Leak detector of liquid sodium

International Nuclear Information System (INIS)

Himeno, Yoshiaki.

1975-01-01

Object: To arrange a cable core connected to a leakage current detector on the outer wall of piping for liquid sodium, devices or the like and apply a voltage to said core and outer wall to quickly and securely detect the leakage of liquid sodium. Structure: A cable, which is composed of metal coating formed of metal material (copper, steel, stainless, etc.) which is apt to be corroded by reaction products of liquid sodium with water and oxygen in air, and metal oxide (such as magnesium oxide, beryllium oxide, aluminum oxide) as an electric insulator is arranged on the outer wall of pipes or devices. In the event sodium is leaked from the pipes or devices, said metal coating and the insulator are corroded, and the leakage of sodium is sensed by a leakage current detector through the core in the cable. (Kamimura, M.)

The reaction between barium and nitrogen in liquid sodium: resistivity studies

International Nuclear Information System (INIS)

Addison, C.C.; Creffield, G.K.; Hubberstey, P.; Pulham, R.J.

1976-01-01

The reaction of nitrogen with solutions of barium (between 0.34 and 6.89 mol % Ba) in liquid sodium at 573 K has been followed by changes in the electrical resistivity of the liquid. The capillary method has been employed, continuous sampling during reaction being achieved by electromagnetic pumping. The initial solution of nitrogen in the metal, followed by precipitation of barium and nitrogen from sodium as the nitride Ba 2 N, are reflected in the resistivity changes. The solubility of nitrogen in the alloy is a linear function of the barium concentration: S(mol % N) = x/4 (0 <= x <= 6.89 mol % Ba). This and the decrease in resistivity which invariably occurs during the solution process, provides additional information on the nature of solvation of nitrogen in solution in the liquid metal. (author)

Sodium-Water Reaction approach and mastering for ASTRID Steam Generator design

International Nuclear Information System (INIS)

Saez, Manuel; Allou, Alexandre; Beauchamp, François; Bertrand, Carole; Rodriguez, Gilles; Menou, Sylvain; Prele, Gérard

2013-01-01

Conclusions: • Modular Steam Generator concept selected for ASTRID: → Brings flexibility for the expertise of failed modules after their removal; → Intrinsically limit the mechanical consequences of a postulated large Sodium-Water Reaction. • Sodium-Water-Air Reaction studies include both prevention and mitigation aspects, with dedicated tools to be developed through R&D. • Regarding Safety analysis, the possibility to move from the scenario of instantaneous failure of the whole Steam Generator tube bundle toward a scenario with sequenced failure needs to be investigated. • The Steam Generator is one of the key components in the Sodium-cooled Fast Reactor system for it provides an interface between sodium and water. The design objective for the Steam Generator is related to the improvement of mastering of Sodium-Water Reaction. • Potential Sodium-Water Reactions can be eliminated by adopting a Gas based Power Conversion System

Experimental study o the sodium-concrete reaction

International Nuclear Information System (INIS)

Goncalves, A.C.; Torres, A.R.; Brito Aghina, L.O. de; Messere e Castro, P.

1986-01-01

Effects and aspects of security are verified during a sodium leakage in high temperatures on liners of contention cells and directly on the concrete. As this kinetic process involves reactions between materials in solid state (such as oxides and carbonates) vapors and gases (such as water and CO 2 )) with liquid sodium, effects of each phase of the heterogeneous mixture are separately analysed. Are still analysed produced products and briefly discussed the kinetic of the sodium-concrete reaction. (Author) [pt

Control of sodium fires and sodium-water reactions in breeder reactors

International Nuclear Information System (INIS)

Foerster, K.; Ruloff, G.; Voss, J.

1985-01-01

The excellent neutronic and thermodynamic properties of sodium as a fast-reactor coolant are somewhat counterbalanced by its high oxygen affinity. Because incidents like sodium fires and sodium-water reactions cannot be absolutely excluded, their effects and preventive measures have to be investigated. Characteristics and counter-measures are discussed. (orig.) [de

Improvement on reaction model for sodium-water reaction jet code and application analysis

International Nuclear Information System (INIS)

Itooka, Satoshi; Saito, Yoshinori; Okabe, Ayao; Fujimata, Kazuhiro; Murata, Shuuichi

2000-03-01

In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.30) and application analysis to the water injection tests for confirmation of code propriety were performed. On the improvement of the code, a gas-liquid interface area density model was introduced to develop a chemical reaction model with a little dependence on calculation mesh size. The test calculation using the improved code (LEAP-JET ver.1.40) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results and the influence to analysis result of a model are reasonable. For the application analysis to the water injection tests, water injection behavior and SWR jet behavior analyses on the new SWAT-1 (SWAT-1R) and SWAT-3 (SWAT-3R) tests were performed using the LEAP-BLOW code and the LEAP-JET code. In the application analysis of the LEAP-BLOW code, parameter survey study was performed. As the results, the condition of the injection nozzle diameter needed to simulate the water leak rate was confirmed. In the application analysis of the LEAP-JET code, temperature behavior of the SWR jet was investigated. (author)

Analysis of the Sodium-Water Reaction Phenomena by Small Water/Steam Leaks

International Nuclear Information System (INIS)

Jeong, J-Y; Kim, T-J; Kim, J-M; Kim, B-H; Park, N-C

2006-01-01

One of the important problems to be solved in the design and construction of a sodium cooled fast reactor is to confirm the safety and reliability of the steam generator which transfers the heat from the sodium to the water. Sodium-water reaction events may occur when material faults such as a pinhole or cracks occur in the heat transfer tube wall. When such a leak occurs, evaporating water or superheated steam enters through a small leak into the sodium. The surface of this steam jet reacts with the surrounding sodium. Due to turbulence, sodium and particles of the reaction products are drawn at a high velocity into the jet. Impingement of these particles on an adjacent tube is followed by a combined process of a corrosion and erosion which results in a local weakening of the affected tube. If there is no reliable detection available in time, wastage will ultimately result in an additional leak in the adjacent tube. Therefore, it is very significant to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. The objective of this study is a basic investigating of the sodium-water reaction phenomena by small water/steam leaks

Acoustic detection for small-leak sodium-water reaction

International Nuclear Information System (INIS)

Nei, Hiromichi; Ohshima, Iwao; Ujihara, Kozaburo; Hori, Masao

1977-01-01

Characteristics of acoustic signal produced by sodium-water reaction due to steam injection and by Ar gas injection into sodium were experimentally investigated. Acoustic signal was measured by using Kistler 808A and 815A5 accelerometers. Root mean square (RMS) measurements and frequency analysis of the signal were conducted. The RMS measurements could detect a small water leakage into sodium, as small as 0.07g/sec, in the present loop. The peaks in a frequency spectrum were caused by the natural vibration of a rod on which the acoustic transducer was mounted. The RMS was approximately proportional to the one-third power of the steam leak rate and increased to some extent with the ambient sodium temperature. RMS values, both for sodium-water reaction and Ar gas injection, were about the same order of magnitude, when the data were plotted against the volumetric flow rates of steam and Argas. (auth.)

Wastage of Steam Generator Tubes by Sodium-Water Reaction

International Nuclear Information System (INIS)

Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

2010-01-01

The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown in Figure 1. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. For this, multi-target wastage tests for modified 9Cr-1Mo steel tube bundle by intermediate leaks are being prepared

A study on sodium-concrete reaction

Energy Technology Data Exchange (ETDEWEB)

Pae, Jae Huem; Min, Byung Hoon; Lee, Joon Sik; Lee, Choong Hui; Chung, Ki Hong; Keum, Choong Ki [Suwon University, Suwon (Korea, Republic of)

1994-07-15

Sodium is commonly used as a coolant in liquid metal reactor. A large amount of its leakage may be possible in hypothetical accidents, even though the possibility is very low. In case that the leaked hot sodium comes in direct contact with structural concrete of liquid metal reactor, the reactor`s integrity can be challenged by the rupture of structure materials, hydrogen generation and its explosion, and release of radioactive aerosols due to sodium-concrete reaction. The knowledge of sodium-concrete reaction is evaluated to be one of the important and indispensable technologies for the establishment of safety measure in liquid metal reactor. In this study, the experimental facility of sodium-concrete reaction is to be designed, constructed and operated. And the reaction phenomena of sodium-concrete reaction is also to be analyzed through the experimental results. The aim of this study is to establish the measure of safety and protection for sodium-related facilities and to secure one of the fundamental technologies of liquid metal reactor safety. 47 refs., 7 figs., 13 tab.

Thermal analysis experiment for elucidating sodium-water chemical reaction mechanism in steam generator of sodium-cooled fast reactor

International Nuclear Information System (INIS)

Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki

2012-01-01

For the purpose of elucidating the mechanism of the sodium-water surface reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature, and decomposition temperature of sodium hydride (NaH) have been identified from DTA curves. Based on the measured reaction temperature, rate constant of sodium monoxide (Na 2 O) generation was obtained. Thermal analysis results indicated that Na 2 O generation at the secondary overall reaction should be considered during the sodium-water reaction. (author)

Liquid sodium technology research

International Nuclear Information System (INIS)

Kim, W.C.; Lee, Y.W.; Nam, H.Y.; Chun, S.Y.; Kim, J.; Won, S.Y.

1982-01-01

This report describes the technology of impurity control and measurement of liquid sodium, problems associated with material degradation and change of heat transfer characteristics in liquid sodium, and the conceptual design of multipurpose sodium test loop. Discussion and the subsequent analysis are also made with regard to the test results for the sodium-H 2 0 reaction and its effects on the system. (author)

Japanese position paper on sodium-water reaction testing and design

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Tanabe, H.; Miyake, O.; Kuroha, M.; Hoshi, Y.

1984-01-01

PNC has been developing the steam generator with helically coiled heat transfer tube bundle and downcommer tubes for the prototype fast reactor Monju since 1968. To establish the safety design against the sodium-water reaction accident was one of the most important R and D items at the start of the development. PNC started the experimental study initially in the large leak region in 1970. Until now, during twelve years, the experimental studies have been performed, which covers the phenomena from a micro leak to a large one, with the use of the SWAT-1 rig, SWAT-2 loop, SWAT-3 loop, and SWAT-4 rigs. The reliable leak detection system is necessary to minimize the damage by the sodium-water reaction. Two groups of efforts have been paid for developing the detection system. One is to develop the leak detector itself, and another is to grasp the hydrogen transport behavior in the sodium in the steam generator and the secondary piping system. Four sodium loops have been used for the development. The development of computer codes has also progressed in parallel with the sodium-water reaction experiments. Three codes have been accomplished for the design tools against the sodium-water reaction. Through the efforts mentioned above, sufficient experiences were obtained for designing and operating the Monju steam generator system

Apparatus for detecting leakage of liquid sodium

Science.gov (United States)

Himeno, Yoshiaki

1978-01-01

An apparatus for detecting the leakage of liquid sodium includes a cable-like sensor adapted to be secured to a wall of piping or other equipment having sodium on the opposite side of the wall, and the sensor includes a core wire electrically connected to the wall through a leak current detector and a power source. An accidental leakage of the liquid sodium causes the corrosion of a metallic layer and an insulative layer of the sensor by products resulted from a reaction of sodium with water or oxygen in the atmospheric air so as to decrease the resistance between the core wire and the wall. Thus, the leakage is detected as an increase in the leaking electrical current. The apparatus is especially adapted for use in detecting the leakage of liquid sodium from sodium-conveying pipes or equipment in a fast breeder reactor.

Experimental Study of Na based Titanium Nanofluid-Water Reaction

Energy Technology Data Exchange (ETDEWEB)

Park, Gunyeop; Kim, Soo Jae; Baek, Jehyun; Kim, Hyun Soo; Oh, Sun Ryung; Park, Hyun Sun; Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of)

2015-10-15

In KALIMER-600, a sodium-cooled fast reactor designed by KAERI, thermal energy is transported from high-temperature liquid Na (526 .deg. C at 0.1 MPa) to low temperature water (230 .deg. C at - 19.5 MPa) through a heat exchanger. If any leakage or rupture occurs during the operation of this heat exchanger, highly pressurized liquid water can penetrate into the liquid Na channels; this contact should instantly cause SWR. As reaction continues, liquid water is soon vaporized by pressure drop and huge amount of reaction heat. This generated water vapor expands large reaction area and increases sodium-water vapor reaction process. Therefore, the rapid generation of reaction product (like H{sub 2}) and water vapor increases the system pressure that can cause the system failure in SFR. To reduce this strong chemical reaction phenomena between Na and water, some we have focused on suppressing the chemical reactivity of liquid Na by dispersing nanoparticles (NPs). For the real application of NaTiNF, the pressure change induced by NaTiNF-water reaction is compared with Na-water reaction in the present study. NaTiNF contains 100nm of Ti NPs at 0.2 vol. %. The reaction rate of NaTiNF-water reaction is also investigated as reaction temperature increases. Sodium-water vapor reaction (SVR) will occur when an SWR accident occurs in SFR. In this manner, NaTiNF-water vapor reaction is experimentally performed for ensuring the suppression of chemical reactivity of NaTiNF in contact with water vapor. In the basic step for reducing risk of an SWR in SFR, we have experimentally verified the suppressed chemical reactivity of liquid sodium using Ti NPs through SWR and SVR experiments. In SWR, Na based titanium nanofluid (NaTiNF) shows lower pressure change than Na. As T{sub R} increases, P{sub max} in Na-water reaction increases while NaTiNF does not. The reaction rate of NaTiNF shows twice slower than that of Na. In SVR, NaTiNF shows slower temperature increase than Na. The distinct

Development of computer code on sodium-water reaction products transport

International Nuclear Information System (INIS)

Arikawa, H.; Yoshioka, N.; Suemori, M.; Nishida, K.

1988-01-01

The LMFBR concept eliminating the secondary sodium system has been considered to be one of the most promissing concepts for offering cost reductions. In this reactor concept, the evaluation of effects on reactor core by the sodium-water reaction products (SWRPs) during sodium-water reaction at primary steam generator becomes one of the major safety issues. In this study, the calculation code was developed as the first step of the processes of establishing the evaluation method for SWRP effects. The calculation code, called SPROUT, simulates the SWRPs transport and distribution in primary sodium system using the system geometry, thermal hydraulic data and sodium-water reacting conditions as input. This code principally models SWRPs behavior. The paper contain the modelings for SWRPs behaviors, with solution, precipation, deposition and so on, and the results and discussions of the demonstration calculation for a typical FBR plant eliminating the secondary sodium system

An Evaluation of the Acoustic Signal processing Techniques for Sodium-Water Reaction Detection in KALIMER-600

International Nuclear Information System (INIS)

Hur, Seop; Seong, S. H.; Kim, T. J.; Kim, S. O.; Lee, M. K.

2005-02-01

KALIMER-600 is a pool type fast breeder reactor using liquid sodium as a coolant. Although it has the several advantages such as long-term fuel cycle and enhanced safety concepts, it is possible to leak the secondary side water/steam into sodium boundary. This event could make the plant abnormal condition. One of the major design issues in KALIMER-600 is, therefore, to develop the system which can early detect the sodium-water reaction to protect the sodium-water reaction event. After evaluating the various signal processing techniques for passive acoustic leak detection, we have proposed the early leak detection logics. the signal processing techniques for evaluation were the spectral estimation using the linear modeling, the estimation error of linear modeling, the system adaptation rate using an adaptive signal processing, and the background noise cancellation using adaptive and fixed filtering. As the analysis results regarding the stationary and the cross-correlation of leak signals and background noises, the two signal systems met a wide-dense stationary process and there was only the week cross correlation relationship between two signals. It is ,therefore, possible to use the linear/harmonic modeling of signal systems, and the leak signal in sensor outputs can be discriminated. As the results of the evaluation of the various spectral estimation methods, the spectral estimation method based on autoregressive modeling was more practical comparing with other methods in the sodium-water reaction detection. The passive acoustic leak detection logics were suggested based on above evaluations. the logics consist of 3 levels; transient identification, leak determination and leak symptom identification. The simulation results using sodium-water reaction signals showed that it was possible to determine the leak at above -3dB of SNR, while between -3 dB and -10 dB of SNR the logics determined the leak symptom identification. The detection sensitivity can be enhanced

Corrosion of Steels in the Vicinity of a Sodium-Water Reaction

Energy Technology Data Exchange (ETDEWEB)

Davies, R. A.; Bray, J. A.; Lyons, J. M. [U.K. Atomic Energy Authority, Dounreay Experimental Reactor Establishment, Thurso, Caithness (United Kingdom)

1967-06-15

Rapid corrosion of steels in the vicinity of a sodium-water reaction could lead to a major reaction in a sodium-water heat exchanger. An investigation of the magnitude of the corrosion problem has been carried out under conditions simulating both a small water leak and a full size pipe burst, and further tube failures have been obtained. These experiments were carried out on a sodium rig which could accommodate simple full-scale models of sections of heat exchanger, and up to 70 lb of water was injected into 700 lb of sodium in 9.0s. The corrosion phenomena have also been investigated on a small scale under more controllable conditions by pumping water at normal pressures into a pot of sodium. With a flow-rate of 1 ml/min corrosion rates in excess of 0.005 in./min have been obtained. The effect of various parameters on the corrosion rate has been studied, and a comparison has been made of the corrosion rates obtained with a variety of steels. The corrosion appears to be a direct result of conditions during the reaction, and the appearance of the specimen and pattern of damage suggests that the main effect is concentrated where the sodium water reaction front impinges on the metal surface. The corrosion rates are very much lower with stainless steel and nickel alloys than with ferritic materials, and suggest that the phenomena are associated with the formation of (Na{sub 2}O){sub 2}FeO. Iron powder has also been observed in the vicinity of the reaction which would suggest that this is reduced at a later stage, either as a result of the hydrogen produced during reaction, or by thermal cycling. (author)

Development of the SPIKE code for analysis of the sodium-water reaction

Energy Technology Data Exchange (ETDEWEB)

Hwang, Sung Tai; Park, Jin Ho; Choi, Jong Hyeun; Kim, Tae Joon [Korea Atomic Energy Research Institute, Taejon (Korea)

1998-08-01

In the secondary loop of liquid metal reactors, including SG, water leak into sodium causes the sudden increase of pressure by the H{sub 2} and heat generated from reaction. At few miliseconds after leak, a sharp and short-lived increase of pressure is generated and its propagation depends on the acoustic constraint characteristics of secondary loop. As increasing leak amount of water, another pressure increase is caused by H{sub 2} and its transients depends on the resistance of pressure opening system, such as rupture disc. For prediction of the transients of initial spike pressure, a code of SPIKE was developed. The code was based on the following simplifications and assumptions: combination of total and half release of H{sub 2} rate, spherical shape of H{sub 2} bubble, compressible and Newtonian fluid for sodium. The program was built in FOTRAN language and consisted of 5 modules. Several sample calculations were performed to test the code and to determine the scale down factor of experimental facilities for experimental verification of the code: parameter study of the variables in chemical reaction model, comparison study with results calculated by superposition methods for simple piping structures, comparison study with results calculated by previous researchers, and calculation for KALIMER models of various size. With these calculation results, the generally predicted phenomena of sodium water reaction can be explained and the calculated ones by SPIKE code were well agreed with the previous study. And the scale down factor can be determined. (author). 88 refs., 99 figs., 39 tabs.

Computer simulation for sodium-concrete reactions

International Nuclear Information System (INIS)

Zhang Bin; Zhu Jizhou

2006-01-01

In the liquid metal cooled fast breeder reactors (LMFBRs), direct contacts between sodium and concrete is unavoidable. Due to sodium's high chemical reactivity, sodium would react with concrete violently. Lots of hydrogen gas and heat would be released then. This would harm the ignorantly of the containment. This paper developed a program to simualte sodium-conrete reactions across-the-board. It could give the reaction zone temperature, pool temperature, penetration depth, penetration rate, hydrogen flux and reaction heat and so on. Concrete was considered to be composed of silica and water only in this paper. The variable, the quitient of sodium hydroxide, was introduced in the continuity equation to simulate the chemical reactions more realistically. The product of the net gas flux and boundary depth was ably transformed to that of penetration rate and boundary depth. The complex chemical kinetics equations was simplified under some hypothesises. All the technique applied above simplified the computer simulation consumedly. In other words, they made the computer simulation feasible. Theoretics models that applied in the program and the calculation procedure were expatiated in detail. Good agreements of an overall transient behavior were obtained in the series of sodium-concrete reaction experiment analysis. The comparison between the analytical and experimental results showed the program presented in this paper was creditable and reasonable for simulating the sodium-concrete reactions. This program could be used for nuclear safety judgement. (authors)

Sodium-water reaction product flow system

Energy Technology Data Exchange (ETDEWEB)

Shirataki, K; Wada, H

1978-11-18

Purpose: To provide the subject equipments wherein thermal insulating layers which neither exfoliate nor react by the impact due to high temperature sodium and hydrogen gas and are used for mitigating the thermal impact are provided on the inner surfaces of the emission system equipments, thereby preventing the destruction of the emission system equipments. Constitution: Thermal insulating layers are formed on the inner surfaces of sodium-water reaction product emission system equipments, that is, the inner surface of the emission system pipeline, that of the accommodation vessel and the surface of the cyclone separator, by film treatment, coating or heat resisting coating, and these surfaces are covered with the layers. Each of the layers is made of a material which does not cause a rapid reaction with high temperature sodium or hydrogen gas nor exfoliates and is withstandable for several seconds in which the thermal impact of at least the emission system comes into question, and its thickness is more than one capable of securing the necessary thermal resistance computed by the thermal impact analysis of the emission system.

Sodium-water reaction product flow system

International Nuclear Information System (INIS)

Shirataki, Koji; Wada, Hozumi.

1978-01-01

Purpose: To provide the subject equipments wherein thermal insulating layers which neither exfoliate nor react by the impact due to high temperature sodium and hydrogen gas and are used for mitigating the thermal impact are provided on the inner surfaces of the emission system equipments, thereby preventing the destruction of the emission system equipments. Constitution: Thermal insulating layers are formed on the inner surfaces of sodium-water reaction product emission system equipments, that is, the inner surface of the emission system pipeline, that of the accommodation vessel and the surface of the cyclone separator, by film treatment, coating or heat resisting coating, and these surfaces are covered with the layers. Each of the layers is made of a material which does not cause a rapid reaction with high temperature sodium or hydrogen gas nor exfoliates and is withstandable for several seconds in which the thermal impact of at least the emission system comes into question, and its thickness is more than one capable of securing the necessary thermal resistance computed by the thermal impact analysis of the emission system. (Yoshihara, H.)

Acoustic signal processing for the detection of sodium boiling or sodium-water reaction in LMFRs. Final report of a co-ordinated research programme 1990-1995

International Nuclear Information System (INIS)

1997-05-01

This report is a summary of the work performed under a co-ordinated research programme entitled Acoustic Signal Processing for the Detection of Sodium Boiling or Sodium-Water Reaction in Liquid Metal Cooled Fast Reactors. The programme was organized by the IAEA and carried out from 1990 to 1995. It was the continuation of an earlier research co-ordination programme entitled Signal Processing Techniques for Sodium Boiling Noise Detection, which was carried out from 1984 to 1989. Refs, figs, tabs

Pressure transients resulting from sodium-water reaction following a large leak in LMFBR steam generator

International Nuclear Information System (INIS)

Rajput, A.K.

1984-01-01

The study of sodium water reaction, following a large leak, concerns primarily with the estimation of pressure/flow transients that are developed in the steam generator and the associated secondary circuit. This paper describes the mathematical formulations used in SWRT (Sodium Water Reaction Transients) code developed to estimate such pressure transients for FBTR plant. The results, obtained using SWRT have been presented for a leak in economiser (20m from bottom water header) and for a leak in super heater portions. A time lag of 50 m sec was considered for rupture disc takes to burst once the pressure experienced by it exceeds the set value. Also described in annexure to this paper is the mathematical formulation for two phase transient flow for the better estimation of leak rate from the ruptured end of the damaged heat transfer tube. This leak model considers slip but assumes thermal equilibrium between the liquid and vapour phases

Analyses of hydrodynamic effects of large sodium-water reactions

International Nuclear Information System (INIS)

Sakano, K.; Shindo, Y.; Koishikawa, A.; Maekawa, I.

1977-01-01

Large leak sodium-water reactions that would occur in a steam generator of LMFBR causes abrupt changes of pressure and velocity of fluid in a secondary sodium system and relief system. This paper describes SOWACS-III together with its model and method. Results of analyses are also given, the comparison with experimental results of initial pressure spike being included. SOWACS-III treats the system which consists of the steam generator, vessel, valve, pump and pipe, and uses the following models and methods. (1) Components are assumed to be one-dimensional. (2) Pressure wave propagation near a reaction zone, where hydrogen is generated, is analyzed with the spherical co-ordinate (sphere-cylinder model). (3) A moving boundary is formed by contact of sodium with other fluid such as hydrogen and nitrogen. The boundary travels without mixing of sodium and another fluid through the boundary (boundary tracking model). The boundary can be treated not to move from the original place (fixed boundary model). (4) Pressure wave propagation is analyzed by the explicit method of characteristics in one-dimensional Eulerian co-ordinate. (5) Flow-induced force is analyzed by momentum balance. (6) The lateral motion of relief piping caused by the force is analyzed by NASTRAN code. Analyses were carried out for large sodium-water reaction experiments in SWAT-3 rig of PNC by using the sphere-cylinder model. The calculated pressure spike in the reaction vessel was compared with the measured one for a few milliseconds after water injection. The calculated value and measured one were 6.4 ata and 6.7 ata for peak pressure and 0.6 ms and 2.8 ms for rising time, respectively

Design of the US-CRBRP sodium/water reaction pressure relief system

International Nuclear Information System (INIS)

Kruger, G.B.; Murdock, T.B.; Rodwell, E.; Sane, J.O.

1976-01-01

Protection against intermediate sodium system overpressure from the sodium/water reaction associated with large leaks within the CRBRP Steam Generators is provided by the sodium/water reaction pressure relief system (SWRPRS). This system consists of rupture disks connected to the intermediate sodium piping adjacent to the inlet to the superheater and outlet from the evaporator modules. The rupture discs relieve into piping that leads to reaction produce separator tanks, which in turn are vented to a centrifugal separator and flare stack arranged to burn hydrogen gas exhausting into the atmosphere. Analyses have been conducted using the TRANSWRAP Computer Code to predict the system pressures and flow rates during the large leak event. Experimental tests to be conducted in the large leak test rig (LLTR) will be used to confirm the analysis techniques used in the design

Implications of small water leak reactions on sodium heated steam generator design

Energy Technology Data Exchange (ETDEWEB)

Smedley, J A

1975-07-01

Various types of sodium water reactions have been looked on as possibly causing hazard conditions in sodium heated steam generator units ranging from the very improbable boiler tube double ended guillotine fracture to the almost certain occurrence of micro-leaks. Within this range small water leaks reactions have attracted particular interest and the present paper looks at the principles of associating the reactions with detection and protection systems for Commercial Fast Reactors. A method is developed for assessing whether adequate protection has been provided against the effects of small water leak reactions in a steam generator unit. (author)

Small leak detection by measuring surface oscillation during sodium-water reaction in steam generator

International Nuclear Information System (INIS)

Nei, Hiromichi; Hori, Masao

1977-01-01

Small leak sodium-water reaction tests were conducted to develop various kinds of leak detectors for the sodium-heated steam generator in FBR. The super-heated steam was injected into sodium in a reaction vessel having a sodium free surface, simulating the steam generator. The level gauge in the reaction vessel generated the most reliable signal among detectors, as long as the leak rates were relatively high. The level gauge signal was estimated to be the sodium surface oscillation caused by hydrogen bubbles produced in sodium-water reaction. Experimental correlation was derived, predicting the amplitude as a function of leak rate, hydrogen dissolution ratio, bubble rise velocity and other parameters concerned, assuming that the surface oscillation is in proportion to the gas hold-up. The noise amplitude under normal operation without water leak was increased with sodium flow rate and found to be well correlated with Froud number. These two correlations predict that a water leak in a ''MONJU'' class (300 MWe) steam generator could possibly be detected by level gauges at a leak rate above 2 g/sec. (auth.)

Prediction of metal wastage produced by sodium-water reaction jets

International Nuclear Information System (INIS)

Payne, J.F.B.

1979-01-01

When a leak occurs in a sodium heated boiler, a sodium-water reaction jet is formed which causes further damage to the boiler, ie wastage. A correlation is developed which predicts wastage of ferritic steel boiler tubes for the range of leak sizes, sodium temperatures and leak to target spacing for which data available. (author)

Wastage Behavior of Modified 9Cr-1Mo Steel Tube Material by Sodium-Water Reaction (II)

International Nuclear Information System (INIS)

Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

2010-01-01

The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. The objective of this study is a basic investigating of the sodium-water reaction phenomena by small water/steam leaks. For this, wastage tests for modified 9Cr-1Mo steel tube material were conducted, and an empirical formula of the wastage rate for this material was obtained from the results

Reactions of oxygen and hydrogen with liquid sodium - a critical survey

International Nuclear Information System (INIS)

Ullmann, H.

1982-01-01

The fundamentals of solvation chemistry are presented with appropriate components formulated. Methods of investigation and kinetics of the reactions are described. The hydrogen equilibrium pressure and saturation solubilities are described. The chemical equilibrium between O and H in solution is presented with detailed tabulation of the saturation solutions of oxygen, hydrogen and hydroxide in liquid sodium. Agreements and differences with the literature are presented

Sodium-concrete reaction model development

International Nuclear Information System (INIS)

Nguyen, D.H.; Muhlestein, L.D.; Postma, A.K.

1982-07-01

Major observations have been formulated after reviewing test results for over 100 sodium-concrete reaction tests. The observations form the basis for developing a mechanistic model to predict the transient behavior of sodium-concrete reactions. The major observations are listed. Mechanisms associated with sodium and water transport to the reaction zone are identified, and represented by appropriate mathematical expressions. The model attempts to explain large-scale, long-term (100 h) test results were sodium-concrete reactions terminated even in the presence of unreacted sodium and concrete

Hydrogen meter for service in liquid sodium

International Nuclear Information System (INIS)

McCown, J.J.

1983-11-01

This standard establishes the requirements for the design, materials, fabrication, quality assurance, examination, and acceptance testing of a hydrogen meter and auxiliary equipment for use in radioactive or nonradioactive liquid sodium service. The meter shall provide a continuous and accurate indication of the hydrogen impurity concentration over the range 0.03 to 10 ppM hydrogen in sodium at temperatures between 800 and 1000 0 F (427 and 538 0 C). The meter may also be used to rapidly monitor changes in hydrogen concentration, over the same concentration range, and, therefore can be used as a sensor for sodium-water reactions in LMFBR steam generators

The Promising Features of New Nano Liquid Metals—Liquid Sodium Containing Titanium Nanoparticles (LSnanop

Directory of Open Access Journals (Sweden)

Toshio Itami

2015-07-01

Full Text Available A new kind of suspension liquid was developed by dispersing Ti nanoparticles (10 nm in liquid Na, which was then determined by TEM (transmission electron microscopy analysis. The volume fraction was estimated to be 0.0088 from the analyzed Ti concentration (2 at. % and the densities of Ti and Na. This suspension liquid, Liquid Sodium containing nanoparticles of titanium (LSnanop, shows, despite only a small addition of Ti nanoparticles, many striking features, namely a negative deviation of 3.9% from the ideal solution for the atomic volume, an increase of 17% in surface tension, a decrease of 11% for the reaction heat to water, and the suppression of chemical reactivity to water and oxygen. The decrease in reaction heat to water seems to be derived from the existence of excess cohesive energy of LSnanop. The excess cohesive energy was discussed based on simple theoretical analyses, with particular emphasis on the screening effect. The suppression of reactivity is discussed with the relation to the decrease of heat of reaction to water or the excess cohesive energy, surface tension, the action as a plug of Ti oxide, negative adsorption on the surface of LSnanop, and percolation.

Analyses for experiment on sodium-water reaction temperature by the CHAMPAGNE code

International Nuclear Information System (INIS)

Yoshioka, Naoki; Kishida, Masako; Yamada, Yumi

2000-03-01

In this work, analyses on sodium-water reaction temperature in the new SWAT-1(SWAT-1R) test were completed by the CHAMPAGNE code in order to understand void and velocity distribution in sodium system, which was difficult to be measured in experiments. The application method of the RELAP5/Mod2 code was investigated to LMFBR steam generator (SG) blow down analysis, too. The following results were obtained. (1) Analyses on sodium-water reaction temperature in the SWAT-1R test. 1) Analyses were carried out for the SWAT-1R test under the condition water leak rate 600 g/s by treating the pressure loss coefficient, the interface friction coefficient and the coefficient related to reaction rate as parameters. The effect and mechanism of each parameter on the shape of reaction zone were well understood by these analyses. 2) The void and velocity distribution in sodium system were estimated by use of the most suitable parameters. These analytical results are expected to be useful for planning of the SWAT-1R test and evaluation of test result. (2) Investigation of the RELAP5/Mod2 code. 1) The items to be improved in the RELAP5/Mod2 code were clarified to apply this code to the FBR SG blow down analysis. 2) One of these items was an addition of the shell-side (sodium-side) model. A sodium-side model was designed and added to the RELAP5/Mod2 code. Test calculations were carried out by this improved code and the basic function of this code was confirmed. (author)

Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-08-15

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

Fundamental study on temperature estimation of steam generator tubes at sodium-water reaction

International Nuclear Information System (INIS)

Furukawa, Tomohiro; Yoshida, Eiichi

2008-11-01

In case of the tube failure in the steam generator of the sodium cooled fast breeder reactor, its adjoined tubes are rapidly heated up by the chemical reaction between sodium and water/steam. And it is known that the tubes have the damage called 'wastage' by the disclosure steam jet. This research is a fundamental study based on the metallography about temperature estimation of the damaged tubes at the sodium-water reaction for the establishment of mechanism analysis technique of the behavior. In the examination, the material which gave the rapid thermal history which imitated sodium-water reaction was produced. And it was investigated whether the thermal history (i.e. maximum temperature and the holding time) of the samples could be presumed from the metallurgical examination of the samples. The major results are as follows: (1) The microstructure of the sample which was given the rapid thermal heating has reserved the influence of the maximum temperature and the time, and the structure can explain by referring to the equilibrium diagram and the continuous cooling transformation diagram. (2) Results of the electrolytic extraction of the samples, the ratio of the remained volume to the electrolyzed volume degreased with the increase of the maximum temperature and the time. Furthermore, it was observed the correlation between the remained volume of each element (Cr, Mo, Fe, V and Nb) and the thermal history. (3) It was obtained that the thermal history of the tubes damaged by sodium-water reaction might be able to be estimated from the metallurgical examinations. (author)

Energy balance and flow in steam generator part with sodium-water reaction

International Nuclear Information System (INIS)

Matal, O.

1980-01-01

Relations were derived for the calculation of heat liberated during the sodium water reaction in a tube failure in different parts of a steam generator. The results are graphically shown in i-T diagrams. Heat removal is described from the reaction zone to water and steam in undisturbed tubes and to the steam generator metal structure. (author)

Some new fatigue tests in high temperature water and liquid sodium environment

International Nuclear Information System (INIS)

Hattori, Takahiro; Yamauchi, Takayoshi; Kanasaki, Hiroshi; Kondo, Yoshiyuki; Endo, Tadayoshi.

1987-01-01

To evaluate the fatigue strength of structural materials for PWR or FBR plants, fatigue test data must be obtained in an environment of simulated primary and secondary water for PWR or of high temperature liquid sodium for FBR. Generally, such tests make it necessary to prepare expensive facilities, so when large amount of fatigue data are required, it is necessary to rationalize and simplify the fatigue tests while maintaining high accuracy. At the Takasago Research Development Center, efforts to rationalize facilities and maintain accuracy in fatigue tests have been made by developing new test methods and improving conventional techniques. This paper introduces a new method of low cycle fatigue test in high temperature water, techniques for automatic measurement of crack initiation and propagation in high temperature water environment and a multiple type fatigue testing machine for high temperature liquid sodium. (author)

The preparation, physicochemical properties, and the cohesive energy of liquid sodium containing titanium nanoparticles

International Nuclear Information System (INIS)

Saito, Jun-ichi; Itami, Toshio; Ara, Kuniaki

2012-01-01

Liquid sodium containing titanium nanoparticles (LSnanop) of 10-nm diameter was prepared by dispersing titanium nanoparticles (2 at.% Ti) into liquid sodium with the addition of stirring and ultrasonic sound wave. The titanium nanoparticles themselves were prepared by the vapor deposition method. This new liquid metal, LSnanop, shows a remarkable stability due to the Brownian motion of nanoparticles in liquid sodium medium. In addition, the difference of measured heat of reaction to water between this LSnanop and liquid sodium indicates the existence of cohesive energy between the liquid sodium medium and dispersed titanium nanoparticles. The origin of the cohesive energy, which serves to stabilize this new liquid metal, was explained by the model of screened nanoparticles in liquid sodium. In this model, negatively charged nanoparticles with transferred electrons from liquid sodium are surrounded by the positively charged screening shell, which may inhibit the gathering of nanoparticles by the “Coulombic repulsion coating.” The atomic volume of LSnanop shows the shrinkage from the linear law, which also suggests the existence of cohesive energy. The viscosity of LSnanop is almost the same as that of liquid sodium. This behavior was explained by the Einstein equation. The surface tension of LSnanop is 17 % larger than that of liquid sodium. The cohesive energy and the negative adsorption may be responsible to this increase. Titanium nanoparticles in liquid sodium seem to be free from the Coulomb fission. This new liquid metal containing nanoparticles suggests the possibility to prepare various stable suspensions with new properties.

