, directly observable via scanning electron microscope imaging, was used to demonstrate the enhanced electrolyte interphase (SEI)-forming ability, that is, fast cathodic reactivity of this class of electrolytes and the suppressed dendrite growth. Finally, the induced inhomogeneous deposition enabled us to witness the SEI cracking and revealed previously unreported bundled Li fibers below the pre-existing SEI and nonrod-shaped protuberances resulting from Li extrusion.

Assessment of fiber optic sensors for aging monitoring of industrial liquid coolants

Science.gov (United States)

Riziotis, Christos; El Sachat, Alexandros; Markos, Christos; Velanas, Pantelis; Meristoudi, Anastasia; Papadopoulos, Aggelos

2015-03-01

Lately the demand for in situ and real time monitoring of industrial assets and processes has been dramatically increased. Although numerous sensing techniques have been proposed, only a small fraction can operate efficiently under harsh industrial environments. In this work the operational properties of a proposed photonic based chemical sensing scheme, capable to monitor the ageing process and the quality characteristics of coolants and lubricants in industrial heavy machinery for metal finishing processes is presented. The full spectroscopic characterization of different coolant liquids revealed that the ageing process is connected closely to the acidity/ pH value of coolants, despite the fact that the ageing process is quite complicated, affected by a number of environmental parameters such as the temperature, humidity and development of hazardous biological content as for example fungi. Efficient and low cost optical fiber sensors based on pH sensitive thin overlayers, are proposed and employed for the ageing monitoring. Active sol-gel based materials produced with various pH indicators like cresol red, bromophenol blue and chorophenol red in tetraethylorthosilicate (TEOS), were used for the production of those thin film sensitive layers deposited on polymer's and silica's large core and highly multimoded optical fibers. The optical characteristics, sensing performance and environmental robustness of those optical sensors are presented, extracting useful conclusions towards their use in industrial applications.

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)

Thermodynamic analysis of chromium solubility data in liquid lithium containing nitrogen: Comparison between experimental data and computer simulation

International Nuclear Information System (INIS)

Krasin, Valery P.; Soyustova, Svetlana I.

2015-01-01

The mathematical formalism for description of solute interactions in dilute solution of chromium and nitrogen in liquid lithium have been applied for calculating of the temperature dependence of the solubility of chromium in liquid lithium with the various nitrogen contents. It is shown that the derived equations are useful to provide understanding of a relationship between thermodynamic properties and local ordering in the Li–Cr–N melt. Comparison between theory and data reported in the literature for solubility of chromium in nitrogen-contaminated liquid lithium, was allowed to explain the reasons of the deviation of the experimental semi-logarithmic plot of chromium content in liquid lithium as a function of the reciprocal temperature from a straight line. - Highlights: • The activity coefficient of chromium in ternary melt can be obtained by means of integrating the Gibbs–Duhem equation. • In lithium with the high nitrogen content, the dependence of a logarithm of chromium solubility as a function of the reciprocal temperature has essentially nonlinear character. • At temperatures below a certain threshold, the process of dissolution of chromium in lithium will be controlled by the equilibrium concentration of nitrogen required for the formation of ternary nitride Li_9CrN_5at a given temperature.

Lithium-modulated conduction band edge shifts and charge-transfer dynamics in dye-sensitized solar cells based on a dicyanamide ionic liquid.

Science.gov (United States)

Bai, Yu; Zhang, Jing; Wang, Yinghui; Zhang, Min; Wang, Peng

2011-04-19

Lithium ions are known for their potent function in modulating the energy alignment at the oxide semiconductor/dye/electrolyte interface in dye-sensitized solar cells (DSCs), offering the opportunity to control the associated multichannel charge-transfer dynamics. Herein, by optimizing the lithium iodide content in 1-ethyl-3-methylimidazolium dicyanamide-based ionic liquid electrolytes, we present a solvent-free DSC displaying an impressive 8.4% efficiency at 100 mW cm(-2) AM1.5G conditions. We further scrutinize the origins of evident impacts of lithium ions upon current density-voltage characteristics as well as photocurrent action spectra of DSCs based thereon. It is found that, along with a gradual increase of the lithium content in ionic liquid electrolytes, a consecutive diminishment of the open-circuit photovoltage arises, primarily owing to a noticeable downward movement of the titania conduction band edge. The conduction band edge displacement away from vacuum also assists the formation of a more favorable energy offset at the titania/dye interface, and thereby leads to a faster electron injection rate and a higher exciton dissociation yield as implied by transient emission measurements. We also notice that the adverse influence of the titania conduction band edge downward shift arising from lithium addition upon photovoltage is partly compensated by a concomitant suppression of the triiodide involving interfacial charge recombination. © 2011 American Chemical Society

International Space Station Active Thermal Control Sub-System On-Orbit Pump Performance and Reliability Using Liquid Ammonia as a Coolant

Science.gov (United States)

Morton, Richard D.; Jurick, Matthew; Roman, Ruben; Adamson, Gary; Bui, Chinh T.; Laliberte, Yvon J.

2011-01-01

The International Space Station (ISS) contains two Active Thermal Control Sub-systems (ATCS) that function by using a liquid ammonia cooling system collecting waste heat and rejecting it using radiators. These subsystems consist of a number of heat exchangers, cold plates, radiators, the Pump and Flow Control Subassembly (PFCS), and the Pump Module (PM), all of which are Orbital Replaceable Units (ORU's). The PFCS provides the motive force to circulate the ammonia coolant in the Photovoltaic Thermal Control Subsystem (PVTCS) and has been in operation since December, 2000. The Pump Module (PM) circulates liquid ammonia coolant within the External Active Thermal Control Subsystem (EATCS) cooling the ISS internal coolant (water) loops collecting waste heat and rejecting it through the ISS radiators. These PM loops have been in operation since December, 2006. This paper will discuss the original reliability analysis approach of the PFCS and Pump Module, comparing them against the current operational performance data for the ISS External Thermal Control Loops.

Status of National Spherical Torus Experiment Liquid Lithium Divertor

Science.gov (United States)

Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

2009-11-01

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

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

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.)

Development of windowless liquid lithium targets for fragmentation and fission of 400-kW uranium beams

CERN Document Server

Nolen, J A; Hassanein, A; Novick, V J; Plotkin, P; Specht, J R

2003-01-01

The driver linac of the proposed rare isotope accelerator facility is designed to deliver 2x10 sup 1 sup 3 uranium ions per second at 400 MeV/u on target for radionuclide production via the fission and fragmentation mechanisms. The ion optics of the large acceptance, high-resolution fragment separators that follow the production target require primary beam spot widths of 1 mm. To cope with the resulting high power densities, windowless liquid lithium targets are being developed. The present designs build on existing experience with liquid lithium and liquid sodium systems that have been used for fusion and fission applications. However, no completely windowless systems have been developed or tested to date. For the beam power indicated above (400 kW), the flow requirements are up to about 20 m/s and 10 l/s linear and volume flow rates, respectively. The required target thickness is 1-1.5 g/cm sup 2 (2-3 cm lithium thickness). At this time a prototype windowless system with a lithium thickness of 1-2 cm is und...

Comparative analysis of coolants for FBR of future nuclear power

International Nuclear Information System (INIS)

Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

2001-01-01

Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

Comparative analysis of coolants for FBR of future nuclear power

Energy Technology Data Exchange (ETDEWEB)

Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

2001-07-01

Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

International Nuclear Information System (INIS)

Casino, William A. Jr.

2006-01-01

The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

Control of the nitrogen concentration in liquid lithium by the hot trap method

International Nuclear Information System (INIS)

Sakurai, Toshiharu; Yoneoka, Toshiaki; Tanaka, Satoru; Suzuki, Akihiro; Muroga, Takeo

2002-01-01

The nitrogen concentration in liquid lithium was controlled by the hot-trap method. Titanium, vanadium and a V-Ti alloy were used as nitrogen gettering materials. Gettering experiments were conducted at 673, 773 and 823 K for 0.4-2.8 Ms. After immersion, the nitrogen concentration increased in titanium and V-Ti were tested at 823 K. Especially the nitrogen gettering effect by the V-10at.%Ti alloy was found to be large. Nitrogen was considered to exist mainly as solid solution in the V-10at.%Ti alloy. The decrease of the nitrogen concentration in liquid lithium by the V-Ti gettering was also confirmed

Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2004-08-01

In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

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

The solid-liquid extraction separation of lithium isotopes by porous composite materials doped with ionic liquids and 2,2'-binaphthyldiyl-17-crown-5

International Nuclear Information System (INIS)

Xiao-Li Sun; Ling Gu; Dan Qiu; Dong-Hong Ren; Zaijun Li; Zhi-Guo Gu; Jiangnan University, Wuxi

2015-01-01

A green and efficient solid-liquid extraction method of lithium isotopes separation by porous composite materials doped with imidazolium ionic liquids and 2,2'-binaphthyldiyl-17-crown-5 has been reported in this paper. The composite materials of mesoporous silica and impregnated resin were synthesized by sol-gel and direct impregnation process, respectively. Various extraction parameters such as the concentration of lithium salt, anion of lithium salt, initial pH, time and temperature were investigated. Under optimized conditions, the maximum single-stage separation factor of 6 Li/ 7 Li was 1.048 ± 0.002, the maximum extraction efficiency was 15.86 %. The sorbents can be regenerated easily with HCl solution and reused repeatedly. (author)

Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

International Nuclear Information System (INIS)

2003-01-01

In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

Research on organic liquids as reactor coolants. Status report from Hungary

International Nuclear Information System (INIS)

Kiss, I.

1967-01-01

The organic-moderated and cooled nuclear reactor concept has stimulated extensive activities in numerous different areas of research. Investigations started in Hungary in 1958 do not cover all topics of interest in organic reactors and so far no projects have been started to build such a reactor. Since OMRE and other organic reactor experiments have already shown the potential use of organic materials as reactor coolants and moderators, efforts have been focused rather on the investigation and solution of certain specific particular problems and also on economic aspects. One of the most important objectives seems to be a better knowledge of the radiolytic heat transfer and neutron physics behaviour of organic liquids. In Hungary the following topics were selected for investigation: Radiation stability of organic compounds and their mixtures; Heat-transfer studies; Investigations on the moderating parameters of organic liquids; Economic analysis of the possible use of organic reactors for process heat application

Research on organic liquids as reactor coolants. Status report from Hungary

Energy Technology Data Exchange (ETDEWEB)

Kiss, I [Central Research Institute for Physics, Budapest (Hungary)

1967-01-01

The organic-moderated and cooled nuclear reactor concept has stimulated extensive activities in numerous different areas of research. Investigations started in Hungary in 1958 do not cover all topics of interest in organic reactors and so far no projects have been started to build such a reactor. Since OMRE and other organic reactor experiments have already shown the potential use of organic materials as reactor coolants and moderators, efforts have been focused rather on the investigation and solution of certain specific particular problems and also on economic aspects. One of the most important objectives seems to be a better knowledge of the radiolytic heat transfer and neutron physics behaviour of organic liquids. In Hungary the following topics were selected for investigation: Radiation stability of organic compounds and their mixtures; Heat-transfer studies; Investigations on the moderating parameters of organic liquids; Economic analysis of the possible use of organic reactors for process heat application.

An investigation into the efficiency of ion-exchange membranes in simulated PWR coolants

International Nuclear Information System (INIS)

Clune, T.

1980-11-01

This report describes an investigation of the retention efficiency of cation-exchange membranes for magnesium, calcium and nickel ions in PWR-coolant type solutions containing 2 ppm lithium (as lithium hydroxide) and 1000 ppm boron (as boric acid). By analysis of the membranes themselves or of the effluent, the retention characteristics of the membranes in various experimental conditions have been examined. (author)

Sudden contact of a hot liquid with a volatile coolant: instability of the created vapour film

International Nuclear Information System (INIS)

Pion, Agnes

1983-01-01

As the sudden contact of a hot body with a coolant which may evaporate, results, after some delay, in an explosive evaporation, this research thesis proposes an interpretation based on the study of the destabilization of the vapour film which forms at the surface of the hot body. The author reports the modelling of the evolution of the average thickness of the film before the explosion. The possible chemical reactions at the surface of the hot body are taken into account. A base flow is obtained which allows the calculation of the evolution of Rayleigh-Taylor instabilities which may occur at the gas-coolant interface. This study is applied to the interaction between liquid sodium and water [fr

Inertia thermonuclear reactor

International Nuclear Information System (INIS)

Imon, Toshiharu; Nakamura, Norio; Oomura, Hiroshi.

1983-01-01

Purpose: To eliminate the requirement of power for controlling the flow velocity of coolants flowing through a porous structure blanket, as well as establish a uniform and stable coolant layer. Constitution: Breeding blanket is made with mesh-like or fiberous porous body, and liquid lithium is introduced into the porous body. The porous body functions as a resistive member to inhibit the free fall of the liquid lithium, so the coolant flowing velocity can be determined to a desired value by appropriately selecting the porosity therein. Further, since liquid lithium flows downwardly at a uniform speed under the effect of the gravitational force, the layer thickness is made uniform to effectively recover neutron energy. Also, while waves are formed at the boundary surface of the liquid lithium layer other than for the porous body due to the collision of fine balls or the likes, they are instantly eliminated by the porous body and the flow can be stabilized. (Yoshino, Y.)

Intermetallic and electrical insulator coatings on high-temperature alloys in liquid-lithium environments

International Nuclear Information System (INIS)

Park, J.H.

1994-06-01

In the design of liquid-metal cooling systems for fusion-reactor blanket, applications, the corrosion resistance of structural materials and the magnetohydrodynamic (MHD) force and its subsequent influence on thermal hydraulics and corrosion are major concerns. When the system is cooled by liquid metals, insulator coatings are required on piping surfaces in contact with the coolant. The objective of this study is to develop stable corrosion-resistant electrical insulator coatings at the liquid-metal/structural-material interface, with emphasis on electrically insulating coatings that prevent adverse MHD-generated currents from passing through the structural wall, and Be-V intermetallic coatings for first-wall components that face the plasma. Vanadium and V-base alloys are leading candidate materials for structural applications in a fusion reactor. Various intermetallic films were produced on V-alloys and on Types 304 and 316 stainless steel. The intermetallic layers were developed by exposure of the materials to liquid Li containing 2 at temperatures of 500--1030 degree C. CaO electrical insulator coatings were produced by reaction of the oxygen-rich layer with <5 at. % Ca dissolved in liquid Li at 400--700 degree C. The reaction converted the oxygen-rich layer to an electrically insulating film. This coating method is applicable to reactor components because the liquid metal can be used over and over; only the solute within the liquid metal is consumed. This paper will discuss initial results on the nature of the coatings and their in-situ electrical resistivity characteristics in liquid Li at high temperatures

Sloshing of coolant in a seismically isolated reactor

International Nuclear Information System (INIS)

Wu, T.S.; Guildys, J.; Seidensticker, R.W.

1988-01-01

During a seismic event, the liquid coolant inside the reactor vessel has sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a synthetic ground motion whose response spectra envelop those specified by the NRC Regulator Guide 1.60, it is found that the maximum sloshing wave height increases from 18 in. to almost 30 in. when the system is seismically isolated. Since higher sloshing wave may introduce severe impact forces and thermal shocks to the reactor closure and other components within the reactor vessel, adequate design considerations should be made either to suppress the wave height or to reduce the effects caused by high waves

Development of small, fast reactor core designs using lead-based coolant

International Nuclear Information System (INIS)

Cahalan, J. E.; Hill, R. N.; Khalil, H. S.; Wade, D. C.

1999-01-01

A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations

Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

OpenAIRE

Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Kozub, Thomas; Berzak, Laura; Hammett, Gregory; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Jacobson, Craig M.; Lucia, Matthew; Jones, Andrew; Lundberg, Daniel; Timberlake, John; Majeski, Richard

2010-01-01

Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas wa...