Interparticle potential of 10 nanometer titanium nanoparticles in liquid sodium: Theoretical approach

Energy Technology Data Exchange (ETDEWEB)

Kim, Soo Jae; Park, Gun Yeop; Park, Hyun Sun; Baek, Je Hyun [POSTECH, Pohang (Korea, Republic of); Kim, Moo Hwan [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2015-10-15

A suspension of titanium nanoparticles (Ti NPs) in liquid sodium (Na) has been proposed as a method to mitigate the violent sodium-water reaction (SWR). The interparticle potential between Ti NPs in liquid Na may play a significant role in the agglomeration of NPs on the reaction surface and in the bulk liquid Na, since the potential contributes to a reduction in the long-term dispersion stability. For the effective control of the SWR with NPs, a physical understanding of the molecular dynamics of NPs in liquid Na is key. Therefore in this study, the nonretarded Van der Waals model and the solvation potential model are employed to analyze the interparticle potential. The ab initio calculation reveals that a strong repulsive force driven by the solvation potential exceeds the interparticle attraction and predicts the agglomeration energy required for two 10-nm Ti NPs to be 4 x 10{sup -17} J. The collision theory suggests that Ti NPs can be effective suppressors of the SWR due to the high energy barrier that prevents significant agglomeration of Ti NPs in quiescent liquid Na.

Numerical simulation code for combustion of sodium liquid droplet and its verification

International Nuclear Information System (INIS)

Okano, Yasushi

1997-11-01

The computer programs for sodium leak and burning phenomena had been developed based on mechanistic approach. Direct numerical simulation code for sodium liquid droplet burning had been developed for numerical analysis of droplet combustion in forced convection air flow. Distributions of heat generation and temperature and reaction rate of chemical productions, such as sodium oxide and hydroxide, are calculated and evaluated with using this numerical code. Extended MAC method coupled with a higher-order upwind scheme had been used for combustion simulation of methane-air mixture. In the numerical simulation code for combustion of sodium liquid droplet, chemical reaction model of sodium was connected with the extended MAC method. Combustion of single sodium liquid droplet was simulated in this report for the verification of developed numerical simulation code. The changes of burning rate and reaction product with droplet diameter and inlet wind velocity were investigated. These calculation results were qualitatively and quantitatively conformed to the experimental and calculation observations in combustion engineering. It was confirmed that the numerical simulation code was available for the calculation of sodium liquid droplet burning. (author)

A study of atomic interaction between suspended nanoparticles and sodium atoms in liquid sodium

International Nuclear Information System (INIS)

Saito, Jun-ichi; Ara, Kuniaki

2010-01-01

A feasibility study of suppression of the chemical reactivity of sodium itself using an atomic interaction between nanoparticles and sodium atoms has been carried out. We expected that the atomic interaction strengthens when the nanoparticle metal is the transition element which has a major difference in electronegativity from sodium. We also calculated the atomic interaction between nanoparticle and sodium atoms. It became clear that the atomic bond between the nanoparticle atom and the sodium atom is larger than that between sodium atoms, and the charge transfer takes place to the nanoparticle atom from the sodium atom. Using sodium with suspended nanoparticles, the fundamental physical properties related to the atomic interaction were investigated to verify the atomic bond. The surface tension of sodium with suspended nanoparticles increased, and the evaporation rate of sodium with suspended nanoparticles also decreased compared with that of sodium. Therefore the presence of the atomic interaction between nanoparticles and sodium was verified from these experiments. Because the fundamental physical property changes by the atomic interaction, we expected changes in the chemical reactivity characteristics. The chemical reaction properties of sodium with suspended nanoparticles with water were investigated experimentally. The released reaction heat and the reaction rate of sodium with suspended nanoparticles were reduced than those of sodium. The influence of the charge state of nanoparticle on the chemical process with water was theoretically investigated to speculate on the cause of reaction suppression. The potential energy in both primary and side reactions changed by the charge transfer, and the free energy of activation of the reaction with water increased. Accordingly, the reaction barrier also increased. This suggests there is a possibility of the reduction in the reaction of sodium by the suspension of nanoparticles. Consequently the possibility of the

Acoustic sodium-water reaction detection of the Phenix steam generators

International Nuclear Information System (INIS)

Carminati, M.; Martin, L.; Sauzaret, A.

1990-01-01

The systems for acoustic sodium-water reaction detection and hydrogen detection of the Phenix steam generators as well as systems for monitoring signals analysis and processing are described. It is reported that the results obtained during operation and calibration phases are very encouraging and that industrial equipment showing the same performance are being examined. 6 figs

Reactive wetting by liquid sodium on thin Au platin

International Nuclear Information System (INIS)

Kawaguchi, Munemichi; Hamada, Hirotsugu

2014-01-01

For practical use of an under-sodium viewer, the behavior of sodium wetting is investigated by modeling the reactive and non-reactive wetting of metallic-plated steels by liquid sodium to simulate sodium wetting. The non-reactive wetting simulation results showed good agreement with Tanner's law, in which the time dependencies of the droplet radius and contact angle are expressed as R N ∝ t 1/10 and θ∝ t -3/10 , respectively; therefore, the model was considered suitable for the simulation. To simulate reactive wetting, the model of fluid flow induced by the interfacial reaction was incorporated into the simulation of non-reactive wetting. The reactive wetting simulation results, such as the behavior of the precursor liquid film and central droplet, showed good agreement with sodium wetting experiments using thin Au plating at 250°C. An important result of the reactive wetting simulation is that the gradient of the reaction energy at the interface appeared on the new interface around the triple line, and that fluid flow was induced. This interfacial reactivity during sodium wetting of thin Au plating was enhanced by the reaction of sodium and nickel oxide through pinholes in the plating. (author)

Wastage Behavior of Modified 9Cr-1Mo Steel Tube Material by Sodium-Water Reaction

International Nuclear Information System (INIS)

Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Park, Nam Cook

2009-01-01

The development of a sodium-heated steam generator with safety and reliability is an essential requirement from the viewpoint of the economic efficiency of a sodium-cooled fast reactor. In most cases, these steam generators, which are in the process of development or operating, are of a shell-in tube type, with a high pressure water/steam inside the tubes and low pressure sodium on the shell-side, with a single wall tube as a barrier between these fluids. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. The objective of this study is a basic investigating of the sodium-water reaction phenomena by small water/steam leaks. For this, wastage tests for modified 9Cr-1Mo steel were conducted

Experimental investigation of solid sodium-water reaction: tests results and phenomenological analysis

International Nuclear Information System (INIS)

Daudin, K.; Beauchamp, F.; Proust, C.

2014-01-01

Sodium-Water Reaction (SWR) is an issue one has to be capable to deal with for the next generation of nuclear reactors (SFR for GEN IV). The background of these experiments is the improvement of safety demonstration regarding SWR in an open volume. This experimental campaign is conducted at the CEA Cadarache inside a cylindrical reactor filled with inert gas. The sodium is inside a loading pot and water comes into contact by immersion. SWR and its physical effects are followed by different pressure and temperature sensors. The results show a limit to the overpressure increasing sodium mass. Global assessment of physical effects of SWR contributes to put forward the relative nature of phenomena with geometric configuration, and the importance of scale effects. (authors)

Large sodium water reaction calculations in a LMFBR steam generator

International Nuclear Information System (INIS)

Finck, P.; Lepareux, M.; Schwab, B.; Blanchet, Y.

1986-05-01

The French approach to the analysis of large and violent sodium water reactions is presented. The basis for choosing the Design Basis Accident is discussed. An energetical analysis of the physical phenomena involved stresses the specific needs for computing tools. The feature of these tools are then described, and a validation test is presented. Finally, industrial applications are described. 8 refs

Study on fine particles influence on sodium sulfite and oxygen gas-liquid reaction

Energy Technology Data Exchange (ETDEWEB)

Tao, Shuchang; Zhao, Bo; Wang, Shujuan; Zhuo, Yuqun; Chen, Changhe [Tsinghua Univ., Beijing (China). Dept. of Thermal Engineering; Ministry of Education, Beijing (China). Key Lab. for Thermal Science and Power Engineering

2013-07-01

Wet limestone scrubbing is the most common flue gas desulfurization process for control of sulfur dioxide emissions from the combustion of fossil fuels, and forced oxidation is a key part of the reaction. During the reaction which controlled by gas-liquid mass transfer, the fine particles' characteristic, size, solid loading and temperature has a great influence on gas-liquid mass transfer. In the present work is to explain how these factors influence the reaction between Na{sub 2}SO{sub 3} and O{sub 2} and find the best react conditions through experiment. The oxidation rate was experimentally studied by contacting pure oxygen with a sodium sulfite solution with active carbon particle in a stirred tank, and the system pressure drop was record by the pressure sensor. At the beginning the pressure is about 215 kPa and Na{sub 2}SO{sub 3} is about 0.5mol/L. The temperature is 40, 50, 60, 70, 80 C. Compare the results of no particles included, we can conclude that high temperature, proper loadings and smaller particles resulting in higher mass transfer coefficients k{sub L}.

An experimental study on impingement wastage of Mod 9Cr 1Mo steel due to sodium water reaction

Energy Technology Data Exchange (ETDEWEB)

Kishore, S., E-mail: skishore@igcar.gov.in [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Ashok Kumar, A.; Chandramouli, S.; Nashine, B.K.; Rajan, K.K.; Kalyanasundaram, P.; Chetal, S.C. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2012-02-15

Highlights: Black-Right-Pointing-Pointer Sodium heated steam generators are crucial components of fast breeder reactors. Black-Right-Pointing-Pointer A leak in steam generator tube will cause sodium water reaction that damages the tubes. Black-Right-Pointing-Pointer Experimental study was conducted to quantify the extent of damage on Mod 9Cr 1Mo tube due to a water leak. - Abstract: Sodium heated steam generator (SG) is a crucial component in the heat transport system of a fast breeder reactor (FBR). In case, one of its water/steam carrying tubes becomes defective, water/steam leaks into sodium, flowing in the shell side, causing sodium-water reaction, which is highly exothermic and producing corrosive NaOH. The reaction jet originating from a leaking tube may impinge on its adjacent tube, resulting in damage of the tube. Impingement wastage refers to this kind of damage, occurring to a tube of sodium heated SG, owing to a small water/steam leak from a neighboring tube. Extensive research works have been conducted all over the world to study various aspects of this phenomenon. Experimental studies were carried out in Indira Gandhi Centre for Atomic Research (IGCAR) to understand the effect of impingement wastage on Mod 9Cr 1Mo, which is the tube material of prototype fast breeder reactor (PFBR) SG. This paper brings out the data and experience gained through the experiments.

Analysis of the sodium concrete interactions with the NABE code

International Nuclear Information System (INIS)

Soule, N.

1989-01-01

Experimental studies have been performed in France to investigate sodium-concrete interactions: thermal decomposition of concrete, specific chemical reactions, experimentation in liquid and vapour phase, sodium-concrete interaction without liner protection. Simultaneously computer codes have been developed in order to study the response of the containment building of a liquid metal fast breeder reactor to a sodium pool fire worsened by a sodium-concrete interaction: the NABE code. This code takes into account: a) sodium combustion; b) thermal decomposition of concrete with associated chemical reactions: (liquid sodium-vapour water reaction, liquid sodium-carbon dioxide reaction, liquid sodium-solid compounds of concrete, hydrogen combustion); c) chemical reactions in vapour phase; d) decay heat; e) gas aerosol inlets/outlets; f) aerosol behaviour (sedimentation, diffusion, leak); g) thermal exchanges. An example of a situation, typical of assessment of beyond design basis situations in LMFBR, is given. (author)

An experimental study on sodium-water reaction in the double pool LMFBR, (4)

International Nuclear Information System (INIS)

Kumagai, Hiromichi; Yoshida, Kazuo; Uotani, Masaki; Akimoto, Tokuzo

1989-01-01

Double Pool type LMFBR set the rectangular cross-sectional steam generator (SGs) inside a secondary vessel. The initial spike pressure rise caused by large sodium-water reaction in SGs might be radiated into a large sodium pool in the secondary vessel. Therefore basic experiments on pressure wave propagation were carried out by generating pressure wave in water by mean of a set of drop hummer and piston. But the experimental apparatus in water was not convenience to simulate the structure near the bottom end of the SGs shell. In this reports, experiments were carried out by generating pulse sound pressure in air, and compared with the results pressure waves in water. (author)

Activity of NaOH buffered by silicate solids in molten sodium acetate-water at 3170C

International Nuclear Information System (INIS)

Weres, O.; Tsao, L.

1988-01-01

Silica and sodium acetate are present in the steam generator tube sheet crevices of many nuclear power plants. Trace solutes in the condensate are tremendously concentrated in the crevices by boiling. Sparingly soluble sodium silicates and other solids precipitate from the crevice liquid leaving an extremely concentrated molten mixture of water, sodium acetate and other salts. The precipitates buffer the activity of sodium hydroxide in the superheated liquid that remains. The activity of NaOH corresponding to the buffers quartz/sodium disilicate and sodium disilicate/sodium metasilicate at 317 0 C has been determined experimentally. The sodium hydroxide content of a sodium acetate-water melt buffered by these reactions was determined by chemical analysis, and the corresponding activity of NaOH at temperature was calculated using the recently published Pitzer-Simonson Model of molten salt-water mixtures. The molten mixture of sodium acetate and water plays the role solvent in these experiments and calculations. The free energies of formation of solid sodium silicates at 317 0 C were also determined. The activity of NaOH corresponding to other silicate and phosphate buffers was calculated using published thermodynamic data and estimated from phase diagrams

Sodium water reaction R and D for French LMFBR

International Nuclear Information System (INIS)

Cambillard, E.; Finck, P.; Lapicore, A.; Simeon, C.

1985-01-01

This paper presents the research and development which is underway for the French LMFBR steam generator safety study. The program comprises three major areas: (1) the analysis of realistic leaks, which includes the leak evolution and its consequences; (2) the response time of leak detection systems compared to leak propagation phenomena; and (3) the guillotine rupture (DBA) studies relative to source term evaluation by experimental/calculational approach and mechanical calculations. This program has provided information for the demonstrations of the steam generator safety in respect to a sodium-water reaction

Preliminary Study of Feasibility of Acoustic Detection of Small Sodium-Water Reactions in Lmfbr Steam Generators.

Energy Technology Data Exchange (ETDEWEB)

Chamberlain, H. V.

1970-06-01

An evaluation of acoustic techniques for the detection of sound resulting from water leaks into liquid sodium was conducted. Acoustic spectra over the range of 100Hz to 80kHz were detected. A leak as small as 0.00008 lb/ sec was detected by the acoustic instrumentation. The indicated leak rate was the smallest that was injected into the liquid sodium, and is not to be interpreted as the smallest leak rate that might be detected by the acoustic method. Detection time was found to be essentially instantaneous.

Comparison of cavitation tests on the SNR 300 prototype sodium pump, carried out using water at room temperature and liquid sodium at 5800C

International Nuclear Information System (INIS)

Fakkel, R.H.; Hoornweg, C.J.; Kamerling, B.; Ten Wolde, Tj.; Heslenfeld, M.W.; Mendte, W.K.; Bunjies, J.H.

1976-01-01

This paper gives results of tests carried out on a centrifugal pump in both a water and a sodium test facility. The pump tested is a prototype of the primary circulation pumps intended for a LMFBR 300 MWe nuclear power station (the SNR reactor) under construction at Kalkar, West Germany. The pump characteristics under various NPSH-conditions were investigated, and a comparison is made in the paper between the results of cavitation tests using water and liquid sodium. An attempt is made to account for differences in pump characteristics in both types of tests by referring to the physical properties of the liquids used (water and sodium). An attempt is also made to correlate the results obtained with the full-scale prototype to those obtained previously with a half-scale model of the impeller. The various test circuits used should be made identical, in order to avoid that differences in geometry should obscure essential features resulting from using different liquids. Yet, it showed that in this respect, the tests did not obey to this key-rule, reasons why the test results did not fully reveal the essential physical properties of either fluid under cavitating conditions. (author)

Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Sang June Ahn

2016-08-01

Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

Sodium/water reaction detection confirmation and location with time domain beam former

International Nuclear Information System (INIS)

Cornu, C.

1997-01-01

The CEA studied the validity of a time beamforming method for the detection and location of Sodium/water reaction in steam generators of breeder reactors. In the context of the RCM, we apply this method on recorded data during a water injection in Sodium in ASB loop, artificially mixed with PFR background. Despite the severity of experiment conditions (the signal to noise ratio is between -6 and -24 dB). We show that the employed method completed with a low frequency pass band filter allows us to locate the injection with a precision of 30% of the diameter of the loop. Using the method in the course of time allows us to coarsely locate the start time and the duration of the leak. The good functioning of the method is however perturbed by uncertainty about the wave celebrity in the sodium about wave propagation in waves guides that are mounted with the sensors and in the structure of the loop. (author). 1 ref., 8 figs

The improvement of the heat transfer model for sodium-water reaction jet code

International Nuclear Information System (INIS)

Hashiguchi, Yoshirou; Yamamoto, Hajime; Kamoshida, Norio; Murata, Shuuichi

2001-02-01

For confirming the reasonable DBL (Design Base Leak) on steam generator (SG), it is necessary to evaluate phenomena of sodium-water reaction (SWR) in an actual steam generator realistically. The improvement of a heat transfer model on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.40) and application analysis to the water injection tests for confirmation of propriety for the code were performed. On the improvement of the code, the heat transfer model between a inside fluid and a tube wall was introduced instead of the prior model which was heat capacity model including both heat capacity of the tube wall and inside fluid. And it was considered that the fluid of inside the heat exchange tube was able to treat as water or sodium and typical heat transfer equations used in SG design were also introduced in the new heat transfer model. Further additional work was carried out in order to improve the stability of the calculation for long calculation time. The test calculation using the improved code (LEAP-JET ver.1.50) were carried out with conditions of the SWAT-IR·Run-HT-2 test. It was confirmed that the SWR jet behavior on the result and the influence to the result of the heat transfer model were reasonable. And also on the improved code (LEAP-JET ver.1.50), user's manual was revised with additional I/O manual and explanation of the heat transfer model and new variable name. (author)

Study on chemical reactivity control of liquid sodium. Research program

International Nuclear Information System (INIS)

Saito, Jun-ichi; Ara, Kuniaki; Sugiyama, Ken-ichiro; Kitagawa, Hiroshi; Oka, Nobuki; Yoshioka, Naoki

2007-01-01

Liquid sodium has the excellent properties as coolant of the fast breeder reactor (FBR). On the other hand, it reacts high with water and oxygen. So an innovative technology to suppress the reactivity is desired. The purpose of this study is to control the chemical reactivity of liquid sodium by dispersing the nanometer-size metallic particles (we call them Nano-particles) into liquid sodium. We focus on the atomic interaction between Nano-particles and sodium atoms. And we try to apply it to suppress the chemical reactivity of liquid sodium. Liquid sodium dispersing Nano-particles is named 'Nano-fluid'. Research programs of this study are the Nano-particles production, the evaluation of reactivity suppression of liquid sodium and the feasibility study to FBR plant. In this paper, the research programs and status are described. The important factors for particle production were understood. In order to evaluate the chemical reactivity of Nano-fluid the research programs were planned. The feasibility of the application of Nano-fluid to the coolant of FBR plant was evaluated preliminarily from the viewpoint of design and operation. (author)

SOCON: a computer model for analyzing the behavior of sodium-concrete reactions

International Nuclear Information System (INIS)

Nguyen, D.G.; Muhlestein, L.D.

1985-03-01

Guided by experimental evidence available to date, ranging from basic laboratory studies to large scale tests, a mechanistic computer model (the SOCON model) has been developed to analyze the behavior of SOdium-CONcrete reactions. The model accounts for the thermal, chemical and mechanical phenomena which interact to determine the consequences of the reactions. Reaction limiting mechanisms could be any process which reduces water release and sodium transport to fresh concrete; the buildup of the inert reaction product layer would increase the resistance to sodium transport; water dry-out would decrease the bubble agitation transport mechanism. However, stress-induced failure of concrete, such as spalling, crushing and cracking, and a massive release of gaseous products (hydrogen, water vapor and CO 2 ) would increase the transport of sodium to the reaction zone. The results of SOCON calculations are in excellent agreement with measurements obtained from large-scale sodium-limestone concrete reaction tests of duration up to 100 hours conducted at the Hanford Engineering Development Laboratory. 8 refs., 7 figs

Development of analysis model for mid and long-term effects of sodium water reaction event in LMR

International Nuclear Information System (INIS)

Eoh, Jae Hyuk; Sim, Yoon Sub; Kim, Seong O; Kim, Yeon Sik; Kim, Eui Kwang; Wi, Myung Hwan

2002-04-01

The Sodium-Water Reaction(SWR) is important in the design consideration of a LMR steam generator. To develop the analysis code for long-term effects of SWR, investigation on the characteristics of various SWR analysis code and the assessment of an analysis model for long term effects were performed. In an event of SWR, pressure spikes of wave propagation occur at its initial stage and last for a very short time, and then bulk motion of fluid and reaction products is progressed and lasts for a long time. In a case SWR occurs, a number of hydrogen bubbles produced and sodium is entrained into the bubbles through the gas-liquid bubble interfaces by evaporation or diffusion. The partial pressure of the sodium in a hydrogen bubble is determined as a function of the bubble size, temperature, and pressure, and is rapidly decreased as its size increased. From this, it can be considered that the bulk motion in the later phase of SWR is an axial motion caused by expansion of a single-phase hydrogen gas bubble produced by a reaction in the vicinity of the leak site. Through this investigation, a preliminary simple analysis model for long-term effects of SWR was set up and sensitivity study using the system design parameters such as pressure and temperature of IHTS for KALIMER was performed. Also, a simpler analysis model using the cover gas pressure change related to the production of a hydrogen bubble in a steam generator was developed from the analyses results. These simple analysis models of the reaction site and the pressure behavior with hydrogen production can be used to develop the mid and long-term analysis code for SWR in the KALIMER steam generator design

Large-leak sodium-water reaction analysis for steam generators

International Nuclear Information System (INIS)

Sakano, K.; Shindo, Y.; Hori, M.

1975-01-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

Large-leak sodium-water reaction analysis for steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakano, K; Shindo, Y; Hori, M

1975-07-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

Large scale sodium-water reaction tests for Monju steam generators

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Hori, M.

1976-01-01

To demonstrate the safe design of the steam generator system of the prototype fast reactor Monju against the postulated large leak sodium-water reaction, a large scale test facility SWAT-3 was constructed. SWAT-3 is a 1/2.5 scale model of the Monju secondary loop on the basis of the iso-velocity modeling. Two tests have been conducted in SWAT-3 since its construction. The test items using SWAT-3 are discussed, and the description of the facility and the test results are presented

High-Throughput Determination of Sodium Danshensu in Beagle Dogs by the LCMS/MS Method, Employing Liquid-Liquid Extraction Based on 96-Well Format Plates

Directory of Open Access Journals (Sweden)

Jingjing Jiang

2017-04-01

Full Text Available Sodium Danshensu (sodium d-(+-β-(3,4-dihydroxyphenyl lactate, one of the water-soluble ingredients in Salvia miltiorrhiza, exhibits potent relaxation of the coronary artery and anticoagulation effection. A high-throughput, rapid, and sensitive method combining liquid chromatography with electrospray ionization tandem mass spectrometry to determine the sodium danshensu in beagle dog plasma was developed and validated, using gallic acid as an internal standard (IS. Acidified plasma samples were extracted using 96-well liquid-liquid extraction, and were eluted on a CNW Athena C18 column (3 μm, 2.1 × 100 mm by using a gradient mobile phase system of methanol and water (containing 0.2% formic acid. The mass spectrometric detection was achieved using negative ion electrospray ionization mode and monitoring the precursor→production combinations of m/z 197→135 for sodium danshensu and 169→125 for IS, in multiple reaction monitoring modes. Good linearity was achieved, and the linear range was 10–1000 ng/mL (R2 > 0.996 with a quantification limit of 10 ng/mL for sodium danshensu in beagle dog plasma. The intra- and inter-day precision (RSD ranged from 2.1% to 9.0%. The accuracy (RE was between −8.6% and 5.7% at all quality control levels. The validated method was successfully applied to the pharmacokinetics study of sodium danshensu in beagle dog plasma after intravenous injection and oral administration of sodium danshensu.

High-Throughput Determination of Sodium Danshensu in Beagle Dogs by the LCMS/MS Method, Employing Liquid-Liquid Extraction Based on 96-Well Format Plates.

Science.gov (United States)

Jiang, Jingjing; Zhao, Xin; Li, Xiuxiu; Wu, Shengyuan; Yu, Shidan; Lou, Yuefen; Fan, Guorong

2017-04-25

Sodium Danshensu (sodium d-(+)-β-(3,4-dihydroxyphenyl) lactate), one of the water-soluble ingredients in Salvia miltiorrhiza , exhibits potent relaxation of the coronary artery and anticoagulation effection. A high-throughput, rapid, and sensitive method combining liquid chromatography with electrospray ionization tandem mass spectrometry to determine the sodium danshensu in beagle dog plasma was developed and validated, using gallic acid as an internal standard (IS). Acidified plasma samples were extracted using 96-well liquid-liquid extraction, and were eluted on a CNW Athena C18 column (3 μm, 2.1 × 100 mm) by using a gradient mobile phase system of methanol and water (containing 0.2% formic acid). The mass spectrometric detection was achieved using negative ion electrospray ionization mode and monitoring the precursor→production combinations of m / z 197→135 for sodium danshensu and 169→125 for IS, in multiple reaction monitoring modes. Good linearity was achieved, and the linear range was 10-1000 ng/mL (R² > 0.996) with a quantification limit of 10 ng/mL for sodium danshensu in beagle dog plasma. The intra- and inter-day precision (RSD) ranged from 2.1% to 9.0%. The accuracy (RE) was between -8.6% and 5.7% at all quality control levels. The validated method was successfully applied to the pharmacokinetics study of sodium danshensu in beagle dog plasma after intravenous injection and oral administration of sodium danshensu.

Experiences on removal of sodium-water reaction products in SWAT-3

International Nuclear Information System (INIS)

Tanabe, H.; Hiroi, H.; Sato, M.; Otaka, J.

2002-01-01

This report summarizes experiences and information concerning the removal of sodium water reaction products (SMRP) obtained through large leak tests of the Steam Generator Safety Test Facility (SWAT-3) at PNC/OEC, which were conducted to validate the safety design of steam generators of a prototype LMFBR Monju. The following three problems are discussed here: (1) drainability of SWRP, (2) removal of SWRP by using a cold trap, and (3) steam cleaning of SWRP. (author)

Outline of sodium-water reaction test in case of large leak with SWAT-3 testing equipments

International Nuclear Information System (INIS)

Sato, Minoru

1978-01-01

The key component in sodium-cooled fast reactors in steam generators, and the sodium-water reaction owing to the break of heating tubes may cause serious damages in equipments and pipings. The main factor controlling this phenomenon is the rate of leak of water. When the rate of water leak is small, the propagation of heating tube breaking may occur owing to ''wastage phenomenon'', on the other hand, when the rate of water leak is large, the phenomena of explosive pressure and flow occur due to the reaction heat and a large quantity of hydrogen generated by the reaction. In PNC, the testing equipments of SWAT-2 for small water leak and SWAT-1 for large leak were constructed, and the development test has been carried out to establish the method of safety design experimentally. The synthetic test equipment for the safety of steam generators, SWAT-3, was constructed to carry out the large water leak test in the scale close to actual plants. The object of the test, the outline of the test equipment, the phenomena of pressure and flow in the water injection test, the confirmation of the occurrence of secondary breaking of adjacent heating tubes, and the disposal of reaction products are described in this paper. This test is till going on, and the final conclusion will be reported later. (Kako, I.)

Effect of liquid sodium thermochemical reactions with stainless steels on mechanical response

International Nuclear Information System (INIS)

Subbaraman, G.; Reifsnider, K.L.

1976-01-01

An analytical approach is presented to study the elastic response of an LMFBR fuel clad subject to strong property degradation in liquid sodium environment and internal poison pressure, at the gas plenum section of the fuel pin. The nature and analysis of the thermochemical reactions and the available experimental data are briefly reviewed. Property variations in the radial direction due to these inservice reactions are represented by a generalized, continuous function, f(r), which is introduced into the constitutive equations of stress equilibrium. By introducing an adjustable constant A into a specific form of f(r), various severities in the variation of the elastic parameters through the thickness of the clad are accommodated and particular analytical solutions are obtained for the displacement. Closed form solutions, which are available for particular values of A, are used as validity checks in computing the more general solutions. The stress-strain responses including the deviatoric stresses used in creep analyses are presented for several cases. The merits of the analytical formulation of the problem and the need to include similar approaches in fuel performance calculations are stated

Wastage-resistant characteristics of 12Cr steel tube material. Small leak sodium-water reaction test

International Nuclear Information System (INIS)

Shimoyama, Kazuhito

2004-03-01

In the water leak accident of a steam generator designed for a sodium cooled reactor in the Feasibility Study, the localization of tube failure propagation by using an advanced water leak detector will be required from the viewpoints of the safety and economical efficiency of the plant. So far, the conventional knowledge and analytical tools have been used in the investigation and evaluation of water leak phenomenon; nevertheless, there was neither test data nor the study of quantitative evaluation on the corrosion behavior, so-called wastage-resistant characteristics, of 12Cr steel tube material in sodium-water reactions. Wastage tests for the 12Cr steel tube material were conducted in small water leaks by use of the Sodium-Water Reaction Test Rig (SWAT-1R), and the data of wastage rate were obtained in the parameter of water leak rate under the constant sodium temperature and distance between leak and target tubes. The test results lead to the following conclusions: (1) The wastage-resistibility of 12Cr steel is 1.6 times greater than that of 9Cr steel and is 2.7 times greater than that of 2.25Cr-1Mo steel. (2)The wastage-resistibility of 12Cr steel increases in smaller water leaks; especially in water leak rates of 1 g/sec or less, it is more excellent than that of SUS321 stainless steel used as Monju superheater tube material. (3) Based on the correlation of wastage rate for the 9Cr steel, the correlation for the 12Cr steel has been obtained to be used for the evaluation of tube failure propagation. As the correlation of wastage rate for the 12Cr steel is based on the correlation for the 9Cr steel, it gives enough conservatism in smaller water leaks. To serve in accurately evaluating the tube failure propagation in smaller water leaks, it is necessary to obtain new correlation of wastage rate for the 12Cr steel based on the data in the wide range of water leak rates. (author)

The improving of sodium cleaning device with atomized water

International Nuclear Information System (INIS)

Guo Huanfang; Zhang Xiufeng; Yu Chunli; Hong Shunzhang; Xing Chaoqing; Yuan Waimei

1999-04-01

The atomized water in N 2 under pressure can react with sodium on the sodium-soiled surfaces under control. This method has the following main advantages: The reaction proceeds at room temperature; high efficiency; fast reaction velocity; small quantity of water; the mechanical and/or corrosion damage by sodium reaction products can be avoided. The results of washing fuel element of CEFR (China Experimental Fast Reactor) were satisfactory

Water quality for liquid wastes

International Nuclear Information System (INIS)

Mizuniwa, Fumio; Maekoya, Chiaki; Iwasaki, Hitoshi; Yano, Hiroaki; Watahiki, Kazuo.

1985-01-01

Purpose: To facilitate the automation of the operation for a liquid wastes processing system by enabling continuous analysis for the main ingredients in the liquid wastes accurately and rapidly. Constitution: The water quality monitor comprises a sampling pipeway system for taking out sample water for the analysis of liquid wastes from a pipeway introducing liquid wastes to the liquid wastes concentrator, a filter for removing suspended matters in the sample water and absorption photometer as a water quality analyzer. A portion of the liquid wastes is passed through the suspended matter filter by a feedpump. In this case, sulfate ions and chloride ions in the sample are retained in the upper portion of a separation color and, subsequently, the respective ingredients are separated and leached out by eluting solution. Since the leached out ingredients form ferric ions and yellow complexes respectively, their concentrations can be detected by the spectrum photometer. Accordingly, concentration for the sodium sulfate and sodium chloride in the liquid wastes can be analyzed rapidly, accurately and repeatedly by which the water quality can be determined rapidly and accurately. (Yoshino, Y.)

GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

Energy Technology Data Exchange (ETDEWEB)

S. A. Eide; S. V. Chmielewski; T. D. Swantz

1990-02-01

A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

Designing solid-liquid interphases for sodium batteries

KAUST Repository

Choudhury, Snehashis

2017-10-06

Secondary batteries based on earth-abundant sodium metal anodes are desirable for both stationary and portable electrical energy storage. Room-temperature sodium metal batteries are impractical today because morphological instability during recharge drives rough, dendritic electrodeposition. Chemical instability of liquid electrolytes also leads to premature cell failure as a result of parasitic reactions with the anode. Here we use joint density-functional theoretical analysis to show that the surface diffusion barrier for sodium ion transport is a sensitive function of the chemistry of solid–electrolyte interphase. In particular, we find that a sodium bromide interphase presents an exceptionally low energy barrier to ion transport, comparable to that of metallic magnesium. We evaluate this prediction by means of electrochemical measurements and direct visualization studies. These experiments reveal an approximately three-fold reduction in activation energy for ion transport at a sodium bromide interphase. Direct visualization of sodium electrodeposition confirms large improvements in stability of sodium deposition at sodium bromide-rich interphases.

Safety Evaluation for IHTS Integrity due to the Steam Generator Sodium-Water Reaction Event in the PGSFR

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang-Jun; Lee, Kwi Lim; Ha, Kwi-Seok; Lee, Seung Won; Jeong, Taekyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this paper, the integrity of the IHTS and SG by the SWR event are evaluated using the SWAAMII code. A sodium has a chemical characteristics to rigorously react the water or steam and produce the high pressure waves and high temperature reaction heat. It has an excellent characteristics as a reactor coolant. But, there is an event to be considered in the sodium cooled fast reactor design. The Sodium-Water Reaction (SWR) event can be occurred by the water or steam leaks due to the break of the steam generator tubes. The propagated high pressure waves threathen the structural integrity of the affected Intermediate Heat Transport System (IHTS) and steam generator. If the IHTS pipes are failed, the sodium of the IHTS can be released to the containment building. To the peak pressure point of view, it is performed to evaluate the integrity of the major components due to the SWR event in the SG. The generated peak pressures due to the five SG tubes simultaneous break event are within the range of the design pressure for the SG, IHX and IHTS including the related pipes.

Analysis of acoustic data from UK sodium/water reaction test facilities

International Nuclear Information System (INIS)

Rowley, R.; Mcknight, J.A.; Airey, J.

1990-01-01

This paper describes acoustic measurements made during a number of sodium/water reaction experiments in the UK. The tests have included water and steam injections through both realistic (fatigue crack) defects and machined orifices and have covered a range of experimental conditions including those appropriate to the inlet and outlet regions of the EFR steam generators. Injection rates were typically in the range 0.1 to 30 g/s. Where possible, gas injections were also included in the test programme for comparison, since it is anticipated that a practical SGU acoustic leak detection system would include a facility for gas injections to allow system calibration, and to confirm transmission properties within the SGU. The test sections were instrumented with accelerometers on waveguides and in some cases included an under-sodium microphone situated about 300mm above the reaction zone. Tape recordings were made during the tests and used for detailed analysis off-line, although an audible output from one of the acoustic channels was used to monitor the progress of the injections and provide information for the rig operators. A comparison of the signal amplitudes measured during the experiments with typical reactor background noise was made and an estimate of the detection sensitivity of an acoustic monitoring system was deduced. 3 refs, 5 figs, 1 tab

A fundamental study on sodium-water reaction in the double-pool-type LMFBR

International Nuclear Information System (INIS)

Yoshida, Kazuo; Akimoto, Tokuzo

1987-01-01

In order to evaluate the pressure rise by large sodium-water reaction in the Double-Pool LMFBR, basic tests on pressure wave celerity in rectangular tube are carried out. The initial spike pressure in rectangular-shelled steam generator of the Double Pool reactor, strongly depends on pressure wave celerity. In this study, celerity was measured as a function of pressure wave rising time and pulse height, and influence of water around the test section on celerity was investigated. (author)

Sodium-water reaction studies for MONJU steam generators

International Nuclear Information System (INIS)

Hori, M.; Sato, M.; Nei, H.; Harasaki, T.; Hishida, M.; Saito, T.

1975-01-01

The R and D results of the PNC's sodium-water reaction project are reviewed. The purposes of the project with the specific object for each test rig and computer code are given. The test items which should be investigated for the safety evaluation of the MONJU steam generators are discussed, and the status of the PNC's work on each item is described. The results on the small-leak wastage measurement are shown and the improved experimental equations to predict the wastage rate from the leak rate and the sodium temperature are given. The preliminary results on the wastage of tube bundle in the intermediate leak range are shown. The depth and the area of the wastage and also the wastage rate for each tube are shown graphically. The measured peak value of the initial pressure spike for the large leak is shown. The scatter of the data and its causes are discussed. The bubble growth rate estimated from the void probe measurement is presented. The results of the simulation experiment on the pressure wave propagation to the secondary circuit are given, comparing them with the prediction by the one-dimensional computer codes SWAC-5K and SWAC-5H. (author)

Sodium-water reaction studies for MONJU steam generators

Energy Technology Data Exchange (ETDEWEB)

Hori, M; Sato, M; Nei, H; Harasaki, T; Hishida, M; Saito, T

1975-07-01

The R and D results of the PNC's sodium-water reaction project are reviewed. The purposes of the project with the specific object for each test rig and computer code are given. The test items which should be investigated for the safety evaluation of the MONJU steam generators are discussed, and the status of the PNC's work on each item is described. The results on the small-leak wastage measurement are shown and the improved experimental equations to predict the wastage rate from the leak rate and the sodium temperature are given. The preliminary results on the wastage of tube bundle in the intermediate leak range are shown. The depth and the area of the wastage and also the wastage rate for each tube are shown graphically. The measured peak value of the initial pressure spike for the large leak is shown. The scatter of the data and its causes are discussed. The bubble growth rate estimated from the void probe measurement is presented. The results of the simulation experiment on the pressure wave propagation to the secondary circuit are given, comparing them with the prediction by the one-dimensional computer codes SWAC-5K and SWAC-5H. (author)

A study on sodium-concrete reaction

Energy Technology Data Exchange (ETDEWEB)

Bae, Jae Heum; Min, Byong Hun [Suwon University, Suwon (Korea, Republic of)

1997-07-01

A small sodium-concrete reaction facility was designed, manufactured and installed. this facility has been operated under inert gas(N{sub 2}) with different experimental variables such as sodium injection temperature, injection amount of sodium, aging period of concrete, sodium reservoir temperature. As a result, it was found that sodium injection temperature and injected amount of sodium has little effect on sodium-concrete reaction. However, sodium reservoir temperature and aging period of concrete has relatively high impact on sodium-concrete reaction. Sodium-concrete reaction model has also been developed and compared with experimental results. (Author) 51 refs., 16 tabs., 64 figs.

The removal of dinitrochlorobenzene from industrial residuals by liquid-liquid extraction with chemical reaction

Directory of Open Access Journals (Sweden)

G. C. M. Ferreira

2007-09-01

Full Text Available Nitrochlorobenzenes (NCBs are very important in the chemical industry since they have been used as raw material for the manufacture of crop protection products, as active ingredients in the pharmaceutical industry, as pigments and as antioxidants as well as for other uses. In industrial processes, NCBs are produced by monochlorobenzene (MCB nitration reactions and one of the main residuals formed is dinitrochlorobenzene (DNCB, which is mainly composed of the isomer 2,4DNCB. This subproduct, although of commercial interest when in its pure state, is generally incinerated due to the high costs of recovery treatment and purification. The objective of this study is to present an alternative to the treatment of industrial residuals containing DNCB. The technique consists of converting DNCB into sodium dinitrophenolate, which is very soluble in water and is also easy to reuse. For this purpose, liquid-liquid extraction with chemical reaction (alkaline hydrolysis with a rotating disc contactor (RDC is used. Experimental data on MCB nitration reactions as well as alkaline hydrolysis using a rotating disc contactor are presented.

Some conditions affecting the definition of design basis accidents relating to sodium/water reactions

International Nuclear Information System (INIS)

Bolt, P.R.

1984-01-01

The possible damaging effects of large sodium/water reactions on the steam generator, IHX and secondary circuit are considered. The conditions to be considered in defining the design basis accidents for these components are discussed, together with some of the assumptions that may be associated with design assessments of the scale of the accidents. (author)

Wastewater reuse in liquid sodium silicate manufacturing in alexandria, egypt.

Science.gov (United States)

Ismail, Gaber A; Abd El-Salam, Magda M; Arafa, Anwar K

2009-01-01

Soluble sodium silicates (waterglass) are liquids containing dissolved glass which have some water like properties. They are widely used in industry as sealants, binders, deflocculants, emulsifiers and buffers. Their most common applications in Egypt are in the pulp and paper industry (where they improve the brightness and efficiency of peroxide bleaching) and the detergent industry, in which they improve the action of the detergent and lower the viscosity of liquid soaps. The survey results showed that the production was carried out batch-wise, in an autoclave (dissolver). Sodium silicate in the state of crushed glass was charged in an autoclave (dissolver) with sodium hydroxide and water. The product is filtered through a press. The left over sludge (mud and silicates impurities) is emptied into the local sewer system. Also, sludge (silica gel) was discharged from the neutralization process of the generated alkaline wastewater and consequently clogging the sewerage system. So this study was carried out to modify the current wastewater management system which eliminates sludge formation, the discharge of higher pH wastewater to the sewer system, and to assess its environmental and economic benefits. To assess the characteristics of wastewater to be reused, physico-chemical parameters of 12 samples were tested using standard methods. The survey results showed that a total capacity of the selected enterprise was 540 tons of liquid sodium silicates monthly. The total amount of wastewater being discharged was 335 m3/month. Reusing of wastewater as feed autoclave water reduced water consumption of 32.1% and reduced wastewater discharge/month that constitutes 89.6% as well as saving in final product of 6 ton/month. It was concluded that reusing of wastewater generated from liquid sodium silicate manufacturing process resulted in cheaper and environmental-friendly product.

Contribution to the prediction of sodium-water reactions effects: application to confinement losses inside a steam generator building of a sodium fast reactor

International Nuclear Information System (INIS)

Daudin, Kevin

2015-01-01

Study of sodium-water reaction (SWR) consequences in open air represents a challenge in the frame of safety assessments of sodium fast reactors (SFR). In case of major accident and to predict consequences of SWR, it is necessary to better appreciate phenomena and especially quantity and rate of the energy release. The objective is thus to strengthen the understanding of such reactions in order to predict with lore accuracy its consequences on mechanical equipment in the surroundings. This work focuses on three areas : research of accidental sequences, experimental investigation, and phenomenological analysis before the explosive contact. At first, a tree structure risk analysis with calculations of dangerous phenomena permitted to suggest how the contact between reactants may happen. Then, demonstrative experimental studies were performed to deepen some practical aspects of the phenomenology, like the influence of the way the reactants get in contact. Data analysis conducted to the development of a phenomenological model, implemented into a software platform for numerical simulations. Although numerous hypothesis, transient heat transfer consideration enables to reproduce experimental observations, especially the influence of mixing conditions (sodium mass and initial temperatures) on the phenomenology. This study of the premixing step of sodium-water explosion is relevant in the frame of current prediction methods of mechanical loadings on structures. (author) [fr

Sodium concrete reaction - Structural considerations

International Nuclear Information System (INIS)

Ferskakis, G.N.

1984-01-01

An overview of the sodium concrete reaction phenomenon, with emphasis on structural considerations, is presented. Available test results for limestone, basalt, and magnetite concrete with various test article configurations are reviewed. Generally, tests indicate reaction is self limiting before all sodium is used. Uncertainties, however, concerning the mechanism for penetration of sodium into concrete have resulted in different theories about a reaction model. Structural behavior may be significant in the progression of the reaction due to thermal-structuralchemical interactions involving tensile cracking, compressive crushing, or general deterioration of concrete and the exposure of fresh concrete surfaces to react with sodium. Structural behavior of test articles and potential factors that could enhance the progression of the reaction are discussed

An Investigation of Tube Wastage Caused by Small Leaks of Water in Sodium-Heated Steam Generators

Energy Technology Data Exchange (ETDEWEB)

Muenchow, H. O.; Chase, W. L. [Atomic Power Development Associates, Inc., Detroit, MI (United States)

1967-06-15

Examination of the results from a sodium-water reaction that occurred in a steam generator in the Enrico Fermi Plant revealed a generalized corrosion in the area of the reaction, producing failure of some of the adjacent tubes. The questions raised by this experience clearly indicated that the problem of small leaks causing general tube wastage should be investigated. An experimental programme was designed and divided into two categories: (1) basic studies to establish the mechanisms of wastage, including wasted surface temperature measurements, chemical attack and hot erosion; (2) studies to establish the effect of design and operating parameters on wastage occurring in a test rig that simulates the dynamics of a commercial steam generator. A small test rig supporting a 4-ft-high reaction vessel containing 5 gal of sodium was constructed. Wasted surface temperatures measured in this rig using thermocouples as the target for the sodium - water reaction have indicated temperatures in excess of 2200 Degree-Sign F. Corrosion studies, using the reaction products NaOH and Na{sub 2}O, have shown severe attack on 24 Cr-1 Mo steel tabs in 1 h under an argon atmosphere from 2200 Degree-Sign F to 2400 Degree-Sign F. Evidence indicates that corrosion is (1) minimal in the gas space above the reaction products, (2) extremely severe at the gas-liquid interface, and (3) less severe in the molten reaction products. A 101-ft-high support structure was constructed to house a test rig designed to simulate the conditions in a commercial sodium-heated steam generator. The test rig circulates 5700 lb of sodium at 400 gal/min through a 16-ft-high reaction vessel. Tests conducted in this rig have shown that the circulating water coil, which served as the target tube during the sodium-water reaction, was wasted to the point of rupturing, causing a large secondary reaction and additional damage to surrounding material. The temperature recorder went off scale at 2200 Degree-Sign F

The compatibility of basalt and MgO with liquid sodium

International Nuclear Information System (INIS)

Jung, J.; Runge, H.

1984-01-01

The interaction of commercially available basalt and MgO-ceramics with liquid sodium has been investigated up to 900 0 C. The two basalt qualities even reacted with sodium at low temperatures and short exposure times. Some inserted MgO-ceramics exhibited the expected good sodium compatibility even at 900 0 C for 100 hours. The reaction mechanisms, the volume changes and the mass balance are discussed. In the thermal shock experiments, the basaltic materials were totally disintegrated while the MgO-qualities remained compact. (author)

Gas-Liquid Separator design of SWRPRS in PGSFR

Energy Technology Data Exchange (ETDEWEB)

Yoon, Jung; Lee, Tae-ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

There is the Sodium-Water Reaction Pressure Relief System (SWRPRS) in PGSFR to prevent the Sodium- Water Reaction (SWR) due to the break of the steam generator tube. The piping to atmosphere includes several components such as gasliquid separator, backpressure rupture disk, and hydrogen igniter. Among these components, gas-liquid separator separates the liquid sodium which is included in gas SWR products not to react sodium and air. In this study, the size of gas-liquid separator, which is based on the hydrogen volume which is exhausted in the sodium dump tank, is determined. To determine the gas-liquid separator for the separation of gas and sodium particle dumped the SDT, Stairmand's model which has high performance among standard cyclone separator models is selected. The body diameter is determined, and other dimensions are determined due to the ratio about the body diameter. Shepherd and Lapple's model is selected as the pressure drop calculation model considering the conservation.

The UK contribution to the sodium-water reaction R and D programme in support of EFR

International Nuclear Information System (INIS)

Currie, R.; McCrindle, K.B.; Wright, P.J.

1990-01-01

The sodium-water reaction R and D programme for European Fast Reactor (EFR), is an integrated European project, as described in another paper at this Specialists Meeting. In this paper the UK contribution to this programme is described. The major sodium-water reaction test facility in the UK is the Super Noah Rig at Dounreay. This Rig is currently being modified to allow the following to be achieved during experiments: Provision of water/steam at the correct EFR Steam Generator Unit (SGU) flowrate, temperature and pressure in the 24 target tubes modelling the EFR SGU in the test section. Provision of sodium at the correct EFR SGU velocity, temperature and pressure through the test section modelling the EFR SGU during the experiments. Experiments in this facility to determine the leakrate at which overheating becomes important in various regions of the EFR SGU and to determine the damage caused by overheating will be the major UK contribution in the future. The modifications to the Rig and the future experimental programme are described. Several intermediate leak experiments have been carried out in the Super Noah Rig in the past, and the main results from them are summarized. These experiments have primarily been at PFR conditions but results have been obtained for 2.25 Cr 1Mo, 9 Cr 1 Mo and A800 steels, with and without internal water flow in the target tubes. During these previous experiments, the sodium was static and was at atmospheric pressure. Results from experiments carried out in the Small Water Leak Rig (SWLR) and the Small-Scale Test Facilities at Dounreay on microleak evolution, wastage of tubes and SGU walls, the behaviour of small leaks at the low temperature, bottom-tubeplate conditions and corrosion at sodium pool surfaces are presented. The future programme for these facilities is discussed. Fundamental experiments which have been carried out at Harwell on the behaviour of blocked microleaks are described. 2 refs, 15 figs, 1 tab

Thermal and chemical interaction of hot liquid sodium with limestone concrete in argon atmosphere

International Nuclear Information System (INIS)

Fakir, Charan Parida; Sanjay, Kumar Das; Anil, Kumar Sharma; Ramesh, S.S.; Somayajulu, P.A.; Malarvizhi, B.; Kasinathan, N.; Rajan, M.

2007-01-01

Sodium cooled fast breeder reactors (FBRs) may experience accidental leakage of hot liquid sodium in the inert equipment cells and reactor cavity. The leaked sodium at temperature ranging from 120degC to 550degC can come in contact with the sacrificial layer of limestone concrete. In order to study the thermal and chemical impact of sodium on the limestone concrete, five experimental runs were carried out under different test conditions simulating accident scenarios as realistically as possible. In each experimental run, a given mass of liquid sodium preheated to a specified temperature was dumped on the surface of concrete specimen housed in a test vessel with argon atmosphere. The sodium pool formed on the concrete was heated with an immersion heater to maintain the pool temperature at pre-selected level. The temperatures at various strategic locations were continuously monitored throughout the test run. Online measurement of pressure, hydrogen gas and oxygen gas in argon atmosphere was conducted. The solid samples of sodium debris were retrieved from the posttest concrete specimen by manual core drilling device for chemical analysis of reacted and un-reacted sodium. After cleaning the sodium debris, a power-drilling machine was employed to collect powder samples at regular depth interval from the concrete block floor to determine residual free and bound water. This paper presents some of the dominant thermal and chemical features related to structural safety of the concrete. Among the thermal parameters, on-set time and residence period for Energetic Thermal Transients (ETT) along with peak and average heat generation rates are evaluated. Chemical parameters such as rate and extent of water release from concrete, sodium consumption, sodium hydroxide production and sodium emission into argon atmosphere are also elucidated. Physicochemical characteristics of post-test sodium and concrete debris were investigated. Moreover spatial distribution of sodium, free and

Metal corrosion in a supercritical carbon dioxide - liquid sodium power cycle.

Energy Technology Data Exchange (ETDEWEB)

Moore, Robert Charles; Conboy, Thomas M.

2012-02-01

A liquid sodium cooled fast reactor coupled to a supercritical carbon dioxide Brayton power cycle is a promising combination for the next generation nuclear power production process. For optimum efficiency, a microchannel heat exchanger, constructed by diffusion bonding, can be used for heat transfer from the liquid sodium reactor coolant to the supercritical carbon dioxide. In this work, we have reviewed the literature on corrosion of metals in liquid sodium and carbon dioxide. The main conclusions are (1) pure, dry CO{sub 2} is virtually inert but can be highly corrosive in the presence of even ppm concentrations of water, (2) carburization and decarburization are very significant mechanism for corrosion in liquid sodium especially at high temperature and the mechanism is not well understood, and (3) very little information could be located on corrosion of diffusion bonded metals. Significantly more research is needed in all of these areas.

Some problems of leaks in sodium-water steam generator

International Nuclear Information System (INIS)

Kozlov, F.A.; Sergeev, G.V.; Sednev, A.R.; Makarov, V.M.

1976-01-01

The paper contains data on wastage of steam generator structural materials and high-nickel alloys in the zone of water leakage into sodium as well as investigation results for self-enlargement of water leaks into sodium through defects in these materials. It is shown that the rate of material damage in the zone of sodium-water reaction and in the channel with water leaking-out decreases with increasing nickel content in steels and strongly depends on sodium temperature. The paper presents experimentally obtained dependences of leakage self-enlargement rates on sodium temperature and leakage size

Characterization of the liquid sodium spray generated by a pipework hole

International Nuclear Information System (INIS)

Torsello, G.; Parozzi, F.; Nericcio, L.; Araneo, L.; Cozzi, F.; Carcassi, M.; Mattei, N.

2012-01-01

Due to its advantageous thermodynamic characteristics at high temperature (550 deg. C), liquid sodium is the main candidate to be the cooling fluid for Generation TV nuclear reactors SFR (Sodium-cooled Fast Reactors). Now, sodium reacts very violently, both with the water and the oxygen of the air. Only few data were known about the liquid sodium behaviour when spread in the environment through micro defects. These are often present in a cooling circuit in welded or sealed joints and more rarely in the pipes. Micro defects, on the other hand, can be also generated in a cooling circuit because of the vibrations always present in a circuit into which a fluid runs. A new set-up, named LISOF, was built for testing high temperature liquid sodium when passing through micro defects and generating sprays or jets. Sprays and jets were generated by means of nozzles embedding sub milli-metric holes the diameter of which was: 0.2 mm, 0.4 mm, 0.5 mm. Tests were performed by pressurizing liquid sodium (550 deg. C) at: 3, 6 and 9 barg. Normal and high speed cinematography were used for the direct observation of the liquid sodium sprays while Phase Doppler Interferometry was used for the measurement of the droplets characteristics and velocity. Tests concerning the behaviour of the high temperature liquid sodium firing in air or in contact with the cement cover applied to a scaled down core catcher simulacrum were also performed. The paper presents the built set-up and the collected results. (authors)

Characterization of the liquid sodium spray generated by a pipework hole

Energy Technology Data Exchange (ETDEWEB)

Torsello, G.; Parozzi, F.; Nericcio, L. [RSE - Nuclear and Industrial Plant Safety Team, Power Generation System Dept., via Rubattino 54, 20134 Milano (Italy); Araneo, L.; Cozzi, F. [Politecnico di Milano, Energy Dept., via Lambruschini 4, 20156 Milano (Italy); Carcassi, M.; Mattei, N. [Universita di Pisa-Facolta d' Ingegneria DIMNP-Mechanical, Nuclear and Production Dep., Largo L. Lazzarino 2, 56126 Pisa (Italy)

2012-07-01

Due to its advantageous thermodynamic characteristics at high temperature (550 deg. C), liquid sodium is the main candidate to be the cooling fluid for Generation TV nuclear reactors SFR (Sodium-cooled Fast Reactors). Now, sodium reacts very violently, both with the water and the oxygen of the air. Only few data were known about the liquid sodium behaviour when spread in the environment through micro defects. These are often present in a cooling circuit in welded or sealed joints and more rarely in the pipes. Micro defects, on the other hand, can be also generated in a cooling circuit because of the vibrations always present in a circuit into which a fluid runs. A new set-up, named LISOF, was built for testing high temperature liquid sodium when passing through micro defects and generating sprays or jets. Sprays and jets were generated by means of nozzles embedding sub milli-metric holes the diameter of which was: 0.2 mm, 0.4 mm, 0.5 mm. Tests were performed by pressurizing liquid sodium (550 deg. C) at: 3, 6 and 9 barg. Normal and high speed cinematography were used for the direct observation of the liquid sodium sprays while Phase Doppler Interferometry was used for the measurement of the droplets characteristics and velocity. Tests concerning the behaviour of the high temperature liquid sodium firing in air or in contact with the cement cover applied to a scaled down core catcher simulacrum were also performed. The paper presents the built set-up and the collected results. (authors)

Research on Liquid Management Technology in Water Tank and Reactor for Propulsion System with Hydrogen Production System Utilizing Aluminum and Water Reaction

Science.gov (United States)

Imai, Ryoji; Imamura, Takuya; Sugioka, Masatoshi; Higashino, Kazuyuki

2017-12-01

High pressure hydrogen produced by aluminum and water reaction is considered to be applied to space propulsion system. Water tank and hydrogen production reactor in this propulsion system require gas and liquid separation function under microgravity condition. We consider to install vane type liquid acquisition device (LAD) utilizing surface tension in the water tank, and install gas-liquid separation mechanism by centrifugal force which swirling flow creates in the hydrogen reactor. In water tank, hydrophilic coating was covered on both tank wall and vane surface to improve wettability. Function of LAD in water tank and gas-liquid separation in reaction vessel were evaluated by short duration microgravity experiments using drop tower facility. In the water tank, it was confirmed that liquid was driven and acquired on the outlet due to capillary force created by vanes. In addition of this, it was found that gas-liquid separation worked well by swirling flow in hydrogen production reactor. However, collection of hydrogen gas bubble was sometimes suppressed by aluminum alloy particles, which is open problem to be solved.

Electrochemical Impedance Response of the surface treated FMS in Liquid Sodium Environment

International Nuclear Information System (INIS)

Lee, Jeong Hyeon; Shin, Sang Hun; Kim, Ji Hyun

2014-01-01

HT9 and Gr.92 are known as compatible in sodium environment because the usual refueling time of SFRs is designed about 54 months. It is very important to investigate the corrosion-related behavior such as surface corrosion rate, carburization, decarburization and mechanical properties for its operation time. SiC and Si 3 N 4 CVD coating for decarburization barrier on the surface of FMS is considered in this study. The decarburization process where dissolved carbon near the specimen surface disused in to the liquid sodium. This process can originate from the difference between dissolved carbon in the material and liquid sodium. A compatibility test the cladding tube revealed that a decrease of the mechanical property instigated by the aging proves governed the whole mechanical property. To monitor the corrosion behavior of these candidate materials in sodium environment, Electrochemical Impedance Spectroscopy (EIS) method is first introduced and investigated in this study. The compatibility of cladding and structural materials with sodium has to be carefully investigated, as sodium could promote corrosion of cladding and structural materials in two ways. One is produced by the dissolution of alloy constituents into the sodium, and the other is produced through a chemical reaction with impurities (especially oxygen and carbon) in the sodium environment. EIS test with pre-oxidized Gr. 92 specimen in 200 .deg. C liquid sodium environment was carried out in this study. A clear Nyquist and Bode plots were obtained in liquid metal environment and the resistance of sodium and the oxide, and the capacitance of the oxide were measured from this result

Liquid sodium pool fires

Energy Technology Data Exchange (ETDEWEB)

Casselman, C [DSN/SESTR, Centre de Cadarache, Saint-Paul-lez-Durance (France)

1979-03-01

Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

Liquid sodium pool fires

International Nuclear Information System (INIS)

Casselman, C.

1979-01-01

Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

Molecular dynamics simulation of sodium aluminosilicate glass structures and glass surface-water reactions using the reactive force field (ReaxFF)

Science.gov (United States)

Dongol, R.; Wang, L.; Cormack, A. N.; Sundaram, S. K.

2018-05-01

Reactive potentials are increasingly used to study the properties of glasses and glass water reactions in a reactive molecular dynamics (MD) framework. In this study, we have simulated a ternary sodium aluminosilicate glass and investigated the initial stages of the glass surface-water reactions at 300 K using reactive force field (ReaxFF). On comparison of the simulated glass structures generated using ReaxFF and classical Buckingham potentials, our results show that the atomic density profiles calculated for the surface glass structures indicate a bond-angle distribution dependency. The atomic density profiles also show higher concentrations of non-bridging oxygens (NBOs) and sodium ions at the glass surface. Additionally, we present our results of formation of silanol species and the diffusion of water molecules at the glass surface using ReaxFF.

SWAAM-LT: The long-term, sodium/water reaction analysis method computer code

International Nuclear Information System (INIS)

Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.

1993-01-01

The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data

Distributed temperature sensor testing in liquid sodium

Energy Technology Data Exchange (ETDEWEB)

Gerardi, Craig, E-mail: cgerardi@anl.gov; Bremer, Nathan; Lisowski, Darius; Lomperski, Stephen

2017-02-15

Highlights: • Distributed temperature sensors measured high-resolution liquid-sodium temperatures. • DTSs worked well up to 400 °C. • A single DTS simultaneously detected sodium level and temperature. - Abstract: Rayleigh-backscatter-based distributed fiber optic sensors were immersed in sodium to obtain high-resolution liquid-sodium temperature measurements. Distributed temperature sensors (DTSs) functioned well up to 400 °C in a liquid sodium environment. The DTSs measured sodium column temperature and the temperature of a complex geometrical pattern that leveraged the flexibility of fiber optics. A single Ø 360 μm OD sensor registered dozens of temperatures along a length of over one meter at 100 Hz. We also demonstrated the capability to use a single DTS to simultaneously detect thermal interfaces (e.g. sodium level) and measure temperature.

Development of fast reactor containment safety analysis code, CONTAIN-LMR. (3) Improvement of sodium-concrete reaction model

International Nuclear Information System (INIS)

Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

2015-01-01

A computer code, CONTAIN-LMR, is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. Because a sodium-concrete reaction behavior is one of the most important phenomena in the accident, a Sodium-Limestone Concrete Ablation Model (SLAM) has been developed and installed into the original CONTAIN code at Sandia National Laboratories (SNL) in the U.S. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically using a three-region model, containing a pool (sodium and reaction debris) region, a dry (boundary layer (B/L) and dehydrated concrete) region, and a wet (hydrated concrete) region, the application is limited to the reaction between sodium and limestone concrete. In order to apply SLAM to the reaction between sodium and siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved to consider the new chemical reactions between sodium and silicon dioxide. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency (JAEA). It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena. (author)

Analysis of self-wastage phenomena of micro leak caused by sodium-water reaction in sodium-cooled fast breeder reactor through simulant experiment

International Nuclear Information System (INIS)

Jang, Sunghyon; Takata, Takashi; Yamaguchi, Akira

2014-01-01

Self-wastage phenomena are an enlargement of a leak on the heat transfer tube caused by a corrosive sodium-water reaction (SWR) in a steam generator (SG) of sodium-cooled fast breeder reactor (SFR). If the steam generator operates for sometimes under this condition, the self-wastage phenomena start from the sodium side and advance through the tube thickness. The leak rate stays almost constant level until the wastage reaches the sodium side, however, when the thin diaphragm of the tube wall is removed, the leak rate sharply increase, and it may bring a secondary failure of the surrounding heat transfer tubes. The design and safety concern is a possibility of the secondary failure of nearby SG tubes that could cause undesirable development of the accidents. One needs to evaluate the increased resultant leak rate due to the self-wastage phenomenon. Therefore, a quantification of the diameter of enlarged leak is needed to estimate the resultant leak rate. For this purpose, a simulant self-wastage experiment was proposed to investigate the self-enlargement of the leak so that evaluate the mechanism of the Self-wastage. In the experiment, high concentrated hydrochloric acid (HCl) is injected to the reaction tank that is filled sodium hydroxide (NaOH) solution through a nozzle made by paraffin wax. The self-enlargement of the leak was evaluated by considering the melted nozzle due to the reaction heat released from the Neutralization reaction. Also, a numerical investigation has been carried out to evaluate the enlarged nozzle and validate the results of experimental methodology. Based on the experimental and computational results, it is found that despite initial leak rate, there is an upper limit in the enlarged nozzle. These results show a similar tendency with the experimental result of SWAT-4 experiment carried out by Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan. Furthermore, the increased resultant leak rate is evaluated using the enlarged

Liquid-liquid equilibrium of water + PEG 8000 + magnesium sulfate or sodium sulfate aqueous two-phase systems at 35°C: experimental determination and thermodynamic modeling

Directory of Open Access Journals (Sweden)

B. D. Castro

2005-09-01

Full Text Available Liquid-liquid extraction using aqueous two-phase systems is a highly efficient technique for separation and purification of biomolecules due to the mild properties of both liquid phases. Reliable data on the phase behavior of these systems are essential for the design and operation of new separation processes; several authors reported phase diagrams for polymer-polymer systems, but data on polymer-salt systems are still relatively scarce. In this work, experimental liquid-liquid equilibrium data on water + polyethylene glycol 8000 + magnesium sulfate and water + polyethylene glycol 8000 + sodium sulfate aqueous two-phase systems were obtained at 35°C. Both equilibrium phases were analyzed by lyophilization and ashing. Experimental results were correlated with a mass-fraction-based NRTL activity coefficient model. New interaction parameters were estimated with the Simplex method. The mean deviations between the experimental and calculated compositions in both equilibrium phases is about 2%.

The chemistry of the liquid alkali metals

International Nuclear Information System (INIS)

Addison, C.C.

1984-01-01

A study of liquid alkali metals. It encourages comparison with molecular solvents in chapter covering the nature and reactivity of dissolved species, solvation, solubility and electrical conductivity of solutions. It demonstrates lab techniques unique to liquid alkali metals. It discusses large-scale applications from storage batteries to sodium-cooled reactors and future fusion reactors, and associated technological problems. Contents: Some Basic Physical and Chemical Properties; Manipulation of the Liquids; The Chemistry of Purification Methods; Species Formed by Dissolved Elements; Solubilities and Analytical Methods; Alkali Metal Mixtures; Solvation in Liquid Metal; Reactions Between Liquid Alkali Metals and Water; Reactions of Nitrogen with Lithium and the Group II Metals in Liquid Sodium; The Formation, Dissociation and Stability of Heteronuclear Polyatomic Anions; Reactions of the Liquid Alkali Metals and Their Alloys with Simple Alipatic Hydrocarbons; Reactions of the Liquid Alkali Metals with Some Halogen Compounds; Hydrogen, Oxygen and Carbon Meters; Surface Chemistry and Wetting; Corrosion of Transition Metals by the Liquid Alkali Metals; Modern Applications of the Liquid Alkali Metals

Study of the influence of liquid sodium on the mechanical behavior of T91 steel in liquid sodium

International Nuclear Information System (INIS)

Hemery, S.

2013-01-01

We studied the sensitivity of T91 steel to embrittlement by liquid sodium. An experimental procedure was set up to proceed to mechanical testing in sodium under an inert atmosphere. The introduction of a liquid sodium pre-exposure step prior to mechanical testing enabled the study of both the wettability of T91 by sodium and the structure of the sodium steel/interface as a function of the exposure parameters. The mechanical properties of T91 steel are significantly reduced in liquid sodium provided the wetting conditions are good. The use of varying oxygen and hydrogen concentrations suggests that oxygen plays a major role in enhancing the wettability of T91. The sensitivity of the embrittlement to strain rate and temperature was characterized. These results showed the existence of a ductile to brittle transition depending on both parameters. Its characterization suggests that a diffusion step is the limiting rate phenomenon of this embrittlement case. TEM and EBSD analysis of arrested cracks enabled us to establish that the fracture mode is inter-lath or intergranular. This characteristic is coherent with the crack path commonly reported in liquid metal embrittlement. A similar procedure was applied to the unalloyed XC10 steel. The results show a behavior which is similar to the one of T91 steel and suggest a common mechanism for liquid sodium embrittlement of body centered cubic steels. Moreover, they confirm that the ductile to brittle transition seems associated with a limited crack propagation rate. The propagation is thermally activated with activation energy of about 50 kJ/mol. Finally, it was shown that 304L austenitic steel is sensitive to liquid sodium embrittlement as well. Some fracture surfaces testify of an intergranular fracture mode, but some questions still remain about the crack path. (author) [fr

Corrosion behaviors of ceramics against liquid sodium. Sodium corrosion characteristics of sintering additives

International Nuclear Information System (INIS)

Tachi, Yoshiaki; Kano, Shigeki; Hirakawa, Yasushi; Yoshida, Eiichi

1998-01-01

It has been progressed as the Frontier Materials Research to research and develop ceramics to apply for several components of fast breeder reactor using liquid sodium as coolant instead of metallic materials. Grain boundary of ceramics has peculiar properties compared with matrix because most of ceramics are produced by hardening and firing their raw powders. Some previous researchers indicated that ceramics were mainly corroded at grain boundaries by liquid sodium, and ceramics could not be used under corrosive environment. Thus, it is the most important for the usage of ceramics in liquid sodium to improve corrosion resistance of grain boundaries. In order to develop the advanced ceramics having good sodium corrosion resistance among fine ceramics, which have recently been progressed in quality and characteristics remarkably, sodium corrosion behaviors of typical sintering additives such as MgO, Y 2 O 3 and AlN etc. have been examined and evaluated. As a result, the followings have been clarified and some useful knowledge about developing advanced ceramics having good corrosion resistance against liquid sodium has been obtained. (1) Sodium corrosion behavior of MgO depended on Si content. Samples containing large amount of Si were corroded severely by liquid sodium, whereas others with low Si contents showed good corrosion resistance. (2) Both Y 2 O 3 and AlN, which contained little Si, showed good sodium corrosion resistance. (3) MgO, Y 2 O 3 and AlN are thought to be corroded by liquid sodium, if they contain some SiO 2 . Therefore, in order to improve sodium corrosion resistance, it is very important for these ceramics to prevent the contamination of matrix with SiO 2 through purity control of their raw powders. (author)

Effect of silicate module of water glass on rheological parameters of poly(sodium acrylate)/sodium silicate hydrogels

Science.gov (United States)

Mastalska-Popiawska, J.; Izak, P.

2017-01-01

The poly(sodium acrylate)/sodium silicate hydrogels were synthesized in the presence of sodium thiosulphate and potassium persulphate as the redox initiators and N,N’-methylene-bisacrylamide as the cross-linking monomer. 20 wt% aqueous solution of sodium acrylate was polymerized together with water glass with different silicate modules (M) from 1.74 to 2.29, in three mass ratio of the monomer solution to the water glass 2:1, 1:1 and 1:2. Such obtained hybrid composites were rheologically tested using the oscillation method. It allowed to designate the crossover point during polymerization, as well as to define the viscoelastic properties of the casted hydrogel samples one week after the reaction. The obtained results of the oscillation measurements showed that cross-linking reaction proceeds very quickly and the lower the silicate module is, the process starts faster. After the completion of the reaction the silicate-polymer hydrogels are strongly elastic materials and the highest elasticity characterizes systems with the mass ratio 1:2, i.e. with the highest water glass content.

Liquid sodium oxygenmeter

International Nuclear Information System (INIS)

Jakes, D.; Fresl, M.; Svoboda, V.

1979-02-01

The results of test runs of two design varieties of liquid sodium oxygenmeter in sodium loops are described. The accuracy and sensitivity are discussed reached using this instrument within 1 and 10 p.p.m. of oxygen concentration. A change in the used reference system is proposed based on practical experiences and thermochemical calculations. Ceramic electrolyte corrosion is analysed and the possible interpretation of the corrosion effect on the galvanic cell electromotive force is suggested. (author)

Study of optimal transformation of liquid effluents resulting from the destruction of radioactive sodium by water into ultimate solid wastes

International Nuclear Information System (INIS)

Rodriguez, G.; Camaro, S.; Fiquet, O.; Bernard, A.; Le Bescop, P.

1997-01-01

In the framework of sodium waste processing, it has been proposed to retain only processes that treat the sodium using water, thus generating the same by-products: hydrogen and sodium hydroxide. As the objective is to minimise radioactive liquid releases and as, moreover, the authorizations with respect to sodium salt releases are highly restrictive, several solutions have been envisaged for transforming the active sodium hydroxide coming from sodium destruction processes into ultimate solid wastes that can be stored on the surface in a storage site approved by the ANDRA (National Radioactive Waste Management Agency): the Aube Storage Site (CSA). Two processes have been considered and compared: immobilisation in concrete (cementation) and immobilisation in ceramic (ceramisation). These two processes are evaluated according to several criteria: the state of advancement of the process, the quantity of sodium hydroxide (and therefore of sodium) that can be treated per package. (author)

Test results of sodium-water reaction testing in near prototypical LMR steam generator

International Nuclear Information System (INIS)

Boardman, C.E.; Hui, M.; Neely, H.H.

1990-01-01

An extensive test program has been performed in the United States to investigate the effects of large sodium-water reaction events in LMFBR steam generators. Tests were conducted in the Large Leak Test Rig (LLTR) located at the Energy Technology Engineering Center (ETEC). The program was divided into two phases, Series I and Series II, for the purpose of satisfying near-term and long-term needs. Series II was further subdivided into large and intermediate leak tests. This paper will emphasize the Series II intermediate leak tests and resulting conclusions for steam generator design and operation. 11 figs, 2 tabs

Sodium removal disassembly and examination of the Fermi secondary sodium pump

International Nuclear Information System (INIS)

Maffei, H.P.; Funk, C.W.; Ballif, J.L.

1974-01-01

The Fermi secondary pump is a centrifugal single stage design. The pump had been operated more than 42,000 hours between 450 and 800 0 F. Sodium was drained from the pump in 1973 and the system was back filled with carbon dioxide. The pump was fabricated for 2.25 Cr-1 Mo Croloy steel. Prior to cleaning the pump was inerted and heated with 150 0 F nitrogen using the pump casing as the containment vessel. The water-vapor-nitrogen process was used in three increasing stages of water concentration. The hydrogen concentration in the discharge line was followed as an indicator of the sodium-water reaction rate. Upon completion of the hydrogen evolution, the pump was rinsed several times with hot water. Six pounds of sodium were removed from the pump during a process cycle of 79 hours including rinsing. The maximum pump temperature recorded was 175 0 F with no variation exceeding 10 0 F. The hydrogen concentration in the effluent provided a very satisfactory index for control of the reaction by adjustment of the water-vapor concentration feed to the system. Rinsing effectiveness was limited by a pool of water in the volute that was not drainable with the available system hook up. Sodium and its compounds were removed from all internal surfaces that could be observed by the first stage of disassembly. All such surfaces were coated with a black deposit. Areas above the sodium liquid level were coated with a vermillion colored oxide. Sodium was found on the (1) threads of the impeller nut lock screw, (2) impeller nut-tapered shaft interface, and (3) vapor deposited sodium was found in the oil seal

Hydrogen generation systems utilizing sodium silicide and sodium silica gel materials

Science.gov (United States)

Wallace, Andrew P.; Melack, John M.; Lefenfeld, Michael

2015-07-14

Systems, devices, and methods combine reactant materials and aqueous solutions to generate hydrogen. The reactant materials can sodium silicide or sodium silica gel. The hydrogen generation devices are used in fuels cells and other industrial applications. One system combines cooling, pumping, water storage, and other devices to sense and control reactions between reactant materials and aqueous solutions to generate hydrogen. Multiple inlets of varied placement geometries deliver aqueous solution to the reaction. The reactant materials and aqueous solution are churned to control the state of the reaction. The aqueous solution can be recycled and returned to the reaction. One system operates over a range of temperatures and pressures and includes a hydrogen separator, a heat removal mechanism, and state of reaction control devices. The systems, devices, and methods of generating hydrogen provide thermally stable solids, near-instant reaction with the aqueous solutions, and a non-toxic liquid by-product.