Soft X-ray emission spectroscopy of liquids and lithium battery materials

International Nuclear Information System (INIS)

Augustsson, Andreas

2004-01-01

Lithium ion insertion into electrode materials is commonly used in rechargeable battery technology. The insertion implies changes in both the crystal structure and the electronic structure of the electrode material. Side-reactions may occur on the surface of the electrode which is exposed to the electrolyte and form a solid electrolyte interface (SEI). The understanding of these processes is of great importance for improving battery performance. The chemical and physical properties of water and alcohols are complicated by the presence of strong hydrogen bonding. Various experimental techniques have been used to study geometrical structures and different models have been proposed to view the details of how these liquids are geometrically organized by hydrogen bonding. However, very little is known about the electronic structure of these liquids, mainly due to the lack of suitable experimental tools. In this thesis examples of studies of lithium battery electrodes and liquid systems using soft x-ray emission spectroscopy will be presented. Monochromatized synchrotron radiation has been used to accomplish selective excitation, in terms of energy and polarization. The electronic structure of graphite electrodes has been studied, before and after lithium intercalation. Changes in the electronic structure upon lithiation due to transfer of electrons into the graphite π-bands have been observed. Transfer of electrons in to the 3d states of transition metal oxides upon lithiation have been studied, through low energy excitations as dd- and charge transfer-excitations. A SEI was detected on cycled graphite electrodes. By the use of selective excitation different carbon sites were probed in the SEI. The local electronic structure of water, methanol and mixtures of the two have been examined using a special liquid cell, to separate the liquid from the vacuum in the experimental chamber. Results from the study of liquid water showed a strong influence on the 3a1 molecular

Pyrrolidinium-based ionic liquid electrolyte with organic additive and LiTFSI for high-safety lithium-ion batteries

International Nuclear Information System (INIS)

Yang, Binbin; Li, Cuihua; Zhou, Junhui; Liu, Jianhong; Zhang, Qianling

2014-01-01

Highlights: • New ionic liquid electrolytes composed by PYR 13 TFSI and EC/DMC-5%VC. • Mixed electrolyte for use in high-safety lithium-ion batteries. • LiTFSI concentration in IL electrolyte greatly affects the rate capability of the cell. • The optimal mixed electrolyte is ideal for applications at high temperature. - Abstract: In this paper, we report on the physicochemical properties of mixed electrolytes based on an ionic liquid N-propyl-N-methylpyrrolidiniumbis (trifluoromethanesulfonyl) imide (PYR 13 TFSI), organic additives, and lithium bis (trifluoromethanesulfonyl) imide (LiTFSI) for high safety lithium-ion batteries. The proposed optimal content of ionic liquid in the mixed electrolyte is 65 vol%, which results in non- flammability, high thermal stability, a wide electrochemical window of 4.8 V, low viscosity, low bulk resistance and the lowest interface resistance to lithium anode. The effects of the concentration of LiTFSI in the above electrolyte are critical to the rate performance of the LiFePO 4 -based battery. We have found the suitable LiTFSI concentration (0.3 M) for good capacity retention and rate capability

Measurement of free-surface of liquid metal lithium jet for IFMIF target

International Nuclear Information System (INIS)

Hiroo Kondo; Nobuo Yamaoka; Takuji Kanemura; Seiji Miyamoto; Hiroshi Horiike; Mizuho Ida; Hiroo Nakamura; Izuru Matsushita; Takeo Muroga

2006-01-01

This reports an experimental study on flow characteristics of a lithium target flow of International Fusion Materials Irradiation Facility (IFMIF). Surface shapes of the target were tried to measure by pattern projection method that is a three dimensional image measurement method. Irregularity of the surface shape caused by surface wakes was successfully measured by the method. IFMIF liquid lithium target is formed a flat plane jet of 25 mm in depth and 260 mm in width, and flows in a flow velocity range of 10 to 20 m/s. Aim of this study is to develop measurement techniques for monitoring of the target when IFMIF is in operation. The lithium target flow is high speed jet and the temperature high is more than 500 K. Also, light is not transmitted into liquid metal lithium. Therefore, almost of all flow measurement techniques developed for water are not used for lithium flow. In this study, pattern projection method was employed to measure the surface irregularity of the target. In the method, stripe patterns are projected onto the flow surface. The projected patterns are deformed according the surface shape. Three-dimensional surface shape is measured by analyzing the deformed patterns recorded using a CCD camera. The method uses the property that lithium dose not transmit visible lights. The experiments were carried out using a lithium loop at Osaka University. In this facility, lithium plane jet of 10 mm in depth and 70 mm width is obtained in the velocity range of less than 15 m/s using a two contractions nozzle. The pattern projection method was used to measure the amplitude of surface irregularity caused by surface wakes. The surface wakes were generated from small damaged at the nozzle edge caused by erosion, and those were successfully measured by the method. The measurement results showed the amplitude of the surface wakes were approximately equal to a size of damage of a nozzle. The amplitude was decreasing with distance to down stream and with decreasing

Characterization of fueling NSTX H-mode plasmas diverted to a liquid lithium divertor

Energy Technology Data Exchange (ETDEWEB)

Kaita, R., E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kugel, H.W.; Abrams, T. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A., E-mail: mjaworsk@pppl.gov [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Kallman, J. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Mansfield, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543 (United States); and others

2013-07-15

Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

Theoretical study and experimental detection of cavitation phenomena in Liquid Lithium Target Facility for IFMIF

International Nuclear Information System (INIS)

Orco, G. Dell; Horiike, H.; Ida, M.; Nakamura, H.

2006-01-01

In the IFMIF (International Fusion Materials Irradiation Facility) testing facility, the required high energy neutrons emission will be produced by reaction of two D + beams with a free surface liquid Lithium jet target flowing along concave back-wall at 20 m/s. The Lithium height in the experimental loop and its relevant static pressure, the high flow velocities and the presence of several devices for the flow control and the pressure reduction increase the risk of cavitation onset in the target system. Special attention has to be taken in the primary pump, in the flow straightener, in the nozzle and their interconnections where the local pressure decreases and/or velocity increases or flow separations could promote the emission of cavitation vapour bubbles. The successive bubble re-implosions, in the higher pressure liquid bulk, could activate material erosion and transportation of activated particulates. These bubbles, if emitted close to the free jet flow, could also procure hydraulic instability and disturbance of the neutron field in the D + beams-Lithium target zone. Therefore, the cavitation risk must be properly foreseen along the whole IFMIF Lithium target circuit and its occurrence at different operating condition should be also monitored by special instrumentation. ENEA, in close cooperation with JAEA, has demonstrated the capability to detect the onset of the cavitation noises in liquid Lithium, by using the ENEA patented accelerometric gauge called CASBA-2000, during hydraulic test campaigns carried-out at Osaka University Lithium facility on a straight mock-up of the IFMIF back plate target. Comparison with the Thoma' cavitation similitude criteria have also determined the critical threshold limit for the estimation of the onset. Theoretical study on the conditions of cavitations generation in the IFMIF Lithium Target Circuit were also launched between ENEA and JAEA aiming at analysing the risk of the cavitation occurrence in the Lithium flow by

A Compact Self-Driven Liquid Lithium Loop for Industrial Neutron Generation

Science.gov (United States)

Stemmley, Steven; Szott, Matt; Kalathiparambil, Kishor; Ahn, Chisung; Jurczyk, Brian; Ruzic, David

2017-10-01

A compact, closed liquid lithium loop has been developed at the University of Illinois to test and utilize the Li-7(d,n) reaction. The liquid metal loop is housed in a stainless steel trench module with embedded heating and cooling. The system was designed to handle large heat and particle fluxes for use in neutron generators as well as fusion devices, solely operating via thermo-electric MHD. The objectives of this project are two-fold, 1) produce a high energy, MeV-level, neutron source and 2) provide a self-healing, low Z, low recycling plasma facing component. The flowing volume will keep a fresh, clean, lithium surface allowing Li-7(d,n) reactions to occur as well as deuterium adsorption in the fluid, increasing the overall neutron output. Expected yields of this system are 107 n/s for 13.5 MeV neutrons and 108 n/s for 2.45 MeV neutrons. Previous work has shown that using a tapered trench design prevents dry out and allows for an increase in velocity of the fluid at the particle strike point. For heat fluxes on the order of 10's MW/m2, COMSOL models have shown that high enough velocities ( 70 cm/s) are attainable to prevent significant lithium evaporation. Future work will be aimed at addressing wettability issues of lithium in the trenches, experimentally determine the velocities required to prevent dry out, and determine the neutron output of the system. The preliminary results and discussion will be presented. DOE SBIR project DE-SC0013861.

Heat transfer in the lithium-cooled blanket of a pulsed fusion reactor

International Nuclear Information System (INIS)

Cort, G.E.; Krakowski, R.A.

1978-01-01

The transient temperature distribution in the lithium-cooled blanket of a pulsed fusion reactor has been calculated using a finite-element heat-conduction computer program. An auxiliary program was used to predict the coolant transient velocity in a network of parallel and series flow passages with constant driving pressure and varying magnetic field. The coolant velocity was calculated by a Runge-Kutta numerical integration of the conservation equations. The lithium coolant was part of the finite-element heat-conduction mesh with the velocity terms included in the total matrix. The matrix was solved implicitly at each time step for the nodal point temperatures. Slug flow was assumed in the coolant passages and the Boussinesq analogy was used to calculate turbulent heat transfer when the magnetic field was not present

Design data, liquid distributors and condenser for a distillation column to enrich tritium in metallic lithium

International Nuclear Information System (INIS)

Barnert, E.

1984-01-01

Tritium, one fuel component of the fusion reactor is bred from the reactors blanket material lithium. Before extracting the tritium from, for instance, metallic lithium by permeation it has to be enriched in the lithium by high temperature distillation. In this report design data for a column for high temperature distillation are given. About the testing of two distributors for small liquid quantities and of a condenser is reported. (orig.) [de

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)

Experimental lithium system. Final report

International Nuclear Information System (INIS)

Kolowith, R.; Berg, J.D.; Miller, W.C.

1985-04-01

A full-scale mockup of the Fusion Materials Irradiation Test (FMIT) Facility lithium system was built at the Hanford Engineering Development Laboratory (HEDL). This isothermal mockup, called the Experimental Lithium System (ELS), was prototypic of FMIT, excluding the accelerator and dump heat exchanger. This 3.8 m 3 lithium test loop achieved over 16,000 hours of safe and reliable operation. An extensive test program demonstrated satisfactory performance of the system components, including the HEDL-supplied electromagnetic lithium pump, the lithium jet target, the purification and characterization hardware, as well as the auxiliary argon and vacuum systems. Experience with the test loop provided important information on system operation, performance, and reliability. This report presents a complete overview of the entire Experimental Lithium System test program and also includes a summary of such areas as instrumentation, coolant chemistry, vapor/aerosol transport, and corrosion

Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

Science.gov (United States)

McLean, A. G.; Gan, K. F.; Ahn, J.-W.; Gray, T. K.; Maingi, R.; Abrams, T.; Jaworski, M. A.; Kaita, R.; Kugel, H. W.; Nygren, R. E.; Skinner, C. H.; Soukhanovskii, V. A.

2013-07-01

Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of Tsurface near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q⊥,peak = 5 MW/m2 inter-ELM and up to 10 MW/m2 during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

Ferrous alloy metallurgy - liquid lithium corrosion and welding. Progress report, January 1-December 31, 1980

International Nuclear Information System (INIS)

Olson, D.L.; Matlock, D.K.

1980-01-01

Fatigue crack growth has been used to evaluate the interaction between liquid lithium and an imposed stress. Fatigue crack growth data on type 304L stainless steel at 700C and 2 1/4Cr-1Mo steel between 500 and 700C show that for all imposed test conditions (i.e. frequency, temperature, and nitrogen content in the lithium) the interaction of lithium with the strain at the crack tip results in enhanced crack growth rates. The enhanced growth rates result from the effects of either enhanced grain boundary penetration or a change in crack propagation mechanism due to liquid metal embrittlement. Auger spectroscopy of grain boundary penetrated specimen shows that a lithium-oxygen compound forms at the grain boundary. Moessbauer evaluations of the ferrite layer of corroded type 304 stainless steel are being used to develop a model for weight loss in liquid lithium. The welding research in progress is directed to characterize the influence of variations of the austenitic weld metal composition on the microstructural and mechanical properties of dissimilar metal weldments. Weldments of 2 1/4Cr-1Mo steel to 316 stainless steel have been investigated for fusion microstructure, thermal expansion impact strength and characterization of specific long time in-service failures. Modification of weld metal microstructures by microalloy additions is being investigated as a concept to improve weld metal properties. The behavior of a strip electrode in a gas metal arc is being investigated to determine the feasibility of gas metal arc weld strip overlay cladding

Application of ionic liquids as an electrolyte additive on the electrochemical behavior of lead acid battery

Energy Technology Data Exchange (ETDEWEB)

Rezaei, Behzad; Mallakpour, Shadpour; Taki, Mahmood [Department of Chemistry, Isfahan University of Technology, Isfahan 84156-83111 (Iran)

2009-02-15

Ionic liquids (ILs) belong to new branch of salts with unique properties which their applications have been increasing in electrochemical systems especially lithium-ion batteries. In the present work, for the first time, the effects of four ionic liquids as an electrolyte additive in battery's electrolyte were studied on the hydrogen and oxygen evolution overpotential and anodic layer formation on lead-antimony-tin grid alloy of lead acid battery. Cyclic and linear sweep voltammetric methods were used for this study in aqueous sulfuric acid solution. The morphology of grid surface after cyclic redox reaction was studied using scanning electron microscopy. The results show that most of added ionic liquids increase hydrogen overpotential and whereas they have no significant effect on oxygen overpotential. Furthermore ionic liquids increase antimony dissolution that might be related to interaction between Sb{sup 3+} and ionic liquids. Crystalline structure of PbSO{sub 4} layer changed with presence of ionic liquids and larger PbSO{sub 4} crystals were formed with some of them. These additives decrease the porosity of PbSO{sub 4} perm selective membrane layer at the surface of electrode. Also cyclic voltammogram on carbon-PbO paste electrode shows that with the presence of ionic liquids, oxidation and reduction peak current intensively increased. (author)

A 20 kw beam-on-target test of a high-power liquid lithium target for RIA

International Nuclear Information System (INIS)

Reed, Claude B.; Nolen, Jerry A.; Specht, James R.; Novick, Vincent J.; Plotkin, Perry

2004-01-01

The high-power heavy-ion beams produced by the Rare Isotope Accelerator (RIA) driver linac have large energy deposition density in solids and in many cases no solid materials would survive the full beam power. Liquid lithium technology has been proposed to solve this problem in RIA. Specifically, a windowless target for the production of radioactive ions via fragmentation, consisting of a jet of about 3 cm thickness of flowing liquid lithium, exposed to the beamline vacuum [1,2] is being developed. To demonstrate that power densities equivalent to a 200-kW RIA uranium beam, deposited in the first 4 mm of a flowing lithium jet, can be handled by the windowless target design, a high power 1 MeV Dynamitron was leased and a test stand prepared to demonstrate the target's capability of absorbing and carrying away a 20kW heat load without disrupting either the 5 mm x 10 mm flowing lithium jet target or the beam line vacuum

Investigation of corrosion, water reaction, polonium evaporation and bismuth resource in liquid metal lead-bismuth technology

Energy Technology Data Exchange (ETDEWEB)

Takano, Hideki; Takizuka, Takakazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kitano, Teruaki [Mitsui Shipbuilding and Engineering Co. Ltd., Tokyo (Japan)

2000-10-01

Lead-bismuth is the first candidate material for liquid metal target find coolant of fueled blanket system in accelerator-driven system (ADS) studied at JAERI. Advantages of the lead-bismuth utilization are non-active material, very low capture cross section, low melting point of 125degC and high boiling point of 1670degC, and beside coolant void reactivity become negative. But problems are due to the high corrosivity to most of the structural materials and the corrosive data are scarcity. In this report, corrosivity, reaction with water, thermal-hydraulics, chemical toxicity etc. are studied by investigating some facilities utilized and researched really for lead or lead-bismuth. And, furthermore, polonium evaporation rate and bismuth resource are investigated. Main results obtained are as follows: (1) In a refinery, there are enough employment experience for liquid Pb-Bi in period of about 17 years and not corrosion for the thermal conductive materials (1Cr-0.5Mo steel) used under the condition of natural convection with temperature around 400degC. (2) In Russia, extensive experience in the use as Russian submarines and in R and D during about 50 years are available. And as a result, it will be able to lead approximately zero corrosion for Cr-Si materials by adjusting oxygen film with oxygen concentration control between 10{sup -7} to 10{sup -5}% mass. However, the corrosion data are not enough systematically collected involving them in radiation dose field. (3) In liquid-dropping experiment, it is shown that interaction between water and high temperature liquid Pb-Bi is reduced steeply with rising of atmosphere pressure. But, in order to design the second circuit removal model of ADS, the interaction should be evaluated by water continuous injection experiment. (4) Polonium forms PbPo in Pb-Bi, and the evaporation rate become less three factor than that of Po, and furthermore, the rate decreases in the atmosphere. The effects of Po on employee and environment

Current-carrying capacity dependence of composite Bi2Sr2CaCu2O8 superconductors on the liquid coolant conditions

International Nuclear Information System (INIS)

Romanovskii, V R; Watanabe, K; Awaji, S; Nishijima, G

2006-01-01

The thermal runaway conditions of the composite Bi 2 Sr 2 CaCu 2 O 8 superconductor cooled by liquid helium or liquid hydrogen are compared. The study based on the static analysis of thermoelectric modes was made when the volume fraction of the superconductor in a composite was varied. Some specific trends underlying the onset of thermal runaway in superconducting composites cooled by liquid coolants are discussed. It is stated that the operating modes of superconducting composites may be characterized by stable states during which the current-carrying capacity of a superconductor is not effectively used even with a high amount of superconductor in the composite. These states are possible due to the corresponding temperature variation of the resistivities of the matrix and the superconductor in the high operating temperature range. They have to be considered in experiments when the critical current of a superconductor is determined or when the optimal stable operating modes of the current-carrying elements based on the Bi 2 Sr 2 CaCu 2 O 8 superconductor, which is cooled by liquid coolant, are defined

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)

Ionic liquids and derived materials for lithium and sodium batteries.

Science.gov (United States)

Yang, Qiwei; Zhang, Zhaoqiang; Sun, Xiao-Guang; Hu, Yong-Sheng; Xing, Huabin; Dai, Sheng

2018-03-21

The ever-growing demand for advanced energy storage devices in portable electronics, electric vehicles and large scale power grids has triggered intensive research efforts over the past decade on lithium and sodium batteries. The key to improve their electrochemical performance and enhance the service safety lies in the development of advanced electrode, electrolyte, and auxiliary materials. Ionic liquids (ILs) are liquids consisting entirely of ions near room temperature, and are characterized by many unique properties such as ultralow volatility, high ionic conductivity, good thermal stability, low flammability, a wide electrochemical window, and tunable polarity and basicity/acidity. These properties create the possibilities of designing batteries with excellent safety, high energy/power density and long-term stability, and also provide better ways to synthesize known materials. IL-derived materials, such as poly(ionic liquids), ionogels and IL-tethered nanoparticles, retain most of the characteristics of ILs while being endowed with other favourable features, and thus they have received a great deal of attention as well. This review provides a comprehensive review of the various applications of ILs and derived materials in lithium and sodium batteries including Li/Na-ion, dual-ion, Li/Na-S and Li/Na-air (O 2 ) batteries, with a particular emphasis on recent advances in the literature. Their unique characteristics enable them to serve as advanced resources, medium, or ingredient for almost all the components of batteries, including electrodes, liquid electrolytes, solid electrolytes, artificial solid-electrolyte interphases, and current collectors. Some thoughts on the emerging challenges and opportunities are also presented in this review for further development.

Measurement of hydrogen solubility and desorption rate in V-4Cr-4Ti and liquid lithium-calcium alloys

Energy Technology Data Exchange (ETDEWEB)

Park, J.H.; Erck, R.; Park, E.T. [Argonne National Lab., IL (United States)] [and others

1997-04-01

Hydrogen solubility in V-4Cr-4Ti and liquid lithium-calcium was measured at a hydrogen pressure of 9.09 x 10{sup {minus}4} torr at temperatures between 250 and 700{degrees}C. Hydrogen solubility in V-4Cr-4Ti and liquid lithium decreased with temperature. The measured desorption rate of hydrogen in V-4Cr-4Ti is a thermally activated process; the activation energy is 0.067 eV. Oxygen-charged V-4Cr-4Ti specimens were also investigated to determine the effect of oxygen impurity on hydrogen solubility and desorption in the alloy. Oxygen in V-4Cr-4Ti increases hydrogen solubility and desorption kinetics. To determine the effect of a calcium oxide insulator coating on V-4Cr-4Ti, hydrogen solubility in lithium-calcium alloys that contained 0-8.0 percent calcium was also measured. The distribution ratio R of hydrogen between liquid lithium or lithium-calcium and V-4Cr-4Ti increased as temperature decreased (R {approx} 10 and 100 at 700 and 250{degrees}C, respectively). However at <267{degrees}C, solubility data could not be obtained by this method because of the slow kinetics of hydrogen permeation through the vanadium alloy.

Measurement of hydrogen solubility and desorption rate in V-4Cr-4Ti and liquid lithium-calcium alloys

International Nuclear Information System (INIS)

Park, J.H.; Erck, R.; Park, E.T.

1997-01-01

Hydrogen solubility in V-4Cr-4Ti and liquid lithium-calcium was measured at a hydrogen pressure of 9.09 x 10 -4 torr at temperatures between 250 and 700 degrees C. Hydrogen solubility in V-4Cr-4Ti and liquid lithium decreased with temperature. The measured desorption rate of hydrogen in V-4Cr-4Ti is a thermally activated process; the activation energy is 0.067 eV. Oxygen-charged V-4Cr-4Ti specimens were also investigated to determine the effect of oxygen impurity on hydrogen solubility and desorption in the alloy. Oxygen in V-4Cr-4Ti increases hydrogen solubility and desorption kinetics. To determine the effect of a calcium oxide insulator coating on V-4Cr-4Ti, hydrogen solubility in lithium-calcium alloys that contained 0-8.0 percent calcium was also measured. The distribution ratio R of hydrogen between liquid lithium or lithium-calcium and V-4Cr-4Ti increased as temperature decreased (R ∼ 10 and 100 at 700 and 250 degrees C, respectively). However at <267 degrees C, solubility data could not be obtained by this method because of the slow kinetics of hydrogen permeation through the vanadium alloy

The use of Zeolite into the controlling of Lithium concentration in the PWR primary water coolant (I) : the influences of Ca, Mg and Boric Acid concentration into the exchanges capacity of Ammonium Zeolite

International Nuclear Information System (INIS)

Sumijanto; Siti-Amini

1996-01-01

In this first part of research, the influences of calsium, magnesium and boric acid concentrations to the zeolite uptake of lithium in the PWR primary water coolant have been studied. The ammonium form of zeolite was found by modification of the natural zeolite which was originated from Bayah. The results showed that the boric acid concentration in the normal condition of PWR operation absolutely did not affects the lithium uptake. The Li uptake efficiency was influenced by the presence of Ca and Mg ions in order to the presence of cations competition which was dominated by Ca ion

Solubility of iron in liquid lead

International Nuclear Information System (INIS)

Ali-Khan, I.

1981-01-01

The use of liquid lead in high temperature chemical and metallurgical processes is well known. The structural materials applied for the containment of these processes are either iron base alloys or possess iron as an alloying element. Besides that, lead itself is alloyed in some steels to achieve some very useful properties. For understanding the effect of liquid lead in such structural materials, it is important to determine the solubility of iron in liquid lead which would also be indicative of the stability of these alloys. At the institute of reactor materials of KFA Juelich, investigations have been conducted to determine the solubility of iron in liquid lead up to a temperature of about 1000 0 C. In this presentation the data concerning the solubility of iron in liquid lead are brought up to date and discussed including the results of our previous investigations. (orig.)

Laboratory studies of H retention and LiH formation in liquid lithium

Energy Technology Data Exchange (ETDEWEB)

Martín-Rojo, A.B. [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain); UC3M Madrid, 126, 28903 Getafe (Spain); Oyarzabal, E. [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain); U.N.E.D. Ciudad Universitaria, S/N, 28040, Madrid Spain (Spain); Tabarés, F.L., E-mail: tabares@ciemat.es [Ass. Euratom-Ciemat, Av. Complutense 22, 28040 Madrid (Spain)

2014-12-15

Highlights: • Absorption and thermal desorption experiments of hydrogen isotopes in liquid lithium have been performed at exposure temperatures up to 400 °C. • The kinetics of the involved processes indicate a two-stage mechanism for hydride production. • TDS peaks at temperatures well below the expected one for thermal decomposition of the hydride were systematically recorded, although only a small fraction of the absorbed gas was released during the TDS cycle. • The absorption of H{sub 2} in a D{sub 2}-loaded sample was investigated at two temperatures, and no obvious influence of the preexisting species in the rate of absorption of H{sub 2} was seen. • Deuterium absorption takes place at a higher rate than that of hydrogen. - Abstract: Laboratory experiments on H/D retention on liquid lithium followed by thermal desorption spectrometry (TDS) have been performed at Ciemat. Two different experimental set ups were used in order to expose liquid Li to hydrogen gas or to hydrogen glow discharge plasmas at temperatures up to 673 K. In the present work the results concerning the gas phase absorption are addressed. Two different kinetics of absorption were identified from the time evolution of the uptake. Alternate exposures to H{sub 2} and D{sub 2} were carried out in order to study the isotope exchange and its possible use for tritium retention control in Fusion Reactor. Although important differences were found in the absorption kinetics of both species, the total retention seems to be governed by the total sum of hydrogenic isotopes, and only small differences were found in the corresponding TDS spectra, on which evidence of some isotope exchange is observed. The results are discussed in relation to the potential use of liquid lithium walls in a Fusion Reactor.

Laboratory studies of H retention and LiH formation in liquid lithium

International Nuclear Information System (INIS)

Martín-Rojo, A.B.; Oyarzabal, E.; Tabarés, F.L.

2014-01-01

Highlights: • Absorption and thermal desorption experiments of hydrogen isotopes in liquid lithium have been performed at exposure temperatures up to 400 °C. • The kinetics of the involved processes indicate a two-stage mechanism for hydride production. • TDS peaks at temperatures well below the expected one for thermal decomposition of the hydride were systematically recorded, although only a small fraction of the absorbed gas was released during the TDS cycle. • The absorption of H 2 in a D 2 -loaded sample was investigated at two temperatures, and no obvious influence of the preexisting species in the rate of absorption of H 2 was seen. • Deuterium absorption takes place at a higher rate than that of hydrogen. - Abstract: Laboratory experiments on H/D retention on liquid lithium followed by thermal desorption spectrometry (TDS) have been performed at Ciemat. Two different experimental set ups were used in order to expose liquid Li to hydrogen gas or to hydrogen glow discharge plasmas at temperatures up to 673 K. In the present work the results concerning the gas phase absorption are addressed. Two different kinetics of absorption were identified from the time evolution of the uptake. Alternate exposures to H 2 and D 2 were carried out in order to study the isotope exchange and its possible use for tritium retention control in Fusion Reactor. Although important differences were found in the absorption kinetics of both species, the total retention seems to be governed by the total sum of hydrogenic isotopes, and only small differences were found in the corresponding TDS spectra, on which evidence of some isotope exchange is observed. The results are discussed in relation to the potential use of liquid lithium walls in a Fusion Reactor

Electromagnetic pumping of liquid lithium in inertial confinement fusion reactors

International Nuclear Information System (INIS)

Baker, R.S.; Blink, J.A.; Tessier, M.J.

1983-01-01

The basic operating principles and geometries of ten electromagnetic pumps are described. Two candidate pumps, the annular-linear-induction pump and the helical-rotor electromagnetic pump, are compared for possible use in a full-scale liquid-lithium inertial confinement fusion reactor. A parametric design study completed for the helical-rotor pump is shown to be valid when applied to an experimental sodium pump. Based upon the preliminary HYLIFE requirements for a lithium flow rate per pump of 8.08 m 3 /s at a head of 82.5 kPa, a complete set of 70 variables are specified for a helical-rotor pump with either a normally conducting or a superconducting winding. The two alternative designs are expected to perform with efficiencies of 50 and 60%, respectively

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

International Nuclear Information System (INIS)

1994-01-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-07-01

The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned.

The use of lithium as a marker for the retention of liquids in the oral cavity after rinsing.

Science.gov (United States)

Hanning, Sara M; Kieser, Jules A; Ferguson, Martin M; Reid, Malcolm; Medlicott, Natalie J

2014-01-01

The aim of this study was to validate the use of lithium as a marker to indicate the retention of simple liquids in the oral cavity and use this to determine how much liquid is retained in the oral cavity following 30 s of rinsing. This is a validation study in which saliva was spiked with known concentrations of lithium. Twenty healthy participants then rinsed their mouths with either water or a 1 % w/v carboxymethylcellulose (CMC) solution for 30 s before expectorating into a collection cup. Total volume and concentration of lithium in the expectorant were then measured, and the percentage of liquid retained was calculated. The mean amount of liquid retained was 10.4 ± 4.7 % following rinsing with water and 15.3 ± 4.1 % following rinsing with 1 % w/v CMC solution. This difference was significant (p < 0.01). Lithium was useful as a marker for the retention of liquids in the oral cavity, and a value for the amount of water and 1 % w/v CMC solution remaining in the oral cavity following a 30-s rinse was established. The present study quantifies the retention of simple fluids in the oral cavity, validating a technique that may be applied to more complex fluids such as mouth rinses. Further, the application of this method to specific population groups such as those with severe xerostomia may assist in developing effective saliva substitutes.

Compatibility of AlN ceramics with molten lithium

Energy Technology Data Exchange (ETDEWEB)

Yoneoka, Toshiaki; Sakurai, Toshiharu; Sato, Toshihiko; Tanaka, Satoru [Tokyo Univ., Department of Quantum Engineering and Systems Science, Tokyo (Japan)

2002-04-01

AlN ceramics were a candidate for electrically insulating materials and facing materials against molten breeder in a nuclear fusion reactor. In the nuclear fusion reactor, interactions of various structural materials with solid and liquid breeder materials as well as coolant materials are important. Therefore, corrosion tests of AlN ceramics with molten lithium were performed. AlN specimens of six kinds, different in sintering additives and manufacturing method, were used. AlN specimens were immersed into molten lithium at 823 K. Duration for the compatibility tests was about 2.8 Ms (32 days). Specimens with sintering additive of Y{sub 2}O{sub 3} by about 5 mass% formed the network structure of oxide in the crystals of AlN. It was considered that the corrosion proceeded by reduction of the oxide network and the penetration of molten lithium through the reduced pass of this network. For specimens without sintering additive, Al{sub 2}O{sub 3} containing by about 1.3% in raw material was converted to fine oxynitride particles on grain boundary or dissolved in AlN crystals. After immersion into lithium, these specimens were found to be sound in shape but reduced in electrical resistivity. These degradation of the two types specimens were considered to be caused by the reduction of oxygen components. On the other hand, a specimen sintered using CaO as sintering additive was finally became appreciably high purity. This specimen showed good compatibility for molten lithium at least up to 823 K. It was concluded that the reduction of oxygen concentration in AlN materials was essential in order to improve the compatibility for molten lithium. (author)

A liquid-nitrogen monitor for lithium-drifted germanium detectors

International Nuclear Information System (INIS)

Andeweg, A.H.

1977-11-01

An instrument has been developed that makes use of a load cell to monitor the liquid nitrogen in the Dewar flask of a lithium-drifted germaniun detector. The contents are recorded on a chart recorder, and an alarm is sounded when the previously set content has been reached. A signal switches off the high-voltage power supply 30 minutes after the alarm is triggered. The calibration of the load-cell monitor is described in an appendix [af

RELAP/SCDAPSIM/MOD4.0 modification for transient accident scenario of Test Blanket Modules in ITER involving helium flows into heavy liquid metal

Energy Technology Data Exchange (ETDEWEB)

Freixa, J.; Pérez, M.; Mas de les Valls, E.; Batet, L.; Sandeep, T.; Chaudhari, V.; Reventós, F.