Distributed temperature sensor testing in liquid sodium

Energy Technology Data Exchange (ETDEWEB)

Gerardi, Craig; Bremer, Nathan; Lisowski, Darius; Lomperski, Stephen

2017-02-01

Rayleigh-backscatter-based distributed fiber optic sensors were immersed in sodium to obtain high-resolution liquid-sodium temperature measurements. Distributed temperature sensors (DTSs) functioned well up to 400°C in a liquid sodium environment. The DTSs measured sodium column temperature and the temperature of a complex geometrical pattern that leveraged the flexibility of fiber optics. A single Ø 360 lm OD sensor registered dozens of temperatures along a length of over one meter at 100 Hz. We also demonstrated the capability to use a single DTS to simultaneously detect thermal interfaces (e.g. sodium level) and measure temperature.

Effectiveness of Chlorinated Water, Sodium Hypochlorite, Sodium ...

African Journals Online (AJOL)

This study evaluated the efficacy of chlorinated water, sodium hypochlorite solution, sodium chloride solution and sterile distilled water in eliminating pathogenic bacteria on the surfaces of raw vegetables. Lettuce vegetables were dipped in different concentrations of chlorinated water, sodium hypochlorite solution, sodium ...

Similarity rules of thermal stratification phenomena for water and sodium

International Nuclear Information System (INIS)

Ohtsuka, M.; Ikeda, T.; Yamakawa, M.; Shibata, Y.; Moriya, S.; Ushijima, S.; Fujimoto, K.

1988-01-01

Similarity rules for thermal stratification phenomena were studied using sodium and water experiments with scaled cylindrical vessels. The vessel dimensions were identical to focus on the effect of differences in fluid properties upon the phenomena. Comparisons of test results between sodium and water elucidated similar and dissimilar characteristics for thermal stratification phenomena which appeared in the scaled vessels. Results were as follows: (1) The dimensionless upward velocity of the thermal stratification interface was proportional to Ri -0.74 for water and sodium during the period when the buoyancy effect was dominant. (2) Dimensionless temperature transient rate at the outlet slit decreased with Ri for sodium and remained constant for water where Ri>0.2. The applicability of the scaled test results to an actual power plant was also studied by using multi-dimensional numerical analysis which was verified by the water and sodium experiments. Water experiments could simulate liquid metal fast breeder reactor flows more accurately than sodium experiments for dimensionless temperature gradient at the thermal stratification interface and dimensionless temperature transient rate at the intermediate heat exchanger inlet

Sodium-water wastage and reactions program performed by general electric in support of the US. AEC LMFBR steam generator development

International Nuclear Information System (INIS)

Greene, D.A.

1975-01-01

This paper constitutes an interim report on the sodium-water reaction programs performed, using the GE-SOWAT, GE-SMALL LEAK BEHAVIOR RIG, and GE-PTTR facilities in support of LMFBR steam generator development and its application to the Clinch River Breeder Reactor Plant. Test data from these rigs are presented, including wastage data as a function of water injection rate, sodium temperature, and orifice geometry. Initial results for self-wastage of defects under prototypical conditions, and from proof-of-principle tests of a protected heat transfer tube concept are also presented. An analytical basis for wastage phenomena is suggested. (author)

Sodium-water wastage and reactions program performed by general electric in support of the US. AEC LMFBR steam generator development

Energy Technology Data Exchange (ETDEWEB)

Greene, D A

1975-07-01

This paper constitutes an interim report on the sodium-water reaction programs performed, using the GE-SOWAT, GE-SMALL LEAK BEHAVIOR RIG, and GE-PTTR facilities in support of LMFBR steam generator development and its application to the Clinch River Breeder Reactor Plant. Test data from these rigs are presented, including wastage data as a function of water injection rate, sodium temperature, and orifice geometry. Initial results for self-wastage of defects under prototypical conditions, and from proof-of-principle tests of a protected heat transfer tube concept are also presented. An analytical basis for wastage phenomena is suggested. (author)

Conclusions from the sodium-water reaction experiments performed with a straight tube bundle model for a steam generator with respect to the calculation method of the accident design pressure

Energy Technology Data Exchange (ETDEWEB)

Ludwig, P W.P.; Ruijterman, C

1975-07-01

This paper will give some conclusions, drawn from the big-leakages sodium-water reaction experiments, on the calculation methods to be used in determining the sodium blow down, the water supply and the bubble pressure. The necessity of taking into account the compressibility of sodium is demonstrated. (author)

Conclusions from the sodium-water reaction experiments performed with a straight tube bundle model for a steam generator with respect to the calculation method of the accident design pressure

International Nuclear Information System (INIS)

Ludwig, P.W.P.; Ruijterman, C.

1975-01-01

This paper will give some conclusions, drawn from the big-leakages sodium-water reaction experiments, on the calculation methods to be used in determining the sodium blow down, the water supply and the bubble pressure. The necessity of taking into account the compressibility of sodium is demonstrated. (author)

Ultrasonic imaging in liquid sodium

Energy Technology Data Exchange (ETDEWEB)

Lubeigt, E. [CEA Cadarache, DEN/DTN/STCP/LIET, 13108 Saint-Paul-Lez-Durance Cedex (France); Laboratoire de Mecanique et d' Acoustique, CNRS UPR 7051, 13402 Marseille Cedex 20 (France); Mensah, S.; Chaix, J.F.; Rakotonarivo, S. [Laboratoire de Mecanique et d' Acoustique, CNRS UPR 7051, 13402 Marseille Cedex 20 (France); Gobillot, G. [CEA Cadarache, DEN/DTN/STCP/LIET, 13108 Saint-Paul-Lez-Durance Cedex (France)

2015-07-01

The fourth generation of nuclear reactor can use liquid sodium as the core coolant. When the reactor is operating, sodium temperatures can reach up to 600 deg. C. During maintenance periods, when the reactor is shut down, the coolant temperature is reduced to 200 deg. C. Because molten sodium is optically opaque, ultrasonic imaging techniques are developed for maintenance activities. Under-sodium imaging aims at i) checking the health of immersed structures. It should also allow ii) to assess component degradation or damage as cracks and shape defects as well as iii) the detection of lost objects. The under-sodium imaging system has to sustain high temperature (up to 300 deg. C) and hostility of the sodium environment. Furthermore, specific constraints such as transducers characteristics or the limited sensor mobility in the reactor vessel have to be considered. This work focuses on developing a methodology for detecting damages such as crack defects with ultrasound devices. Surface-breaking cracks or deep cracks are sought in the weld area, as welds are more subject to defects. Traditional methods enabled us to detect emerging cracks of submillimeter size with sodium-compatible high-temperature transducer. The presented approach relies on making use of prior knowledge about the environment through the implementation of differential imaging and time-reversal techniques. Indeed, this approach allows to detect a change by comparison with a reference measurement and by focusing back to any change in the environment. It is a means of analysis and understanding of the physical phenomena making it possible to design more effective inspection strategies. Difference between the measured signals reveals the acoustic field scattered by a perturbation (a crack for instance), which may occur between periodical measurements. The imaging method relies on the adequate combination of two computed ultrasonic fields, one forward and one adjoint. The adjoint field, which carries the

Ultrasonic imaging in liquid sodium

International Nuclear Information System (INIS)

Lubeigt, E.; Mensah, S.; Chaix, J.F.; Rakotonarivo, S.; Gobillot, G.

2015-01-01

The fourth generation of nuclear reactor can use liquid sodium as the core coolant. When the reactor is operating, sodium temperatures can reach up to 600 deg. C. During maintenance periods, when the reactor is shut down, the coolant temperature is reduced to 200 deg. C. Because molten sodium is optically opaque, ultrasonic imaging techniques are developed for maintenance activities. Under-sodium imaging aims at i) checking the health of immersed structures. It should also allow ii) to assess component degradation or damage as cracks and shape defects as well as iii) the detection of lost objects. The under-sodium imaging system has to sustain high temperature (up to 300 deg. C) and hostility of the sodium environment. Furthermore, specific constraints such as transducers characteristics or the limited sensor mobility in the reactor vessel have to be considered. This work focuses on developing a methodology for detecting damages such as crack defects with ultrasound devices. Surface-breaking cracks or deep cracks are sought in the weld area, as welds are more subject to defects. Traditional methods enabled us to detect emerging cracks of submillimeter size with sodium-compatible high-temperature transducer. The presented approach relies on making use of prior knowledge about the environment through the implementation of differential imaging and time-reversal techniques. Indeed, this approach allows to detect a change by comparison with a reference measurement and by focusing back to any change in the environment. It is a means of analysis and understanding of the physical phenomena making it possible to design more effective inspection strategies. Difference between the measured signals reveals the acoustic field scattered by a perturbation (a crack for instance), which may occur between periodical measurements. The imaging method relies on the adequate combination of two computed ultrasonic fields, one forward and one adjoint. The adjoint field, which carries the

Effects of the presence of core debris on the behavior of sodium-concrete reactions

International Nuclear Information System (INIS)

Nguyen, D.H.; Muhlestein, L.D.

1984-01-01

Calculations using the SOCON model indicated the following: the temperature was increased throughout the concrete and the reaction product layer. Temperature could be raised to above sodium bp. Rate of release and accumulation of water and CO 2 gas were increased. The sodium mass transport to the reaction surface was also increased. As a consequence, more hydrogen and chemical heat were produced. The probability of concrete mechanical failure was higher. Sodium boiling inside the reaction product layer would not significantly alter the course of the reaction, unless it could reduce the rate of sodium transport. Although the chemical heat dominated during the early period, the decay heat could become the main source later. The reactions were driven by three main heat sources: the chemical heat, core debris heat and conduction heat from the hot sodium pool. The latter could become a heat sink. Even with the presence of core debris, the chemical reaction penetration was self-limiting and eventually, the reaction penetration rate decreased to a small value

Modeling of under-expanded reactive CO2-into-sodium jets, in the frame of sodium fast reactors

International Nuclear Information System (INIS)

Vivaldi, D.

2013-01-01

This PhD work was motivated by the investigations in the frame of supercritical CO 2 Brayton cycles as possible energy conversion cycles for the Sodium-cooled Fast nuclear Reactors (SFRs). This technology represents an alternative to conventional steam Rankine cycles, with the main advantage represented by the elimination of the accidental sodium-water reaction scenario. Nevertheless, CO 2 chemically reacts with sodium, through an exothermic reaction leading to solid reaction products, mainly sodium carbonate. Following an accidental leakage inside the sodium-CO 2 heat exchanger of a SFR, the CO 2 , having an operating pressure of about 200 bars, would be injected into the low-operating pressure liquid sodium, creating an under-expanded reactive CO 2 -into-sodium jet. The under-expanded jet features a sonic gas injection velocity and an under-expansion in the first region downstream the leakage, where the CO 2 is accelerated to supersonic velocities. The exothermic reaction between the CO 2 and the sodium causes an increasing of the temperature inside the heat exchanger. An experimental facility was built at CEA Cadarache, for the realization of CO 2 -into-sodium jets: this facility has provided preliminary results in terms of temperature variations inside the jet due to the exothermic reaction. However, this type of experimental tests are complicated to realize and to analyse, due to the technical difficulties of realizing the contact between CO 2 and sodium, and to the incertitude of temperature measurement inside a two-phase high velocity jet. It follows that a numerical model of this kind of jets is required, in order to understand the CO 2 -sodium kinetics of reaction inside the jet and being able to transpose the phenomenon to relevant SFR sodium-CO 2 heat exchangers. This would allow to understand the consequences of a leakage inside a sodium-CO 2 heat exchanger, in terms of, for instance, temperature profiles inside the heat exchanger and on tube surfaces

Steam and sodium leak simulation in a fluidized-bed steam generator

International Nuclear Information System (INIS)

Vaux, W.G.; Keeton, A.R.; Keairns, D.L.

1977-01-01

A fluidized-bed steam generator for the liquid metal fast breeder reactor enhances plant availability and minimizes the probability of a water/sodium reaction. An experimental test program was conceived to assess design criteria and fluidized-bed operation under conditions of water, steam, and sodium leaks. Sodium, steam, and water were leaked into helium-fluidized beds of metal and ceramic particles at 900 F. Test results show the effects of leaks on the heat transfer coefficient, quality of fluidization, leak detection, and cleanup procedures

Reaction of water vapor with a clean liquid uranium surface

International Nuclear Information System (INIS)

Siekhaus, W.

1985-01-01

To study the reaction of water vapor with uranium, we have exposed clean liquid uranium surfaces to H 2 O under UHV conditions. We have measured the surface concentration of oxygen as a function of exposure, and determined the maximum attainable surface oxygen concentration X 0 /sup s/ as a function of temperature. We have used these measurements to estimate, close to the melting point, the solubility of oxygen (X 0 /sup b/, -4 ) and its surface segregation coefficient β/sup s/(> 10 3 ). 8 refs., 5 figs., 1 tab

Cleaning of the equipment of residual sodium by means of water-vacuum technology

International Nuclear Information System (INIS)

Klykov, B.P.; Lednev, A.I.

1997-01-01

Results of investigation into a problem of equipment decontamination from sodium, that have been conducted in OKBM since 1960 are given. The investigations performed have shown that a water-vacuum washing process is the most optimal method for equipment decontamination from sodium residues. The essence of the method is in conduction of sodium-water reaction under reduced pressure in a leak-tight tank. Boundary conditions are selected experimentally which not allow sodium to be melted during the process, that gives possibility to control the sodium-water reaction. Continuous removal of H 2 and reaction products creates safe conditions for the process conduction. More that 20-year period of operation of a stationary water-vacuum facility and washing the electromagnetic pump for BN-350 fast nuclear reactor directly at is test rig are the best proofs of the proposed method. This method is well suitable for washing the equipment contaminated by radioactive sodium, because by-products of the process are simply utilized. The method is used in a number of Russian enterprises, and recommended for implementation at BN-350 and BN-600 reactor plants. (author)

Experimental and theoretical investigations on safety of the SNR - straight-tube design steam generator with sodium-water reactions

International Nuclear Information System (INIS)

Dumm, K.; Sauermann, F.; Schnitker, W.; Welter, A.

A number of large sodium-water reaction tests has been performed in a steam generator model in order to verify the layout criteria of the SNR straight-tube design steam generators under accident conditions. The experimental setup is described. The test results and their applicability to the SNR steam generators are given and discussed. (U.S.)

Photoionization of Sodium Salt Solutions in a Liquid Jet

International Nuclear Information System (INIS)

Grieves, G. A.; Petrik, Nikolay G.; Herring-Captain, J.; Olanrewaju, B.; Aleksandrov, A.; Tonkyn, Russell G.; Barlow, Stephan E.; Kimmel, Gregory A.; Orlando, Thomas M.

2008-01-01

A liquid microjet was employed to examine the gas/liquid interface of aqueous sodium halide (Na+X-, X=Cl, Br, I) salt solutions. Laser excitation at 193 nm produced and removed cations of the form H+(H2O)n and Na+(H2O)m from liquid jet surfaces containing either NaCl, NaBr or NaI. The protonated water cluster yield varied inversely with increasing salt concentration, while the solvated sodium ion cluster yield varied by anion type. The distribution of H+(H2O)n at low salt concentration is identical to that observed from low-energy electron irradiated amorphous ice and the production of these clusters can be accounted for using a localized ionization/Coulomb expulsion model. Production of Na+(H2O)m is not accounted for by this model but requires ionization of solvation shell waters and a contact ion/Coulomb expulsion mechanism. The reduced yields of Na+(H2O)m from high concentration (10-2 and 10-1 M) NaBr and NaI solutions indicate a propensity for Br- and I- at the solution surfaces and interfaces. This is supported by the observation of multiphoton induced production and desorption of Br+ and I+ from the 10-2 and 10-1 M solution surfaces

Photoionization of Sodium Salt Solutions in a Liquid Jet

Energy Technology Data Exchange (ETDEWEB)

Grieves, G. A.; Petrik, Nikolay G.; Herring-Captain, J.; Olanrewaju, B.; Aleksandrov, A.; Tonkyn, Russell G.; Barlow, Stephan E.; Kimmel, Gregory A.; Orlando, Thomas M.

2008-06-05

A liquid microjet was employed to examine the gas/liquid interface of aqueous sodium halide (Na+X-, X=Cl, Br, I) salt solutions. Laser excitation at 193 nm produced and removed cations of the form H+(H2O)n and Na+(H2O)m from liquid jet surfaces containing either NaCl, NaBr or NaI. The protonated water cluster yield varied inversely with increasing salt concentration, while the solvated sodium ion cluster yield varied by anion type. The distribution of H+(H2O)n at low salt concentration is identical to that observed from low-energy electron irradiated amorphous ice and the production of these clusters can be accounted for using a localized ionization/Coulomb expulsion model. Production of Na+(H2O)m is not accounted for by this model but requires ionization of solvation shell waters and a contact ion/Coulomb expulsion mechanism. The reduced yields of Na+(H2O)m from high concentration (10-2 and 10-1 M) NaBr and NaI solutions indicate a propensity for Br- and I- at the solution surfaces and interfaces. This is supported by the observation of multiphoton induced production and desorption of Br+ and I+ from the 10-2 and 10-1 M solution surfaces.

Initial pressure spike and its propagation phenomena in sodium-water reaction tests for MONJU steam generators

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Tanaka, N.; Hori, M.

1977-01-01

With the objective of demonstrating the safe design of steam generators for prototype LMFBR MONJU against the postulated large-leak accident, a number of large-leak sodium-water reaction tests have been conducted using the SWAT-1 and SWAT-3 rigs. Investigation of the potential effects of pressure load on the system is one of the major concerns in these tests. This paper reports the behavior of initial pressure spike in the reaction vessel, its propagation phenomena to the simulated secondary cooling system, and the comparisons with the computer code for one-dimensional pressure wave propagation problems. Both rigs used are the scaled-down models of the helically coiled steam generators of MONJU. The SWAT-1 rig is a simplified model and consists of a reaction vessel (1/8 scale of MONJU evaporator with 0.4 m dia. and 2.5 m height) and a pressure relief system i.e., a pressure relief line and a reaction products tank. On the other hand, the SWAT-3 rig is a 1/2.5 scale of MONJU SG system and consists of an evaporator (reaction vessel with 1.3 m dia. and 6.35 m height), a superheater, an intermediate heat exchanger (IHX), a piping system simulating the secondary cooling circuit and a pressure relief system. The both water injection systems consist of a water injection line with a rupture disk installed in front of injection hole and an electrically heated water tank. Choice of water injection rates in the scaled-down models is made based on the method of iso-velocity modeling. Test results indicated that the characteristics of the initial pressure spike are dominated by those of initial water injection which are controlled by the conditions of water heater and the size of water injection hole, etc

Water Mock-up for the Sodium Waste Treatment Process

Energy Technology Data Exchange (ETDEWEB)

Nam, Ho Yun; Kim, Jong Man; Kim, Byung Ho; Lee, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

It is important to safely treat the waste sodium which was produced from the sodium cooled fast reactors and the sodium facilities. About 1.3 tons of sodium waste has accumulated at KAERI from the sodium experiments which have been carried out since 1990. Also, large scaled sodium experiments are scheduled to verify the design of the sodium cooled fast reactor. As a treatment method for the waste sodium produced at the sodium facility, an investigation of the reaction procedure of the waste sodium with the sodium hydroxide aqueous has been developed. The NOAH process was developed in France for the treatment of waste sodium produced from sodium facilities and reactors. In the NOAH process, a small amount of sodium waste is continuously injected into the upper space which is formed on the free surface of the aqueous and slowly reacted with sodium hydroxide aqueous. Since the density of the sodium is lower than that of the aqueous, the injected sodium waste sometimes accumulates above the free surface of the sodium hydroxide aqueous, and its reaction rate becomes slow or suddenly increases. In the improved process, the sodium was injected into a reaction vessel filled with a sodium hydroxide aqueous through an atomizing nozzle installed on a lower level than that of the aqueous to maintain the reaction uniformly. Fig.1 shows the sodium waste process which was proposed in KAERI. The aqueous is composed of 60% sodium hydroxide, and its temperature is about 60 .deg. C. The process is an exothermic reaction. The hydrogen gas is generated, and the concentration of the sodium hydroxide increases in this process. It needs several systems for the process, i.e. a waste sodium injection, a cooling of the aqueous, hydrogen ventilation, and neutralization with nitric acid. The atomizing nozzle was designed to inject the sodium with the nitrogen gas which supplies a heat to the sodium to prevent its solidification and to uniformly mix the sodium with the aqueous. There are

Corrosion by liquid metals - Application to liquid sodium

International Nuclear Information System (INIS)

Lavielle, Lisette.

1978-10-01

In this bibliographic review on the corrosion by liquid metals, the first part is devoted to the theoretical aspects of the problem and the second part concerns the corrosion of steels by liquid sodium, as example. Obvious the numerous works now published, the mechanisms are still leaving bad known [fr

Towards the Characterization of the Bubble Presence in Liquid Sodium of Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Cavaro, M.; Jeannot, J.P.; Payan, C.

2013-06-01

In a Sodium cooled Fast Reactors (SFR), different phenomena such as gas entrainment or nucleation can lead to gaseous micro-bubbles presence in the liquid sodium of the primary vessel. Although this free gas presence has no direct impact on the core neutronics, the French Atomic Energy and Alternative Energies Commission (CEA) currently works on its characterization to, among others, check the absence of risk of large gas pocket formation and to assess the induced modifications of the sodium acoustic properties. The main objective is to evaluate the void fraction values (volume fraction of free gas) and the radii histogram of the bubbles present in liquid sodium. Acoustics and electromagnetic techniques are currently developed at CEA: - The low-frequency speed of sound measurement, which allows us to link - thanks to Wood's model - the measured speed of sound to the actual void fraction. - The nonlinear mixing of two frequencies, based on the nonlinear resonance behavior of a bubble. This technique allows knowing the radius histogram associated to a bubble cloud. Two different mixing techniques are presented in this paper: the mixing of two high frequencies and the mixing of a high and a low frequency. - The Eddy-current flowmeter (ECFM), the output signal of which is perturbed by free gas presence and in consequence allows detecting bubbles. For each technique, initial results are presented. Some of them are really promising. So far, acoustic experiments have been led with an air-water experimental set-up. Micro-bubbles clouds are generated with a dissolved air flotation device and monitored by an optical device which provides reference measurements. Generated bubbles have radii range from few micrometers to several tens of micrometers. Present and future air/water experiments are presented. Furthermore, a development plan of in-sodium tests is presented in terms of a device set-up, instrumentation, modeling tools and experiments. (authors)

Reactions between rocks and the hydroxides of calcium, sodium and potassium: progress report no. 1

International Nuclear Information System (INIS)

Van Aardt, J.H.P.; Visser, S.

1982-01-01

The reaction between the hydroxides of calcium, sodium and potassium, and clay minerals, feldspars, and some rocks (aggregates for use in concrete) was investigated. The reaction products were examined by means of x-ray diffraction and chemical analysis. The solid reaction products identified were hydrated calcium silicates,hydrated calcium aluminates, and hydrated calcium alumina silicates. It was found that, in the presence of water, calcium hydroxide liberated alkali into solution if the rocks and minerals contained alkali metals in their structure. Two crystalline hydrated sodium calcium silicates (12A and 16A) were prepared in the system Na 2 O-CaO-SiO 2 -H 2 O at 80 degrees Celsius. The one compound (12A) was also observed when sodium hydroxide plus calcium hydroxide and water reacted with silica- or silicate-containing rocks

Physical properties of liquid sodium

International Nuclear Information System (INIS)

Alberdi Primicia, J.; Martinez Piquer, T.A.

1977-01-01

The molten sodium has been the more accepted coolant for the first generation of FBR, by this reason the knowledge of its technology is needed for the development of the next LMFBR. A series of necessary data for designing sodium liquid systems are given. Tables and graphics about the most important physical sodium properties between 1200-1400 degC are gathered. The results have been obtained from equations that relate the properties with temperature using a Fortran IV program. (author) [es

Effect of Sodium Carboxymethyl Celluloses on Water-catalyzed Self-degradation of 200-degree C-heated Alkali-Activated Cement

Energy Technology Data Exchange (ETDEWEB)

Sugama T.; Pyatina, T.

2012-05-01

We investigated the usefulness of sodium carboxymethyl celluloses (CMC) in promoting self-degradation of 200°C-heated sodium silicate-activated slag/Class C fly ash cementitious material after contact with water. CMC emitted two major volatile compounds, CO2 and acetic acid, creating a porous structure in cement. CMC also reacted with NaOH from sodium silicate to form three water-insensitive solid reaction products, disodium glycolate salt, sodium glucosidic salt, and sodium bicarbonate. Other water-sensitive solid reaction products, such as sodium polysilicate and sodium carbonate, were derived from hydrolysates of sodium silicate. Dissolution of these products upon contact with water generated heat that promoted cement’s self-degradation. Thus, CMC of high molecular weight rendered two important features to the water-catalyzed self-degradation of heated cement: One was the high heat energy generated in exothermic reactions in cement; the other was the introduction of extensive porosity into cement.

Korea advanced liquid metal reactor development - Development of measuring techniques of the sodium two-phase flow

Energy Technology Data Exchange (ETDEWEB)

Kim, Moo Hwan; Cha, Jae Eun [Pohang University of Science and Technology, Pohang (Korea)

2000-04-01

The technology which models and measures the behavior of bubble in liquid sodium is very important to insure the safety of the liquid metal reactor. In this research, we designed/ manufactured each part and loop of experimental facility for sodium two phase flow, and applied a few possible methods, measured characteristic of two phase flow such as bubbly flow. A air-water loop similar to sodium loop on each measuring condition was designed/manufactured. This air-water loop was utilized to acquire many informations which were necessary in designing the two phase flow of sodium and manufacturing experimental facility. Before the manufacture of a electromagnetic flow meter for sodium, the experiment using each electromagnetic flow mete was developed and the air-water loop was performed to understand flow characteristics. Experiments for observing the signal characteristics of flow were performed by flowing two phase mixture into the electromagnetic flow mete. From these experiments, the electromagnetic flow meter was designed and constructed by virtual electrode, its signal processing circuit and micro electro magnet. It was developed to be applicable to low conductivity fluid very successfully. By this experiment with the electromagnetic flow meter, we observed that the flow signal was very different according to void fraction in two phase flow and that probability density function which was made by statistical signal treatment is also different according to flow patterns. From this result, we confirmed that the electromagnetic flow meter could be used to understand the parameters of two phase flow of sodium. By this study, the experimental facility for two phase flow of sodium was constricted. Also the new electromagnetic flow meter was designed/manufactured, and experimental apparatus for two phase flow of air-water. Finally, this study will be a basic tool for measurement of two phase flow of sodium. As the fundamental technique for the applications of sodium at

Improvement and test calculation on basic code or sodium-water reaction jet

Energy Technology Data Exchange (ETDEWEB)

Saito, Yoshinori; Itooka, Satoshi [Advanced Reactor Engineering Center, Hitachi Works, Hitachi Ltd., Hitachi, Ibaraki (Japan); Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo [Consulting Engineering Dept., Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

1999-03-01

In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3{center_dot}Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

Improvement and test calculation on basic code or sodium-water reaction jet

International Nuclear Information System (INIS)

Saito, Yoshinori; Itooka, Satoshi; Okabe, Ayao; Fujimata, Kazuhiro; Sakurai, Tomoo

1999-03-01

In selecting the reasonable DBL (design basis water leak rate) on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on the basic code for sodium-water reaction (SWR) jet was performed for an actual scale SG. The improvement points of the code are as follows; (1) introduction of advanced model such as heat transfer between the jet and structure (tube array), cooling effect of the structure, heat transfer between analytic cells, and (2) model improvement for heat transfer between two-phase flow and porous-media. The test calculation using the improved code (LEAP-JET ver.1.30) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results is reasonable and Influence to analysis result of a model. Code integration with the blow down analytic code (LEAP-BLOW) was also studied. It is suitable that LEAP-JET was improved as one of the LEAP-BLOW's models, and it was integrated into this. In addition to above, the improvement for setting of boundary condition and the development of the interface program to transfer the analytical results of LEAP-BLOW have been performed in order to consider the cooling effect of coolant in the tube simply. However, verification of the code by new SWAT-1 and SWAT-3 test data planned in future is necessary because LEAP-JET is under development. And furthermore advancement needs to be planned. (author)

Development of experimental method for self-wastage behavior in sodium-water reaction. Development of test rig (SWAT-2R) and study for experimental procedure

International Nuclear Information System (INIS)

Abe, Yuta; Shimoyama, Kazuhito; Kurihara, Akikazu

2014-07-01

In case of water leak from a penetrated crack on a tube of steam generator in the sodium cooled fast reactor (SFR), self-wastage, that increases the size of leak, may take place by corrosion related to chemical reaction between sodium and water. If the self-wastage continues in a certain period of time, the intact tube bundle may be damaged as a result of enlarged leak. For the safety evaluation of the accident, JAEA has been developing the analytical method of self-wastage using the multi-dimensional sodium-water reaction code. Experiments conducted so far used mainly crack-type test pieces. However, reproducibility was limited and it was difficult to evaluate individual effects of the phenomena in detail. This report describes the development of new experimental rig (SWAT-2R). SWAT-2R enables to examine corrosion effecting factors that were ambiguous in the previous studies. The report includes description of development of micro-leak test piece, examination of experimental procedure. The results will provide fundamental data for validation of the self-wastage analytical method. (author)

austenitic steel corrosion by oxygen-containing liquid sodium

International Nuclear Information System (INIS)

Rivollier, Matthieu

2017-01-01

France is planning to construct the 4. generation of nuclear reactors. They will use liquid sodium as heat transfer fluid and will be made of 316L(N) austenitic steel as structural materials. To guarantee optimal operation on the long term, the behavior of this steel must be verified. This is why corrosion phenomena of 316L(N) steel by liquid sodium have to be well-understood. Literature points out that several corrosion phenomena are possible. Dissolved oxygen in sodium definitely influences each of the corrosion phenomenon. Therefore, the austenitic steel corrosion in oxygen-containing sodium is proposed in this study. Thermodynamics data point out that sodium chromite formation on 316L(N) steel is possible in sodium containing roughly 10 μg.g -1 of oxygen for temperature lower than 650 C (reactor operating conditions).The experimental study shows that sodium chromite is formed at 650 C in the sodium containing 200 μg.g -1 of oxygen. At the same concentration and at 550 C, sodium chromite is clearly observed only for long immersion time (≥ 5000 h). Results at 450 C are more difficult to interpret. Furthermore, the steel is depleted in chromium in all cases.The results suggest the sodium chromite is dissolved in the sodium at the same time it is formed. Modelling of sodium chromite formation - approached by chromium diffusion in steel (in grain and grain boundaries -, and dissolution - assessed by transport in liquid metal - show that simultaneous formation and dissolution of sodium chromite is a possible mechanism able to explain our results. (author) [fr

Simulation experiments for a large leak sodium-water reaction analysis. Volume 4. IHTS/relief system simulation tests

International Nuclear Information System (INIS)

Ploeger, D.W.

1978-09-01

Tests were performed in which a simplified 1/8-scale model of the intermediate heat transfer system and relief system of a LMFBR was subjected to a simulated sodium-water reaction in a steam generator. Pressures in the intermediate heat exchanger (IHX) and in the pipe were measured. The flow of water through the relief system was photographed and its velocity was measured. The forces on the relief system elbows resulting from the fluid flow were also measured. The tests were performed primarily to validate pulse propagation codes used for design and for direct use as design data

SNR-steam generator design with respect to large sodium water reactions

International Nuclear Information System (INIS)

Jong, J.J. de; Kellner, A.; Florie, C.J.L.

1984-01-01

This paper deals with the experiences gained during the licensing procedure for the steam generators for the SNR 300 LMFBR regarding large sodium-water reactions. A description is given of the different calculations executed to investigate the effects of large leaks on the 85 MW helical coiled and straight tube steam generators. The investigations on the helical coiled steam generators are divided in the formulations of fluid behaviour, dynamic force calculations, dynamic response calculation and finally stress analyses. Several results are shown. The investigations on the straight tube steam generators are performed using models describing fluid-structure interaction, coupled with stress analyses. Several results are presented. A description is given of the problems and necessary construction changes during the licensing process. Advises are given for future analyses and design concepts for second generation commercial size LMFBR steam generators with respect to large leaks; based on the experience, gained with SNR 300, and using some new calculations for SNR 2. (author)

Role of aquaporin and sodium channel in pleural water movement.

Science.gov (United States)

Jiang, Jinjun; Hu, Jie; Bai, Chunxue

2003-12-16

The role of the ENaC sodium channel and aquaporin-1 (AQP1) water channel on pleural fluid dynamics in mice was investigated. 0.25 ml of hypertonic or isosmolar fluid was infused into the pleural space in anesthetized wildtype and AQP1 null mice. Pleural fluid was sampled at specified times to quantify the osmolality and volume. The sodium channel activator terbutaline increased isosmolar fluid clearance by 90% while the sodium channel inhibitor amiloride decreased it by 15%, but had no effect on osmotically driven water transport. AQP1 deletion significantly decreased osmotic water transport in pleural space by twofold, but it had no effect on isosmolar fluid clearance. Pretreatment with dexamethasone increased pleural osmotic fluid entry by 25%, while intravenous injection of HgCl2 decreased osmotic pleural water movement by 43%. These results provided evidence for a role of a sodium channel in pleural fluid absorption; AQP1 plays a major role in osmotic liquid transport but it does not affect isosmolar fluid clearance.

Quantum chemical evaluation for the stability of liquid sodium containing titanium nanoparticles

International Nuclear Information System (INIS)

Suzuki, Ai; Inaba, Kenji; Ishizawa, Yukie; Miura, Ryuji; Hatakeyama, Nozomu; Miyamoto, Akira; Saito, Jun-ichi; Ara, Kuniaki

2015-01-01

Recently, liquid sodium containing titanium nanoparticles (LSnanop) have attracted considerable attention. In this study, suspension state of Ti nanoparticle in liquid sodium was quantum chemically evaluated. The atomic interaction between Ti nanoparticles and sodium atoms in the liquid sodium medium was investigated. There were some literatures which gained quantum chemical insight into a nanoparticle with the surrounding sodium atom. However, liquid sodium medium itself together with a Ti nanoparticle under the realistic temperature has not yet been investigated theoretically. To overcome the problem of conventional theoretical method, we applied computationally low-load Tight Binding Quantum Chemical Molecular Dynamics (TB-QCMD) calculation method to investigate the suspension state of the Ti nanoparticle in liquid sodium metal. (author)

Deterioration of limestone aggregate mortars by liquid sodium in fast breeder reactor environment

Energy Technology Data Exchange (ETDEWEB)

Mohammed Haneefa, K., E-mail: mhkolakkadan@gmail.com [Department of Civil Engineering, IIT Madras, Chennai (India); Santhanam, Manu [Department of Civil Engineering, IIT Madras, Chennai (India); Parida, F.C. [Radiological Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2014-08-15

Highlights: • Limestone mortars were exposed to liquid sodium exposure at 550 °C. • Micro-analytical techniques were used to characterize the exposed specimens. • The performance of limestone mortar was greatly influenced by w/c. • The fundamental degradation mechanisms of limestone mortars were identified. - Abstract: Hot liquid sodium at 550 °C can interact with concrete in the scenario of an accidental spillage of sodium in liquid metal cooled fast breeder reactors. To protect the structural concrete from thermo-chemical degradation, a sacrificial layer of limestone aggregate concrete is provided over it. This study investigates the fundamental mechanisms of thermo-chemical interaction between the hot liquid sodium and limestone mortars at 550 °C for a duration of 30 min in open air. The investigation involves four different types of cement with variation of water-to-cement ratios (w/c) from 0.4 to 0.6. Comprehensive analysis of experimental results reveals that the degree of damage experienced by limestone mortars displayed an upward trend with increase in w/c ratios for a given type of cement. Performance of fly ash based Portland pozzolana cement was superior to other types of cements for a w/c of 0.55. The fundamental degradation mechanisms of limestone mortars during hot liquid sodium interactions include alterations in cement paste phase, formation of sodium compounds from the interaction between solid phases of cement paste and aggregate, modifications of interfacial transition zone (ITZ), decomposition of CaCO{sub 3}, widening and etching of rhombohedral cleavages, and subsequent breaking through the weakest rhombohedral cleavage planes of calcite, staining, ferric oxidation in grain boundaries and disintegration of impurity minerals in limestone.

Deterioration of limestone aggregate mortars by liquid sodium in fast breeder reactor environment

International Nuclear Information System (INIS)

Mohammed Haneefa, K.; Santhanam, Manu; Parida, F.C.