2015-07-01

The Institute for Plasma Research (IPR), India, is currently involved in the design and development of its Test Blanket Module (TBM) for testing in ITER (International Thermo nuclear Experimental Reactor). The Indian TBM concept is a Lead-Lithium cooled Ceramic Breeder (LLCB), which utilizes lead-lithium eutectic alloy (LLE) as tritium breeder, neutron multiplier and coolant. The first wall facing the plasma is cooled by helium gas. In preparation of the regulatory safety files of ITER-TBM, a number of off-normal event sequences have been postulated. Thermal hydraulic safety analyses of the TBM system will be carried out with the system code RELAP/SCDAPSIM/MOD4.0 which was initially designed to predict the behavior of light water reactor systems during normal and accidental conditions. In order to analyze some of the postulated off-normal events, there is the need to simulate the mixing of Helium and Lead-Lithium fluids. The Technical University of Catalonia is cooperating with IPR to implement the necessary changes in the code to allow for the mixing of helium and liquid metal. In the present study, the RELAP/SCDAPSIM/MOD4 two-phase flow 6-equations structure has been modified to allow for the mixture of LLE in the liquid phase with dry Helium in the gas phase. Practically obtaining a two-fluid 6-equation model where each fluid is simulated with a set of energy, mass and momentum balance equations. A preliminary flow regime map for LLE and helium flow has been developed on the basis of numerical simulations with the OpenFOAM CFD toolkit. The new code modifications have been verified for vertical and horizontal configurations. (Author)

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

Power reactors and sub-critical blanket systems with lead and lead-bismuth as coolant and/or target material. Utilization and transmutation of actinides and long lived fission products

International Nuclear Information System (INIS)

2003-05-01

High level radioactive waste disposal is an issue of great importance in the discussion of the sustainability of nuclear power generation. The main contributors to the high radioactivity are the fission products and the minor actinides. The long lived fission products and minor actinides set severe demands on the arrangements for safe waste disposal. Fast reactors and accelerator driven systems (ADS) are under development in Member States to reduce the long term hazard of spent fuel and radioactive waste, taking advantage of their incineration and transmutation capability. Important R and D programmes are being undertaken in many Member States to substantiate this option and advance the basic knowledge in this innovative area of nuclear energy development. The conceptual design of the lead cooled fast reactor concept BREST-OD-300, as well as various other conceptual designs of lead/lead-bismuth cooled fast reactors have been developed to meet enhanced safety and non-proliferation requirements, aiming at both energy production and transmutation of nuclear waste. Some R and D studies indicate that the use of lead and lead-bismuth coolant has some advantages in comparison with existing sodium cooled fast reactor systems, e.g.: simplified design of fast reactor core and BOP, enhanced inherent safety, and easier radwaste management in related fuel cycles. Moreover, various ADS conceptual designs with lead and lead-bismuth as target material and coolant also have been pursued. The results to date are encouraging, indicating that the ADS has the potential to offer an option for meeting the challenges of the back end fuel cycle. During the last decade, there have been substantial advances in several countries with their own R and D programme in the fields of lead/lead-bismuth cooled critical and sub-critical concepts. coolant technology, and experimental validation. In this context, international exchange of information and experience, as well as international

Liquid jet experiments: relevance to inertial confinement fusion reactors

International Nuclear Information System (INIS)

Hoffman, M.A.

1981-01-01

In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

Measurement and modeling of surface temperature dynamics of the NSTX liquid lithium divertor

Energy Technology Data Exchange (ETDEWEB)

McLean, A.G., E-mail: mclean@fusion.gat.com [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Gan, K.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Ahn, J.-W.; Gray, T.K.; Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Abrams, T.; Jaworski, M.A.; Kaita, R.; Kugel, H.W. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R.E. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Skinner, C.H. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

2013-07-15

Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of T{sub surface} near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q{sub ⊥,peak} = 5 MW/m{sup 2} inter-ELM and up to 10 MW/m{sup 2} during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

Improved liquid-lithium target for the FMIT facility

International Nuclear Information System (INIS)

Miles, R.R.; Greenwell, R.K.; Hassberger, J.A.; Ingham, J.G.

1982-11-01

An improved target for the Fusion Materials Irradiation Testing Facility was designed. The purpose of the target is to produce a high neutron flux (10 19 n/m 2 sec) for testing of candidate first wall materials for fusion reactors. The neutrons are produced through a Li(d,n) stripping reaction between accelerated deuterons (35 MeV, 0.1A) and a thin jet of flowing liquid lithium. The target consists of a high speed (approx. 17 m/s), free surface wall jet which is exposed to the high (10 -4 Pa) accelerator vacuum. The energy deposited by the deuteron beam in the lithium is sufficient to heat the jet internally to a maximum temperature of roughly 740 0 C, 430 0 C greater than the saturation temperature at the jet free surface. For this reason, the jet flows along a curved wall which provides the pressurization required to prevent sperheat internal to the jet. Supporting hardware for the jet and a drain line which controls the jet beyond the beam intercept region

Liquid lithium surface control and its effect on plasma performance in the HT-7 tokamak

Energy Technology Data Exchange (ETDEWEB)

Zuo, G.Z.; Ren, J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, J.S., E-mail: hujs@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, Z.; Yang, Q.X.; Li, J.G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zakharov, L.E. [Princeton University Plasma Physics Laboratory Princeton, NJ 08543 (United States); Ruzic, David N. [University of Illinois, Urbana, IL 61801 (United States)

2014-12-15

Highlights: • Strong interaction between plasma and Li would cause strong Li emission and lead to disruptive plasmas, and probable reasons were analyzed. • Serious Li would be emitted from the free statics surface mainly due to J × B force leading to plasma instable and disruptions. • CPS surface would partially suppress the emission and be beneficial for plasma operation. • Li emission from flowing LLLs on free surfaces on SS trenches and on SS plate were compared. - Abstract: Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.

Liquid lithium loop system to solve challenging technology issues for fusion power plant

Science.gov (United States)

Ono, M.; Majeski, R.; Jaworski, M. A.; Hirooka, Y.; Kaita, R.; Gray, T. K.; Maingi, R.; Skinner, C. H.; Christenson, M.; Ruzic, D. N.

2017-11-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s-1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s-1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s-1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s-1 of tritium is needed to maintain the fusion fuel cycle

Experimental study of gaseous lithium deuterides and lithium oxides. Implications for the use of lithium and Li2O as breeding materials in fusion reactor blankets

International Nuclear Information System (INIS)

Ihle, H.R.; Wu, C.H.; Kudo, H.

1980-01-01

In addition to LiH, which has been studied extensively by optical spectroscopy, the existence of a number of other stable lithium hydrides has been predicted theoretically. By analysis of the saturated vapour over dilute solutions of the hydrogen isotopes in lithium, using Knudsen effusion mass spectrometry, all lithium hydrides predicted to be stable were found. Solutions of deuterium in lithium were used predominantly because of practical advantages for mass spectrometric measurements. The heats of dissociation of LiD, Li 2 D, LiD 2 and Li 2 D 2 , and the binding energies of their singly charged positive ions were determined, and the constants of the gas/liquid equilibria were calculated. The existence of these lithium deuterides in the gas phase over solutions of deuterium in lithium leads to enrichment of deuterium in the gas above 1240 K. The enrichment factor, which increases exponentially with temperature and is independent of concentration for low concentrations of deuterium in the liquid, was determined by Rayleigh distillation experiments. It was found that it is thermodynamically possible to separate deuterium from lithium by distillation. One of the alternatives to the use of lithium in (D,T)-fusion reactors as tritium-breeding blanket material is to employ solid lithium oxide. This has a high melting point, a high lithium density and still favourable tritium-breeding properties. Because of its rather high volatility, an experimental study of the vaporization of Li 2 O was undertaken by mass spectrometry. It vaporizes to give lithium and oxygen, and LiO, Li 2 O, Li 3 O and Li 2 O 2 . The molecule Li 3 O was found as a new species. Heats of dissociation, binding energies of the various ions and the constants of the gas/solid equilibria were determined. The effect of using different materials for the Knudsen cells and the relative thermal stabilities of lithium-aluminium oxides were also studied. (author)

Knight shift of 23Na and 7Li nuclei in liquid sodium-lithium alloys

International Nuclear Information System (INIS)

Feitsma, P.D.

1977-01-01

The Knight shift of 23 Na and 7 Li nuclei in liquid sodium-lithium alloys has been measured. Some aspects of the theoretical interpretation of the Knight shift within the diffraction model, are clarified

A green strategy for lithium isotopes separation by using mesoporous silica materials doped with ionic liquids and benzo-15-crown-5

International Nuclear Information System (INIS)

Wen Zhou; Xiao-Li Sun; Lin Gu; Fei-Fei Bao; Xin-Xin Xu; Chun-Yan Pang; Zaijun Li; Zhi-Guo Gu; Jiangnan University, Wuxi

2014-01-01

Three new mesoporous silica materials IL15SGs (HF15SG, TF15SG and DF15SG) doped with benzo-15-crown-5 and imidazolium based ionic liquids (C 8 mim + PF 6 - , C 8 mim + BF 4 - or C 8 mim + NTf 2 - ) have been prepared by a simple approach to separating lithium isotopes. The formed mesoporous structures of silica gels have been confirmed by transmission electron microscopy image and N 2 gas adsorption-desorption isotherm. Imidazolium ionic liquids acted as templates to prepare mesoporous materials, additives to stabilize extractant within silica gel, and synergetic agents to separate the lithium isotopes. Factors such as lithium salt concentration, initial pH, counter anion of lithium salt, extraction time, and temperature on the lithium isotopes separation were examined. Under optimized conditions, the extraction efficiency of HF15SG, TF15SG and DF15SG were found to be 11.43, 10.59 and 13.07 %, respectively. The heavier isotope 7 Li was concentrated in the solution phase while the lighter isotope 6 Li was enriched in the gel phase. The solid-liquid extraction maximum single-stage isotopes separation factor of 6 Li- 7 Li in the solid-liquid extraction was up to 1.046 ± 0.002. X-ray crystal structure analysis indicated that the lithium salt was extracted into the solid phase with crown ether forming [(Li 0.5 ) 2 (B 15 ) 2 (H 2 O)] + complexes. IL15SGs were also easily regenerated by stripping with 20 mmol L -1 HCl and reused in the consecutive removal of lithium ion in five cycles. (author)

New Ether-functionalized Morpholinium- and Piperidinium-based Ionic Liquids as Electrolyte Components in Lithium and Lithium-Ion Batteries.

Science.gov (United States)

Navarra, Maria Assunta; Fujimura, Kanae; Sgambetterra, Mirko; Tsurumaki, Akiko; Panero, Stefania; Nakamura, Nobuhumi; Ohno, Hiroyuki; Scrosati, Bruno

2017-06-09

Here, two ionic liquids, N-ethoxyethyl-N-methylmorpholinium bis(trifluoromethanesulfonyl)imide (M 1,2O2 TFSI) and N-ethoxyethyl-N-methylpiperidinium bis(trifluoromethanesulfonyl)imide (P 1,2O2 TFSI) were synthesized and compared. Fundamental relevant properties, such as thermal and electrochemical stability, density, and ionic conductivity were analyzed to evaluate the effects caused by the presence of the ether bond in the side chain and/or in the organic cation ring. Upon lithium salt addition, two electrolytes suitable for lithium batteries applications were found. Higher conducting properties of the piperidinium-based electrolyte resulted in enhanced cycling performances when tested with LiFePO 4 (LFP) cathode in lithium cells. When mixing the P 1,2O2 TFSI/LiTFSI electrolyte with a tailored alkyl carbonate mixture, the cycling performance of both Li and Li-ion cells greatly improved, with prolonged cyclability delivering very stable capacity values, as high as the theoretical one in the case of Li/LFP cell configurations. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

77 FR 2437 - Special Conditions: Gulfstream Aerospace Corporation, Model GVI Airplane; Rechargeable Lithium...

Science.gov (United States)

2012-01-18

... delivery of the affected aircraft. In addition, the substance of these special conditions has been subject... Ni-Cd and lead-acid cells, some types of lithium-battery cells use flammable liquid electrolytes. The... lithium batteries. The flammable-fluid fire-protection requirements of Sec. 25.863. In the past, this rule...

A conceptual composite blanket design for the Tokamak type of thermonuclear reactor incorporating thermoelectric pumping of liquid lithium

International Nuclear Information System (INIS)

Dutta Gupta, P.B.

1981-01-01

The conceptual liquid lithium blanket design for the tokamak type of thermonuclear reactor put forward is a modification of the initial simple but novel design concept enunciated earlier that exploits the availability of suitably oriented magnetic fields and temperature gradients within the blanket to pump the liquid as has been shown feasible by laboratory model experiments. The modular construction of the blanket cells is retained but the earlier simple back to back double spiralling channel module is replaced by a composite unit of three radially nested layer-structures to optimise heat and tritium extraction from the blanket. The layer-structure at the first wall generates liquid lithium circulation by thermoelectric magnetohydrodynamic forces and the segregated double spiralling channels serve as inlet-outlet driving devices. The outermost layer-structure is cooled by helium. Liquid lithium in the intermediate layer-structure is pumped at a very slow rate. The choice of the relative dimensional proportions of the three layer-structure and the channel cross-section, material property and the spiralling contour is of critical importance for the design. This paper presents the design data for a conceptual design of such a blanket with a 5000 MW (th) rating. (author)

High power accelerator-based boron neutron capture with a liquid lithium target and new applications to treatment of infectious diseases

Energy Technology Data Exchange (ETDEWEB)

Halfon, S. [Soreq NRC, Yavne 81800 (Israel); Racah Institute of Physics, Hebrew University, Jerusalem 91904 (Israel)], E-mail: halfon@phys.huji.ac.il; Paul, M. [Racah Institute of Physics, Hebrew University, Jerusalem 91904 (Israel); Steinberg, D. [Biofilm Laboratory, Institute of Dental Sciences, Faculty of Dentistry, Hebrew University-Hadassah (Israel); Nagler, A.; Arenshtam, A.; Kijel, D. [Soreq NRC, Yavne 81800 (Israel); Polacheck, I. [Clinical Microbiology and Infectious Diseases, Hadassah-Hebrew University Medical Center (Israel); Srebnik, M. [Department of Medicinal Chemistry and Natural Products, School of Pharmacy, Hebrew University, Jerusalem 91120 (Israel)

2009-07-15

A new conceptual design for an accelerator-based boron neutron capture therapy (ABNCT) facility based on the high-current low-energy proton beam driven by the linear accelerator at SARAF (Soreq Applied Research Accelerator Facility) incident on a windowless forced-flow liquid-lithium target, is described. The liquid-lithium target, currently in construction at Soreq NRC, will produce a neutron field suitable for the BNCT treatment of deep-seated tumor tissues, through the reaction {sup 7}Li(p,n){sup 7}Be. The liquid-lithium target is designed to overcome the major problem of solid lithium targets, namely to sustain and dissipate the power deposited by the high-intensity proton beam. Together with diseases conventionally targeted by BNCT, we propose to study the application of our setup to a novel approach in treatment of diseases associated with bacterial infections and biofilms, e.g. inflammations on implants and prosthetic devices, cystic fibrosis, infectious kidney stones. Feasibility experiments evaluating the boron neutron capture effectiveness on bacteria annihilation are taking place at the Soreq nuclear reactor.

Conductivity of liquid lithium electrolytes with dispersed mesoporous silica particles

International Nuclear Information System (INIS)

Sann, K.; Roggenbuck, J.; Krawczyk, N.; Buschmann, H.; Luerßen, B.; Fröba, M.; Janek, J.

2012-01-01

Highlights: ► The conductivity of disperse lithium electrolytes with mesoporous fillers is studied. ► In contrast to other investigations in literature, no conductivity enhancement could be observed for standard battery electrolytes and typical mesoporous fillers in various combinations. ► Disperse electrolytes can become relevant in terms of battery safety. ► Dispersions of silicas and electrolyte with LiPF 6 as conducting salt are not stable, although the silicas were dried prior to preparation and the electrolyte water content was controlled. Surface modification of the fillers improved the stability. ► The observed conductivity decrease varied considerably for various fillers. - Abstract: The electrical conductivity of disperse electrolytes was systematically measured as a function of temperature (0 °C to 60 °C) and filler content for different types of fillers with a range of pore geometry, pore structure and specific surface area. As fillers mesoporous silicas SBA-15, MCM-41 and KIT-6 with pore ranges between 3 nm and 15 nm were dispersed in commercially available liquid lithium electrolytes. As electrolytes 1 M of lithium hexafluorophosphate (LiPF 6 ) in a mixture of ethylene carbonate (EC) and diethylene carbonate (DEC) at the ratio 3:7 (wt/wt) and the same solvent mixture with 0.96 M lithium bis(trifluoromethanesulfon)imide (LiTFSI) were used. No conductivity enhancement could be observed, but with respect to safety aspects the highly viscous disperse pastes might be useful. The conductivity decrease varied considerably for the different fillers.

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.)

Loss of coolant accident mitigation for liquid metal cooled space reactors

International Nuclear Information System (INIS)

Georgevich, Vladimir; Best, Frederick; Erdman, Carl

1989-01-01

A loss of coolant accident (LOCA) in a liquid metal-cooled space reactor system has been considered as a possible accident scenario. Development of new concepts that will prevent core damage by LOCA caused elevated temperatures is the primary motivation of this work. Decay heat generated by the fission products in the reactor core following shutdown is sufficiently high to melt the fuel unless energy can be removed from the pins at a sufficiently rapid rate. There are two major reasons that prevent utilization of traditional emergency cooling methods. One is the absence of gravity and the other is the vacuum condition outside the reactor vessel. A concept that overcomes both problems is the Saturated Wick Evaporation Method (SWEM). This method involves placing wicking structures at specific locations in the core to act as energy sinks. One of its properties is the isothermal behaviour of the liquid in the wick. The absorption of energy by the surface at the isothermal temperature will direct the energy into an evaporation process and not in sensible heat addition. The use of this concept enables establishment of isothermal positions within the core. A computer code that evaluates the temperature distribution of the core has been developed and the results show that this design will prevent fuel meltdown. (author)

K2 Mn4 O8 /Reduced Graphene Oxide Nanocomposites for Excellent Lithium Storage and Adsorption of Lead Ions.

Science.gov (United States)

Hao, Shu-Meng; Qu, Jin; Yang, Jing; Gui, Chen-Xi; Wang, Qian-Qian; Li, Qian-Jie; Li, Xiaofeng; Yu, Zhong-Zhen

2016-03-01

Ion diffusion efficiency at the solid-liquid interface is an important factor for energy storage and adsorption from aqueous solution. Although K 2 Mn 4 O 8 (KMO) exhibits efficient ion diffusion and ion-exchange capacities, due to its high interlayer space of 0.70 nm, how to enhance its mass transfer performance is still an issue. Herein, novel layered KMO/reduced graphene oxide (RGO) nanocomposites are fabricated through the anchoring of KMO nanoplates on RGO with a mild solution process. The face-to-face structure facilitates fast transfer of lithium and lead ions; thus leading to excellent lithium storage and lead ion adsorption. The anchoring of KMO on RGO not only increases electrical conductivity of the layered nanocomposites, but also effectively prevents aggregation of KMO nanoplates. The KMO/RGO nanocomposite with an optimal RGO content exhibits a first cycle charge capacity of 739 mA h g -1 , which is much higher than that of KMO (326 mA h g -1 ). After 100 charge-discharge cycles, it still retains a charge capacity of 664 mA h g -1 . For the adsorption of lead ions, the KMO/RGO nanocomposite exhibits a capacity of 341 mg g -1 , which is higher than those of KMO (305 mg g -1 ) and RGO (63 mg g -1 ) alone. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Rock-salt structure lithium deuteride formation in liquid lithium with high-concentrations of deuterium: a first-principles molecular dynamics study

Science.gov (United States)

Chen, Mohan; Abrams, T.; Jaworski, M. A.; Carter, Emily A.

2016-01-01

Because of lithium’s possible use as a first wall material in a fusion reactor, a fundamental understanding of the interactions between liquid lithium (Li) and deuterium (D) is important. We predict structural and dynamical properties of liquid Li samples with high concentrations of D, as derived from first-principles molecular dynamics simulations. Liquid Li samples with four concentrations of inserted D atoms (LiDβ , β =0.25 , 0.50, 0.75, and 1.00) are studied at temperatures ranging from 470 to 1143 K. Densities, diffusivities, pair distribution functions, bond angle distribution functions, geometries, and charge transfer between Li and D atoms are calculated and analyzed. The analysis suggests liquid-solid phase transitions can occur at some concentrations and temperatures, forming rock-salt LiD within liquid Li. We also observe formation of some D2 molecules at high D concentrations.

CaO insulator coatings on a vanadium-base alloy in liquid 2 at.% calcium-lithium

International Nuclear Information System (INIS)

Park, J.H.; Kassner, T.F.

1996-01-01

The electrical resistance of CaO coatings produced on V-4%Cr-4%Ti and V-15%Cr-5%Ti by exposure of the alloy (round bottom samples 6-in. long by 0.25-in. dia.) to liquid lithium that contained 2 at.% dissolved calcium was measured as a function of time at temperatures between 300-464 degrees C. The solute element, calcium in liquid lithium, reacted with the alloy substrate at these temperatures for 17 h to produce a calcium coating ∼7-8 μm thick. The calcium-coated vanadium alloy was oxidized to form a CaO coating. Resistance of the coating layer on V-15Cr-5Ti, measured in-situ in liquid lithium that contained 2 at.% calcium, was 1.0 x 10 10 Ω-cm 2 at 300 degrees C and 400 h, and 0.9 x 10 10 Ω-cm 2 at 464 degrees C and 300 h. Thermal cycling between 300 and 464 degrees C changed the resistance of the coating layer, which followed insulator behavior. Examination of the specimen after cooling to room temperature revealed no cracks in the CaO coating. The coatings were evaluated by optical microscopy, scanning electron microscopy (SEM), electron dispersive spectroscopy (EDS), and X-ray analysis. Adhesion between CaO and vanadium alloys was enhanced as exposure time increased

A open-quotes zero wasteclose quotes coolant management strategy

International Nuclear Information System (INIS)

Kennicott, M.A.

1994-01-01

In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant's Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant's liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study's goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a open-quotes zero wasteclose quotes machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs

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.)

Reaction rates and electrical resistivities of the hydrogen isotopes with, and their solubilities in, liquid lithium

International Nuclear Information System (INIS)

Pulham, R.J.; Adams, P.F.; Hubberstey, P.; Parry, G.; Thunder, A.E.

1976-01-01

The rate of reaction, k, of hydrogen and of deuterium with liquid lithium have been determined up to pressures of 20kNm -2 and at temperatures between 230 and 270 0 C. The reaction is first order with an apparent activation energy of 52.8 and 55.2 kJmol -1 for hydrogen and deuterium, respectively. The deuterium isotope effect, k/sub H/k/sub D/, decreases from 2.95 at 230 to 2.83 at 270 0 C. Tritium is predicted to react even more slowly than deuterium. The freezing point of lithium is depressed by 0.082 and 0.075 0 C, respectively, by dissolved hydride and deuteride giving eutectics at 0.016 mol percent H and 0.012 mol percent D in the metal-salt phase diagrams. The depression and eutectic concentration are expected to be less for tritium. The increase in the resistivity of liquid lithium caused by dissolved hydrogen isotopes is linear and relatively large, 5 x 10 -8 Ωm (mol percent H or D) -1 . The solubility of lithium hydride and deuteride was determined from the marked change in resistivity on saturation. The liquidus of the metal-salt phase diagram rises steeply from the eutectic point to meet the two-immiscible liquid region. Tritium is expected to be less soluble than deuterium. The partial molar enthalpies of solution are 44.2 and 55.0 kJmol -1 for hydrogen and deuterium, respectively. These values are used to calculate the solvation enthalpies of the isotope anions in the metal

Lithium polymer cell assembled by in situ chemical cross-linking of ionic liquid electrolyte with phosphazene-based cross-linking agent

International Nuclear Information System (INIS)

Choi, Ji-Ae; Kang, Yongku; Kim, Dong-Won

2013-01-01

Highlights: ► Ionic liquid-based cross-linked gel polymer electrolytes were synthesized and their electrochemical properties were investigated. ► Lithium polymer cells with in situ cross-linked gel polymer electrolytes exhibited reversible cycling behavior with good capacity retention. ► The use of ionic liquid-based cross-linked gel polymer electrolytes significantly improved the thermal stability of the cells. -- Abstract: Ionic liquid-based cross-linked gel polymer electrolytes were prepared with a phosphazene-based cross-linking agent, and their electrochemical properties were investigated. Lithium polymer cells composed of lithium anode and LiCoO 2 cathode were assembled with ionic liquid-based cross-linked gel polymer electrolyte and their cycling performance was evaluated. The interfacial adhesion between the electrodes and the electrolyte by in situ chemical cross-linking resulted in stable capacity retention of the cell. A reduction in the ionic mobility in both the electrolyte and the electrode adversely affected discharge capacity and high rate performance of the cell. DSC studies demonstrated that the use of ionic liquid-based cross-linked gel polymer electrolytes provided a significant improvement in the thermal stability of the cell

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)

MES lead bismuth forced circulation loop and test results

International Nuclear Information System (INIS)

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

2003-01-01

Liquid lead-bismuth is a promising material as future reactor coolant or intensive neutron source material for accelerator driven system (ADS). Mitsui Engineering and Shipbuilding Co., Ltd. (MES) completed lead-bismuth coolant (LBC) forced circulation loop in May 2001 and acquired engineering data on economizer, electro magnetic pump, electro magnetic flow meter and so on. For quality control of LBC, oxygen sensor and filtering element are developing using some hydrogen and moisture mixed gases. Structural materials corrosion test for accelerator driver system (ADS) will start soon. And thermal hydraulic test for ADS will start in tree years. (author)

Lithium

Science.gov (United States)

Bradley, Dwight C.; Stillings, Lisa L.; Jaskula, Brian W.; Munk, LeeAnn; McCauley, Andrew D.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

2017-12-19

Lithium, the lightest of all metals, is used in air treatment, batteries, ceramics, glass, metallurgy, pharmaceuticals, and polymers. Rechargeable lithium-ion batteries are particularly important in efforts to reduce global warming because they make it possible to power cars and trucks from renewable sources of energy (for example, hydroelectric, solar, or wind) instead of by burning fossil fuels. Today, lithium is extracted from brines that are pumped from beneath arid sedimentary basins and extracted from granitic pegmatite ores. The leading producer of lithium from brine is Chile, and the leading producer of lithium from pegmatites is Australia. Other potential sources of lithium include clays, geothermal brines, oilfield brines, and zeolites. Worldwide resources of lithium are estimated to be more than 39 million metric tons, which is enough to meet projected demand to the year 2100. The United States is not a major producer at present but has significant lithium resources.

Evaluation of organic moderator/coolants for fusion breeder blankets

International Nuclear Information System (INIS)

Romero, J.B.

1980-03-01

Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

First experiments with a liquid-lithium based high-intensity 25-keV neutron source

International Nuclear Information System (INIS)

Paul, M.

2014-01-01

A high-intensity neutron source based on a Liquid-Lithium Target (LiLiT) and the 7 Li(p,n) reaction was developed at SARAF (Soreq Applied Research Accelerator Facility, Israel) and is used for nuclear astrophysics experiments. The setup was commissioned with a 1.3 mA proton beam at 1.91 MeV, producing a neutron yield of ~ 2 ×10 10 n/s, more than one order of magnitude larger than conventional 7 Li(p,n)-based neutron sources and peaked at ~25 keV. The LiLiT device consists of a high-velocity (> 4 m/s) vertical jet of liquid lithium (~200 °C) whose free surface is bombarded by the proton beam. The lithium jet acts both as the neutron-producing target and as a power beam dump. The target dissipates a peak power areal density of 2.5 kW/cm 2 and peak volume density of 0.5 MW/cm 3 with no change of temperature or vacuum regime in the vacuum chamber. Preliminary results of Maxwellian-averaged cross section measurements for stable isotopes of Zr and Ce, performed by activation in the neutron flux of LiLiT, and nuclear-astrophysics experiments in planning will be described. (author)

Thermal hydraulic and power cycle analysis of liquid lithium blanket designs

International Nuclear Information System (INIS)

Misra, B.; Stevens, H.C.; Maroni, V.A.

1977-01-01

Thermal hydraulic and power cycle analyses were performed for the first-wall and blanket systems of tokamak-type fusion reactors under a typical set of design and operating conditions. The analytical results for lithium-cooled blanket cells show that with stainless steel as construction material and with no divertor present, the maximum allowable neutron wall loading is approximately 2 MW/m 2 and is limited by thermal stress criteria. With vanadium alloy as construction material and no divertor present, the maximum allowable neutron wall loading is approximately 8 MW/m 2 and is limited by an interplay of constraints imposed on the maximum allowable structural temperature and the minimum allowable coolant inlet temperature. With a divertor these wall loadings can be increased by from 40 to 90 percent. The cost of the vanadium system is found to be competitive with the stainless steel system because of the higher allowable structural temperatures and concomitant higher thermal efficiencies afforded by the vanadium alloys

Gelled Electrolyte Containing Phosphonium Ionic Liquids for Lithium-Ion Batteries

Directory of Open Access Journals (Sweden)

Mélody Leclère

2018-06-01

Full Text Available In this work, new gelled electrolytes were prepared based on a mixture containing phosphonium ionic liquid (IL composed of trihexyl(tetradecylphosphonium cation combined with bis(trifluoromethanesulfonimide [TFSI] counter anions and lithium salt, confined in a host network made from an epoxy prepolymer and amine hardener. We have demonstrated that the addition of electrolyte plays a key role on the kinetics of polymerization but also on the final properties of epoxy networks, especially thermal, thermo-mechanical, transport, and electrochemical properties. Thus, polymer electrolytes with excellent thermal stability (>300 °C combined with good thermo-mechanical properties have been prepared. In addition, an ionic conductivity of 0.13 Ms·cm−1 at 100 °C was reached. Its electrochemical stability was 3.95 V vs. Li0/Li+ and the assembled cell consisting in Li|LiFePO4 exhibited stable cycle properties even after 30 cycles. These results highlight a promising gelled electrolyte for future lithium ion batteries.

Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

International Nuclear Information System (INIS)

Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

2004-11-01

Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

Health physics aspects of processing EBR-I coolant

International Nuclear Information System (INIS)

Burke, L.L.; Thalgott, J.O.; Poston, J.W. Jr.

1998-01-01

The sodium-potassium reactor coolant removed from the Experimental Breeder Reactor Number One after a partial reactor core meltdown had been stored at the Idaho National Engineering and Environmental Laboratory for 40 years. The State of Idaho considered this waste the most hazardous waste stored in the state and required its processing. The reactor coolant was processed in three phases. The first phase converted the alkali metal into a liquid sodium-potassium hydroxide. The second phase converted this caustic to a liquid sodium-potassium carbonate. The third phase solidified the sodium-potassium carbonate into a form acceptable for land disposal. Health physics aspects and dose received during each phase of the processing are discussed

Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

Energy Technology Data Exchange (ETDEWEB)

Ono, M., E-mail: mono@pppl.gov [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Jaworski, M.A.; Kaita, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Hirooka, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Gray, T.K. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

2017-04-15

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety concerns (Federici et al., 2001) . It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance (Ono et al., 2013, 2014) . The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium (LL) divertor (RLLD) concept (Ono et al., 2013) and its variant, the active liquid lithium divertor concept (ARLLD) (Ono et al., 2014) , taking advantage of the enhanced non-coronal Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/s of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤450 °C than the first wall ∼600–700 °C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ∼1 l/s (l/s) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust/impurities are removed by relatively simple filter and cold/hot trap systems. Using a

Current Status on the Korean Test Blanket Module Development for testing in the ITER

International Nuclear Information System (INIS)

Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Jung, Ki Sok

2010-01-01

Korea has proposed and designed a Helium Cooled Molten Lithium (HCML) Test Blanket Module (TBM) to be tested in the International Thermonuclear Experimental Reactor (ITER). Ferrite Martensitic (FM) steel is used as the structural material and helium (He) is used as a coolant to cool the first wall (FW) and breeding zone. Liquid lithium (Li) is circulated for a tritium breeding, not for a cooling purpose. Main purpose for developing the TBM is to develop the design technology for DEMO and fusion reactor and it should be proved through the experiment in the ITER with TBM. Therefore, we have developed the design scheme and related codes including the safety analysis for obtain the license to be tested in the ITER. In order to develop and install at the ITER, several technologies were developed in parallel; fabrication, breeder, He cooling, tritium extraction and so on. Figure 1 shows the overall TBM development scheme. In Korea, official strategy for developing the TBM is to participate to other parties' concept such as US and EU ones, in which PbLi (lead lithium eutectic), He, and FM steel were used for liquid breeder, coolant, and structural material, respectively

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

Investigation of chloride-release of nuclear grade resin in PWR primary system coolant

International Nuclear Information System (INIS)

Cao Xiaoning; Li Yunde; Li Jinghong; Lin Fangliang

1997-01-01

A new preparation technique is developed for making the low-chloride nuclear-grade resin by commercial resin. The chloride remained in nuclear grade resin may release to PWR primary coolant. The amount of released chloride is depended on the concentration of boron, lithium, other anion impurities, and remained chloride concentration in resin

The impact of lithium wall coatings on NSTX discharges and the engineering of the Lithium Tokamak eXperiment (LTX)

International Nuclear Information System (INIS)

Majeski, R.; Kugel, H.; Kaita, R.; Avasarala, S.; Bell, M.G.; Bell, R.E.; Berzak, L.; Beiersdorfer, P.; Gerhardt, S.P.; Gransted, E.; Gray, T.; Jacobson, C.; Kallman, J.; Kaye, S.; Kozub, T.; LeBlanc, B.P.; Lepson, J.; Lundberg, D.P.; Maingi, R.; Mansfield, D.; Paul, S.F.; Pereverzev, G.V.; Schneider, H.; Soukhanovskii, V.; Strickler, T.; Stotler, D.; Timberlake, J.; Zakharov, L.E.