2014-01-01

Highlights: • Limestone mortars were exposed to liquid sodium exposure at 550 °C. • Micro-analytical techniques were used to characterize the exposed specimens. • The performance of limestone mortar was greatly influenced by w/c. • The fundamental degradation mechanisms of limestone mortars were identified. - Abstract: Hot liquid sodium at 550 °C can interact with concrete in the scenario of an accidental spillage of sodium in liquid metal cooled fast breeder reactors. To protect the structural concrete from thermo-chemical degradation, a sacrificial layer of limestone aggregate concrete is provided over it. This study investigates the fundamental mechanisms of thermo-chemical interaction between the hot liquid sodium and limestone mortars at 550 °C for a duration of 30 min in open air. The investigation involves four different types of cement with variation of water-to-cement ratios (w/c) from 0.4 to 0.6. Comprehensive analysis of experimental results reveals that the degree of damage experienced by limestone mortars displayed an upward trend with increase in w/c ratios for a given type of cement. Performance of fly ash based Portland pozzolana cement was superior to other types of cements for a w/c of 0.55. The fundamental degradation mechanisms of limestone mortars during hot liquid sodium interactions include alterations in cement paste phase, formation of sodium compounds from the interaction between solid phases of cement paste and aggregate, modifications of interfacial transition zone (ITZ), decomposition of CaCO 3 , widening and etching of rhombohedral cleavages, and subsequent breaking through the weakest rhombohedral cleavage planes of calcite, staining, ferric oxidation in grain boundaries and disintegration of impurity minerals in limestone

Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

International Nuclear Information System (INIS)

Jang, Sung Hyun; Takata, Takashi; Yamaguchi, Akira; Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

2015-01-01

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Jang, Sung Hyun; Takata, Takashi [Graduate School of Engineering, Osaka University, Osaka (Japan); Yamaguchi, Akira [Graduate School of Engineering, The University of Tokyo, Ibaraki (Japan); Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki [Japan Atomic Energy Agency, Ibaraki (Japan)

2015-10-15

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

Experimental study of the attenuation waves oriented to transients caused by the sodium-water explosive reaction in fast reactors

International Nuclear Information System (INIS)

Pedroso, L.J.

1990-01-01

One of the problems related to fluid-structure interaction that can compromise the structural integrity of components of a fast reactor is the explosion caused by the sodium-water reaction, in the case of a flood at the level of the thermic exchange wall at the steam generator. In this paper we have considered the aspects of the pressure-waves damping caused by the reaction, when these waves transverse certain perforated structures. In order to solve this problem, we also adopted a parametric experimental approach, using a scale model (RIO test rig). (author)

Evaluation of heat transfer tube failure propagation due to sodium-water reaction in steam generator

International Nuclear Information System (INIS)

Nei, Hiromichi

1978-01-01

An evaluation was made of heat transfer tube failure propagation due to sodium-water reaction wastage in a sodium heated steam generator, by comparing an empirically derived wastage equation with leak detector responses. The experimental data agreed well with the wastage equation even for different values of distance-to-nozzle diameter ratio, though the formula had been based on wastage data obtained for only one given distance. The time taken for failure propagation was estimated for a prototype steam generator, and compared with the responses characteristics of acoustic detectors and level gages. It was found that there exists a range of leak rate between 0.5 and 100 g/sec, where the level gage can play a useful role in leak detection. The acoustic detector can be expected to respond more rapidly than the cover gas pressure gage, if noise is kept below ten times the value observed in an experimental facility, SWAT-2. (auth.)

The kinetics of hydrogen removal from liquid sodium

International Nuclear Information System (INIS)

Gwyther, J.R.; Whittingham, A.C.

1981-01-01

The rates of hydrogen removal from liquid sodium-sodium hydride mixtures have been measured as a function of sodium stirring rate at temperatures up to 420 0 C. Two techniques have been employed - removal under continuous evacuation in which hydrogen flow rates were measured using a capillary flow technique and by argon purging in which hydrogen concentrations in the argon carrier gas were measured by gas chromatography. The results have been used to assess the feasibility of thermal decomposition of sodium hydride for the regeneration of hydride-laden LMFBR cold traps. Studies on the kinetics of desorption of hydrogen from solution in liquid sodium at temperatures up to 400 0 C are also presented and possible kinetic mechanisms discussed. (orig.)

Computational studies of atmospherically-relevant chemical reactions in water clusters and on liquid water and ice surfaces.

Science.gov (United States)

Gerber, R Benny; Varner, Mychel E; Hammerich, Audrey D; Riikonen, Sampsa; Murdachaew, Garold; Shemesh, Dorit; Finlayson-Pitts, Barbara J

2015-02-17

CONSPECTUS: Reactions on water and ice surfaces and in other aqueous media are ubiquitous in the atmosphere, but the microscopic mechanisms of most of these processes are as yet unknown. This Account examines recent progress in atomistic simulations of such reactions and the insights provided into mechanisms and interpretation of experiments. Illustrative examples are discussed. The main computational approaches employed are classical trajectory simulations using interaction potentials derived from quantum chemical methods. This comprises both ab initio molecular dynamics (AIMD) and semiempirical molecular dynamics (SEMD), the latter referring to semiempirical quantum chemical methods. Presented examples are as follows: (i) Reaction of the (NO(+))(NO3(-)) ion pair with a water cluster to produce the atmospherically important HONO and HNO3. The simulations show that a cluster with four water molecules describes the reaction. This provides a hydrogen-bonding network supporting the transition state. The reaction is triggered by thermal structural fluctuations, and ultrafast changes in atomic partial charges play a key role. This is an example where a reaction in a small cluster can provide a model for a corresponding bulk process. The results support the proposed mechanism for production of HONO by hydrolysis of NO2 (N2O4). (ii) The reactions of gaseous HCl with N2O4 and N2O5 on liquid water surfaces. Ionization of HCl at the water/air interface is followed by nucleophilic attack of Cl(-) on N2O4 or N2O5. Both reactions proceed by an SN2 mechanism. The products are ClNO and ClNO2, precursors of atmospheric atomic chlorine. Because this mechanism cannot result from a cluster too small for HCl ionization, an extended water film model was simulated. The results explain ClNO formation experiments. Predicted ClNO2 formation is less efficient. (iii) Ionization of acids at ice surfaces. No ionization is found on ideal crystalline surfaces, but the process is efficient on

Application of the subchannel analysis code COBRA III C for liquid sodium

International Nuclear Information System (INIS)

Nissen, K.L.

1981-01-01

The subchannel-analysis code COBRA III C was developed to gain knowledge of mass flow and temperature distribution in rod bundles of light water reactors. A comparison of experimental results for the temperature distribution in a 19 rod bundle with calculations done by the computer program shows the capability of COBRA III C to handle liquid sodium cooling. The code needs sodium properties as well as changed correlations for turbulent mixing and heat transfer at the rod. (orig.) [de

Effect of liquid sodium on long-term properties of austenitic steels

International Nuclear Information System (INIS)

Svoboda, V.; Merta, J.; Slach, J.

The effect is discussed of liquid sodium on the long-term properties of austenitic steels corresponding to the ASI 304 and ASI 316 types, mainly of steel CSN 17348. The choice is described of test specimens and of the experimental sodium test equipment. Testing was carried out using the so-called indirect method, i.e., the liquid sodium effect was assessed using the results of creep tests of two groups of specimens, one exposed to sodium and the other to the inert argon atmosphere. Otherwise the tests proceeded under identical conditions. The sodium stand had been manufactured for exposure of test specimens to liquid sodium. The morphology of specimen surfaces was studied by the JSN-50A electron microscope. The results of testing steel CSN 17348-AKV EXTRA S exposed to liquid sodium containing 10 ppm of oxygen at a temperature of 550 degC showed a significant sodium effect on the basic mechanical properties, on long-term creep strength and on the metallurgical properties. (Oy)

Mechanical behavior and fracture characterization of the T91 martensitic steel in liquid sodium environment

International Nuclear Information System (INIS)

Hamdane, Ouadie

2012-01-01

The T91 martensitic steel is designed to constitute structural material of future sodium fast reactors of fourth generation, where it will be subjected to stresses in presence of liquid sodium. This study presents a qualitative and quantitative estimate of the sensitivity of T91 steel towards the phenomenon of liquid metal embrittlement. The effect of liquid sodium on T91 steel was studied and quantified according to the temperature and the cross head rate displacement, by using a set-up of Small Punch Test, three and four bending test, developed in laboratory. Mechanical tests in sodium environment are carried out inside a Plexiglas cell, conceived and developed at the laboratory. The atmosphere inside this cell is severely purified and controlled, in order to avoid on the one hand an explosive reaction of sodium with moisture, or an ignition with oxygen, and on the other hand to minimize the presence of impurities in liquid sodium used. The presence of sodium accelerates T91 steel fracture at low temperature, without modifying its ductile character. The T91 pre-immersion in sodium makes it possible to dissolve the protective layer of chromium oxide, and to obtain an intimate contact with the molten metal. However, pre-immersion generates a surface defects which cause a partial embrittlement by sodium. The hardening of T91 steel by heat treatment with a tempering temperature of 550 C (T91-TR550) causes a total embrittlement of steel in presence of sodium, with and without pre-immersion. The rupture of the T91-TR550 steel takes then place by intergranular de-cohesion, corresponding to the crack initiation phase, followed by laths de-cohesion, corresponding to the phase of propagation of these cracks. The mechanism suggested in this study is based on the intergranular penetration of sodium, supported by the presence of segregated impurities such phosphorus, and by the plastic deformation [fr

Effect of sodium bicarbonate and varying concentrations of sodium chloride in brine on the liquid retention of fish (Pollachius virensL.) muscle

DEFF Research Database (Denmark)

Åsli, Magnus; Ofstad, Ragni; Böcker, Ulrike

2016-01-01

BACKGROUND Negative health effects associated with excessive sodium (Na) intake have increased the demand for tasty low-Na products (<2% NaCl) rather than traditional heavily salted fish products (∼20% NaCl). This study investigates the causes of improved yield and liquid retention of fish muscle...... in greater intracellular space at 30 and 60 g kg−1 NaCl. CONCLUSION Sodium bicarbonate addition to low-salt solutions can improve yield and flesh quality of fish muscle owing to altered water mobility and wider space between the muscle cells......BACKGROUND Negative health effects associated with excessive sodium (Na) intake have increased the demand for tasty low-Na products (fish products (∼20% NaCl). This study investigates the causes of improved yield and liquid retention of fish muscle...

Water leaks in sodium-heated fast reactor boilers

International Nuclear Information System (INIS)

Hayes, D.J.

1978-01-01

Constraints on plant design which may result from considerations of leak behaviour and leak detection limits are briefly considered. The sodium-water interface and reactions, the behaviour of small leaks, hydrogen bubbles and detection methods, including galvanic cell methods, are included. (UK)

Modeling and analysis of liquid deuterium-water reactions

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1995-01-01

This Presentation highlights the following: Overview of LD 2 -water reactions their connections to research reactors with cold sources; some key features and ingredients of vapor explosions in general; Examination of results of 1970 experiment at Grenoble Nuclear Research Center; Thermodynamic evaluations of energetics of explosive LD 2 -D 2 O reactions. This presentation concentrates only on the technical aspects of LD 2 /LH 2 - water reactions; it is not intended to draw/imply safety-related conclusions for research reactors

Development of Sodium Technology for LMR

Energy Technology Data Exchange (ETDEWEB)

Kim, B. H.; Nam, H. Y.; Kim, T. J. (and others)

2007-03-15

In the experiments to investigate the characteristics of the free surface fluctuation in a vessel, the experimental correlation was modeled to describe the free surface fluctuation in the upper plenum of a liquid metal reactor within 95% reliability and 2.4% error. The correlation was used to verify the computational model. The new conceptual flowmeters were suggested to measure the sodium flow for the reliability enhancement. The electromagnetic flowmeter with permanent magnet showed a good linearity and repeatability. For reuse of the sodium contaminated component, CO{sub 2} bubbling method was developed. Sodium in 0.3mm crevice specimen was removed completely. The optimum condition for the used sodium treatment was deduced to estimate which reaction is more safe and adequate for operation condition by analyzing the reactivity alleviation condition and the reaction rate with the control of sodium hydroxide concentration A series of tests were carried out to investigate the enlargement rate of the nozzle hole itself and the sodium-water reaction temperature associated with needle-like jets of a high-pressure water/steam into the sodium side of a steam generator. The size of the nozzle hole became larger with an increased duration of the steam injection both for the 2.25Cr-1Mo and M9Cr-1Mo steels by a self-wastage phenomenon. For developing the SWR acoustic leak detection technology, the tool prepared by the LabVIEW was installed with the system, and confirmed the performance of the on-line acoustic leak detection tool using the SWR leak signal acquired in the KAERI facility.

Study of the moderating effect of salts on the sodium-water reaction on the cleaning of irradiated fuel assemblies from fast neutron reactors, using fluid sodium heat transfer

International Nuclear Information System (INIS)

Lacroix, Marie

2014-01-01

Within the framework of the development of generation IV reactors one of the research tracks is related to the development of fast neutron reactors using fluid sodium heat transfer. The CEA (French Alternative Energies and Atomic Energy Commission) plans to build a prototype of reactor of this type called 'ASTRID'. To address development requirements for this prototype, research is in progress on the reactor's availability and in particular on the reduction of the washing duration for residual sodium fuel assemblies during their discharge. In fact, because sodium is very reactive with water (presently the only available process), the washing is done, for example, by very gradual addition. A solution currently being studied at the CEA and which is the subject of this thesis report consists of the addition of an aqueous salts solutions to the washing water in order to slow down the kinetic reaction. This doctoral dissertation describes the various salts, which have been evaluated and aims to explain their action mode. (author) [fr

Plan for support of large-plant (post-CRBR) needs in large-leak sodium-water reaction area

International Nuclear Information System (INIS)

Whipple, J.C.

1980-03-01

Work in the large leak test and analysis area of steam generator development has been carried out at GE-ARSD under 189a SG037 since 1973. The currently planned master schedule for the SG037 program is shown. Principal activities are the large leak testing program being carried out at the Large Leak Test Rig and the analysis methods development. The plan for supporting the large plant (post-CRBR) needs in the large leak sodium-water reaction area is outlined. Most of the needs will be answered in the current SG037 large leak program

Combined convective heat transfer of liquid sodium flowing across tube banks

International Nuclear Information System (INIS)

Ma, Ying; Sugiyama, Ken-ichiro; Ishiguro, Ryoji

1989-01-01

In order to clarify the heat transfer characteristics of combined convection of liquid sodium, a numerical analysis is performed for liquid sodium which flows through a single horizontal row of tubes in the direction of gravity. The correlation of heat transfer characteristics between liquid sodium and ordinary fluids is also discussed. The heat transfer characteristics at large Reynolds numbers are improved when the Richardson number is increased, and the improvement rate is enlarged with increase in p/d value, since convection effect is relatively large. However heat transfer coefficients do not differ from those of forced convection at small Reynolds numbers even when the Richardson number reaches a high value because of conduction effect. A good consistence of heat transfer characteristics of combined convection between liquid sodium and air is obtained at the same Peclet number and Richardson number. This means that the fundamental heat transfer characteristics of combined convection of liquid sodium can be investigated with ordinary fluids. (author)

Acoustic Leak Detection Testing Using KAERI Sodium-Water Reaction Signals for a SFR Steam Generator

International Nuclear Information System (INIS)

Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man; Kim, Byung-Ho; Hahn, Do-Hee; Yugay, Valeriy S.

2009-01-01

The results of an experimental study of water in a sodium leak noise spectrum formation at 0.004-0.54 g/sec, various rates of water into a sodium leak, smaller than 1.0 g/sec, are presented. We focused on studying a micro leak detection with an increasing rate of water into sodium. On the basis of the experimental leak noise data manufactured in KAERI the simple dependency of an acoustic signal level from the rate of a micro and small leak at different frequency bands is presented to understand the principal analysis for the development of an acoustic leak detection methodology used in a K- 600 steam generator

Liquid-Crystalline Ionic Liquids as Ordered Reaction Media for the Diels-Alder Reaction.

Science.gov (United States)

Bruce, Duncan W; Gao, Yanan; Canongia Lopes, José Nuno; Shimizu, Karina; Slattery, John M

2016-11-02

Liquid-crystalline ionic liquids (LCILs) are ordered materials that have untapped potential to be used as reaction media for synthetic chemistry. This paper investigates the potential for the ordered structures of LCILs to influence the stereochemical outcome of the Diels-Alder reaction between cyclopentadiene and methyl acrylate. The ratio of endo- to exo-product from this reaction was monitored for a range of ionic liquids (ILs) and LCILs. Comparison of the endo:exo ratios in these reactions as a function of cation, anion and liquid crystallinity of the reaction media, allowed for the effects of liquid crystallinity to be distinguished from anion effects or cation alkyl chain length effects. These data strongly suggest that the proportion of exo-product increases as the reaction media is changed from an isotropic IL to a LCIL. A detailed molecular dynamics (MD) study suggests that this effect is related to different hydrogen bonding interactions between the reaction media and the exo- and endo-transition states in solvents with layered, smectic ordering compared to those that are isotropic. © 2016 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Ultrasonic inspection for wastage in the LMFBR steam generator due to sodium--water reactions

International Nuclear Information System (INIS)

Neely, H.H.; Renger, L.

1977-01-01

As part of a program to study the results of large sodium-water reactions in the LMFBR Steam Generator, a boreside ultrasonic inspection device was developed to measure the wall thickness and diameter of the 2- 1 / 4 Cr-1 Mo, 0.397 in. I.D. steam tubes. The reaction was created in a near prototype steam generator by guillotine-type rupture of a steam tube, while the generator was at operating conditions. Wastage occurred on the surrounding tubes due to the high temperature reaction. The UT test instrument was designed to operate with a 15 MHz transducer in the pulse-echo shear-wave mode, with a sampling rate of 10 4 /sec. System outputs are diameter, wall thickness, attitude and axial position of the transducer. All are displayed digitally and may be recorded. Measurements are fed into a computer for later retrieval, and/or cascaded outputs into an x-y recorded displaying either out-of-limit or thickness data. The UT data taken in this experiment were consistent with physical measurements on a tube which was removed from the generator after the test. A machined flat 1 / 8 -inch long and 0.002-inch deep could readily be detected

Process of transformation of radioactive waste of metal sodium into soda solution by reaction with an alcohol followed by hydrolysis

International Nuclear Information System (INIS)

Chevalier, Gerard; Mathurin, Rene.

1981-09-01

Reviews of the literature and of the laboratory tests are followed by a presentation of the results obtained during experiments carried out on a model with some ten grams of sodium contaminated by radioactive materials and on an industrial pre-pilot with several kilograms of non-contaminated sodium. Sodium is converted into alcoholate through the action of ethylcarbitol (CH 3 CH 2 OCH 2 CH 2 OCH 2 OH) on liquid sodium in suspension in xylene at 110 deg C. Once the reaction is complete, xylene is distillated and the alcoholate is in solution in an axcess of alcohol. Hydrolysis by water gives out the initial alcohol which is then extracted from the aqueous phase by toluene. All these operations are carried out in inert atmosphere (nitrogen). Sodium is thus converted into a sodium hydroxide aqueous solution with emission of hydrogen, the intermediate products (alcohol, xylene, toluene) being recyled. The process is reliable and recycling of organic products is favourable economically. The advantage of the method is to concentrate nearly all the radioactivity of the contaminated sodium in the aqueous phase, thus avoiding the dispersion of activity especially with the gaseous effluents. Finally, data are given allowing to consider the realization of a pilot with a weekly capacity of 100 kg of sodium, in semi-continuous operation [fr

Interaction of Liquid Sodium With 304 Stainless Steel

National Research Council Canada - National Science Library

Moberly, John

1968-01-01

The effect of a liquid sodium environment on 304 stainless steel has important engineering significance because of the potential use of this liquid-metal solid-metal system in fast breeder reactors...

Thermodynamic and transport properties of sodium liquid and vapor

International Nuclear Information System (INIS)

Fink, J.K.; Leibowitz, L.

1995-01-01

Data have been reviewed to obtain thermodynamically consistent equations for thermodynamic and transport properties of saturated sodium liquid and vapor. Recently published Russian recommendations and results of equation of state calculations on thermophysical properties of sodium have been included in this critical assessment. Thermodynamic properties of sodium liquid and vapor that have been assessed include: enthalpy, heat capacity at constant pressure, heat capacity at constant volume, vapor pressure, boiling point, enthalpy of vaporization, density, thermal expansion, adiabatic and isothermal compressibility, speed of sound, critical parameters, and surface tension. Transport properties of liquid sodium that have been assessed include: viscosity and thermal conductivity. For each property, recommended values and their uncertainties are graphed and tabulated as functions of temperature. Detailed discussions of the analyses and determinations of the recommended equations include comparisons with recommendations given in other assessments and explanations of consistency requirements. The rationale and methods used in determining the uncertainties in the recommended values are also discussed

Report of sodium cavitation

International Nuclear Information System (INIS)

Murai, Hitoshi; Shima, Akira; Oba, Toshisaburo; Kobayashi, Ryoji; Hashimoto, Hiroyuki

1975-01-01

The damage of components for LMFBRs due to sodium cavitation is serious problem. This report summarizes the following items, (1) mechanism of the incipience of sodium cavitation, (2) damage due to sodium cavitation, (3) detection method for sodium cavitation, and (4) estimation method for sodium cavitation by the comparison with water cavitation. Materials were collected from the reports on liquid metal cavitation, sodium cavitation and water cavitation published from 1965 to now. The mechanism of the incipience of sodium cavitation cavitation parameters (mean location, distributed amount or occurrence aspect and stability), experiment of causing cavitation with Venturi tube, and growth of bubbles within superheated sodium. The sodium cavitation damage was caused by magnetostriction vibration method and with Venturi tube. The state of damage was investigated with the cavitation performance of a sodium pump, and the damage was examined in view of the safety of LMFBR plants. Sodium cavitation was detected with acoustic method, radiation method, and electric method. The effect of physical property of liquid on incipient cavitation was studied. These are thermodynamic effect based on quasistatic thermal equilibrium condition and the effect of the physical property of liquid based on bubble dynamics. (Iwase, T.)

How Sodium Chloride Salt Inhibits the Formation of CO2 Gas Hydrates.

Science.gov (United States)

Holzammer, Christine; Finckenstein, Agnes; Will, Stefan; Braeuer, Andreas S

2016-03-10

We present an experimental Raman study on how the addition of sodium chloride to CO2-hydrate-forming systems inhibits the hydrate formation thermodynamically. For this purpose, the molar enthalpy of reaction and the molar entropy of reaction for the reaction of weakly hydrogen-bonded water molecules to strongly hydrogen bonded water molecules are determined for different salinities from the Raman spectrum of the water-stretching vibration. Simultaneously, the influence of the salinity on the solubility of CO2 in the liquid water-rich phase right before the start of hydrate formation is analyzed. The results demonstrate that various mechanisms contribute to the inhibition of gas hydrate formation. For the highest salt concentration of 20 wt % investigated, the temperature of gas hydrate formation is lowered by 12 K. For this concentration the molar enthalpy and entropy of reaction become smaller by 50 and 20%, respectively. Concurrently, the solubility of carbon dioxide is reduced by 70%. These results are compared with data in literature for systems of sodium chloride in water (without carbon dioxide).

A fundamental study on sodium-water reaction in the double pool LMFBR, (3)

International Nuclear Information System (INIS)

Uotani, Masaki; Kumagai, Hiromichi; Nishi, Yoshihisa; Yoshida, Kazuo

1989-01-01

The double pool LMFBR is an innovative reactor that Central Research Institute of Electric Power Industry proposed for the purpose of reducing the construction cost of FBRs, and it is characterized by immersing steam generators in the annular plenum formed between the primary vessel and the outer secondary vessel. Therefore, it is expected that the pressure behavior at the time of sodium-water reaction due to the breaking of heating tubes is largely different from the case of steam generators of conventional FBRs. In order to ensure the soundness of the primary vessel that containes the reactor core, it is necessary to sufficiently grasp the pressure behavior in the plenum, and this basic experiment and analysis are related to the pressure behavior due to piston motion that arises in the initial period of quasi-steady pressure. About 1/10 scale annular plenum was used, and the generation of reaction product gas was simulated by the release of nitrogen. When gas was released in the plenum, the highest pressure rise occurred in the initial period of release, and thereafter, periodic variation arose. The pressure waveform and the value of pressure rise as the results of the model analysis agreed well with the measured results. (K.I.)

A simple method to prepare ZnO and Al(OH)3 nanorods by the reaction of the metals with liquid water

International Nuclear Information System (INIS)

Panchakarla, L.S.; Shah, M.A.; Govindaraj, A.; Rao, C.N.R.

2007-01-01

Reaction of liquid water with Zn and Al powders and foils have been investigated in the 25-75 deg. C range. The reaction of Zn metal powder with water in this temperature range yields ZnO nanorods. The diameter of the nanorods decreases slightly with the increase in the reaction temperature, accompanied by an increase in the relative intensity of UV emission band. Zn metal foils also yield ZnO nanorods on reaction with water in the 25-75 deg. C range. Reaction of Al metal powder or foil with water in the 25-75 deg. C range yields Al(OH) 3 nanorods. The formation of ZnO and Al(OH) 3 nanorods by the reaction of the metals with water is suggested to occur because of the decomposition of water by the metal giving hydrogen. - Graphical abstract: The reaction of water at a temperature in the 25-27 deg. C range with zinc metal gives rise to ZnO nanorods; with Al metal water gives Al(OH) 3 nanorods

Reaction velocity of sodium hydration in humid air and sodium carbonation in humid carbon dioxide atmosphere. Fundamental study on sodium carbonate process in FBR bulk sodium coolant disposal technology

International Nuclear Information System (INIS)

Tadokoro, Yutaka; Yoshida, Eiichi

1999-11-01

A sodium carbonate processing method, which changes sodium to sodium carbonate and/or sodium bicarbonate by humid carbon dioxide, has been examined and about to be applied to large test loops dismantling. However, that the basic data regarding the progress of the reaction is insufficient on the other hand, is a present condition. The present report therefore aims at presenting basic data regarding the reaction velocity of sodium hydration in humid air and sodium carbonation in humid carbon dioxide atmosphere, and observing the reaction progress, for the application to large test loops dismantling. The test result is summarized as follows. (1) Although the reaction velocity of sodium varied with sodium specimen sizes and velocity measurement methods, the reaction velocity of sodium hydration was in about 0.16 ∼ 0.34 mmh -1 (0.016 ∼ 0.033g cm -2 h -1 , 6.8x10 -4 ∼ 1.4x10 -3 mol cm -2 h -1 ) and that of sodium carbonation was in about 0.16 ∼ 0.27mmh -1 (0.016 ∼ 0.023g cm -2 h -1 , 6.8x10 -4 ∼ 1.1x10 -3 mol cm -2 h -1 ) (26 ∼ 31degC, RH 100%). (2) The reaction velocity of sodium in carbon dioxide atmosphere was greatly affected by vapor partial pressure (absolutely humidity). And the velocity was estimated in 0.08 ∼ 0.12mmh -1 (0.008 ∼ 0.012g cm -2 h -1 , 3.4x10 -4 ∼ 5.2x10 -4 mol cm -2 h -1 ) in the carbon dioxide atmosphere, whose temperature of 20degC and relative humidity of 80% are assumed real sodium carbonate process condition. (3) By the X-ray diffraction method, NaOH was found in humid air reaction product. Na 2 CO 3 , NaHCO 3 were found in carbon dioxide atmosphere reaction product. It was considered that Sodium changes to NaOH, and subsequently to NaHCO 3 through Na 2 CO 3 . (4) For the application to large test loops dismantling, it is considered possible to change sodium to a target amount of sodium carbonate (or sodium bicarbonate) by setting up gas supply quantity and also processing time appropriately according to the surface area

Experimental and analytical studies of iodine mass transfer from xenon-iodine mixed gas bubble to liquid sodium pool

International Nuclear Information System (INIS)

Miyahara, S.; Sagawa, N.; Shimoyama, K.

1996-01-01

In the fuel pin failure accident of a liquid metal fast reactor, volatile fission products play an important role in the assessment of radiological consequences. Especially the radioisotopes of elemental iodine are important because of their high volatility and of the low permissible dose to human thyroid. The released iodines are known to be retained in the coolant sodium as sodium iodide due to the chemical affinity between alkali metals and halogens. However, the xenon and krypton released with iodines into the sodium pool as bubbles may influence the reaction rate of iodine with sodium during the bubble rising. So far, the only few experimental results have been available concerning the decontamination factor (DF: the ratio of the initial iodine mass in the mixed gas bubble to the released mass into the cover gas) of iodine in this phenomenon. Therefore, experimental and analytical studies were carried out to study the mass transfer of iodine from a xenon-iodine mixed gas bubble to the liquid sodium pool. In the experiments, the bubble was generated in the sodium pool by cracking a quartz ball which contains the xenon-iodine mixed gas and then, the mixed gas released into the argon cover gas was collected to determine the transferred iodine mass into the pool. A rising velocity of the bubble was measured by Chen-type void sensors arranged vertically in the pool. From the measured rising velocity and another observation of bubble behavior in simulated water experiments, it is found that the generated bubble breaks up into several smaller bubbles of spherical cap type during the rising period. Transferred iodine mass per unit initial bubble volume from the bubble to the sodium pool shows increases with increasing time and the initial iodine concentration. A mass transfer rate obtained by differentiating the transferred iodine mass with respect to the time indicates a rapid decrease just after the bubble generation and a slow decrease for the successive period

Hydrogen generation systems and methods utilizing sodium silicide and sodium silica gel materials

Energy Technology Data Exchange (ETDEWEB)

Wallace, Andrew P.; Melack, John M.; Lefenfeld, Michael

2017-12-19

Systems, devices, and methods combine thermally stable reactant materials and aqueous solutions to generate hydrogen and a non-toxic liquid by-product. The reactant materials can sodium silicide or sodium silica gel. The hydrogen generation devices are used in fuels cells and other industrial applications. One system combines cooling, pumping, water storage, and other devices to sense and control reactions between reactant materials and aqueous solutions to generate hydrogen. Springs and other pressurization mechanisms pressurize and deliver an aqueous solution to the reaction. A check valve and other pressure regulation mechanisms regulate the pressure of the aqueous solution delivered to the reactant fuel material in the reactor based upon characteristics of the pressurization mechanisms and can regulate the pressure of the delivered aqueous solution as a steady decay associated with the pressurization force. The pressure regulation mechanism can also prevent hydrogen gas from deflecting the pressure regulation mechanism.

Exchange reaction between tritiated hydrogen and water vapor

International Nuclear Information System (INIS)

Yamada, Koichi; Takano, Kenichi; Watanabe, Tamaki.

1979-01-01

Exchange reaction of tritiated hydrogen to water vapor under the condition of tritium gas concentration between 1 μCi/l and 1 mCi/l was studied. Tritium gas with hydrogen gas of 5 Torr and water of 20 mg were enclosed in a Pyrex glass ampule with volume of about 100 ml. The mixed gas with water vapor was heated with electric furnace. The heating time was between 2 and 100 hr, and the temperature was 776, 725, 675, 621, and 570.5 0 K. After heating, tritiated water was trapped with liquid nitrogen, and counted with a liquid scintillation counter. The radioactive concentration of initial tritiated hydrogen was measured with a calibrated ionization chamber. The main results obtained are as follows; 1) the concentration of produced tritiated water is well proportioned to that of initial tritiated hydrogen, 2) the activation energy of exchange reaction from tritiated hydrogen to tritiated water is 26.2 kcal/mol and that of inverse reaction is 27.4 kcal/mol, 3) the reaction rate at room temperature which calculated with activation energy is 1.04 x 10 -13 day -1 , and then exchange reaction at room temperature is negligible. (author)

Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors

International Nuclear Information System (INIS)

Vega R, A. K.; Espinosa P, G.; Gomez T, A. M.

2016-09-01

The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

Behaviour of oxygen in liquid sodium

International Nuclear Information System (INIS)

Torre Cabezas, M. de la

1975-01-01

In this work, the vacuum distillation method has been used for the determination of oxygen concentration in liquid sodium. During this investigation, more than 800 analyses have been made and a fluctuation of between 15 and 20$ has been noted in the results. The performance of a cold trap to remove oxygen from sodium has been studied and the corresponding mass transfer coefficient evaluated. The value of this coefficient was in good agreement with those achieved by other workers. (Authors) 69 refs

Review of the sodium fire experiments including sodium-concrete-reactions and summary of the results

International Nuclear Information System (INIS)

Cherdron, W.

1996-01-01

In the technical and design concept of containment systems of sodium cooled breeder reactors it has to be considered, that leakages in sodium pipes lead to sodium fires. The temperature and pressure rise caused by sodium fires makes it indispensable to analyse these accidents to be able to assess the safety of the whole system. Generally sodium leakages may lead to three different types of fires with different consequences. The main influences are the geometry of the leakage, shape, size, location, and the sodium conditions, such as temperature, flow rate and velocity. It must be also considered the reaction of sodium with surfaces like concrete. The paper gives an overview over all the sodium fire experiments performed in the FAUNA-facility (220 m 3 ) of the Forschungszentrum Karlsruhe in the years 1979 to 1993. The experimental program started with the investigation of pool fires on burning areas between 2 and 12 m 2 with up to 500 kg of Sodium. The experiments had been continued with 3 combined fires and 40 experiments on spray fires. 7 experiments on sodium-concrete reactions completed the program. (author)

Sodium safety manual

International Nuclear Information System (INIS)

Hayes, D.J.; Gardiner, R.L.

1980-09-01

The sodium safety manual is based upon more than a decade of experience with liquid sodium at Berkeley Nuclear Laboratories (BNL). It draws particularly from the expertise and experience developed in the course of research work into sodium fires and sodium water reactions. It draws also on information obtained from the UKAEA and other sodium users. Many of the broad principles will apply to other Establishments but much of the detail is specific to BNL and as a consequence its application at other sites may well be limited. Accidents with sodium are at best unpleasant and at worst lethal in an extremely painful way. The object of this manual is to help prevent sodium accidents. It is not intended to give detailed advice on specific precautions for particular situations, but rather to set out the overall strategy which will ensure that sodium activities will be pursued safely. More detail is generally conveyed to staff by the use of local instructions known as Sodium Working Procedures (SWP's) which are not reproduced in this manual although a list of current SWP's is included. Much attention is properly given to the safe design and operation of larger facilities; nevertheless evidence suggests that sodium accidents most frequently occur in small-scale work particularly in operations associated with sodium cleaning and special care is needed in all such cases. (U.K.)

A Simple Quantitative Synthesis: Sodium Chloride from Sodium Carbonate.

Science.gov (United States)

Gold, Marvin

1988-01-01

Describes a simple laboratory procedure for changing sodium carbonate into sodium chloride by adding concentrated HCl to cause the reaction and then evaporating the water. Claims a good stoichiometric yield can be obtained in one three-hour lab period. Suggests using fume hood for the reaction. (ML)

Water simulation of sodium reactors

International Nuclear Information System (INIS)

Grewal, S.S.; Gluekler, E.L.

1981-01-01

The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

Ionic conductivity of ternary electrolyte containing sodium salt and ionic liquid

International Nuclear Information System (INIS)

Egashira, Minato; Asai, Takahito; Yoshimoto, Nobuko; Morita, Masayuki

2011-01-01

Highlights: ► Ternary electrolyte containing NaBF 4 , polyether and ionic liquid has been prepared. ► The conductivity of the electrolytes has been evaluated toward content of ionic liquid. ► The conductivity shows maximum 1.2 mS cm −1 and is varied in relation to solution structure. - Abstract: For the development of novel non-aqueous sodium ion conductor with safety of sodium secondary cell, non-flammable ionic liquid is attractive as electrolyte component. A preliminary study has been carried out for the purpose of constructing sodium ion conducting electrolyte based on ionic liquid. The solubility of sodium salt such as NaBF 4 in ionic liquid is poor, thus the ternary electrolyte has been prepared where NaBF 4 with poly(ethylene glycol) dimethyl ether (PEGDME) as coordination former is dissolved with ionic liquid diethyl methoxyethyl ammonium tetrafluoroborate (DEMEBF 4 ). The maximum conductivity among the prepared solutions, ca. 1.2 mS cm −1 at 25 °C, was obtained when the molar ratio (ethylene oxide unit in PEGDME):NaBF 4 :DEMEBF 4 was 8:1:2. The relationship between the conductivity of the ternary electrolyte and its solution structure has been discussed.

Gas-Liquid Precipitation of water dissolved heavy metal ions using hydrogen sulfide gas

NARCIS (Netherlands)

Al Tarazi, M.Y.M.

2004-01-01

Precipitation of solids promoted by gas-liquid reactions is applied in many industrial processes such as the production of ammonium phosphate, ammonium sulphate, barium carbonate, calcium carbonate, calcium fluoride, ypsum (calcium sulphate), goethite, sodium bicarbonate, strontium carbonate and

Liquid-liquid extraction of uranium (VI) using Cyanex 272 in kerosene from sodium salicylate medium

International Nuclear Information System (INIS)

Kamble, Pravin N.; Mohite, Baburao S.; Suryavanshi, Vishal J.; Salunkhe, Suresh T.

2015-01-01

Liquid-liquid extraction of uranium (VI) from sodium salicylate media using Cyanex 272 in kerosene has been carried out. Uranium (VI) was quantitatively extracted from 1x10 -4 M sodium salicylate with 5x10 -4 M Cyanex 272 in kerosene. It was stripped quantitatively from the organic phase with 4M HCl and determined spectrophotometrically with arsenazo(III) at 600 nm. The effects of concentrations of sodium salicylate, metal ions and strippants have been studied. Separation of uranium (VI) from other elements was achieved from binary as well as from multicomponent mixtures. The method is simple, rapid and selective with good reproducibility (approximately ±2%). (author)

Intermediate-scale sodium-concrete reaction tests with basalt and limestone concrete

International Nuclear Information System (INIS)

Hassberger, J.A.; Muhlestein, L.D.