2010-01-01

Recent experiments on the National Spherical Torus eXperiment (NSTX) have shown the benefits of solid lithium coatings on carbon PFC's to diverted plasma performance, in both L- and H-mode confinement regimes. Better particle control, with decreased inductive flux consumption, and increased electron temperature, ion temperature, energy confinement time, and DD neutron rate were observed. Successive increases in lithium coverage resulted in the complete suppression of ELM activity in H-mode discharges. A liquid lithium divertor (LLD), which will employ the porous molybdenum surface developed for the LTX shell, is being installed on NSTX for the 2010 run period, and will provide comparisons between liquid walls in the Lithium Tokamak eXperiment (LTX) and liquid divertor targets in NSTX. LTX, which recently began operations at the Princeton Plasma Physics Laboratory, is the world's first confinement experiment with full liquid metal plasma-facing components (PFCs). All materials and construction techniques in LTX are compatible with liquid lithium. LTX employs an inner, heated, stainless steel-faced liner or shell, which will be lithium-coated. In order to ensure that lithium adheres to the shell, it is designed to operate at up to 500-600 degrees C to promote wetting of the stainless by the lithium, providing the first hot wall in a tokamak to Operate at reactor-relevant temperatures. The engineering of LTX will be discussed.

Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren,R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov L. E.

2008-09-26

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW~1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with longpulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Spectroscopic diagnostics for liquid lithium divertor studies on National Spherical Torus Experiment

International Nuclear Information System (INIS)

Soukhanovskii, V. A.; Roquemore, A. L.; Bell, R. E.; Kaita, R.; Kugel, H. W.

2010-01-01

The use of lithium-coated plasma facing components for plasma density control is studied in the National Spherical Torus Experiment (NSTX). A recently installed liquid lithium divertor (LLD) module has a porous molybdenum surface, separated by a stainless steel liner from a heated copper substrate. Lithium is deposited on the LLD from two evaporators. Two new spectroscopic diagnostics are installed to study the plasma surface interactions on the LLD: (1) A 20-element absolute extreme ultraviolet (AXUV) diode array with a 6 nm bandpass filter centered at 121.6 nm (the Lyman-α transition) for spatially resolved divertor recycling rate measurements in the highly reflective LLD environment, and (2) an ultraviolet-visible-near infrared R=0.67 m imaging Czerny-Turner spectrometer for spatially resolved divertor D I, Li I-II, C I-IV, Mo I, D 2 , LiD, CD emission and ion temperature on and around the LLD module. The use of photometrically calibrated measurements together with atomic physics factors enables studies of recycling and impurity particle fluxes as functions of LLD temperature, ion flux, and divertor geometry.

Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

International Nuclear Information System (INIS)

Kugel, W.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren, R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov, L. E.

2008-01-01

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW∼1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization

A {open_quotes}zero waste{close_quotes} coolant management strategy

Energy Technology Data Exchange (ETDEWEB)

Kennicott, M.A.

1994-04-01

In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant`s Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant`s liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study`s goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a {open_quotes}zero waste{close_quotes} machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs.

Natural convection in enclosures containing lead-bismuth and lead

International Nuclear Information System (INIS)

Dzodzo, M.; Cuckovic-Dzodzo, D.

2001-01-01

The design of liquid metal reactors such as Encapsulated Nuclear Heat Source (ENHS) which are based predominantly on the flow generated by natural convection effects demands knowledge of velocity and temperature fields, distribution of the local Nusselt numbers and values of the average Nusselt numbers for small coolant velocity regimes. Laminar natural convection in rectangular enclosures with different aspect ratios, containing lead-bismuth and lead is studied numerically in this paper. The numerical model takes into account variable properties of the liquid metals. The developed correlation for average Nusselt numbers is presented. It is concluded that average Nusselt numbers are lower than in 'normal' fluids (air, water and glycerol) for the same values of Rayleigh numbers. However, the heat flux, which can be achieved, is greater due to the high thermal conductivity of liquid metals. Some specific features of the flow fields generated by natural convection in liquid metals are presented. Their consequences on the design of heat exchangers for liquid metals are discussed. An application of the obtained results to the design of a new type of steam generator, which integrates the intermediate heat exchanger and secondary pool functions of the ENHS reactor, is presented. (authors)

Numerical analysis of high-speed liquid lithium free-surface flow

International Nuclear Information System (INIS)

Gordeev, Sergej; Heinzel, Volker; Stieglitz, Robert

2012-01-01

Highlights: ► The free surface behavior of a high speed lithium jet is investigated by means of a CFD LES analysis. ► The study is aiming to validate adequate LES technique. ► The Osaka University experiments with liquid lithium jet have been simulated. ► Four cases with jet flow velocities of 4, 9, 13 and 15 m/s are analyzed. ► Calculation results show a good qualitative and a quantitative agreement with the experimental data. - Abstract: The free-surface stability of the target of the International Fusion Material Irradiation Facility (IFMIF) is one of the crucial issues, since the spatio-temporal behavior of the free-surface determines the neutron flux to be generated. This article investigates the relation between the evolution of a wall boundary layer in a convergent nozzle and the free surface shape of a high speed lithium jet by means of a CFD LES analysis using the Osaka University experiments. The study is aiming to validate adequate LES technique to analyze the individual flow phenomena observed. Four cases with jet flow velocities of 4, 9, 13 and 15 m/s are analyzed. First analyses of calculation results show that the simulation exhibits a good qualitative and a quantitative agreement with the experimental data, which allows in the future a more realistic prediction of the IFMIF target behavior.

Influence of coolant motion on structure of hardened steel element

Directory of Open Access Journals (Sweden)

A. Kulawik

2008-08-01

Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

Examination results on reaction of lithium

International Nuclear Information System (INIS)

Asada, Takashi

2000-12-01

Before the material corrosion tests in lithium, the reactions of lithium with air and ammonia that will be used for lithium cleaning were examined, and the results were as follows. 1. When lithium put into air, surface of lithium changes to black first but soon to white, and the white layer becomes gradually thick. The first black of lithium surface is nitride (Li 3 N) and it changes to white lithium hydroxide (LiOH) by reaction with water in air, and it grows. The growth rate of the lithium hydroxide is about 1/10 in the desiccator (humidity of about 10%) compare with in air. 2. When lithium put into nitrogen, surface of lithium changes to black, and soon changes to brown and cracks at surface. At the same time with this cracking, weight of lithium piece increases and nitridation progresses respectively rapidly. This nitridation completed during 1-2 days on lithium rod of 10 mm in diameter, and increase in weight stopped. 3. Lithium melts in liquid ammonia and its melting rate is about 2-3 hour to lithium of 1 g. The liquid ammonia after lithium melting showed dark brown. (author)

The current status of fusion reactor blanket thermodynamics

International Nuclear Information System (INIS)

Veleckis, E.; Yonco, R.M.; Maroni, V.A.

1980-01-01

The available thermodynamic information is reviewed for three categories of materials that meet essential criteria for use as breeding blankets in D-T fuelled fusion reactors: liquid lithium, solid lithium alloys, and lithium-containing ceramics. The leading candidate, liquid lithium, which also has potential for use as a coolant, has been studied more extensively than have the solid alloys or ceramics. Recent studies of liquid lithium have concentrated on its sorption characteristics for hydrogen isotopes and its interaction with common impurity elements. Hydrogen isotope sorption data (P-C-T relations, activity coefficients, Sieverts' constants, plateau pressures, isotope effects, free energies of formation, phase boundaries, etc.) are presented in a tabular form that can be conveniently used to extract thermodynamic information for the α-phases of the Li-LiH, Li-LiD and Li-LiT systems and to construct complete phase diagrams. Recent solubility data for Li 3 N, Li 2 O, and Li 2 C 2 in liquid lithium are discussed with emphasis on the prospects for removing these species by cold-trapping methods. Current studies on the sorption of hydrogen in solid lithium alloys (e.g. Li-Al and Li-Pb), made using a new technique (the hydrogen titration method), have shown that these alloys should lead to smaller blanket-tritium inventories than are attainable with liquid lithium and that the P-C-T relationships for hydrogen in Li-M alloys can be estimated from lithium activity data for these alloys. There is essentially no refined thermodynamic information on the prospective ceramic blanket materials. The kinetics of tritium release from these materials is briefly discussed. Research areas are pointed out where additional thermodynamic information is needed for all three material categories. (author)

Liquid-metal pin-fin pressure drop by correlation in cross flow

International Nuclear Information System (INIS)

Wang, Zhibi; Kuzay, T.M.; Assoufid, L.

1994-01-01

The pin-fin configuration is widely used as a heat transfer enhancement method in high-heat-flux applications. Recently, the pin-fin design with liquid-metal coolant was also applied to synchrotron-radiation beamline devices. This paper investigates the pressure drop in a pin-post design beamline mirror with liquid gallium as the coolant. Because the pin-post configuration is a relatively new concept, information in literature about pin-post mirrors or crystals is rare, and information about the pressure drop in pin-post mirrors with liquid metal as the coolant is even more sparse. Due to this the authors considered the cross flow in cylinder-array geometry, which is very similar to that of the pin-post, to examine the pressure drop correlation with liquid metals over pin fins. The cross flow of fluid with various fluid characteristics or properties through a tube bank was studied so that the results can be scaled to the pin-fin geometry with liquid metal as the coolant. Study lead to two major variables to influence the pressure drop: fluid properties, viscosity and density, and the relative length of the posts. Correlation of the pressure drop between long and short posts and the prediction of the pressure drop of liquid metal in the pin-post mirror and comparison with an existing experiment are addressed

Extraction of lanthanide elements and bismuth in molten lithium chloride-liquid bismuth-lithium alloy system

International Nuclear Information System (INIS)

Harada, Makoto; Adachi, Motonari; Kai, Yuichi; Koike, Kenichi

1987-01-01

The equilibrium distributions of neodymium and samarium between molten LiCl and liquid Bi-Li alloy were measured in a wide range of Li-mole fraction in the alloy phase, X Li . These lanthanide elements were extracted through redox reactions. In high X Li range, X Li > 0.03, the distributions of neodymium and bismuth in the salt phase increased markedly. The anomalous increase is attributed to the formation of the compound comprized of Nd, Li, Bi and oxygen in the salt phase. The reaction processes in samarium and neodymium were very fast and the extraction rates are controlled by the diffusion processes of the solutes and metallic lithium. (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.)

A circulating loop tester for liquid alloyed metal of lead-bismuth

International Nuclear Information System (INIS)

Kitano, Teruaki; Ono, Mikinori; Kamata, Kinya

2002-01-01

Mitsui Engineering and Shipbuilding Co., Ltd. (MES) had focused to merits of this lead-bismuth alloy, to actively carry out many works on this field such as an experience of development of heat exchanger at industrial level of intercourse with IPPE (Institute of Physics and Power Engineering) in Russia with an experience of using results for 80 years on coolant for nuclear reactor. Before about 20 years, MES developed a heat exchanger for installation at a lead-zinc separation process in a refinery in Japan under cooperation of the Mitsui Metal and Mine Co., Ltd., to deliver it for a power generation system at the Hachinohe refinery. As the heat exchanger aims at control of cooling in the separation process, it also contributes to power generation of about 4,300 kW, and now it continues to separate and contribute to self-power generation in the refinery. The heat exchanger is filled with the liquid alloyed metal of lead-bismuth for an intermediate thermal medium in its casing. The metal has some merits such as inactivity to air and water, high boiling point (1,700 centigrade), almost no volume change at its coagulation, and its minus reactivity coefficient. However, the metal has some problems to be solved, such as its steel corrosion, its purification, and control technology. To grow up lead-bismuth technology to a nuclear energy technology in Japan, the lead-bismuth circulating loop tester was produced on May, 2001, to establish application technology on this system to nuclear energy technology in Japan. (G.K.)

Properties and Structure of the LiCl-films on Lithium Anodes in Liquid Cathodes

DEFF Research Database (Denmark)

Mogensen, Mogens Bjerg; Hennesø, Erik

2016-01-01

Lithium anodes passivated by LiCl layers in different types of liquid cathodes (catholytes) based on LiAlCl4 in SOCl2 or SO2 have been studied by means of impedance spectroscopy. The impedance spectra have been fitted with two equivalent circuits using a nonlinear least squares fit program...

Improved Cyclability of Liquid Electrolyte Lithium/Sulfur Batteries by Optimizing Electrolyte/Sulfur Ratio

Directory of Open Access Journals (Sweden)

Sheng S. Zhang

2012-12-01

Full Text Available A liquid electrolyte lithium/sulfur (Li/S cell is a liquid electrochemical system. In discharge, sulfur is first reduced to highly soluble Li2S8, which dissolves into the organic electrolyte and serves as the liquid cathode. In solution, lithium polysulfide (PS undergoes a series of complicated disproportionations, whose chemical equilibriums vary with the PS concentration and affect the cell’s performance. Since the PS concentration relates to a certain electrolyte/sulfur (E/S ratio, there is an optimized E/S ratio for the cyclability of each Li/S cell system. In this work, we study the optimized E/S ratio by measuring the cycling performance of Li/S cells, and propose an empirical method for determination of the optimized E/S ratio. By employing an electrolyte of 0.25 m LiSO3CF3-0.25 m LiNO3 dissolved in a 1:1 (wt:wt mixture of dimethyl ether (DME and 1,3-dioxolane (DOL in an optimized E/S ratio, we show that the Li/S cell with a cathode containing 72% sulfur and 2 mg cm−2 sulfur loading is able to retain a specific capacity of 780 mAh g−1 after 100 cycles at 0.5 mA cm−2 between 1.7 V and 2.8 V.

Sound velocity in the coolant of boiling nuclear reactors

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

2009-01-01

To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

Some observations on simulated molten debris-coolant layer dynamics

International Nuclear Information System (INIS)

Greene, G.A.; Klein, J.; Klages, J.; Schwarz, E.; Sanborn, Y.

1983-04-01

Experiments are being performed to investigate high temperature liquid-liquid film boiling between a pool of liquid metal and an overlying coolant pool of R-11 or water. Film boiling has been observed to be stable for R-11; however, considerable liquid-liquid contact has been observed with water well beyond the minimum film boiling temperature. Unstable liquid-liquid film boiling of water has been observed to escalate into dispersive, non-energetic vapor explosions when the interface contact temperature exceeded the spontaneous nucleation temperature. Other parametric trends in the data are discussed

Design of the FMIT lithium target

International Nuclear Information System (INIS)

Hassberger, J.A.; Annese, C.E.; Greenwell, R.K.; Ingham, J.G.; Miles, R.R.; Miller, W.C.

1981-01-01

Development of the liquid lithium target for the Fusion Materials Irradiation Test (FMIT) Facility is described. The target concept, major design goals and design requirements are presented. Progress made in the research and development areas leading to detailed design of the target is discussed. This progress, including experimental and analytic results, demonstrates that the FMIT target design is capable of meeting its major design goals and requirements

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

International Nuclear Information System (INIS)

Kugel, H.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.P.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.K.; Menard, J.; Stotler, D.; Zakharov, L.E.; Maingi, Rajesh; Nygren, R.E.; Soukhanovskii, V.; Wakeland, P.

2009-01-01

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15 25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW 1), to enable ne scan capability (2) in the H-mode, to test the ability to operate at significantly lower density (e.g., ne/nGW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Rapid and simple pretreatment of human body fluids using electromembrane extraction across supported liquid membrane for capillary electrophoretic determination of lithium.

Science.gov (United States)

Strieglerová, Lenka; Kubáň, Pavel; Boček, Petr

2011-05-01

Electromembrane extraction was used for simultaneous sample cleanup and preconcentration of lithium from untreated human body fluids. The sample of a body fluid was diluted 100 times with 0.5 mM Tris solution and lithium was extracted by electromigration through a supported liquid membrane composed of 1-octanol into 100 mM acetic acid acceptor solution. Matrix compounds, such as proteins, red blood cells, and other high-molecular-weight compounds were efficiently retained on the supported liquid membrane. The liquid membrane was anchored in pores of a short segment of a polypropylene hollow fiber, which represented a low cost, single use, disposable extraction unit and was discarded after each use. Acceptor solutions were analyzed using capillary electrophoresis with capacitively coupled contactless conductivity detection (CE-C(4) D) and baseline separation of lithium was achieved in a background electrolyte solution consisting of 18 mM L-histidine and 40 mM acetic acid at pH 4.6. Repeatability of the electromembrane extraction-CE-C(4) D method was evaluated for the determination of lithium in standard solutions and real samples and was better than 0.6 and 8.2% for migration times and peak areas, respectively. The concentration limit of detection of 9 nM was achieved. The developed method was applied to the determination of lithium in urine, blood serum, blood plasma, and whole blood at both endogenous and therapeutic concentration levels. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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.)

Design, calculation and experimental studies for liquid metal system main parameters in support of the liquid lithium fusion reactor

International Nuclear Information System (INIS)

Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.

1999-01-01

A new concept of a Liquid Lithium Fusion Reactor and the first experimental results were presented at the 16th IAEA Conference on Fusion Energy. During the past two years theoretical estimations have been made, and calculated and experimental results have been obtained in confirmation of this concept and supporting its progress. The main results of this work are given in the paper. (author)

Design, calculation and experimental studies for liquid metal system main parameters in support of the liquid lithium fusion reactor

International Nuclear Information System (INIS)

Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.

2001-01-01

A new concept of a Liquid Lithium Fusion Reactor and the first experimental results were presented at the 16th IAEA Conference on Fusion Energy. During the past two years theoretical estimations have been made, and calculated and experimental results have been obtained in confirmation of this concept and supporting its progress. The main results of this work are given in the paper. (author)

Module of lithium divertor for KTM tokamak

Energy Technology Data Exchange (ETDEWEB)

Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

2012-10-15

Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

Feasibility study on the type of KALIMER coolant circulation pump

International Nuclear Information System (INIS)

Nam, H. Y.; Kim, Y. K.; Lee, Y. B.; Hwang, J. S.; Choi, S. K.

1997-07-01

The characteristics of mechanical pump and electromagnetic (EM) pump for liquid sodium coolant in a liquid metal reactor are compared and analysed as a design concept of KALIMER coolant pumps. The type of coolant circulation pump affects the selection of reactor type, economics, and reliability of reactor. Though the mechanical pump has much application experience and give satisfaction to the reliability of developed reactor type, the possibility of development is limited and its large weight and volume have a negative effect on the design of the economical liquid metal reactor. The large scale electromagnetic pump has not been verified yet, but it is expected to be demonstrated in time. Because the size of EM pump is small relative to the mechanical pump, the compact reactor design is possible. Therefore the selection of EM pump can be one of the methods to improve the economics. Since the shape of EM pump can be varied according to the arrangement of electromagnet coils, a new or unique reactor type can be developed easily in the process of KALIMER development. In the view point of economic LMR development, it is desirable to adopt the electromagnetic pump. (author). 50 refs., 11 tabs., 24 figs

Progress of liquid metal technology and application in energy industries

International Nuclear Information System (INIS)

Miyazaki, Keiji; Kamei, Mitsuru; Nei, Hiromichi.

1990-01-01

Liquid metals are excellent energy transport media, and recently remarkable development has been observed in the technology of handling sodium and the machinery and equipment. In nuclear fusion, the development of the use of lithium as the coolant is advanced. For space technology, attention has been paid from the early stage to various liquid metals. For general industries, liquid metals have been used for high temperature heat pipes and the utilization of solar heat, and mercury vapor turbines were manufactured for trial. Besides, attention is paid anew to liquid metal MHD electric power generation. The development of the NaS batteries for electric cars and electric power storage and the interchange of liquid metal technology with the fields of iron and steel, metallurgy and so on advance. It is expected that liquid metal technology bears future advanced energy engineering while deepening the interchange with other advanced fields also in order to reactivate atomic energy technology. Liquid metals have the features of high electric and thermal conductivities, chemical activity and opaque property as metals, and fluidity and relatively high boiling point and melting point as liquids. FBRs, fusion reactors and the power sources for space use are described. (K.I.)

Fragmentation of suddenly heated liquids in ICF reactors. Revision 1

International Nuclear Information System (INIS)

Blink, J.A.; Hoover, W.G.

1985-01-01

Fragmentation of free liquids in Inertial Confinement Fusion reactors could determine the upper bound on reactor pulse rate because increased surface area will enhance the cooling and condensation of coolant ablated by the fusion x rays. Relaxation from the suddenly (neutron) heated state will move a liquid into the negative pressure region under the liquid-vapor P-V dome. The resulting expansion in a diverging geometry will hydrodynamically force the liquid to fragment, with vapor then forming from the new surfaces to fill the cavities. An energy minimization model is used to determine the fragment size that produces the least amount of non-fragment-center-of-mass energy; i.e., the sum of the surface and dilational kinetic energies. This model predicts fragmentation dependence on original system size and amount of isochoric heating as well as liquid density, Grueneisen parameter, surface tension, and sound speed. A two dimensional molecular dynamics code was developed to test the model at a microscopic scale for the Lennard-Jones fluid with its two adjustable constants chosen to represent lithium

Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal-hydraulics and technologies

International Nuclear Information System (INIS)

2007-01-01

As part of the development of advanced nuclear systems, including accelerator-driven systems (ADS) proposed for high-level radioactive waste transmutation and generation IV reactors, heavy liquid metals such as lead (Pb) or lead-bismuth eutectic (LBE) are under evaluation as reactor core coolant and ADS neutron target material. Heavy liquid metals are also being envisaged as target materials for high-power neutron spallation sources. The objective of this handbook is to collate and publish properties and experimental results on Pb and LBE in a consistent format in order to provide designers with a single source of qualified properties and data and to guide subsequent development efforts. The handbook covers liquid Pb and LBE properties, materials compatibility and testing issues, key aspects of the thermal-hydraulics and system technologies, existing test facilities, open issues and perspectives. (author)

Liquid metal cooled nuclear reactor

International Nuclear Information System (INIS)

Leigh, K.M.

1980-01-01

A liquid metal cooled nuclear reactor is described, wherein coolant is arranged to be flowed upwardly through a fuel assembly and having one or more baffles located above the coolant exit of the fuel assembly, the baffles being arranged so as to convert the upwardly directed motion of liquid metal coolant leaving the fuel assembly into a substantially horizontal motion. (author)

Diagnostics of high-speed liquid lithium jet for IFMIF/EVEDA lithium test loop

International Nuclear Information System (INIS)

Kanemura, Takuji; Kondo, Hiroo; Furukawa, Tomohiro; Sugiura, Hirokazu; Horiike, Hiroshi; Yamaoka, Nobuo; Ida, Mizuho; Nakamura, Kazuyuki; Matsushita, Izuru

2011-01-01

Regarding R and Ds on the International Fusion Materials Irradiation Facility (IFMIF), hydraulic stability of the liquid Li jet simulating the IFMIF Li target is planned to be validated using EVEDA Li Test Loop (ELTL). IFMIF is an accelerator-based deuteron-lithium (Li) neutron source for research and development of fusion reactor materials. The stable Li target is required in IFMIF to maintain the quality of the neutron fluence and integrity of the Li target itself. This paper presents diagnostics of the Li jet to be implemented in validation tests of the jet stability in ELTL, and those specifications and methodologies are introduced. In the tests, the following physical parameters need to be measured; thickness of the jet; surface structure (height, length/width and frequency of free-surface waves); local flow velocity at the free surface; and Li evaporation rate. With regard to measurement of jet thickness and the surface wave height, a contact-type liquid level sensor is to be used. As for measurement of wave velocity and visual understanding of detailed free-surface structure, a high-speed video camera is to be leveraged. With respect to Li evaporation measurement, weight change of specimens installed near the free surface and frequency change of a crystal quartz are utilized. (author)

Recent experimental results on solutions of deuterium in lithium

International Nuclear Information System (INIS)

Ihle, H.R.; Wu, C.H.

1976-01-01

The existence of a number of stable molecules containing lithium and hydrogen isotopes in the saturated vapor over dilute solutions of hydrogen isotopes in lithium causes an unexpectedly high density of hydrogen isotopes in the vapor at high temperature. An evaluation of the partial pressures of the gas species Li, Li 2 , LiD, Li 2 D, LiD 2 and D 2 over solutions of deuterium in lithium measured in the temperature range 770 to 970 0 K, and extrapolation to higher temperatures, leads to the conclusion that the ratio of the atom fraction of deuterium in the gas to its atom fraction in the liquid exceeds unity above approximately 1240 0 K; this ratio is independent of the deuterium atom fraction in the liquid at low concentrations. Therefore the thermodynamic supposition that hydrogen isotopes can be separated from lithium by fractional distillation even at extremely low concentration exists. A direct verification of this phenomenon was made by Rayleigh distillation of Li-D solutions in the temperature range 970 to 1600 0 K. These measurements yield also the ratio of the deuterium atom fraction in the gas to that in the liquid and are in good agreement with the data obtained by extrapolation of partial pressures. The enrichment and depletion of deuterium in dependence on the number of theoretical plates of a distillation column at total reflux is calculated using the results

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.)

Overhauser effect in metallic lithium; Effet Overhauser dans le lithium metallique

Energy Technology Data Exchange (ETDEWEB)

Gueron, J.; Ryter, Ch. [Commissariat a l' energie atomique et aux energies alternatives - CEA, Centre d' Etudes Nucleaires de Saclay (France)

1960-07-01

The Overhauser effect has been observed: a) at ordinary temperatures, by measuring the increase in the nuclear resonance signal of Li{sup 7}; b) at the temperature of liquid helium, by observing the electron resonance shift due to the secular part of the electron-nucleus coupling. The metallic lithium particles are produced by irradiating lithium hydride with thermal neutrons. Reprint of a paper published in Physical Review Letters, vol. 3, no. 7, 1959, p. 338-340 [French] L'effet Overhauser est mis en evidence: a) a la temperature ordinaire, en mesurant l'augmentation du signal de resonance nucleaire du Li{sup 7}; b) a la temperature de l'helium liquide, en observant le deplacement de la raie de resonance electronique du a la partie seculaire du couplage electron-noyau. Les particules de lithium metallique sont produites par irradiations aux neutrons thermiques de l'hydrure de lithium Li{sup 7}. Reproduction d'un article publie dans Physical Review Letters, vol. 3, no. 7, 1959, p. 338-340.

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

Development of liquid type TBM technology for ITER

International Nuclear Information System (INIS)

Lee, Dong Won; Kim, S. K.; Yoon, J. S.

2012-03-01

The final objectives of this project are as follows; Development of the key techniques for the liquid type TBM for ITER: Developing plan for leading and participating liquid TBM concepts; Estimating cost and schedule according to development schedule and managing technologies; Developing integrated design system and completing the engineering design for liquid TBM; Developing the key technologies for the liquid TBM; Construction of performance test systems for liquid TBM and verification of the performance. We are technically surveying the ITER system design data, the insufficient part of ITER design, and required R and D items and so on. In Korea, HCML TBM, liquid type breeder with lithium or lead lithium, has been studied during the past years to develop a tritium breeding technology for tritium self-sufficiency of nuclear fusion reactor and the TBM was proposed to be tested in ITER. In this study, we can obtain the key technology of nuclear fusion reactor especially on the TBM design, analysis and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea

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.

Vapor-liquid equilibria of the water + 1,3-propanediol and water + 1,3-propanediol + lithium bromide systems

Energy Technology Data Exchange (ETDEWEB)

Mun, S Y; Lee, H

1999-12-01

Vapor-liquid equilibrium data of the water + 1,3-propanediol and water + 1,3-propanediol + lithium bromide systems were measured at 60, 160, 300, and 760 mmHg at temperatures ranging from 315 to 488 K. The apparatus used in this work is a modified still especially designed for the measurement of low-pressure VLE, in which both liquid and vapor are continuously recirculated. For the analysis of salt-containing solutions, a method incorporating refractometry and gravimetry was used. From the experimental measurements, the effect of lithium bromide on the VLE behavior of water + 1,3-propanediol was investigated. The experimental data of the salt-free system were successfully correlated using the Wilson, NRTL, and UNIQUAC models. In addition, the extended UNIQUAC model of Sander et al. was applied to the VLE calculation of salt-containing mixtures.

Heteroaromatic-based electrolytes for lithium and lithium-ion batteries

Science.gov (United States)

Cheng, Gang; Abraham, Daniel P.

2017-04-18

The present invention provides an electrolyte for lithium and/or lithium-ion batteries comprising a lithium salt in a liquid carrier comprising heteroaromatic compound including a five-membered or six-membered heteroaromatic ring moiety selected from the group consisting of a furan, a pyrazine, a triazine, a pyrrole, and a thiophene, the heteroaromatic ring moiety bearing least one carboxylic ester or carboxylic anhydride substituent bound to at least one carbon atom of the heteroaromatic ring. Preferred heteroaromatic ring moieties include pyridine compounds, pyrazine compounds, pyrrole compounds, furan compounds, and thiophene compounds.

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)

Scientific opportunities at SARAF with a liquid lithium jet target neutron source

Science.gov (United States)

Silverman, Ido; Arenshtam, Alex; Berkovits, Dan; Eliyahu, Ilan; Gavish, Inbal; Grin, Asher; Halfon, Shlomi; Hass, Michael; Hirsh, T. Y.; Kaizer, Boaz; Kijel, Daniel; Kreisel, Arik; Mardor, Israel; Mishnayot, Yonatan; Palchan, Tala; Perry, Amichay; Paul, Michael; Ron, Guy; Shimel, Guy; Shor, Asher; Tamim, Noam; Tessler, Moshe; Vaintraub, Sergey; Weissman, Leo

2018-05-01

SARAF (Soreq Applied Research Accelerator Facility) is based on a 5 mA, 40 MeV, proton/deuteron accelerator. Phase-I, operational since 2010, provides proton and deuteron beams up to 4 and 5 MeV, respectively, for basic and applied research activities. The high power Liquid-Lithium jet Target (LiLiT), with 1.912 MeV proton beam, provides high flux quasi-Maxwellian neutrons at kT 30 keV (about 2 × 1010 n/s/cm2/mA on the irradiated sample, about 1 cm from the target), enabling studies of s-process reactions relevant to nucleo-synthesis of the heavy elements in giant AGB stars. With higher energy proton beams and with deuterons, LiLiT can provide higher fluxes of high energy neutrons up to 20 MeV. The experimental program with SARAF phase-I will be enhanced shortly with a new target room complex which is under construction. Finally, SARAF phase-II, planned to start operation at 2023, will enable full capabilities with proton/ deuteron beams at 5 mA and 40 MeV. Liquid lithium targets will then be used to produce neutron sources with intensities of 1015 n/s, which after thermalization will provide thermal neutron (25 meV) fluxes of about 1012 n/s/cm2 at the entrance to neutron beam lines to diffraction and radiography stations.

Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

Science.gov (United States)

Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

2017-10-01

The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

Corrosion behavior of materials selected for FMIT lithium system

International Nuclear Information System (INIS)

Bazinet, G.D.; Down, M.G.; Matlock, D.K.

1983-01-01

The corrosion program consisted of a multi-disciplinary approach utilizing the liquid lithium test resources and capabilities of several laboratories. Specific concerns associated with the overall objective of materials corrosion behavior were evaluated at each laboratory. Testing conditions included: approx. 3700 hours of exposure to flowing lithium at temperatures from 230 0 C to 270 0 C and approx. 6500 hours of exposure to flowing lithium at an isothermal temperature of 270 0 C. Principal areas of investigation, to be discussed here briefly, included lithium corrosion effects on the following: (1) types 304 and 304L austenitic stainless steels, which are specified as reference materials for the FMIT lithium system; (2) type 304 stainless steel weldments (w/type 308 stainless steel filler) typical of specified tube and butt welds in the lithium system design; (3) titanium, zirconium and yttrium, which represent potential hot trap getter materials; (4) BNi4 braze alloy, used as a potential attachement method in the plug/seat fabrication of liquid lithium valves; and (5) type 321 stainless steel bellows, typical of bellows used in potential liquid lithium valve designs

Compositional depth profiles of the type 316 stainless steel undergone the corrosion in liquid lithium using laser-induced breakdown spectroscopy

Science.gov (United States)

Li, Ying; Ke, Chuan; Liu, Xiang; Gou, Fujun; Duan, Xuru; Zhao, Yong

2017-12-01

Liquid metal lithium cause severe corrosion on the surface of metal structure material that used in the blanket and first wall of fusion device. Fast and accurate compositional depth profile measurement for the boundary layer of the corroded specimen will reveal the clues for the understanding and evaluation of the liquid lithium corrosion process as well as the involved corrosion mechanism. In this work, the feasibility of laser-induced breakdown spectroscopy for the compositional depth profile analysis of type 316 stainless steel which was corroded by liquid lithium in certain conditions was demonstrated. High sensitivity of LIBS was revealed especially for the corrosion medium Li in addition to the matrix elements of Fe, Cr, Ni and Mn by the spectral analysis of the plasma emission. Compositional depth profile analysis for the concerned elements which related to corrosion was carried out on the surface of the corroded specimen. Based on the verified local thermodynamic equilibrium shot-by-shot along the depth profile, the matrix effect was evaluated as negligible by the extracted physical parameter of the plasmas generated by each laser pulse in the longitudinal depth profile. In addition, the emission line intensity ratios were introduced to further reduce the impact on the emission line intensity variations arise from the strong inhomogeneities on the corroded surface. Compositional depth profiles for the matrix elements of Fe, Cr, Ni, Mn and the corrosion medium Li were constructed with their measured relative emission line intensities. The distribution and correlations of the concerned elements in depth profile may indicate the clues to the complicated process of composition diffusion and mass transfer. The results obtained demonstrate the potentiality of LIBS as an effective technique to perform spectrochemical measurement in the research fields of liquid metal lithium corrosion.

Method of producing spherical lithium aluminate particles

International Nuclear Information System (INIS)

Yang, L.; Medico, R.R.; Baugh, W.A.

1983-01-01

Spherical particles of lithium aluminate are formed by initially producing aluminium hydroxide spheroids, and immersing the spheroids in a lithium ion-containing solution to infuse lithium ions into the spheroids. The lithium-infused spheroids are rinsed to remove excess lithium ion from the surface, and the rinsed spheroids are soaked for a period of time in a liquid medium, dried and sintered to form lithium aluminate spherical particles. (author)

Lithium bis(fluorosulfonyl)imide-PYR14TFSI ionic liquid electrolyte compatible with graphite

Czech Academy of Sciences Publication Activity Database

Nádherná, Martina; Reiter, Jakub; Moškon, J.; Dominko, R.

2011-01-01

Roč. 196, č. 18 (2011), s. 7700-7706 ISSN 0378-7753 R&D Projects: GA AV ČR(CZ) KJB200320801; GA AV ČR KJB200320901; GA MŠk(CZ) LC523 Institutional research plan: CEZ:AV0Z40320502 Keywords : Graphite * Ionic liquid * Bis(fluorosulfonyl)imide * Lithium -ion battery * Solid electrolyte interface Subject RIV: CA - Inorganic Chemistry OBOR OECD: Inorganic and nuclear chemistry Impact factor: 4.951, year: 2011

Design and safety aspect of lead and lead-bismuth cooled long-life small safe fast reactors for various core configurations

International Nuclear Information System (INIS)

Zaki, S.; Sekimoto, Hiroshi

1995-01-01

Design and safety aspects of long-life small safe fast reactors using liquid lead or lead-bismuth coolant with metallic or nitride fuel are discussed. Neutronic analyses are performed to investigate the effect of core height to diameter ratio (H/D) on design performance of the proposed reactors. All reactors are subjected to the constraint of 12 years operation without refueling and shuffling with constant 150 MWt reactor power and also to the requirement of maximum excess reactivity during burnup to be less than 0.1%Δk. The results show that the pancake design with H/D of ∼2/3 gives the most negative coolant void coefficient under the requirements for excess reactivity. Modified designs with the central region axially fulfilled with fertile material are proposed to improve the coolant void coefficient. Thermal-hydraulic analysis results show the possibility to operate the reactors up to the end of life without changing their orifice pattern, necessary pumping power for the proposed design smaller than the conventional large sodium cooled FBR, and the natural circulation contribution of 25-40% at the normal operating condition. The reactivity feedback coefficients are also estimated and appeared to be negative for all the components including the coolant density coefficient. (author)

Conceptual design of module fast reactor of ultimate safety cooled by lead-bismuth alloy

International Nuclear Information System (INIS)

Myasnikov, V.O.; Stekolnikov, V.V.; Stepanov, V.S.; Gorshkov, V.T.; Kulikov, M.L.; Shulyndin, V.A.; Gromov, B.F.; Kalashnikov, A.G.; Pashkin, Yu.G.

1993-01-01

During past time all basic problems arisen during working-out of NPP with lead-bismuth coolant were solved: physics and thermal physics of the cores, heat transfer and hydrodynamics, corrosion resistance of the structural materials and coolant technology, radiation and nuclear safety, investigations of emergency situations, development of fuel elements and absorbing elements of the reactor, equipment of the primary circuit and other circuits. A powerful experimental base equpped by unique rigs is made. A series of ship and test NPP has been constructed whereat repair of the plants and reactor refuelling are developed. Highly-skilled groups of investigators, designers and operation personnel capable of performing the development of the reactor plant with MFR within short terms have been formed. In this case MFR with lead-bismuth coolant may become the initial step in development of large-scale nuclear power engineering with fast reactors cooled by liquid lead

Cation effect on small phosphonium based ionic liquid electrolytes with high concentrations of lithium salt

Science.gov (United States)

Chen, Fangfang; Kerr, Robert; Forsyth, Maria

2018-05-01

Ionic liquid electrolytes with high alkali salt concentrations have displayed some excellent electrochemical properties, thus opening up the field for further improvements to liquid electrolytes for lithium or sodium batteries. Fundamental computational investigations into these high concentration systems are required in order to gain a better understanding of these systems, yet they remain lacking. Small phosphonium-based ionic liquids with high concentrations of alkali metal ions have recently shown many promising results in experimental studies, thereby prompting us to conduct further theoretical exploration of these materials. Here, we conducted a molecular dynamics simulation on four small phosphonium-based ionic liquids with 50 mol. % LiFSI salt, focusing on the effect of cation structure on local structuring and ion diffusional and rotational dynamics—which are closely related to the electrochemical properties of these materials.

Ionic diffusion and salt dissociation conditions of lithium liquid crystal electrolytes.

Science.gov (United States)

Saito, Yuria; Hirai, Kenichi; Murata, Shuuhei; Kishii, Yutaka; Kii, Keisuke; Yoshio, Masafumi; Kato, Takashi

2005-06-16

Salt dissociation conditions and dynamic properties of ionic species in liquid crystal electrolytes of lithium were investigated by a combination of NMR spectra and diffusion coefficient estimations using the pulsed gradient spin-echo NMR techniques. Activation energies of diffusion (Ea) of ionic species changed with the phase transition of the electrolyte. That is, Ea of the nematic phase was lower than that of the isotropic phase. This indicates that the aligned liquid crystal molecules prepared efficient conduction pathways for migration of ionic species. The dissociation degree of the salt was lower compared with those of the conventional electrolyte solutions and polymer gel electrolytes. This is attributed to the low concentration of polar sites, which attract the dissolved salt and promote salt dissociation, on the liquid crystal molecules. Furthermore, motional restriction of the molecules due to high viscosity and molecular oriented configuration in the nematic phase caused inefficient attraction of the sites for the salt. With a decreased dissolved salt concentration of the liquid crystal electrolyte, salt dissociation proceeded, and two diffusion components attributed to the ion and ion pair were detected independently. This means that the exchange rate between the ion and the ion pair is fairly slow once the salt is dissociated in the liquid crystal electrolytes due to the low motility of the medium molecules that initiate salt dissociation.

Experimental testing facilities for ultrasonic measurements in heavy liquid metal

International Nuclear Information System (INIS)

Cojocaru, V.; Ionescu, V.; Nicolescu, D.; Nitu, A.

2016-01-01

The thermo-physical properties of Heavy Liquid Metals (HLM), like lead or its alloy, Lead Bismuth Eutectic (LBE), makes them attractive as coolant candidates in advanced nuclear systems. The opaqueness, that is common to all liquid metals, disables all optical methods. For this reason ultrasound waves are used in different applications in heavy liquid metal technology, for example for flow and velocity measurements and for inspection techniques. The practical use of ultrasound in heavy liquid metals still needs to be demonstrated by experiments. This goal requires heavy liquid metal technology facility especially adapted to this task. In this paper is presented an experimental testing facility for investigations of Heavy Liquid Metals acoustic properties, designed and constructed in RATEN ICN. (authors)

Experiments on 18-8 stainless steels exposed to liquid lithium. I. 1,100-hour corrosion tests in lithium of 400, 500 and 6000C in natural circulation type testing apparatus

International Nuclear Information System (INIS)

Nihei, I.; Sumiya, I.; Fukaya, Y.; Yamazaki, Y.

The Japan Atomic Energy Research Institute has planned and started to carry out a series of experiments concerning fusion reactor materials. This report gives the results of the first experiments. The first test materials selected were 18-8 stainless steels, and the experiments were designed to test their behavior when exposed to liquid lithium. Natural circulation type corrosion testing devices (pots) were used as the testing apparatus, and the tests were conducted with lithium temperatures up to 600 0 C

Assessment of alkali metal coolants for the ITER blanket

International Nuclear Information System (INIS)

Natesan, K.; Reed, C.B.; Mattas, R.F.

1994-01-01

The blanket system is one of the most important components of a fusion reactor because it has a major impact on both the economics and safety of fusion energy. The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert the fusion energy into sensible heat and to breed tritium for the fuel cycle. The Blanket Comparison and Selection Study, conducted earlier, described the overall comparative performance of different blanket concepts, including liquid metal, molten salt, water, and helium. This paper will discuss the ITER requirements for a self-cooled blanket concept with liquid lithium and for indirectly cooled concepts that use other alkali metals such as NaK. The paper will address the thermodynamics of interactions between the liquid metals (i.e., lithium and NaK) and structural materials (e.g., V-base alloys), together with associated corrosion/compatibility issues. Available experimental data will be used to assess the long-term performance of the first wall in a liquid metal environment

Design and Characterisation of Solid Electrolytes for All-Solid-State Lithium Batteries

DEFF Research Database (Denmark)

Sveinbjörnsson, Dadi Þorsteinn

The development of all-solid-state lithium batteries, in which the currently used liquid electrolytes are substituted for solid electrolyte materials, could lead to safer batteries offering higher energy densities and longer cycle lifetimes. Designing suitable solid electrolytes with sufficient...... chemical and electrochemical stability, high lithium ion conduction and negligible electronic conduction remains a challenge. The highly lithium ion conducting LiBH4-LiI solid solution is a promising solid electrolyte material. Solid solutions with a LiI content of 6.25%-50% were synthesised by planetary......-rich microstructures during ball milling is found to significantly influence the conductivity of the samples. The long-range diffusion of lithium ions was measured using quasi-elastic neutron scattering. The solid solutions are found to exhibit two-dimensional conduction in the hexagonal plane of the crystal structure...

Development of technology for fabrication of lithium CPS on basis of CNT-reinforced carboxylic fabric

Energy Technology Data Exchange (ETDEWEB)

Tazhibayeva, Irina, E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Baklanov, Viktor; Ponkratov, Yuriy [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Abdullin, Khabibulla [Institute of Experimental and Theoretical Physics of Kazakh National University, Almaty (Kazakhstan); Kulsartov, Timur; Gordienko, Yuriy; Zaurbekova, Zhanna [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Lyublinski, Igor [JSC «Red Star», Moscow (Russian Federation); NRNU «MEPhI», Moscow (Russian Federation); Vertkov, Alexey [JSC «Red Star», Moscow (Russian Federation); Skakov, Mazhyn [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan)

2017-04-15

Highlights: • Preliminary study of carboxylic fabric wettability with liquid lithium is presented. • Preliminary studies of carboxylic fabric wettability with liquid lithium consist in carrying out of experiments at temperatures 673,773 and 873 К in vacuum during long time. • A scheme of experimental device for manufacturing of lithium CPS and matrix filling procedure with liquid lithium are presented. • The concept of lithium limiter with CPS on basis of CNT-reinforced carboxylic fabric is proposed. - Abstract: The paper describes the analysis of liquid lithium interaction with materials based on carbon, the manufacture technology of capillary-porous system (CPS) matrix on basis of CNT-reinforced carboxylic fabric. Preliminary study of carboxylic fabric wettability with liquid lithium is presented. The development of technology includes: microstructural studies of carboxylic fabric before its CNT-reinforcing; validation of CNT-reinforcing technology; mode validation of CVD-method for CNT synthesize; study of synthesized carbon structures. Preliminary studies of carboxylic fabric wettability with liquid lithium consist in carrying out of experiments at temperatures 673, 773 and 873 К in vacuum during long time. The scheme of experimental device for manufacturing of lithium CPS and matrix filling procedure with liquid lithium are presented. The concept of lithium limiter with CPS on basis of CNT-reinforced carboxylic fabric is proposed.

Development of technology for fabrication of lithium CPS on basis of CNT-reinforced carboxylic fabric

International Nuclear Information System (INIS)

Tazhibayeva, Irina; Baklanov, Viktor; Ponkratov, Yuriy; Abdullin, Khabibulla; Kulsartov, Timur; Gordienko, Yuriy; Zaurbekova, Zhanna; Lyublinski, Igor; Vertkov, Alexey; Skakov, Mazhyn

2017-01-01

Highlights: • Preliminary study of carboxylic fabric wettability with liquid lithium is presented. • Preliminary studies of carboxylic fabric wettability with liquid lithium consist in carrying out of experiments at temperatures 673,773 and 873 К in vacuum during long time. • A scheme of experimental device for manufacturing of lithium CPS and matrix filling procedure with liquid lithium are presented. • The concept of lithium limiter with CPS on basis of CNT-reinforced carboxylic fabric is proposed. - Abstract: The paper describes the analysis of liquid lithium interaction with materials based on carbon, the manufacture technology of capillary-porous system (CPS) matrix on basis of CNT-reinforced carboxylic fabric. Preliminary study of carboxylic fabric wettability with liquid lithium is presented. The development of technology includes: microstructural studies of carboxylic fabric before its CNT-reinforcing; validation of CNT-reinforcing technology; mode validation of CVD-method for CNT synthesize; study of synthesized carbon structures. Preliminary studies of carboxylic fabric wettability with liquid lithium consist in carrying out of experiments at temperatures 673, 773 and 873 К in vacuum during long time. The scheme of experimental device for manufacturing of lithium CPS and matrix filling procedure with liquid lithium are presented. The concept of lithium limiter with CPS on basis of CNT-reinforced carboxylic fabric is proposed.

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.)