1981-01-01

Ten tests were performed to investigate the chemical reactions and rate and extent of attack between sodium and basalt and limestone concretes. Test temperatures ranged from 510 to 870 0 C (950 to 1600 0 F) and test times from 2 to 24 hours. Sodium hydroxide was added to some of the tests to assess the impact of a sodium hydroxide-aided reaction on the overall penetration characteristics. Data suggest that the sodium penetration of concrete surfaces is limited. Penetration of basalt concrete in the presence of sodium hydroxide is shown to be less severe than attack by the metallic sodium alone. Presence of sodium hydroxide changes the characteristics of sodium penetration of limestone concrete, but no major differences in bulk penetration were observed as compared to penetration by metallic sodium

Optimization of the cold trap design for the KASOLA sodium facility

International Nuclear Information System (INIS)

Onea, Alexandru; Lux, Martin; Hering, Wolfgang

2012-01-01

The KASOLA (KArlsruhe SOdium LAboratory) experimental facility is currently under construction at Karlsruhe Institute of Technology. The facility serves for research activities on thermal-hydraulics for liquid metal operated systems for transmutation (fast systems, normal operation, transient behaviour, testing of emergency cooling systems), accelerator target development, applications and development of free surface liquid metal targets for accelerators, as well as feasibility studies of liquid metals for solar applications. Supporting heat transfer studies regarding the development of turbulent liquid metal heat transfer models for CFD tools are also foreseen. In sodium operated facilities several impurities can be released during operation, e.g. argon, oxygen, hydrogen, carbon etc., with several adverse effects such as reducing the thermal performance and/or damaging structural materials. The major impurities monitored are sodium oxide Na 2 O and sodium hydride NaH. Hydrogen can diffuse through the steel pipes of the sodium-air heat exchanger or, in a worse case can be generated by a sodium-water reaction, denoting therefore a leak in the tubes of the heat exchanger. Oxygen may origin from the contact with air during maintenance or from the oxide layer of metallic structures initially exposed to sodium during set into operation procedures. The oxygen as an impurity leads to the corrosion of the steel surfaces, therefore values < 2 ppm have to be ensured, while for hydrogen the accepted amount is about 50 ppb (Hemanath et al.). The sodium purification is performed in a cold trap that allows the agglomeration of sodium oxide and sodium hydride on the large surface of a wire mesh. (orig.)

Sodium vapor deposition onto a horizontal flat plate above liquid sodium surface, 2

International Nuclear Information System (INIS)

Kudo, Kazuhiko; Hirata, Masaru.

1977-01-01

The sodium vapor deposition onto a horizontal flat plate above liquid sodium surface was studied. The analysis was performed by assuming that the sodium mist is emitted into the main flow without condensation and then grows up in the main flow and drops on the sodium surface. The effects of growth of sodium mist to the system were investigated. The model of the phenomena is the sodium deposition onto a horizontal flat plate which is placed above the sodium surface with the medium cover gas. One-dimensional analysis can be done. The rate of deposition is greatly reduced when the temperature of the flat plate is lowered. For the analysis of this phenomena, it is assumed that the sodium mist grows by condensation. One of results is that the real state may be the state between the state that the condensation of mist is made in the boundary layer and the state that the mist is condensed in the main flow. Others are that there is no effect of sodium mist condensation on the rate of deposition, and that the rate of the vaporization of sodium is given by the original and the modified model. (Kato, T.)

Study of mechanisms and kinetics of Sodium-CO2 interactions. Contribution to the evaluation of an energy conversion system with supercritical CO2 for sodium fast breeder reactors

International Nuclear Information System (INIS)

Gicquel, L.

2010-01-01

reactions: 4 Na + 3 CO 2 → 2 Na 2 CO 3 + C (7) 4 Na + Na 2 CO 3 → 3 Na 2 O + C (8). This last reaction has been observed in calorimetry. Reaction has also been studied with the development of a model of a reactive CO 2 jet in liquid sodium. This model is based on an hydrodynamics that enables the calculation of speed and flow rates within the jet. It does not take into account sliding phenomenon between liquid and gaseous phases, and propose an homogeneous description of the jet. This model has been validated with sodium-water system during the years 1980-1990. Chemical reactions and associated kinetics, of an Arrhenius type, have been introduced. Pre-exponential factors and activation energies are the parameters to identify. Values taken by temperatures in every point of the jet depend on those parameters, and, according to their values, it is possible to define three areas where the reaction is slow, moderated or rapid. An experimental bench, called DISCO2 (Determination of Sodium-CO 2 interactions), that enable the estimation of kinetics parameters, has been built in the CEA of Cadarache. DISCO2 enables to realize reactive jets of CO 2 in liquid sodium and to record temperatures within the jet, thanks to a comb of thermocouples. Tests carried out in the two above mentioned ranges of temperature have enabled to find again the temperature threshold seen in calorimetric studies. Experimental campaigns have enabled to estimate parameters in both fields of temperature and to strengthen the reactive mechanism. Two series of parameters have been estimated, each one in both fields of temperature. The reactional mechanism introduced into the model was adapted to each area of temperature. The reactions (5) and (7) were initially regarded as dominating in each area. Then mechanism has been improved with the addition of reactions (6) and (8), successive and competitor to reactions (5) and (7). The second option appeared better in both areas. Sodium-CO 2 system is exothermic and

Sodium removal from CSRDM lower part by water vapour - CO2 process

International Nuclear Information System (INIS)

Sundar Raj, S.I.; Sreedhar, B.K.; Gurumoorthy, K.; Rajan, K.K.; Kalyanasundaram, P.; Rajan, M.; Vaidyanathan, G.

2006-01-01

Sodium is the primary and secondary coolant in fast reactors. Primary and secondary circuits components like Control and Safety Rod Drive Mechanism (CSRDM), pumps, heat exchangers etc. handle liquid sodium. Sodium has good affinity to oxygen and reacts vigorously with water. Hence sodium cleaning is the first and important activity in the maintenance of the components. In reactor components this cleaning process also helps in removing a major part of radioactive contaminants after which they are subjected to chemical decontamination. There are several methods available for removing sodium from components. Out of these, the water vapour-CO 2 process is selected for large components such as pumps, heat exchangers etc. while steam cleaning is used for the core sub assemblies. The cleaning processes are to be closely monitored to ensure safety because the release of hydrogen is to be kept below 4 % during the process. This paper discusses the in house facility and the experience in the successful use of the process in the cleaning of CSRDM. (author)

Effect of sodium bicarbonate and varying concentrations of sodium chloride in brine on the liquid retention of fish (Pollachius virens L.) muscle.

Science.gov (United States)

Åsli, Magnus; Ofstad, Ragni; Böcker, Ulrike; Jessen, Flemming; Einen, Olai; Mørkøre, Turid

2016-03-15

Negative health effects associated with excessive sodium (Na) intake have increased the demand for tasty low-Na products (<2% NaCl) rather than traditional heavily salted fish products (∼20% NaCl). This study investigates the causes of improved yield and liquid retention of fish muscle brined with a combination of salt (NaCl) and sodium bicarbonate (NaHCO3 ). Water characteristics and microstructure of saithe (Pollachius virens L.) muscle brined in solutions of NaCl and NaHCO3 or NaCl alone were compared using low-field nuclear magnetic resonance (LF-NMR) T2 relaxometry, microscopy, salt content, liquid retention and colorimetric measurements. Saithe muscle was brined for 92 h in 0, 30, 60, 120 or 240 g kg(-1) NaCl or the respective solutions with added 7.5 g kg(-1) NaHCO3 . NaHCO3 inclusion improved the yield in solutions ranging from 0 to 120 g kg(-1) NaCl, with the most pronounced effect being observed at 30 g kg(-1) NaCl. The changes in yield were reflected in water mobility, with significantly shorter T2 relaxation times in all corresponding brine concentrations. Salt-dependent microstructural changes were revealed by light microscopy, where NaHCO3 supplementation resulted in greater intracellular space at 30 and 60 g kg(-1) NaCl. Sodium bicarbonate addition to low-salt solutions can improve yield and flesh quality of fish muscle owing to altered water mobility and wider space between the muscle cells. © 2015 Society of Chemical Industry.

A Small-Scale Capsule Test for Investigating the Sodium-Carbon Dioxide Reaction

International Nuclear Information System (INIS)

Kim, B. H.; Choi, J. H.; Suk, S. D.; Kim, J. M.; Choi, B. H.; Kim, B. H.; Hahn, D. H.

2007-01-01

The utilization of modular sodium-to-supercritical CO 2 heat exchangers may yield significant improvements for an overall plant energy utilization. The consequences of a failure of the sodium CO 2 heat exchanger boundary, however, would involve the blowdown and intermixing of high-pressure CO 2 in a sodium pool, causing a pressurization which may threaten the structural integrity of the heat exchanger. Available data seems to indicate that the chemical reaction between sodium and CO 2 would likely produce sodium oxides, sodium carbonate, carbon and carbon monoxide. Information on the kinetics of the sodium-CO 2 reaction is virtually non-existent

Characterization of Sodium Spray Aerosols

Energy Technology Data Exchange (ETDEWEB)

Nelson, C. T.; Koontz, R. L.; Silberberg, M. [Atomics International, North American Rockwell Corporation, Canoga Park, CA (United States)

1968-12-15

The consequences of pool and spray fires require evaluation in the safety analysis of liquid metal-cooled fast breeder reactors. Sodium spray fires are characterized by high temperature and pressure, produced during the rapid combustion of sodium in air. Following the initial energy release, some fraction of the reaction products are available as aerosols which follow the normal laws of agglomeration, growth, settling, and plating. An experimental study is underway at Atomics International to study the characteristics of high concentration sprays of liquid sodium in reduced oxygen atmospheres and in air. The experiments are conducted in a 31.5 ft{sup 3} (2 ft diam. by 10 ft high) vessel, certified for a pressure of 100 lb/in{sup 2} (gauge). The spray injection apparatus consists of a heated sodium supply pot and a spray nozzle through which liquid sodium is driven by nitrogen pressure. Spray rate and droplet size can be varied by the injection velocity (nozzle size, nitrogen pressure, and sodium temperature). Aerosols produced in 0, 4, and 10 vol. % oxygen environments have been studied. The concentration and particle size distribution of the material remaining in the air after the spray injection and reaction period are measured. Fallout rates are found to be proportional to the concentration of aerosol which remains airborne following the spray period. (author)

Liquid-liquid extraction of beryllium (II) using Cyanex 923 in kerosene from sodium acetate medium

International Nuclear Information System (INIS)

Kamble, Pravin N.; Madane, Namdev S.; Mohite, Baburao S.

2013-01-01

Liquid-liquid extraction of beryllium(II) from sodium acetate media using cyanex 923 in kerosene has been carried out. Beryllium(II) was quantitatively extracted from 1x10 -1 M sodium acetate with 1x10 -2 M cyanex 923 in kerosene. It was stripped quantitatively from the organic phase with 1M H 2 SO 4 and determined spectrophotometrically with eriochrome cyanine R at 525 nm. The effect of concentrations of sodium acetate, metal ions and strippants have been studied. Separation of beryllium(II) from other elements was achieved from binary as well as from multicomponent mixtures. The method is simple, rapid and selective with good reproducibility (approximately±2%). (author)

Experimental study of the tritium distribution in the effluents resulting from the sodium hydrolysis

Energy Technology Data Exchange (ETDEWEB)

Chassery, A. [CEA, DEN, Centre de Cadarache, Saint-Paul-lez-Durance (France); Universite de Toulouse, Laboratoire de Genie Chimique, Toulouse (France); CNRS, Laboratoire de Genie Chimique, Toulouse (France); Lorcet, H.; Godlewski, J; Liger, K.; Latge, C. [CEA, DEN, Centre de Cadarache, Saint-Paul-lez-Durance (France); Joulia, X. [Universite de Toulouse, Laboratoire de Genie Chimique, Toulouse (France); CNRS, Laboratoire de Genie Chimique, Toulouse (France)

2015-03-15

Within the framework of the dismantling of fast breeder reactors in France several processes are under investigation regarding sodium disposal. One of them, called ELA (radioactive sodium waste treatment process), is based on the implementation of the sodium-water reaction, in a controlled and progressive way, to remove residual sodium. This sodium contains impurities such as sodium hydride, sodium oxide and tritiated sodium hydride. The hydrolysis of these various chemical species leads to the production of a liquid effluent, mainly composed of an aqueous solution of sodium hydroxide, and a gaseous effluent, mainly composed of nitrogen (inert gas), hydrogen and steam. The tritium is distributed between these effluents, and, within the gaseous effluent, according to its forms HT and HTO (tritiated water). HTO being 10,000 times more radio-toxic than HT, a precise knowledge of the mechanisms governing the phase distribution of tritium is necessary. This paper presents the first experimental results from a parametric study on the tritium distribution between the various effluents generated during hydrolysis operations. A series of experiments have been performed in order to study the influence of water flow rate, argon flow rate, initial mass and specific activity of the hydrolyzed sodium sample. An important influence of the total tritium concentration in the hydrolyzed sample has been highlighted. As for the phenomena suspected to be responsible for the phase change of tritiated water, in the studied range of parameters, vaporization induced by the heat of reactions seems to be dominant over the evaporation induced by the inert gas flow rate.

Ab initio molecular dynamics study of the properties of cerium in liquid sodium at 1000 K temperature

Energy Technology Data Exchange (ETDEWEB)

Samin, Adib; Li, Xiang; Zhang, Jinsuo [Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, 201 W 19th Avenue, Columbus, Ohio 43210 (United States); Mariani, R. D. [Idaho National Laboratory, Materials and Fuels Complex, Idaho Falls, Idaho 83415 (United States); Unal, Cetin [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States)

2015-12-21

For liquid-sodium-cooled fast nuclear reactor systems, it is crucial to understand the behavior of lanthanides and other potential fission products in liquid sodium or other liquid metal solutions such as liquid cesium-sodium. In this study, we focus on lanthanide behavior in liquid sodium. Using ab initio molecular dynamics, we found that the solubility of cerium in liquid sodium at 1000 K was less than 0.78 at. %, and the diffusion coefficient of cerium in liquid sodium was calculated to be 5.57 × 10{sup −9} m{sup 2}/s. Furthermore, it was found that cerium in small amounts may significantly alter the heat capacity of the liquid sodium system. Our results are consistent with the experimental results for similar materials under similar conditions.

Contribution to the study of hydrogenated and oxygenated impurities in liquid sodium; Contribution a l'etude des impuretees hydrogenees et oxygenees dans le sodium liquide

Energy Technology Data Exchange (ETDEWEB)

Naud, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-15

This study is made up essentially of two parts. The first is devoted to the development of dosage methods for selectively determining the oxygenated and hydrogenated impurities present in sodium, that is the oxide, the hydride and the hydroxide. The second makes use of these methods for a study of the Na-H{sub 2}-O{sub 2} system, as well as of the related problem of the attack of pyrex glass by molten sodium. The conventional method for dosing oxygen by amalgamation was first adapted to the simultaneous measurement of the hydride. We then developed a method for dosing the total hydrogen by measuring successively the concentrations of gas present in the hydride and hydroxide form. This method is based on the thermal decomposition of the hydride and the reaction between sodium and the hydroxide. Our contribution to the study of the Na-H{sub 2}-O{sub 2} System consists first of all in the study of the reaction between hydrogen and sodium in the temperature range from 150 to 250 deg C and at a pressure of about 150 mm of mercury. The study of the thermal decomposition of the hydride in sodium was then studied. It was possible to make some qualitative observations concerning the reaction between sodium and sodium hydroxide. Finally some complementary tests made it possible to define the nature of the phenomena occurring during the attack of pyrex glass by sodium. (author) [French] Cette etude comporte essentiellement deux parties. La premiere est consacree a la mise au point de methodes de dosages permettant de determiner selectivement les impuretes oxygenees et hydrogenees presentes dans le sodium, a savoir l'oxyde, l'hydrure et l'hydroxyde. La seconde met a profit ces methodes en vue de l'etude du systeme Na-H{sub 2}-O{sub 2}. ainsi que du phenomene connexe d'attaque du verre pyrex par le sodium fondu. La methode classique de dosage de l'oxygene par amalgamation a d'abord ete adaptee au dosage simultane de l'hydrure. Nous avons ensuite mis au point le dosage de l

Sodium-concrete reactions experiments and code development

International Nuclear Information System (INIS)

Casselman, C.; Malet, J.C.; Dufresne, J.; Bolvin, M.

1988-01-01

Hypothesis of hot sodium leak in a fast breeder reactor implies, for the safety organism to consider spillage of sodium on concrete. This safety analysis involves the understanding of sodium-concrete reactions, the knowledge of their consequences and to test the choiced preventive solutions. In association with EDF, the nuclear safety department had carried out an extensive experimental program, the different parts of which are connected with each aspect of this problem: - firstly, interaction between sodium and bare surface of usual concrete; - secondly, the case of a sodium spillage on a concrete surface covered with a defected liner; - thirdly, special concrete tests for a comparison with usual concrete behavior, in direct contact with hot sodium; - at last, a test which concerns a new design with a layer of the selected concrete protected with a defected liner. On the same time, theoretical work leads to elaborate a physical model to describe temporal evolution of thermal and chemical decomposition of a concrete slab under hot sodium action. SORBET-REBUS system will use quoted above test results to its validation

A Small-Scale Capsule Test for Investigating the Sodium-Carbon Dioxide Reaction

Energy Technology Data Exchange (ETDEWEB)

Kim, B. H.; Choi, J. H.; Suk, S. D.; Kim, J. M.; Choi, B. H.; Kim, B. H.; Hahn, D. H

2007-01-15

The utilization of modular sodium-to-supercritical CO{sub 2} heat exchangers may yield significant improvements for an overall plant energy utilization. The consequences of a failure of the sodium CO{sub 2} heat exchanger boundary, however, would involve the blowdown and intermixing of high-pressure CO{sub 2} in a sodium pool, causing a pressurization which may threaten the structural integrity of the heat exchanger. Available data seems to indicate that the chemical reaction between sodium and CO{sub 2} would likely produce sodium oxides, sodium carbonate, carbon and carbon monoxide. Information on the kinetics of the sodium-CO{sub 2} reaction is virtually non-existent.

Comparative Effects of Water, Acid and Sodium Benzoate as ...

African Journals Online (AJOL)

The relative effects of water, sulphuric acid and sodium benzoate as additives on the micelle-catalyzed aquation reactions of the complexes:Fe(Me2Phen)3 2+ and FE(Me4Phen) were studied in acetone using Triton X-100 (TX-100), as the surfactant-catalyst. FE(Me4Phen)2+ equates faster than FE(Me2Phen)2+ in the ...

Solubility of xenon in liquid sodium

International Nuclear Information System (INIS)

Veleckis, E.; Cafasso, F.A.; Feder, H.M.

1976-01-01

The solubility of xenon in liquid sodium was measured as a function of pressure (2-8 atm) and temperature (350-600 0 C). Henry's law was obeyed with the value of the Henry's law constant, K/sub H/ = N/sub Xe//P, ranging from 1.38 x 10 -10 atm -1 at 350C, to 1.59 x 10 -8 atm -1 at 600 0 C where N/sub Xe/ and P are the atom fraction and the partial pressure of xenon, respectively. The temperature dependence of solubility may be represented by log 10 lambda = (0.663 +- 0.01) - (4500 +- 73) T -1 , where lambda is the Ostwald coefficient (the volume of xenon dissolved per unit volume of sodium at the temperature of the experiment). The heat of solution of xenon in sodium was 20.6 +- 0.7 kcal/mole, where the standard state of xenon is defined as that of 1 mole of an ideal gas, confined to a volume equal to the molar volume of sodium

Liquid-liquid extraction of uranium(VI) using Cyanex 272 in toluene from sodium salicylate medium

International Nuclear Information System (INIS)

Madane, Namdev S.; Nikam, Gurunath H.; Jadhav, Deepali V.; Mohite, Baburao S.

2011-01-01

Liquid-liquid extraction of U(VI) from sodium salicylate media using Cyanex 272 in toluene has been carried out. Uranium(VI) was quantitatively extracted from 1 x 10 -3 M sodium salicylate with 5 x 10 -4 M Cyanex 272 in toluene. It was stripped quantitatively from the organic phase with 1M HCl and determined spectrophotometrically with arsenazo(III) at 660 nm. The effect of concentrations of sodium salicylate, extractant, diluents, metal ion and strippants have been studied. Separation of uranium(VI) from other elements was achieved from binary as well as from multicomponent mixtures. The method was extended to determination of uranium(VI) in geological samples. The method is simple, rapid and selective with good reproducibility (approximately ± 2%). (author)

Concentrated Aqueous Sodium Tosylate as Green Medium for Alkene Oxidation and Nucleophilic Substitution Reactions.

Science.gov (United States)

Sela, Tal; Lin, Xiaoxi; Vigalok, Arkadi

2017-11-03

A hydrotropic solution of highly concentrated sodium tosylate (NaOTs) can be used as a recyclable medium for the environmentally benign oxidation of conjugated alkenes with H 2 O 2 . Both uncatalyzed and metal-catalyzed reactions provided the corresponding oxidation products in higher yields than in pure water or many common organic solvents.

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)

Peroxide formation and kinetics of sodium dissolution in alcohols

International Nuclear Information System (INIS)

Muralidaran, P.; Chandran, K.; Ganesan, V.; Periaswami, G.

1997-01-01

Suitable techniques for sodium removal and decontamination of sodium wetted components of Liquid Metal Fast Reactors (LMFRs) are necessary both for repair, reuse and decommissioning of such components. Among the methods followed for sodium removal, alcohol dissolution is usually employed for small components like bellow sealed valves, gripping tools to handle core components and sodium sampling devices (primary and secondary). One of the concerns in the alcohol dissolution method is the possible role of peroxide formation in the ethoxy group during storage and handling leading to explosion. This paper describes the study of peroxide formation in ethyl carbitol and butyl cellosolve as well as some of the results of dissolution kinetic studies carried out in our laboratory using different alcohols. The peroxide formation of ethyl carbitol and butyl cellosolve were studied by iodometric technique. It has been found that the peroxide formation is less in sodium containing alcohol than in pure one. Ethyl carbitol, butyl cellosolve and Jaysol-SS (mixture of ethyl alcohol, methyl alcohol, isopropyl alcohol and methyl isobutyl ketone) were used in dissolution kinetics studies. The effects due to area and orientation of the fresh sodium surface have also been investigated. The reaction rates were studied in the temperature range of 303-343 K. The rate of dissolution was estimated by measuring the sodium content of alcohol at periodic intervals. It is found that the reaction rate varies in the order of ethyl alcohol-water mixture > Jaysol-SS > butyl cellosolve > ethyl carbitol. While cleaning sodium using alcohol, the concentration of alcohol is held essentially constant throughout the process. The rate of reaction depends only on the amount of sodium and follows pseudo-first order kinetics. Increase in surface area has a marked impact on the dissolution rate at lower temperatures while at higher temperatures, the temperature factor overrides the effect due to surface area

Passive vibro-acoustic detection of a sodium-water reaction in a steam generator of a sodium-cooled fast neutrons nuclear reactor by beam forming

International Nuclear Information System (INIS)

Moriot, Jeremy

2013-01-01

This thesis deals with a new method to detect a sodium-water reaction in a steam generator of a fast sodium-cooled nuclear reactor. More precisely, the objective is to detect a micro-leak of water (flow ≤ 1 g/s) in less than 10 seconds by measuring the external shell vibrations of the component. The strong background noise in operation makes impossible the use of a detection system based on a threshold overrun. A beam forming method applied to vibrations measured by a linear array of accelerometers is developed in this thesis to increase the signal-to-noise ratio and to detect and locate the leak in the steam generator. A numerical study is first realized. Two models are developed in order to simulate the signals measured by the accelerometers of the array. The performances of the beam forming are then studied in function of several parameters, such as the source location and frequency, the damping factor, the background noise considered. The first model consists in an infinite plate in contact with a heavy fluid, excited by an acoustic monopole located in this fluid. Analyzing the transverse displacements in the wavenumber domain is useful to establish a criterion to sample correctly the vibration field of the plate. A second model, more representative of the system is also proposed. In this model, an elastic infinite cylindrical shell, filled with a heavy fluid is considered. The finite dimensions in the radial and circumferential directions lead to a modal behavior of the system which impacts the beam forming. Finally, the method is tested on an experimental mock-up which consists in a cylindrical pipe made in stainless steel and filled with water connected to hydraulic circuit. The water flow speed can be controlled by varying the speed of the pump. The acoustic source is generated by a hydro-phone. The performances of the beam forming are studied for different water flow speeds and different amplitude and frequencies of the source. (author) [fr

Sodium-water reaction data needed by a utility for design assessment purposes

Energy Technology Data Exchange (ETDEWEB)

Bolt, P R [Fast Reactor Engineering, Barnwood, Gloucester (United Kingdom)

1978-10-01

Worldwide LMFBR experience has shown that LMFBR steam generator water/steam leakage into sodium can severely reduce steam generator availability and cause lengthy plant outages. Utility assessment LMFBR designs prior to placing an order to construct are likely to give emphasis to matters that can affect steam generator integrity, reliability and total operating cost. The data needed in carrying out such assessments is described. (author)

Sodium-water reaction data needed by a utility for design assessment purposes

International Nuclear Information System (INIS)

Bolt, P.R.

1978-01-01

Worldwide LMFBR experience has shown that LMFBR steam generator water/steam leakage into sodium can severely reduce steam generator availability and cause lengthy plant outages. Utility assessment LMFBR designs prior to placing an order to construct are likely to give emphasis to matters that can affect steam generator integrity, reliability and total operating cost. The data needed in carrying out such assessments is described. (author)

PG BN 1600 sodium fire protection system

International Nuclear Information System (INIS)

Bar, J.; Urbancik, L.

1978-12-01

A design was developed of a fire protection system for steam generator of a 1600 MW sodium cooled fast reactor (BN-1600). Chemical reactions are described of liquid sodium with atmospheric components and solid materials coming into contact with sodium in its release from the steam generator, and in safeguarding protection against sodium fires. The requirements for the purity of nitrogen as an atmosphere inert to liquid sodium are given. Characteristics and basic parameters are shown of level and spray fires, elementary terms are explained concerning the properties of aerosols formed during fires, the methods and means of release signalling and fire alarm are described as are fire precautions using fire-fighting equipment, modifying the support tank and the cell bottom and building sewage pits. The design of the system comprises an alarm system for liquid sodium using point and line electric contact sensors and flame photometer based aerosol sensors as well as a fire-fighting system based on the system of channelling liquid sodium into emergency discharge tanks filled with an inert gas, a set of fire extinguishers and other fire fighting material, and measures for the elimination of sodium fire consequences. (J.B.)

Experimental design for reflection measurements of highly reactive liquid or solid substances with application to liquid sodium

International Nuclear Information System (INIS)

Chan, S.H.; Gossler, A.A.

1980-01-01

A versatile goniometer system with associated electronic components and mechanical instruments has been assembled. It is designed to measure spectral, specular reflectances of highly reactive liquid or solid substances over a spectral range of 0.3 to 9 μ and incidence angles of 12 to 30 0 off the normal direction. The capability of measuring reflectances of liquid substances clearly distinguishes this experimental design from conventional systems which are applicable only to solid substances. This design has been used to measure the spectral, specular reflectance of liquid sodium and preliminary results obtained are compared with those of solid sodium measured by other investigators

Rate theory of ion pairing at the water liquid-vapor interface: A case of sodium iodide

Science.gov (United States)

Dang, Liem X.; Schenter, Gregory K.

2018-06-01

Studies on ion pairing at interfaces have been intensified recently because of their importance in many chemical reactive phenomena, such as ion-ion interactions that are affected by interfaces and their influence on kinetic processes. In this study, we performed simulations to examine the thermodynamics and kinetics of small polarizable sodium iodide ions in the bulk and near the water liquid-vapor interface. Using classical transition state theory, we calculated the dissociation rates and corrected them with transmission coefficients obtained from the reactive flux formalism and Grote-Hynes theory. Our results show that in addition to affecting the free energy of ions in solution, the interfacial environments significantly influence the kinetics of ion pairing. The results on the relaxation time obtained using the reactive flux formalism and Grote-Hynes theory present an unequivocal picture that the interface suppresses ion dissociation. The effects of the use of molecular models on the ion interactions as well as the ion-pair configurations at the interface are also quantified and discussed.

Reactions between sodium and various carbon bearing compounds

Energy Technology Data Exchange (ETDEWEB)

Raine, A C; Thorley, A W [UKAEA, Risley, Warrington, Cheshire (United Kingdom)

1980-05-01

The presence of carbon bearing materials in liquid sodium is undesirable because of their ability to carburise stainless steel components. It has been demonstrated for example that carbon taken up by stainless steels can affect their mechanical properties and that thinner sectioned material such as fuel cladding and the tubing of intermediate heat exchanger may be more sensitive to such effects. Generally speaking, there are a number of potential carbon sources in reactor systems. Some of the sources such as the graphite in neutron shield rods, boron carbide in control rods and carbide fuels are part of the reactor designs while others such as oil in mechanical pumps arid 'coupling-fluids' used to inspect plant components are associated with the respective operation arid inspection of the plant. In this paper it is intended to discuss in general terms the way these various compounds behave in liquid sodium and to assess what effect their presence will have on the materials of construction in fast reactor systems. The paper also reviews the chemistry of the environment in relation to the types of carburizing species which may exist in sodium systems.

Phase equilibrium in systems with ionic liquids: An example for the downstream process of the Biphasic Acid Scavenging utilizing Ionic Liquids (BASIL) process. Part I: Experimental data

International Nuclear Information System (INIS)

Sahandzhieva, Katya; Maurer, Gerd

2012-01-01

Highlights: ► Phase equilibrium for a downstream process in sustainable chemical technology. ► Biphasic Acid Scavenging Utilizing Ionic Liquids (BASIL) Process. ► SLE, LLE, and SLLE of (NaCl + water + 1-propanol + 1-MIM) and its ternary subsystems. ► Experimental phase equilibrium data at temperatures between 298 K and 333 K. - Abstract: Experimental results are presented for the (liquid + liquid), (solid + liquid) and (solid + liquid + liquid) equilibria occurring in the downstream process of a typical example for the Biphasic Acid Scavenging Utilizing Ionic Liquids (BASIL)-processes. In a BASIL process an organic base is used to catalyze a chemical reaction and, at the same time, to scavenge an acid that is an undesired side product of that reaction. The particular example of a BASIL process treated here is the reaction of 1-butanol and acetylchloride to butylacetate and hydrochloric acid, where the acid is scavenged by the organic base 1-methyl imidazole (1-MIM) resulting in the ionic liquid 1-methyl imidazolium chloride. The reaction results in a two-phase system as butylacetate and the ionic liquid reveal a large liquid–liquid miscibility gap. The organic base has to be recovered. This is commonly achieved by treating the ionic liquid–rich liquid phase with an aqueous solution of sodium hydroxide (i.e., converting the ionic liquid to the organic base) and extracting the organic base by an appropriate organic solvent (e.g., 1-propanol). The work presented here deals in experimental work with the (liquid + liquid), (solid + liquid) and (solid + liquid + liquid) phase equilibria that are encountered in such extraction processes. Experimental results are reported for temperatures between about 298 K and 333 K: for the solubility of NaCl in several solvents (1-propanol, 1-MIM), (water + 1-MIM), (1-propanol + 1-MIM), (water + 1-propanol), and (water + 1-propanol + 1-MIM) and for the (liquid + liquid) equilibrium as well as for the (solid + liquid

Modeling of Hydrodynamic Processes at a Large Leak of Water into Sodium in the Fast Reactor Coolant Circuit

Directory of Open Access Journals (Sweden)

Sergey Perevoznikov

2016-10-01

Full Text Available In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium–water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional thermally nonequilibrium three-component gas–liquid flow model (sodium–hydrogen–sodium hydroxide. Consideration is given to the results of a mathematical modeling of experiments involving steam injection into the sodium loop of a circulation test facility. That was done by means of the computer code in which the proposed model had been implemented.

Method of conversion of caustic liquid radioactive wastes containing sodium nitrates into solid insoluble products

International Nuclear Information System (INIS)

Barney, G.S.; Brownell, L.E.

1975-01-01

A proposal is made to convert caustic, liquid, radioactive wastes containing sodium nitrate into a solid product by reaction with powdered aluminium silicate at temperatures between 30 0 and 100 0 C, which is practically insoluble (10 -7 to 10 -10 g/cm 2 -day) and is thermally stable. A cancrinite is formed which binds the radioactive salts in the cage-like structure of its crystal lattice. The method is also suitable for liquid wastes from the Purex method as well as for wastes containing fission products of Cs 137 and Sr 90 in concentrations of 0.37 M to 0.01 M. Numerous detailed examples explain the invention. (UW/LH) [de

(Liquid + liquid) equilibria for benzene + cyclohexane + N,N-dimethylformamide + sodium thiocyanate

International Nuclear Information System (INIS)

Dong, Hongxing; Yang, Xiaoguang; Yue, Guojun; Zhang, Wei; Zhang, Jin

2013-01-01

Graphical abstract: On the left, the figure was phase diagram about the LLE date. On the right, the figure was about the effects of mass fraction of benzene in the raffinate phase to the selectivity(S) coefficient under different salt concentration. ■, the NaSCN and DMF in ratio of 5/95; • , the NaSCN and DMF in ratio of 10/90; ▴, the NaSCN and DMF in ratio of 15/85; ★, the NaSCN and DMF in ratio of 20/80; ▾, the NaSCN and DMF in ratio of 23/77. ♦, only DMF was used extractant (the selectivity coefficient was calculated by literature 17). w 22 , refer to the mass fraction of benzene in the raffinate phase (cyclohexane-rich phase). Highlights: • (Liquid + liquid) equilibrium for quaternary system was measured. • The components include benzene, cyclohexane, N,N-dimethylformamide, sodium thiocyanate. • The (liquid + liquid) equilibrium data can be well correlated by the NRTL model. • Separation of benzene and cyclohexane by NaSCN + DMF was discussed. -- Abstract: (Liquid + liquid) equilibrium (LLE) data for benzene + cyclohexane + N,N-dimethylformamide (DMF) + sodium thiocyanate (NaSCN) were measured experimentally at atmospheric pressure and 303.15 K. The selectivity coefficients from these LLE data were calculated and compared to those previously reported in the literature for the systems (benzene + cyclohexane + DMF) and (benzene + cyclohexane + DMF + KSCN). The NRTL equation was used to correlate the experimental data. The agreement between the predicted and experimental results was good. It was found that the selectivity coefficients of DMF + NaSCN for benzene ranged from 2.45 to 11.99. Considering the relatively high extraction capacity and selectivity for benzene, DMF + NaSCN may be used as a potential extracting solvent for the separation of benzene from cyclohexane

Experimental evaluation of permanent magnet probe flowmeter measuring high temperature liquid sodium flow in the ITSL

Energy Technology Data Exchange (ETDEWEB)

Jeong, Uiju; Kim, Yun Ho [Nuclear engineering Department, Hanyang University, 17 Haengdang-dong, Seongdong-gu, Seoul 133-791 (Korea, Republic of); Kim, Jong-Man; Kim, Tae-Joon [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Sung Joong, E-mail: sungjkim@mit.edu [Nuclear engineering Department, Hanyang University, 17 Haengdang-dong, Seongdong-gu, Seoul 133-791 (Korea, Republic of)

2013-12-15

Highlights: • An Instrument Test Sodium Loop (ITSL) has been built and tested in various conditions at KAERI. • Free fall of liquid sodium was conducted experimentally and numerically. • A Permanent Magnet Probe Flowmeter (PMPF) was experimented in the ITSL. • Excellent linearity of the PMPF was achieved under high temperature condition. - Abstract: The Instrument Test Sodium Loop (ITSL) installed at Korea Atomic Energy Research Institute (KAERI) is a medium-size experimental facility dedicated to obtaining relevant experimental data of liquid sodium flow characteristics under various thermal hydraulic conditions and sodium purification. The ITSL has been utilized to perform thermal flow measurement of the liquid sodium and to calibrate a Permanent Magnet Probe Flowmeter (PMPF). The primary objective of this study is to obtain liquid sodium flow rate given a wide temperature range using the PMPF. Non-stationary method was adopted for the calibration of the probe given the liquid sodium temperature range of 150–415 °C. A relationship between the measured voltage signal and flow rate was obtained successfully. It is observed that the calibration experiments result in excellent linear relationships between measured voltage and volumetric flow rate at various temperature conditions. Also a computational analysis using FlowMaster, is employed to facilitate the calibration process by predicting the liquid sodium flow rate. Finally the effect of the fluid temperature on thermal flow measurements is discussed in light of the obtained experimental data.

Advancement of compressible multiphase flows and sodium-water reaction analysis program SERAPHIM. Validation of a numerical method for the simulation of highly underexpanded jets

International Nuclear Information System (INIS)

Uchibori, Akihiro; Ohshima, Hiroyuki; Watanabe, Akira

2010-01-01

SERAPHIM is a computer program for the simulation of the compressible multiphase flow involving the sodium-water chemical reaction under a tube failure accident in a steam generator of sodium cooled fast reactors. In this study, the numerical analysis of the highly underexpanded air jets into the air or into the water was performed as a part of validation of the SERAPHIM program. The multi-fluid model, the second-order TVD scheme and the HSMAC method considering a compressibility were used in this analysis. Combining these numerical methods makes it possible to calculate the multiphase flow including supersonic gaseous jets. In the case of the air jet into the air, the calculated pressure, the shape of the jet and the location of a Mach disk agreed with the existing experimental results. The effect of the difference scheme and the mesh resolution on the prediction accuracy was clarified through these analyses. The behavior of the air jet into the water was also reproduced successfully by the proposed numerical method. (author)

Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

International Nuclear Information System (INIS)

Geraldo, Issa Cherif; Bose, Tanmoy; Pekpe, Komi Midzodzi; Cassar, Jean-Philippe; Mohanty, A.R.; Paumel, Kévin

2014-01-01

Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

Behavior of supercooled aqueous solutions stemming from hidden liquid-liquid transition in water.

Science.gov (United States)

Biddle, John W; Holten, Vincent; Anisimov, Mikhail A

2014-08-21

A popular hypothesis that explains the anomalies of supercooled water is the existence of a metastable liquid-liquid transition hidden below the line of homogeneous nucleation. If this transition exists and if it is terminated by a critical point, the addition of a solute should generate a line of liquid-liquid critical points emanating from the critical point of pure metastable water. We have analyzed thermodynamic consequences of this scenario. In particular, we consider the behavior of two systems, H2O-NaCl and H2O-glycerol. We find the behavior of the heat capacity in supercooled aqueous solutions of NaCl, as reported by Archer and Carter [J. Phys. Chem. B 104, 8563 (2000)], to be consistent with the presence of the metastable liquid-liquid transition. We elucidate the non-conserved nature of the order parameter (extent of "reaction" between two alternative structures of water) and the consequences of its coupling with conserved properties (density and concentration). We also show how the shape of the critical line in a solution controls the difference in concentration of the coexisting liquid phases.

Heat transfer characteristics around a single heated rod immersed in sodium pool with gas jet injection

International Nuclear Information System (INIS)

Hideto Niikura; Kazuo Soga; Ken-ichiro Sugiyama; Akira Yamaguchi

2005-01-01

In a steam generator using liquid sodium, water intensely reacts with sodium when it leaks out from a heat transfer tube. It is important to evaluate the influence of sodium-water reaction to surrounding tubes and the shell. Hence, it has been desired to develop the simulation code for the evaluation of sodium-water reaction. From this viewpoint, the Japan Nuclear Cycle is now developing the SERAPHIM code. We reported a preliminary study to establish an experimental method for a single heated rod immersed in sodium pool with steam jet impingement planned in the near future as well as to obtain a preliminary data to verify the adequacy of SERAPHIM code. We first measured local and mean heat transfer coefficients around a horizontal single heated rod immersed in a water pool and a sodium pool with a limited volume in the experimental apparatus. It was confirmed that the mean heat transfer coefficients fairly agreed with the existing data for natural convection in water and sodium. Secondary we measured local and mean heat transfer coefficients around a horizontal single heated rod with Ar gas jet impingement immersed in the limited water pool and in the limited sodium pool. It was clearly observed that the local heat transfer coefficients in the sodium pool keep almost the same values in every angle regardless of increase in Ar gas jet velocity varied from about 8.7m/s to about 78m/s. On the other hand, it was confirmed in the water pool that local heat transfer coefficients on the forward stagnation side exposed in the Ar gas jet impingement increase with increasing the jet velocity while the local heat transfer coefficients on the opposite surface keep almost same values regardless of increase in the velocity. (authors)

Fatigue-crack propagation response of two nickel-base alloys in a liquid sodium environment

International Nuclear Information System (INIS)

Mills, W.J.; James, L.A.

1979-01-01

The elevated temperature fatigue-crack propagation response of Inconel 600 and Inconel 718 was characterized within a linear-elastic fracture mechanics framework in air and low-oxygen liquid sodium environments. The crack growth rates of both nickel-base alloys tested in liquid sodium were found to be considerably lower than those obtained in air. This enhanced fatigue resistance in sodium was attributed to the very low oxygen content in the inert sodium environment. Electron fractographic examination of the Inconel 600 and Inconel 718 fatigue fracture surfaces revealed that operative crack growth mechanisms were dependent on the prevailing stress intensity level. Under low growth rate conditions, Inconel 600 and Inconel 718 fracture surfaces exhibited a faceted, crystallographic morphology in both air and sodium environments. In the higher growth rate regime, fatigue striations were observed; however, striations formed in sodium were rather ill-defined. These indistinct striations were attributed to the absence of oxygen in the liquid sodium environment. Striation spacing measurements were found to be in excellent agreement with macroscopic growth rates in both environments

Development of an electrical connector for liquid sodium environment. Final Report

International Nuclear Information System (INIS)

Kataoka, Hajime; Noguchi, Koichi; Takatsudo, Hiroshi; Miyakawa, Shun-ichi

1998-07-01

The INstrumented irradiation Test Assembly (INTA) has been used to conduct precision on-line instrumented irradiation tests in the experimental fast reactor JOYO. In INTA, direct instrumentation wiring between the irradiation test section in the core and the upper structure section in the rotating plug makes INTA structurally complex and expensive. Instead of direct wiring, if an electrical connector capable of withstanding a heated liquid sodium environment could be used between the irradiation test section and the upper structure section, the upper mechanism of INTA could be reused and testing costs would be drastically reduced. Moreover, the reactor load factor would be improved because of reduced handling time for INTA. In an attempt to gain this advantage, research and development of an electric connector in a sodium environment was carried out from 1988 to 1996 at PNC. As no previous R and D had been conducted in this area, this development activity was conducted in a boot strap manner. The first test was carried out for a small model fabrication, the second was for a water partial model, and the third was for a sodium partial model. Based on those tests, a prototype design specification of the connector was determined. In the sodium partial model test, the resilience of the electrical connector insulation to the sodium environment was investigated. However, severe cracking in the ceramic insulator caused by the high temperature sodium environment was discovered at the junction of ceramic insulator and metallic electrode. Although additional sodium partial tests were performed for various material combinations of ceramic insulators, metallic electrodes, brazing materials and metallization materials, the results of the tests were unsatisfactory. Therefore, it was decided that the development of the connector in sodium should cease at PNC in 1997. (J.P.N.)

Measurements of the viscosity of sodium tetraborate (borax)-UO2 and of sodium metaborate-UO2 liquid solutions

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Roth, A.

1983-01-01

Adding UO 2 produces an increase of viscosity of borax and sodium metaborate. For temperatures below 920 0 C the measurements with the borax-UO 2 solution show a phase separation. Contrary to borax the sodium metaborate solutions indicate a well defined melting point. At temperatures slightly below the melting point a solid phase is formed. The tested sodium-borates-UO 2 mixtures are in liquid form. (DG)

Niobium corrosion in flowing liquid sodium between 400 and 600 degrees C

International Nuclear Information System (INIS)

Sannier, J.; Champeix, L.; Darras, R.; Graff, W.

1966-10-01

The corrosion of niobium and two of its alloys was studied under temperature, flow rate, and purity conditions of liquid sodium similar to those likely to occur in a fast neutron reactor. The results are discussed with reference to the following parameters: purification method used for the sodium, temperature, metallurgical condition of the structural metal. Generally speaking, an important role is played by the oxygen content of the liquid metal towards the corrosion of the niobium: although the metal behaves very satisfactorily when a hot trap purification is used, it undergoes corrosion in the presence of sodium which has been purified only by a cold trap. (authors) [fr

Compatibility of sodium with ceramic oxides employed in nuclear reactors

International Nuclear Information System (INIS)

Acena, V.

1981-01-01

A review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors, is presented. The results of a thermo-dynamic/liquid sodium reactions are included. The exercise has been conducted with a view to effecting experimental studies in the future. (author) [es

Compatibility of sodium with ceramic oxides employed in nuclear reactors

International Nuclear Information System (INIS)

Acena Moreno, V.

1981-01-01

This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

Correlation between hydrogen release and degradation of limestone concrete exposed to hot liquid sodium in inert atmosphere

International Nuclear Information System (INIS)

Parida, F.C.; Das, S.K.; Sharma, A.K.; Ramesh, S.S.; Somayajulu, P.A.; Kannan, S.E.

2005-01-01

Full text of publication follows: Concrete is used as a structural material in a Fast Breeder Reactor (FBR) plant for the construction of its foundation, containment, radiation shield and equipment support structures. An accidental leakage of hot sodium on these civil structures can bring about thermo-chemical reactions, with concrete producing hydrogen gas and causing structural degradation. The concrete damage and hydrogen generation take place concurrently due to conduction of heat from sodium into the concrete and migration of steam / moisture in counter current direction towards sodium. In a series of experiments conducted with limestone concrete for two different types of design corresponding to composition and geometry, were exposed to liquid sodium (∼2 kg) at initial temperatures varying from 180 deg. C to 500 deg. C in an inerted test vessel (Capacity = 203 L). Immersion heater was employed to heat the sodium pool on the concrete cavity during the test period in some test runs. On-line continuous measurement of pressure, temperature, hydrogen gas and oxygen gas was carried out. Pre- and post- test nondestructive testing such as colour photography, spatial profiling of ultrasonic pulse velocity and measurement of dimensions were also conducted. Solid samples were collected from sodium debris by manual core drilling machine and from concrete block by hand held electric drilling machine. These samples were subjected to chemical analysis for the determination of free and bound water along with unburnt and burnt sodium. The hydrogen generation parameters such as average and peak release rate as well as release efficiency are derived from measured test variables. These test variables include temperature, pressure and hydrogen concentration in the argon atmosphere contained in the test vessel. The concrete degradation parameters encompass percentage reduction in ultrasonic pulse velocity, depth of physical and chemical dehydration and sodium penetration. These

Water and sodium balance in space

DEFF Research Database (Denmark)

Drummer, C; Norsk, P; Heer, M

2001-01-01

, cumulative water balance and total body water content are stable during flight if hydration, nutritional energy supply, and protection of muscle mass are at an acceptable level. Recent water balance data disclose that the phenomenon of an absolute water loss during space flight, which has often been reported...... and an exaggerated extravasation very early in-flight. The mechanisms for the increased vascular permeability are not known. Evaporation, oral hydration, and urinary fluid excretion, the major components of water balance, are generally diminished during space flight compared with conditions on Earth. Nevertheless...... in the past, is not a consequence of the variable microG. The handling of sodium, however, is considerably affected by microG. Sodium-retaining endocrine systems, such as renin-aldosterone and catecholamines, are much more activated during microG than on Earth. Despite a comparable oral sodium supply, urinary...

THERMOCHEMISTRY OF INTERACTION REACTIONS FOR SODIUM AND ALUMINUM SULPHATES WITH COMPONENTS OF HYDRATING PORTLAND CEMENT

Directory of Open Access Journals (Sweden)

P. I. Yukhnevskiy

2018-01-01

Full Text Available Chemical additives are widely used in the technology of concrete with the purpose to solve various problems and sulphate-containing additives-electrolytes are also used as accelerators for setting and hardening of cement. Action mechanism of additive accelerators for setting and hardening of cement is rather complicated and can not be considered as well-established. An influence of sulfate-containing additives such as sodium sulfate is reduced to acceleration of cement silicate phase hydration by increasing ionic strength of the solution. In addition to it, exchange reactions of anion additive with portlandite phase (Ca(OH2 and aluminate phases of hardening cement have a significant effect on hardening process that lead to formation of readily soluble hydroxides and hardly soluble calcium salts. The influence of sulfate-containing additives on properties of water cement paste and cement stone is quite diverse and depends on salt concentration and cation type. For example, the action of the aluminum sulphate additive becomes more complicated if the additive is subjected to hydrolysis in water, which is aggravated in an alkaline medium of the water cement paste. Formation of hydrolysis products and their reaction with aluminate phases and cement portlandite lead to a significant acceleration of setting. Thus, despite the similarity of additives ensuring participation of anions in the exchange reactions, the mechanism of their influence on cement setting and hardening varies rather significantly. The present paper considers peculiar features concerning the mechanism of interaction of sodium and aluminum sulfate additives in cement compositions from the viewpoint of thermochemistry. Thermochemical equations for reactions of sulfate-containing additives with phases of hydrated cement clinker have been given in the paper. The paper contains description how to calculate thermal effects of chemical reactions and determine an influence of the formed

Operation of the water-to-sodium leak detection system at the experimental breeder reactor II

International Nuclear Information System (INIS)

Osterhout, M.M.

1978-01-01

A water-to-sodium leak detection system was installed at the Experimental Breeder Reactor II in April 1975. The system is designed for early detection of steam generator leaks, using hydrogen meters at the sodium outlets of the evaporators and superheaters. The leak detectors operate by measuring the rate of diffusion of hydrogen from the liquid sodium through a nickel membrane into a dynamic vacuum system. The advantages of this detection system are rapid response time, high sensitivity, stability, and reliability. The system was operated on an experimental basis for the first two years. During this period, data were obtained on detector stability, reliability, maintenance needs, computer interface requirements, calibration, and background hydrogen-level fluctuations. A generic defect in the original detectors was also discovered, requiring redesign of the units. When the new units were installed and proven to be reliable, the system was made fully operational. The data from the hydrogen meters are now used as the primary basis for detection of water-to-sodium leaks

Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

International Nuclear Information System (INIS)

Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

2012-09-01

selection of chemistry controls is vital for NPPs with liquid metal cooled reactors. This paper highlights principles and approaches to chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors. The recommendations on how to arrange chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors are based taking into account: - the experience with operation of fossil power industry; - secondary side water chemistry of lead-bismuth eutectics cooled nuclear reactors at submarines; - steam/water cycles of NPPs with sodium cooled fast breeders BN-350 and BN-600; - secondary water chemistry at conventional NPPs with WER, RBMK and some other reactors. (authors)

Reaction between barium and nitrogen in liquid sodium

International Nuclear Information System (INIS)

Addison, C.C.; Pulham, R.J.; Trevillion, E.A.

1975-01-01

Nitrogen in increasing amounts has been added to separate solutions of barium in sodium of constant composition (ca.4.40 mol % Ba) at 300 0 C. After rendering each mixture homogenous using an electromagnetic pump, filtration, and nitrogen analysis, all the N 2 added has been found in solution up to a solution composition approximating to Ba 4 N (i.e. 1.1 mol % N) beyond which the quantity of dissolved N 2 decreases progressively due to precipitation of the nitride Ba 2 N. The solubilization is interpreted in terms of strong preferential solvation of the nitride ion by barium cations. (author)

Experimental design for reflection measurements of highly reactive liquid or solid substances with application to liquid sodium

International Nuclear Information System (INIS)

Chan, S.H.; Gossler, A.A.

1980-01-01

This technical report describes the experimental part of a program on thermal radiation properties of reactor materials. A versatile goniometer system with associated electronic components and mechanical instruments has been assembled. It is designed to measure spectral, specular reflectances of highly reactive liquid or solid substances over a spectral range of 0.3 μ to 9 μ and incidence angles of 12 0 to 30 0 off the normal direction. The capability of measuring reflectances of liquid substances clearly distinguishes this experimental design from conventional systems which are applicable only to solid substances. This design has been used to measure the spectral, specular reflectance of liquid sodium and preliminary results obtained are compared with those of solid sodium measured by other investigators

Temperature field in concrete when in contact with hot liquids

International Nuclear Information System (INIS)

Andrade Lima, F.R. de.

1981-09-01

In an HCDA (Hypothetical Core Disruptive Accident) it is postulated that liquid metal coolants and core materials come in contact with the retaining concrete structure. A mathematical model and an associated computer program was previously developed to describle the transient heat and mass transfer in the concrete. Implementations on the original program-USINT- are included to consider the variations of the thermal conductivity as a function of the temperature. Also a subroutine - PLOTTI - is incorporated to the program for the plotting of the results. The new program - USINTG - is used to calculate the temperature and pressure fields and the water released from concrete structures during a sodium leak simulation and with the concrete structures in contact with liquid sodium. No consideration about chemical reactions involving the sodium when in contact with concrete is considered. (Author) [pt

Experimental results of the consequences of sodium water reactions at the bottom tube plate region of straight tube steam generators

International Nuclear Information System (INIS)

Ruloff, G.

1990-01-01

Experience with sodium water reactions has shown, that the course of such a steam generator accident depends strongly on its place in the steam generator. For the EFR steam generators we have to differentiate between: weld region at the upper tube plate (gas space); bundle region; weld region at the bottom tube plate. This paper describes results of a running tests program simulating the bottom tube plate area. One main part of these tests is the investigation of the influence of wastage protection shrouds between the tubes in the weld region to avoid a fast leak propagation and to give time for leak detection and mastering of the accidents. (author). 10 figs, 2 tabs

Calculation of wastage by small water leaks in sodium heated steam generators

International Nuclear Information System (INIS)

Tregonning, K.

1976-01-01

On the basis of mechanistic arguments it is suggested that the temperature of the wasting surface would provide a single physically meaningful parameter with which to correlate wastage data. A lumped parameter model is developed which predicts reaction temperature as a function of the major variables in the small water leak situation (Leak rate, tube spacing, sodium temperature). The calculated temperatures explain much of the observed behaviour of wastage rate with these variables and compare well with the limited temperature data available. Wastage rates are correlated with predicted temperature on a total activation energy basis. The results are encouraging and a first conservative method for the calculation of wastage by small water leaks in sodium-heated steam generators is produced

Preliminary experimental study of liquid lithium water interaction

International Nuclear Information System (INIS)

You, X.M.; Tong, L.L.; Cao, X.W.

2015-01-01

Highlights: • Explosive reaction occurs when lithium temperature is over 300 °C. • The violence of liquid lithium water interaction increases with the initial temperature of liquid lithium. • The interaction is suppressed when the initial water temperature is above 70 °C. • Steam explosion is not ignorable in the risk assessment of liquid lithium water interaction. • Explosion strength of liquid lithium water interaction is evaluated by explosive yield. - Abstract: Liquid lithium is the best candidate for a material with low Z and low activation, and is one of the important choices for plasma facing materials in magnetic fusion devices. However, liquid lithium reacts violently with water under the conditions of loss of coolant accidents. The release of large heats and hydrogen could result in the dramatic increase of temperature and pressure. The lithium–water explosion has large effect on the safety of fusion devices, which is an important content for the safety assessment of fusion devices. As a preliminary investigation of liquid lithium water interaction, the test facility has been built and experiments have been conducted under different conditions. The initial temperature of lithium droplet ranged from 200 °C to 600 °C and water temperature was varied between 20 °C and 90 °C. Lithium droplets were released into the test section with excess water. The shape of lithium droplet and steam generated around the lithium were observed by the high speed camera. At the same time, the pressure and temperature in the test section were recorded during the violent interactions. The preliminary experimental results indicate that the initial temperature of lithium and water has an effect on the violence of liquid lithium water interaction.

Preliminary experimental study of liquid lithium water interaction

Energy Technology Data Exchange (ETDEWEB)

You, X.M.; Tong, L.L.; Cao, X.W., E-mail: caoxuewu@sjtu.edu.cn

2015-10-15

Highlights: • Explosive reaction occurs when lithium temperature is over 300 °C. • The violence of liquid lithium water interaction increases with the initial temperature of liquid lithium. • The interaction is suppressed when the initial water temperature is above 70 °C. • Steam explosion is not ignorable in the risk assessment of liquid lithium water interaction. • Explosion strength of liquid lithium water interaction is evaluated by explosive yield. - Abstract: Liquid lithium is the best candidate for a material with low Z and low activation, and is one of the important choices for plasma facing materials in magnetic fusion devices. However, liquid lithium reacts violently with water under the conditions of loss of coolant accidents. The release of large heats and hydrogen could result in the dramatic increase of temperature and pressure. The lithium–water explosion has large effect on the safety of fusion devices, which is an important content for the safety assessment of fusion devices. As a preliminary investigation of liquid lithium water interaction, the test facility has been built and experiments have been conducted under different conditions. The initial temperature of lithium droplet ranged from 200 °C to 600 °C and water temperature was varied between 20 °C and 90 °C. Lithium droplets were released into the test section with excess water. The shape of lithium droplet and steam generated around the lithium were observed by the high speed camera. At the same time, the pressure and temperature in the test section were recorded during the violent interactions. The preliminary experimental results indicate that the initial temperature of lithium and water has an effect on the violence of liquid lithium water interaction.

Hydrogen isotope separation in hydrophobic catalysts between hydrogen and liquid water

Energy Technology Data Exchange (ETDEWEB)

Ye, Linsen, E-mail: yls2005@mail.ustc.edu.cn [China Academy of Engineering Physics, Mianyang 621900 (China); Luo, Deli [Science and Technology on Surface Physics and Chemistry Laboratory, Jiangyou 621907 (China); Tang, Tao; Yang, Wan; Yang, Yong [China Academy of Engineering Physics, Mianyang 621900 (China)

2015-11-15

Hydrogen isotope catalytic exchange between hydrogen and liquid water is a very effective process for deuterium-depleted potable water production and heavy water detritiation. To improve the characteristics of hydrophobic catalysts for this type of reaction, foamed and cellular structures of hydrophobic carbon-supported platinum catalysts were successfully prepared. Separation of deuterium or tritium from liquid water was carried out by liquid-phase catalytic exchange. At a gas–liquid ratio of 1.53 and exchange temperature of 70 °C, the theoretical plate height of the hydrophobic catalyst (HETP = 34.2 cm) was slightly lower than previously reported values. Changing the concentration of the exchange column outlet water yielded nonlinear changes in the height of the packing layer. Configurations of deuterium-depleted potable water and detritiation of heavy water provide references for practical applications.

Ionic-liquid-induced microfluidic reaction for water-soluble Ce1-xTbxF3 nanocrystal synthesis

International Nuclear Information System (INIS)

Xie Nan; Luan Weiling

2011-01-01

Luminescent lanthanide nanocrystals (NCs) are proposed to be a promising new class of fluorescent labeling agents due to their attractive optical and chemical features including low toxicity, wide photoluminescence (PL) emission and high resistance to photobleaching. In this paper, an ionic-liquid-induced synthesis of Ce 1-x Tb x F 3 nanoparticle was investigated via utilizing a capillary microreactor. Ionic liquid-[bmim]BF 4 acts as both a fluoride source and stabilizing solvent during the reaction, which was shown to be a key factor that governs luminescence intensity of the obtained nanoparticles. The luminescent properties can be greatly improved by optimizing the volume percentage of [bmim]BF 4 . Furthermore, the reaction temperature exerts an influence on the properties of the prepared samples. Experimental results show that the colloidal solutions of Tb 3+ -doped CeF 3 NCs exhibit the characteristic emission of Ce 3+ 5d-4f and Tb 3+5 D 4 - 7 F J (J = 6-3) transitions with 5 D 4 - 7 F 5 green emission at 542 nm as the strongest peak. The as-prepared samples are found dispersible in water with the quantum yield (in aqueous solution) as 12%, which indicates a potential application on biolabels, light-emitting diodes (LEDs) and redox luminescent switches.

Transient mixed convection in a cavity. Comparison between water and sodium

International Nuclear Information System (INIS)

Garnier, J.

1983-01-01

The basic problem studied is the interaction between a vortex and a thermal stratification. The experiments are done in a parallelepipedic cavity which bottom communicates with a rectangular channel. A forced flow in this channel induces a recirculating flow in the cavity. The transient condition is a decrease (step wise or slope) of the inlet temperature at a constant flowrate. This problem is studied with two different approaches: experiments in water or in sodium. In the sodium experiments, the dimension of the cavity ensures large values of the Peclet number (about 10 4 ) and a wide range of values for the Richardson number (from 0.1 to 3). With these experiment, all the regimes of mixed convection, from forced convection to complete stratification can be covered. These results are compared with the other approach using a water model. This comparison is very helpful for studies on the thermalhydraulic behavior of Liquid Metal Fast Breeder Reactors. (author)

Assessment of flow induced vibration in a sodium-sodium heat exchanger

Energy Technology Data Exchange (ETDEWEB)

Prakash, V. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)], E-mail: prakash@igcar.gov.in; Thirumalai, M.; Prabhakar, R.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

2009-01-15

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam. It is a liquid metal sodium cooled pool type fast reactor with all primary components located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to the secondary sodium in a sodium to sodium Intermediate Heat Exchanger (IHX), which in turn is transferred to water in the steam generator. PFBR IHX is a shell and tube type heat exchanger with primary sodium on shell side and secondary sodium in the tube side. Since IHX is one of the critical components placed inside the radioactive primary sodium, trouble-free operation of the IHX is very much essential for power plant availability. To validate the design and the adequacy of the support system provided for the IHX, flow induced vibration (FIV) experiments were carried out in a water test loop on a 60 deg. sector model. This paper discusses the flow induced vibration measurements carried out in 60 deg. sector model of IHX, the modeling criteria, the results and conclusion.

Synthesis and characterization of an octaimidazolium-based polyhedral oligomeric silsesquioxanes ionic liquid by an ion-exchange reaction.

Science.gov (United States)

Tan, Jinglin; Ma, Depeng; Sun, Xingrong; Feng, Shengyu; Zhang, Changqiao

2013-04-07

Preparation of POSS-min-DS, an octaimidazolium-based polyhedral oligomeric silsesquioxanes (POSS) room temperature ionic liquid, by an ion-exchange reaction between POSS and sodium dodecyl sulfate was reported. Octaimidazolium-based POSS was synthesized with more than 98% yield within 3 h. POSS-min-DS and octaimidazolium-based POSS were confirmed by (1)H, (13)C, and (29)Si NMR, FT-IR and elemental analysis.

Contribution to the study of hydrogenated and oxygenated impurities in liquid sodium

International Nuclear Information System (INIS)

Naud, G.

1964-07-01

This study is made up essentially of two parts. The first is devoted to the development of dosage methods for selectively determining the oxygenated and hydrogenated impurities present in sodium, that is the oxide, the hydride and the hydroxide. The second makes use of these methods for a study of the Na-H 2 -O 2 system, as well as of the related problem of the attack of pyrex glass by molten sodium. The conventional method for dosing oxygen by amalgamation was first adapted to the simultaneous measurement of the hydride. We then developed a method for dosing the total hydrogen by measuring successively the concentrations of gas present in the hydride and hydroxide form. This method is based on the thermal decomposition of the hydride and the reaction between sodium and the hydroxide. Our contribution to the study of the Na-H 2 -O 2 System consists first of all in the study of the reaction between hydrogen and sodium in the temperature range from 150 to 250 deg C and at a pressure of about 150 mm of mercury. The study of the thermal decomposition of the hydride in sodium was then studied. It was possible to make some qualitative observations concerning the reaction between sodium and sodium hydroxide. Finally some complementary tests made it possible to define the nature of the phenomena occurring during the attack of pyrex glass by sodium. (author) [fr

Destruction of contaminated metallic sodium wastes by reaction on alcohol and hydrolysis

International Nuclear Information System (INIS)

Brault, Auguste; Bruneau, Christian; Chevalier, Gerard; Kerfanto, Michel.

1977-02-01

The reactions of metallic sodium with organic compounds have been reviewed in the light of the problem. An experimental investigation is then described. It shows that metallic sodium can be changed into an alcoholate, then into a soda aqueous solution with conditions allowing to master the reaction velocity. Sodium reacts on the chosen alcohol, monoethyl ether diethylene glycol in the presence of xylene. The alcoholate thus formed is hydrolysed on removal of xylene by distillation. The alcohol set free is separated from soda aqueous phase by addition of an organic solvent and decantation. The alcohol and the solvents are regenerated and recycled [fr

Process and system for stirring liquid sodium flowing through the primary circuit of a steam generator

International Nuclear Information System (INIS)

Fabregue, J.P.

1982-01-01

The invention concerns the stirring of the liquid sodium of a steam generator comprising a primary circuit composed of an elongated vessel through which the liquid sodium flows, a secondary circuit composed of a number of tubes extending inside the long cyclindrical vessel. The process consists in imparting simultaneously to the liquid sodium, during its passage through the cylindrical vessel, a movement of continuous rotation about the longitudinal axis of the cylindrical vessel and an alternating series of radial movements, centripetal and centrifugal, in relation to the longitudinal axis, so that each unit quantity of the sodium comes into contact with a large number of tubes. The application particularly concerns steam generators for nuclear power stations [fr

A computer code SPHINCS for sodium fire safety evaluation

International Nuclear Information System (INIS)

Yamaguchi, Akira

2000-01-01

A computer code SPHINCS solves coupled phenomena of thermal-hydraulics and sodium fire based on a multi-zone model. It deals with arbitrary number of rooms each of which is connected mutually by doorway and penetrations. With regard to the combustion phenomena, flame sheet model and liquid droplet combustion model are used for pool and spray fire, respectively, with the chemical equilibrium model using Gibbs free energy minimization method. The chemical reaction and mass and heat transfer are solved interactively. A specific feature of SPHINCS is detailed representation of thermal-hydraulics of a sodium pool and a steel liner, which is placed on the floor to prevent sodium-concrete contact. The author analyzed a series of pool combustion experiments, in which gas and liner temperatures are measured in detail. It has been found that good agreement is obtained and the SPHINCS has been validated with regard to the pool combustion phenomena. Further research needs are identified for the pool spreading modeling considering thermal deformation of liner and measurement of pool fluidity property of a mixture of liquid sodium and reaction products. SPHINCS code is to be used mainly in the safety evaluation of the consequence of sodium fire accident of liquid metal cooled fast reactor. (author)

Preliminary Results on a Contact between 4 kg of Molten UO2 and Liquid Sodium

International Nuclear Information System (INIS)

Amblard, M.

1976-01-01

The CORECT II Experiment consists in simulating the penetration of sodium into an assembly when the fuel is molten. In other words, it is a shock-tube type of experiment with dimensions representative of a full-scale assembly. the experiment consists in dropping a 100 litre column of sodium onto partially molten UO 2 . The following measurements are carried out in transient regime: - sodium velocity in the column; - pressure in the interaction chamber; - pressures at the bottom and at the top of a 5 m tube; - pressure in the argon blanket. The experimental parameters are: - the mass of UO 2 involved (about 4 or 7 kg of 80% molten UO 2 ); - the initial temperature of the sodium (up to 700 deg. C); - the pressure of the residual gas in the interaction chamber during the fall of the sodium; - the dimensions of the interaction chamber and the sodium supply tube; - the form of contact between the UO 2 and the sodium (the sodium may fall on partially liquid and settled UO 2 or on UO 2 pre-dispersed by forced trapping of sodium). To date, 6 tests have been performed. These tests have always resulted in fine fragmentation without any violent interaction. Since no knowledge is available on the change of grain size distribution with time, on the temperature of grain formation, and on the grain movement in the sodium, it is very difficult to interpret these UO 2 -Na tests. We intend to carry out more severe interaction tests on this experimental set-up, by eliminating as much as possible the non-condensable gas which cushions the mechanical impact of the sodium on the UO 2 (tests have shown that by strongly de-pressurizing the liquid UO 2 the fuel could be dispersed by boiling, and this effect should also improve the possibilities of a liquid/liquid contact). - by injecting a little sodium into the UO 2 to facilitate its dispersion in the coolant

Uranium oxidation: characterization of oxides formed by reaction with water

International Nuclear Information System (INIS)

Fuller, E.L. Jr.; Smyrl, N.R.; Condon, J.B.; Eager, M.H.

1983-01-01

Three different uranium oxide samples have been characterized with respect to the different preparation techniques. Results show that the water reaction with uranium metal occurs cyclically forming laminar layers of oxide which spall off due to the strain at the oxide/metal interface. Single laminae are released if liquid water is present due to the prizing penetration at the reaction zone. The rate of reaction of water with uranium is directly proportional to the amount of adsorbed water on the oxide product. Rapid transport is effected through the open hydrous oxide product. Dehydration of the hydrous oxide irreversibly forms a more inert oxide which cannot be rehydrated to the degree that prevails in the original hydrous product of uranium oxidation with water. 27 figures

Feasibility study of solidification for low-level liquid waste generated by sulfuric acid elution treatment of spent ion exchange resin

International Nuclear Information System (INIS)

Asano, Takashi; Kawasaki, Tooru; Higuchi, Natsuko; Horikawa, Yoshihiko

2007-01-01

Low-level liquid waste with relatively high levels of radioactivity is generated by the sulfuric acid elution treatment of spent ion exchange resin used in water purification systems of nuclear power plants. We studied cement-like solidification process for this type waste that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete, and this leads to cracking during repository storage. It is also preferable that the pH of pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Our solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO 4 precipitation occurs with addition of Ba(OH) 2 ·8H 2 O to the liquid waste when the Ba/SO 4 molar ratio > 1. In the FLS procedure, silica fume and blast furnace slag are added to the liquid wastes containing Ba S O 4 precipitate. The CIS reaction yield is over 98% and the pH of pore water of the solid waste is 11.5 or less. Therefore, we think that our solidification process is one of the best methods for treating liquid waste that contains a high concentration of sodium sulfate. (author)

Analysis of the dynamic response of a double rupture disc assembly to simulated sodium-water reaction pressure pulses

International Nuclear Information System (INIS)

Leonard, J.R.

1980-03-01

A series of double rupture disc experiments were conducted in 1979 to evaluate the dynamic response characteristics of this pressure relief apparatus. The tests were performed in a facility with water simulating sodium and rising pressure pulses representative of the pressure increase resulting from a water/steam leak from a steam generator into sodium in the intermediate heat transport system of a breeder reactor power plant. Maximum source pressures ranged in magnitude from 50 psi to 800 psi. Dynamic response characteristics of each of the two rupture discs were similar to those observed in larger scale sodium-water experiments conducted in the Series I and Series II Large Leak Test Program at the Energy Technology Engineering Center. The SRI double rupture disc dynamic behavior was found to be consistent and amendable to modelling in the TRANSWRAP II computer code. A series of correlations which represent rupture disc buckling parameters were developed for use in the TRANSWRAP II code. The semi-empirical modeling of the rupture discs in the TRANSWRAP II code showed very good agreement with the experimental results

Validation of a CFD model simulating charge and discharge of a small heat storage test module based on a sodium acetate water mixture

DEFF Research Database (Denmark)

Dannemand, Mark; Fan, Jianhua; Furbo, Simon

2014-01-01

Experimental and theoretical investigations are carried out to study the heating of a 302 x 302 x 55 mm test box of steel containing a sodium acetate water mixture. A thermostatic bath has been set up to control the charging and discharging of the steel box. The charging and discharging has been...... for a Computational Fluid Dynamics (CFD) model. The CFD calculated temperatures are compared to measured temperatures internally in the box to validate the CFD model. Four cases are investigated; heating the test module with the sodium acetate water mixture in solid phase from ambient temperature to 52˚C; heating...... the module starting with the salt water mixture in liquid phase from 72˚C to 95˚C; heating up the module from ambient temperature with the salt water mixture in solid phase, going through melting, ending in liquid phase at 78˚C/82˚C; and discharging the test module from liquid phase at 82˚C, going through...

Heat exchanger, particularly liquid sodium heated steam generator

International Nuclear Information System (INIS)

Robin, Marcel; Tillequin, Jean.

1977-01-01

This invention relates to a liquid sodium heated steam generator the characteristic of which is an annular distribution chamber fed by two independent and diametrically opposed manifolds on a common horizontal axis, issuing respectively into two adjacent compartments made in the chambers on both sides of a vertical transversal partition containing the axis of the casing and extending perpendicularly to the manifolds, each compartment being itself divided into a number of adjacent sectors marked by folded metal sheets fixed to the distributor and shaped so as to present in pairs and with the chamber opposite the manifold issuing into a compartment two independent ducts for distributing the sodium flow [fr

Treatment of Residual Sodium and Sodium Potassium from Fast Reactors. Review of Recent Accomplishments, Challenges and Technologies

International Nuclear Information System (INIS)

2015-08-01

In addition to the usual radiation and conventional hazards present during the decommissioning of disused nuclear installations, the presence of residual sodium or the alloy sodium potassium — used in primary, secondary and support systems in reactors using liquid metal as a coolant — presents additional technical, safety and cost challenges for decommissioning. This results from the propensity of these materials to react exothermically with water and moisture in the air potentially resulting in toxic and explosive reactionsThis publication discusses a variety of treatment methods to be considered when dismantling components that still contain residual quantities of sodium or sodium potassium, several of which were presented as contributed papers to the IAEA session during the 5. International Conference and Exhibition on Decommissioning Challenges, in Avignon, France, 7–11 April 2013. The publication provides a synthesis of information presented during the session, which was developed further at a consultants meeting held in Vienna, 2–6 December 2013. Decommissioning challenges faced at eight different facilities in five different countries are discussed, as well as the achievements and lessons learned that are of value to the worldwide decommissioning community

Measurement strategy of the water leakage into a low pressure sodium boundary for a liquid metal reactor

International Nuclear Information System (INIS)

Hur, S.; Kim, D.H.; Seong, S.H.; Kim, S.O.

2004-01-01

This paper deals with the measurement strategy of a water leakage into sodium boundary for a liquid metal reactor. There are several methods including the chemical sensing method, pressure sensing methods, and non-destructive method including the acoustic monitoring technique to measure the leakage. As for the results of the analysis with respect to the event propagation characteristics, it has been recommended that the acoustic method has a capability to detect small and intermediate leaks within required response time of 10 seconds. A leak of one gram/sec could be currently detected within the required response time of 10 seconds with a high reliability. In the case of less than a one gram/sec leakage, the response time could not meet our requirements due to a complicated signal processing logic. Thus, the system configuration for a fast processing of a leak detection has been recommended. It is expected that this configuration of the leak detection system could reduce the response time due to the distributed and parallel processing scheme. (orig.)

Innovating analytical spectroscopies for the improvement of liquid sodium cooled fast neutron reactors safety

International Nuclear Information System (INIS)

Maury, C.

2012-01-01

In the context of the project of sodium fast reactor ASTRID, CEA is currently developing new analytical techniques to monitor the chemical purity of liquid sodium. Indeed, incidental situations occurring in the reactor, such as fuel clad failures, leakages in the steam generator or in the coolant pumps, and accelerated corrosion, might release several elements in the sodium. Analytical techniques based on laser ablation and emission spectroscopy are well suited for this application. They do not require any sample preparation, and can perform direct on-line analysis. Amongst them, Laser-Induced Breakdown Spectroscopy (LIBS) and Laser-Ablation coupled to Laser-Induced Fluorescence (LA-LIF) have been selected for this study. The objective of this work was to characterize the sensitivity of those two techniques for the detection of impurities in liquid sodium. Their limits of detection were calculated for model analytes using calibration lines. Then results were theoretically extrapolated to other analytes of interest. This study shows the feasibility of the detection of steel corrosion products in liquid sodium. However, the LIBS technique is more robust and easier to implement, and would therefore be more suited to nuclear conditions. (author) [fr

In-pile study of the reaction between breeder fuel and sodium

International Nuclear Information System (INIS)

Hugot, J.P.

1982-10-01

Studies carried out until now show that the determinant parameter of fuel can failure evolution is the development of the reaction between mixed uranium and plutonium dioxide and sodium. The parameters of the reaction are presented from results of an out of pile study, as also results obtained from examination on pins failed in reactors. The best way to study in pile the development of the reaction was to irradiate at a constant power a fuel pin containing sodium. In the experiment, the pin was equipped with a central thermocouple. It shows, that the reaction is developing intergranularly, from cracks and interpellet spaces, in an internal fringe of the fuel before spreading to the periphery. An overheating of the pin is associated to the development of the reaction as also a modification of the fuel pin geometry and a reduction of the oxide [fr

Accurate numerical simulation of reaction-diffusion processes for heavy oil recovery

Energy Technology Data Exchange (ETDEWEB)

Govind, P.A.; Srinivasan, S. [Society of Petroleum Engineers, Richardson, TX (United States)]|[Texas Univ., Austin, TX (United States)

2008-10-15

This study evaluated a reaction-diffusion simulation tool designed to analyze the displacement of carbon dioxide (CO{sub 2}) in a simultaneous injection of carbon dioxide and elemental sodium in a heavy oil reservoir. Sodium was used due to the exothermic reaction of sodium with in situ that occurs when heat is used to reduce oil viscosity. The process also results in the formation of sodium hydroxide that reduces interfacial tension at the bitumen interface. A commercial simulation tool was used to model the sodium transport mechanism to the reaction interface through diffusion as well as the reaction zone's subsequent displacement. The aim of the study was to verify if the in situ reaction was able to generate sufficient heat to reduce oil viscosity and improve the displacement of the heavy oil. The study also assessed the accuracy of the reaction front simulation tool, in which an alternate method was used to model the propagation front as a moving heat source. The sensitivity of the simulation results were then evaluated in relation to the diffusion coefficient in order to understand the scaling characteristics of the reaction-diffusion zone. A pore-scale simulation was then up-scaled to grid blocks. Results of the study showed that when sodium suspended in liquid CO{sub 2} is injected into reservoirs, it diffuses through the carrier phase and interacts with water. A random walk diffusion algorithm with reactive dissipation was implemented to more accurately characterize reaction and diffusion processes. It was concluded that the algorithm modelled physical dispersion while neglecting the effect of numerical dispersion. 10 refs., 3 tabs., 24 figs.

Alternative solvent-based methyl benzoate vortex-assisted dispersive liquid-liquid microextraction for the high-performance liquid chromatographic determination of benzimidazole fungicides in environmental water samples.

Science.gov (United States)

Santaladchaiyakit, Yanawath; Srijaranai, Supalax

2014-11-01

Vortex-assisted dispersive liquid-liquid microextraction using methyl benzoate as an alternative extraction solvent for extracting and preconcentrating three benzimidazole fungicides (i.e., carbendazim, thiabendazole, and fluberidazole) in environmental water samples before high-performance liquid chromatographic analysis has been developed. The selected microextraction conditions were 250 μL of methyl benzoate containing 300 μL of ethanol, 1.0% w/v sodium acetate, and vortex agitation speed of 2100 rpm for 30 s. Under optimum conditions, preconcentration factors were 14.5-39.0 for the target fungicides. Limits of detection were obtained in the range of 0.01-0.05 μg/L. The proposed method was then applied to surface water samples and the recovery evaluations at three spiked concentration levels of 5, 30, and 50 μg/L were obtained in the range of 77.4-110.9% with the relative standard deviation water samples. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Selection of steam generator materials for sodium cooled fast breeders

International Nuclear Information System (INIS)

Berge, P.

1977-01-01

The sodium water heat exchangers are now considered as the stumbling block in the development of liquid metal cooled fast breeders, due to the risk of sodium-water reactions. The selection of the materials for these tube-bundles has been very broad, for the different existing, or in-project, reactors in the world: low alloy 2 1/4 Cr - 1 Mo steels (unstabilized or stabilized); 9 Cr - 1 Mo ferritic steel; 18 Cr - 10 Ni austenitic stainless steels; alloy 800. On can also add other ferritic steels, as 9 Cr - 2 Mo stabilized, which are studied for this application. In the framework of the E.D.F.-C.E.A. working group a major effort was undertaken to study the characteristics of these various materials with respect to the main criteria governing construction of the tube bundles and their performance in service: mechanical characteristics at high temperature; fabrication and welding; behavior with respect to mass transfer in sodium; carburization and decarburization; corrosion resistance. The main lines and results of this program are described [fr

Propargyl organometallic compounds. II. Alkylation of sodium derivatives of 1-alkyl-1-aryl-2-alkynes in liquid ammonia

International Nuclear Information System (INIS)

Libman, N.M.; Sevryukov, Yu.P.

1987-01-01

In most cases the alkylation of the sodium derivatives of 1-phenyl-1-alkyl-2-alkynes by methyl, ethyl, isopropyl, and tert-butyl bromides in liquid ammonia takes place preferentially at the sp 2 -hybridized carbon atom, and this leads to the formation of the corresponding acetylenes, The regioselectivity of the reaction is explained by the greater softness of the trigonal atom of the ambient propargyl anion and its smaller screening by the solvate shell compared with the diagonal atom

Novel catalysts for isotopic exchange between hydrogen and liquid water

International Nuclear Information System (INIS)

Butler, J.P.; Rolston, J.H.; Stevens, W.H.

1978-01-01

Catalytic isotopic exchange between hydrogen and liquid water offers many inherent potential advantages for the separation of hydrogen isotopes which is of great importance in the Canadian nuclear program. Active catalysts for isotopic exchange between hydrogen and water vapor have long been available, but these catalysts are essentially inactive in the presence of liquid water. New, water-repellent platinum catalysts have been prepared by: (1) treating supported catalysts with silicone, (2) depositing platinum on inherently hydrophobic polymeric supports, and (3) treating platinized carbon with Teflon and bonding to a carrier. The activity of these catalysts for isotopic exchange between countercurrent streams of liquid water and hydrogen saturated with water vapor has been measured in a packed trickle bed integral reactor. The performance of these hydrophobic catalysts is compared with nonwetproofed catalysts. The mechanism of the overall exchange reaction is briefly discussed. 6 figures

Experimental and numerical analysis of the combustor for a cogeneration system based on the aluminum/water reaction

International Nuclear Information System (INIS)

Milani, Massimo; Montorsi, Luca; Paltrinieri, Fabrizio; Stefani, Matteo

2014-01-01

Highlights: • Aluminum reaction with water is studied as a technology for hydrogen production. • A test rig is developed for the analysis of aluminum/water reaction. • The system is the core component of a cogeneration plant for hydrogen/power production. • The interaction of liquid aluminum jet and water steam stream is investigated. • The main capabilities of the injection system are assessed. - Abstract: The paper focuses on the design of the experimental apparatus aimed at analyzing the performance of the combustion chamber of a cogeneration system based on the reaction of liquid aluminum and water steam. The cogeneration system exploits the heat released by the oxidation of aluminum with water for super-heating the vapor of a steam cycle and simultaneously producing hydrogen. The only by-product is alumina, which in a closed loop can be recycled back and transformed again into aluminum. Therefore, aluminum is used as an energy carrier to transport the energy from the alumina reduction plant to the location of the proposed system. The water is also used in a closed loop since the amount of water produced employing the hydrogen obtained by the proposed system corresponds to the oxidizing water for the Al/H 2 O reaction. This study investigates the combustor where the liquid aluminum–steam reaction takes place. In particular, the design of the combustion chamber and the interaction between the liquid aluminum jet and the water steam flow are evaluated using a numerical and an experimental approach. The test rig is specifically designed for the analysis of the liquid aluminum injection in a slightly super-heated steam stream. The first experiments are carried out to verify the correct behavior of the test rig. Thermography is employed to qualitatively assess the steam entrainment of the liquid aluminum jet. Finally, the experimental measurements are compared with the multi-dimension multi-phase flow simulations in order to estimate the influence of

Market reaction to bank liquidity regulation

OpenAIRE

Bruno, Brunella; Onali, Enrico; Schaeck, Klaus

2018-01-01

We measure market reactions to announcements concerning liquidity regulation, a key innovation in the Basel framework. Our initial results show that liquidity regulation attracts negative abnormal returns. However, the price responses are less pronounced when coinciding announcements concerning capital regulation are backed out, suggesting that markets do not consider liquidity regulation to be binding. Bank- and country-specific characteristics also matter. Liquid balance sheets and high cha...

Perspective: Chemical reactions in ionic liquids monitored through the gas (vacuum)/liquid interface.

Science.gov (United States)

Maier, F; Niedermaier, I; Steinrück, H-P

2017-05-07

This perspective analyzes the potential of X-ray photoelectron spectroscopy under ultrahigh vacuum (UHV) conditions to follow chemical reactions in ionic liquids in situ. Traditionally, only reactions occurring on solid surfaces were investigated by X-ray photoelectron spectroscopy (XPS) in situ. This was due to the high vapor pressures of common liquids or solvents, which are not compatible with the required UHV conditions. It was only recently realized that the situation is very different when studying reactions in Ionic Liquids (ILs), which have an inherently low vapor pressure, and first studies have been performed within the last years. Compared to classical spectroscopy techniques used to monitor chemical reactions, the advantage of XPS is that through the analysis of their core levels all relevant elements can be quantified and their chemical state can be analyzed under well-defined (ultraclean) conditions. In this perspective, we cover six very different reactions which occur in the IL, with the IL, or at an IL/support interface, demonstrating the outstanding potential of in situ XPS to gain insights into liquid phase reactions in the near-surface region.

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

Science.gov (United States)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

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

Analysis of aromatic amines in water samples by liquid-liquid-liquid microextraction with hollow fibers and high-performance liquid chromatography.

Science.gov (United States)

Zhao, Limian; Zhu, Lingyan; Lee, Hian Kee

2002-07-19

Liquid-liquid-liquid microextraction (LLLME) with hollow fibers in high-performance liquid chromatography (HPLC) has been applied as a rapid and sensitive quantitative method for the detection of four aromatic amines (3-nitroaniline, 4-chloroaniline, 4-bromoaniline and 3,4-dichloroaniline) in environmental water samples. The preconcentration procedure was induced by the pH difference inside and outside the hollow fiber. The target compounds were extracted from 4-ml aqueous sample (donor solution, pH approximately 13) through a microfilm of organic solvent (di-n-hexyl ether), immobilized in the pores of a hollow fiber (1.5 cm length x 0.6 mm I.D.), and finally into 4 microl of acid acceptor solution inside the fiber. After a prescribed period of time, the acceptor solution inside the fiber was withdrawn into the microsyringe and directly injected into the HPLC system for analysis. Factors relevant to the extraction procedure were studied. Up to 500-fold enrichment of analytes could be obtained under the optimized conditions (donor solution: 0.1 M sodium hydroxide solution with 20% sodium chloride and 2% acetone; organic phase: di-n-hexyl ether; acceptor solution: 0.5 M hydrochloric acid and 500 mM 18-crown-6 ether; extraction time of 30 min; stirring at 1,000 rev./min). The procedure also served as a sample clean-up step. The influence of humic acid on the extraction efficiency was also investigated, and more than 85% relative recoveries of the analytes at two different concentrations (20 and 100 microg/l) were achieved at various concentration of humic acid. This technique is a low cost, simple and fast approach to the analysis of polar compounds in aqueous samples.

The treatment of contaminated sodium: a literature study

Energy Technology Data Exchange (ETDEWEB)

Van Alsenoy, V; Rahier, A.

1996-07-01

At the Belgian Nuclear Research Centre SCK-CEN, several experiments concerning safety aspects of Liquid Metal Fast Breeder reactors were carried out. During these experiments, an important amount of sodium containing waste was produced. In view of the treatment of this waste, a literature study was performed and third parties were contacted to find a solution. This document summarizes the results of this study. The sodium waste has been characterized by a theoretical study and by radiological measurements. The waste consists mainly of metallic sodium contaminated with corrosion activation products, fission products and even fuel particles. The sodium might also be contaminated with oxidation and reduction products like Na{sub 2}O and NaH. The most important contaminant is {sup 137}Cs. Several third parties, with experience in treating sodium, were contacted and they proposed a treatment of the sodium based on its reaction with water or alcohol. From a safety point of view, these reactions are not satisfactory because they are all exothermic and lead to flammable products or even make use of flammable reactants. Therefore, all the parties foresee extensive and expensive studies prior to the treatment. The urgent nature of the issues together with the important safety aspects were the incentives for the Research and Development group of the Radioactive Waste and Cleanup to look for alternatives. For this purpose, a research programme has been started with the aim to define, test, demonstrate and finally apply a safe process for the treatment of contaminated sodium by oxidation on a fluidized bed followed by vitrification. The collected information confirms that the oxidation of sodium vapour can be carried out safely, leading to the formation of sodium peroxide and oxide.

The treatment of contaminated sodium: a literature study

International Nuclear Information System (INIS)

Van Alsenoy, V; Rahier, A.

1996-07-01

At the Belgian Nuclear Research Centre SCK-CEN, several experiments concerning safety aspects of Liquid Metal Fast Breeder reactors were carried out. During these experiments, an important amount of sodium containing waste was produced. In view of the treatment of this waste, a literature study was performed and third parties were contacted to find a solution. This document summarizes the results of this study. The sodium waste has been characterized by a theoretical study and by radiological measurements. The waste consists mainly of metallic sodium contaminated with corrosion activation products, fission products and even fuel particles. The sodium might also be contaminated with oxidation and reduction products like Na 2 O and NaH. The most important contaminant is 137 Cs. Several third parties, with experience in treating sodium, were contacted and they proposed a treatment of the sodium based on its reaction with water or alcohol. From a safety point of view, these reactions are not satisfactory because they are all exothermic and lead to flammable products or even make use of flammable reactants. Therefore, all the parties foresee extensive and expensive studies prior to the treatment. The urgent nature of the issues together with the important safety aspects were the incentives for the Research and Development group of the Radioactive Waste and Cleanup to look for alternatives. For this purpose, a research programme has been started with the aim to define, test, demonstrate and finally apply a safe process for the treatment of contaminated sodium by oxidation on a fluidized bed followed by vitrification. The collected information confirms that the oxidation of sodium vapour can be carried out safely, leading to the formation of sodium peroxide and oxide

Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

Energy Technology Data Exchange (ETDEWEB)

Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

Measurement of the activity coefficient of carbon in steels in liquid sodium

International Nuclear Information System (INIS)

Surville, G.

1983-06-01

In sodium cooled fast reactors carbon is both a carbon impurity and element of structural materials. Carbon transfert through liquid sodium can produce carburization or decarburization of structural materials. Carbon content in sodium is determined with thin foils of austenitic alloys, when equilibrium is reached thermodynamic activity of carbon in sodium is deduced from carbon activity in alloys. Studied alloys are FeMn 20%, FeNi 30%, Z2CN 18-10 and Z3CND17-13. Carbon activity of alloys in sodium was between 5.10 -3 and 10 -1 at 600 and 650 0 C. Calibration was obtained with the alloys FeNi 30% in gaseous mixtures He-CO-CO 2 of known activity [fr

Triggered fragmentation experiment with sodium, silicone oil and pentane

International Nuclear Information System (INIS)

Morita, T.

1990-12-01

Within the analysis of severe hypothetical fast breeder accidents the consequences of a fuel-coolant-interaction have to be considered, i.e. the thermal interaction between hot molten fuel and sodium. For the detailed understanding of the fragmentation during the thermal interaction of a hot liquid droplet with a cold fluid series of experiments were performed with sodium and solicone oil as hot liquid and pentane as cold easily volatile fluid. For the precise observation of the reaction an efficient high speed camera with a maximum recording frequency of 1x105 f/s was used. So the fragmentation caused by boiling phenomena could be observed. The pictures were used to estimate quantitatively e.g. the volume of the reaction zone and its expansion rate. By a special measuring device for the first time results on the time dependent portion of the liquid within the reaction zone could be gained. Based on the measured results of the experiments the course of a typical reaction, which can be devided into six phases, is presented and physically explained in this report. The influence of experimental parameters, as pressure of the external trigger and temperature of the hot liquid droplet, was investigated and from this the role of the homogeneous nucleation temperature and the external trigger for the reaction was deduced. (orig.) [de

Predictive modeling of liquid-sodium thermal–hydraulics experiments and computations

International Nuclear Information System (INIS)

Arslan, Erkan; Cacuci, Dan G.

2014-01-01

Highlights: • We applied the predictive modeling method of Cacuci and Ionescu-Bujor (2010). • We assimilated data from sodium flow experiments. • We used computational fluid dynamics simulations of sodium experiments. • The predictive modeling method greatly reduced uncertainties in predicted results. - Abstract: This work applies the predictive modeling procedure formulated by Cacuci and Ionescu-Bujor (2010) to assimilate data from liquid-sodium thermal–hydraulics experiments in order to reduce systematically the uncertainties in the predictions of computational fluid dynamics (CFD) simulations. The predicted CFD-results for the best-estimate model parameters and results describing sodium-flow velocities and temperature distributions are shown to be significantly more precise than the original computations and experiments, in that the predicted uncertainties for the best-estimate results and model parameters are significantly smaller than both the originally computed and the experimental uncertainties

Experimental study and phenomenological modeling of the hydrolysis of tritiated sodium: influence of experimental conditions on the tritium distribution in the effluents

International Nuclear Information System (INIS)

Chassery, Aurelien

2014-01-01

Within the framework of the decommissioning of fast reactors, several processes are under investigation regarding sodium disposal. One of them rests on the implementation of the sodium-water reaction (SWR), in a controlled and progressive way, to remove residual sodium containing impurities such as sodium hydrides, sodium oxides and tritiated sodium hydrides. Such a hydrolysis releases some amount of energy and produces a liquid effluent, composed of a solution of soda, and a gaseous effluent, composed of hydrogen, steam and an inert gas. The tritium, originally into the sodium as a soluble (T - ) or precipitate form (NaT), will be distributed between the liquid and gaseous effluent, and according to two chemical forms, the tritium hydride HT and the tritiated water HTO. HTO being 10,000 times more radio-toxic than HT, a precise knowledge of the mechanisms governing the distribution of tritium is necessary in order to estimate the exhaust gas releases and design the process needed to treat the off-gas before its release into the environment. An experimental study has been carried out in order to determine precisely the phenomena involved in the hydrolysis. The influence of the experimental conditions on the tritium distribution has been tested. The results of this study leaded to a phenomenological description of the tritiated sodium hydrolysis that will help to predict the composition of the effluents, regarding tritium. (author) [fr

Acidic ionic liquids for n-alkane isomerization in a liquid-liquid or slurry-phase reaction mode

Energy Technology Data Exchange (ETDEWEB)

Meyer, C.; Hager, V.; Geburtig, D.; Kohr, C.; Wasserscheid, P. [Erlangen-Nuernberg Univ. (Germany). Lehrstuhl fuer Chemische Reaktionstechnik; Haumann, M. [Chemical Reaction Engineering, FAU Busan Campus, Korea (Korea, Republic of)

2011-07-01

Highly acidic ionic liquid (IL) catalysts offer the opportunity to convert n-alkanes at very low reaction temperatures. The results of IL catalyzed isomerization and cracking reactions of pure n-octane are presented. Influence of IL composition, [C{sub 4}C{sub 1}Im]Cl / AlCl{sub 3} / H{sub 2}SO{sub 4} and [C{sub 4}C{sub 1}Im]Cl / AlCl{sub 3} / 1-chlorooctane, on catalyst activity and selectivities to branched alkanes was investigated. Acidic chloroaluminate IL catalysts form liquid-liquid biphasic systems with unpolar organic product mixtures. Thus, recycling of the acidic IL is enabled by simple phase separation in the liquid-liquid biphasic reaction mode or the IL can be immobilized on an inorganic support with a large specific surface area. These supported ionic liquid phase (SILP) catalysts offer the advantage to get a macroscopically heterogeneous system while still preserving all benefits of the homogeneous catalyst which can be used for the slurry-phase n-alkane isomerization. The interaction of the solid support and acidic IL influences strongly the catalytic activity. (orig.)

Experimental study of sodium fires on concrete based on the sodium-concrete reaction and its consequences: study of the behavior of various concretes under metallic sheaths

International Nuclear Information System (INIS)

Berlin; Colome, J.; Malet, J.C.

The problem created by the violent reaction between hot sodium and concrete has only recently been recognized. Its importance was evidenced during experiments in which the sodium-barium oxide concrete reactions led to violent explosions. SESR approached this question during its experimental programs Cassandre and Lucifer. The Cassandre 01 experiment demonstrated the sodium-ordinary concrete reaction, where sodium was burned directly in a concrete vat. The consequences of this fire, pulverization of sodium particles, explosions and deterioration of the concrete led to consideration of protecting the concrete. Among possible shieldings sheath metal appeared to be the safest solution. The Cassandre 08, Lucifer 01 and Lucifer 04 experiments were used to study the behavior of various qualities of concrete protected from fire by a metal wall. The results show that a metal cladding efficiently protects concrete from sodium leaks

Non-Fermi Liquids as Highly Active Oxygen Evolution Reaction Catalysts.

Science.gov (United States)

Hirai, Shigeto; Yagi, Shunsuke; Chen, Wei-Tin; Chou, Fang-Cheng; Okazaki, Noriyasu; Ohno, Tomoya; Suzuki, Hisao; Matsuda, Takeshi

2017-10-01

The oxygen evolution reaction (OER) plays a key role in emerging energy conversion technologies such as rechargeable metal-air batteries, and direct solar water splitting. Herein, a remarkably low overpotential of ≈150 mV at 10 mA cm -2 disk in alkaline solutions using one of the non-Fermi liquids, Hg 2 Ru 2 O 7 , is reported. Hg 2 Ru 2 O 7 displays a rapid increase in current density and excellent durability as an OER catalyst. This outstanding catalytic performance is realized through the coexistence of localized d-bands with the metallic state that is unique to non-Fermi liquids. The findings indicate that non-Fermi liquids could greatly improve the design of highly active OER catalysts.

Development of sodium technology

International Nuclear Information System (INIS)

Hwang, Sung Tai; Nam, H. Y.; Choi, Y. D.

2000-05-01

The objective of present study is to produce the experimental data for development and verification of computer codes for development of LMR and to develop the preliminary technologies for the future large scale verification experiments. A MHD experimental test loop has been constructed for the quantitative analysis of the effect of magnetic field on the sodium flow and experiments are carried out for three EM pumps. The previous pressure drop correlations are evaluated using the experimental data obtained from the pressure drop experiment in a 19-pin fuel assembly with wire spacer. An dimensionless variable is proposed to describe the amplitude and frequency of the fluctuation of free surface using the experimental data obtained from free surface experimental apparatus and an empirical correlation is developed using this dimensionless variable. An experimental test loop is constructed to measure the flow characteristics in IHX shell side and the local pressure drop in fuel assembly, and to test the vibration behaviour of fuel pins due to flow induced vibration. The sodium two-phase flow measuring technique using the electromagnetic flowmeter is developed and the sodium differential pressure drop measuring technique using the method of direct contact of sodium and oil is established. The work on the analysis of sodium fire characteristics and produce data for vlidation of computer code is performed. Perfect reopen time of self plugged leak path was observed to be about 130 minutes after water leak initiation. Reopen shape of a specimen appeared to be double layer of circular type, and reopen size of this specimen surface was about 2mm diameter on sodium side. In small water leakage experiments, the following correlation equation about the reopen time between sodium temperature and initial leak rate was obtained, τ c = δ·g -0.83 ·10 (3570/T Na -3.34) , in 400-500 deg C of liquid sodium atmosphere. The characteristics of pressure propagation and gas flow, and

Development of sodium technology

Energy Technology Data Exchange (ETDEWEB)

Hwang, Sung Tai; Nam, H Y; Choi, Y D [and others

2000-05-01

The objective of present study is to produce the experimental data for development and verification of computer codes for development of LMR and to develop the preliminary technologies for the future large scale verification experiments. A MHD experimental test loop has been constructed for the quantitative analysis of the effect of magnetic field on the sodium flow and experiments are carried out for three EM pumps. The previous pressure drop correlations are evaluated using the experimental data obtained from the pressure drop experiment in a 19-pin fuel assembly with wire spacer. An dimensionless variable is proposed to describe the amplitude and frequency of the fluctuation of free surface using the experimental data obtained from free surface experimental apparatus and an empirical correlation is developed using this dimensionless variable. An experimental test loop is constructed to measure the flow characteristics in IHX shell side and the local pressure drop in fuel assembly, and to test the vibration behaviour of fuel pins due to flow induced vibration. The sodium two-phase flow measuring technique using the electromagnetic flowmeter is developed and the sodium differential pressure drop measuring technique using the method of direct contact of sodium and oil is established. The work on the analysis of sodium fire characteristics and produce data for vlidation of computer code is performed. Perfect reopen time of self plugged leak path was observed to be about 130 minutes after water leak initiation. Reopen shape of a specimen appeared to be double layer of circular type, and reopen size of this specimen surface was about 2mm diameter on sodium side. In small water leakage experiments, the following correlation equation about the reopen time between sodium temperature and initial leak rate was obtained, {tau}{sub c} = {delta}{center_dot}g{sup -0.83}{center_dot}10{sup (3570/T{sub Na}-3.34)}, in 400-500 deg C of liquid sodium atmosphere. The characteristics

Determination of oxygen in liquid sodium

International Nuclear Information System (INIS)

Torre, M. de la; Lapena, J.; Galindo, F.; Couchoud, M.; Celis, B. de; Lopez-Araquistain, J.L.

1976-01-01

The behaviour is analysed of a device for 'in-line' sampling and vacuum distillation. With this procedure 95 results were obtained for the solubility of oxygen in liquid sodium at temperatures between 125 0 and 300 0 C. The correlation between the concentration of oxygen in a saturation state and the corresponding temperature is represented by: 1g C = 6,17 - 2398/T, where C expressed ppm of oxygen by weight and T is the saturation temperature in 0 K. Reference is also made to the first results obtained with the electrochemical oxygen meter and the system for taking and recording data. (author)

Aspects and mechanisms of austenitic stainless steel corrosion in case of sodium leaks under mineral wool insulation

International Nuclear Information System (INIS)

Bertrand, C.; Ardellier, A.

1996-01-01

Sodium pipe rupture tests representative of Fast Reactors Accidents have been carried out on austenitic stainless steel surfaces. These tests improve our knowledge of small sodium leakage propagation in mineral wool insulation. They explain the new and unexpected aspects of the crevice corrosion phenomenon which has been observed on austenitic stainless steel pipe surfaces. Experimental results show that corrosion is limited to a peripheral annular zone, which extends out in concentric waves. The diameter of this corrosion zone is practically constant. Tests show that sodium does not expand directly on the pipe surface. Sodium sprays through mineral wool insulation, where chemical reaction between silica fibers, occluded oxygen and water vapor occur at the same time. Simultaneously, there is a diffusion phenomenon of liquid Na droplets on the mineral wool fibers. The study allows to prove the electrochemical nature of the corrosion. The excess liquid Na, spraying as droplets induces an anodic dissolution mechanism by differential aeration. This phenomenon explains the random microscopic and macroscopic aspects of material removal. (authors). 1 ref., 16 figs

Fertile assembly for a fast neutron nuclear reactor cooled by liquid sodium, with regulation of the cooling rate

International Nuclear Information System (INIS)

Pradal, L.; Berte, M.; Chiarelli, C.

1985-01-01

The assembly has a casing in which are arranged the fertile elements, the liquid sodium flowing through the casing along these elements. It includes several apertured diaphragms transverse to the rods to regulate the liquid sodium flow rate. At least one diaphragm, in its central part around its aperture, of a material soluble in liquid sodium, such as copper. The invention applies, more particularly, to fast neutron nuclear reactor having a heterogeneous core. The coolant flow can increase with time to match the increased power generated by the fertile assembly along its life [fr

Thermophysical properties of sodium in the liquid and gaseous states

International Nuclear Information System (INIS)

Thurnay, K.

1981-02-01

A system of temperature and density dependent thermal-property-functions has been developed and checked for the sodium, using all of the accessible experimental data and the mutual relationships of the properties. In extending these properties beyond the range of the measurements substance independent physical relations have been used. The property-descriptions are valid for all temperatures above the melting point of the sodium and for all densities below the melting density of the liquid. The system consists of the following thermal properties: pressure, heat capacity at constant volume, thermal conductivity. (orig.) [de

Mechanistic approach to the sodium leakage and fire analysis

International Nuclear Information System (INIS)

Yamaguchi, Akira; Muramatsu, Toshiharu; Ohira, Hiroaki; Ida, Masao

1997-04-01

In December 1995, a thermocouple well was broken and liquid sodium leaked out of the intermediate heat transport system of the prototype fast breeder reactor Monju. In the initiating process of the incident, liquid sodium flowed out through the hollow thermocouple well, nipple and connector. As a result, liquid sodium, following ignition and combustion, was dropping from the connector to colide with the duct and grating placed below. The collision may cause fragmentation and scattering of the sodium droplet that finally was piled up on the floor. This report deals with the development of computer programs for the phenomena based on mechanistics approach. Numerical analyses are also made for fundamental sodium leakage and combustion phenomenon, sodium combustion experiment, and Monju incident condition. The contents of this report is listed below: (1) Analysis of chemical reaction process based on molecular orbital method, (2) Thermalhy draulic analysis of the sodium combustion experiment II performed in 1996 at O-arai Engineering Center, PNC, (3) Thermalhy draulic analysis of room A-446 of Monju reactor when the sodium leakage took place, (4) Direct numerical simulation of sodium droplet, (5) Sodium leakage and scattering analysis using three dimensional particle method, (6) Multi-dimensional combustion analysis and multi-point approximation combustion analysis code. Subsequent to the development work of the programs, they are to be applied to the safety analysis of the Fast Breeder Reactor. (author)

Speciation of mercury in water samples by dispersive liquid-liquid microextraction combined with high performance liquid chromatography-inductively coupled plasma mass spectrometry

International Nuclear Information System (INIS)

Jia Xiaoyu; Han Yi; Liu Xinli; Duan Taicheng; Chen Hangting

2011-01-01

The dispersive liquid-liquid microextraction (DLLME) combined with high performance liquid chromatography-inductively coupled plasma mass spectrometry for the speciation of mercury in water samples was described. Firstly methylmercury (MeHg + ) and mercury (Hg 2+ ) were complexed with sodium diethyldithiocarbamate, and then the complexes were extracted into carbon tetrachloride by using DLLME. Under the optimized conditions, the enrichment factors of 138 and 350 for MeHg + and Hg 2+ were obtained from only 5.00 mL sample solution. The detection limits of the analytes (as Hg) were 0.0076 ng mL -1 for MeHg + and 0.0014 ng mL -1 for Hg 2+ , respectively. The relative standard deviations for ten replicate measurements of 0.5 ng mL -1 MeHg + and Hg 2+ were 6.9% and 4.4%, respectively. Standard reference material of seawater (GBW(E)080042) was analyzed to verify the accuracy of the method and the results were in good agreement with the certified values. Finally, the developed method was successfully applied for the speciation of mercury in three environmental water samples.

The Performance test of Mechanical Sodium Pump with Water Environment

International Nuclear Information System (INIS)

Cho, Chungho; Kim, Jong-Man; Ko, Yung Joo; Jeong, Ji-Young; Kim, Jong-Bum; Ko, Bock Seong; Park, Sang Jun; Lee, Yoon Sang

2015-01-01

As contrasted with PWR(Pressurized light Water Reactor) using water as a coolant, sodium is used as a coolant in SFR because of its low melting temperature, high thermal conductivity, the high boiling temperature allowing the reactors to operate at ambient pressure, and low neutron absorption cross section which is required to achieve a high neutron flux. But, sodium is violently reactive with water or oxygen like the other alkali metal. So Very strict requirements are demanded to design and fabricate of sodium experimental facilities. Furthermore, performance testing in high temperature sodium environments is more expensive and time consuming and need an extra precautions because operating and maintaining of sodium experimental facilities are very difficult. The present paper describes performance test results of mechanical sodium pump with water which has been performed with some design changes using water test facility in SAM JIN Industrial Co. To compare the hydraulic characteristic of model pump with water and sodium, the performance test of model pump were performed using vender's experimental facility for mechanical sodium pump. To accommodate non-uniform thermal expansion and to secure the operability and the safety, the gap size of some parts of original model pump was modified. Performance tests of modified mechanical sodium pump with water were successfully performed. Water is therefore often selected as a surrogate test fluid because it is not only cheap, easily available and easy to handle but also its important hydraulic properties (density and kinematic viscosity) are very similar to that of the sodium. Normal practice to thoroughly test a design or component before applied or installed in reactor is important to ensure the safety and operability in the sodium-cooled fast reactor (SFR). So, in order to estimate the hydraulic behavior of the PHTS pump of DSFR (600 MWe Demonstraion SFR), the performance tests of the model pump such as performance

Effects of solvation shells and cluster size on the reaction of aluminum clusters with water

Directory of Open Access Journals (Sweden)

Weiwei Mou

2011-12-01

Full Text Available Reaction of aluminum clusters, Aln (n = 16, 17 and 18, with liquid water is investigated using quantum molecular dynamics simulations, which show rapid production of hydrogen molecules assisted by proton transfer along a chain of hydrogen bonds (H-bonds between water molecules, i.e. Grotthuss mechanism. The simulation results provide answers to two unsolved questions: (1 What is the role of a solvation shell formed by non-reacting H-bonds surrounding the H-bond chain; and (2 whether the high size-selectivity observed in gas-phase Aln-water reaction persists in liquid phase? First, the solvation shell is found to play a crucial role in facilitating proton transfer and hence H2 production. Namely, it greatly modifies the energy barrier, generally to much lower values (< 0.1 eV. Second, we find that H2 production by Aln in liquid water does not depend strongly on the cluster size, in contrast to the existence of magic numbers in gas-phase reaction. This paper elucidates atomistic mechanisms underlying these observations.

Numerical study of the underexpanded nitrogen jets submerged into liquid sodium in the frame of Sodium-cooled Fast Reactor (SFRs)

International Nuclear Information System (INIS)

Chen, F.; Allou, A.; Parisse, J.D.

2017-01-01

The study of the consequences of a gas leakage in the secondary/ tertiary heat exchangers is one of the essential points in the safety analysis of Sodium-cooled Fast nuclear Reactors (SFRs). This work is in the frame of the technology of the Compact plates Sodium-Gas heat Exchangers (ECSG) which is an alternative to conventional steam Rankine cycles. The overpressure of the tertiary nitrogen loop causes the formation of underexpanded gas jets submerged in the liquid sodium. In order to establish a safety evaluation, it would be an asset to be able to estimate the leakage. The gas leak detection by the acoustic method based on the bubbles field has been proposed. It requires then a delicate knowledge of the bubble field. This work contributes to development a numerical tool and its validation to model the transport and the production of bubbles in the downstream of underexpanded gas jets. The code CANOP modeling bi-phasic compressible flow is investigated under the actual condition of the underexpanded nitrogen jets submerged in the liquid sodium in an ECSG channel. Expensive computational cost is limited by using an Adaptive Mesh Refinement. (author)

Small liquid sodium leaks

International Nuclear Information System (INIS)

Dufresne, J.; Rochedereux, Y.; Antonakas, D.; Casselman, C.; Malet, J.C.

1986-05-01

Usually, pessimistic considerations inassessing the safety of secondary sodium loops in LMFBR reactor lead to assume guillotine rupture releasing a large amount of sodium estimate the consequences of large sodium fires. In order to reduce these consequences, one has to detect the smallest leak as soon as possible and to evaluate the future of an initial small leak. Analysis of the relationship between crack size and sodium outflow rate; Analysis of a sodium pipe with a small open crack

Sodium voiding analysis in Kalimer

International Nuclear Information System (INIS)

Chang, Won-Pyo; Jeong, Kwan-Seong; Hahn, Dohee

2001-01-01

A sodium boiling model has been developed for calculations of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. The sodium boiling in liquid metal reactors using sodium as coolant should be modeled because of phenomenon difference observed from that in light water reactor systems. The developed model is a multiple -bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbles that fill the whole cross section of the coolant channel except for liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble. The present study is focused on not only demonstration of the sodium voiding behavior predicted by the developed model, but also confirmation on qualitative acceptance for the model. In results, the model catches important phenomena for sodium boiling, while further effort should be made for the complete analysis. (author)

Study on flow characteristics of chemically reacting liquid jet

International Nuclear Information System (INIS)

Hong Seon Dae; Okamoto, Koji; Takata, Takashi; Yamaguchi, Akira

2004-07-01

Tube rupture accidents in steam generators of sodium-cooled fast breeder reactors are important for safety because the rupture may propagates to neighboring tubes due to sodium-water reaction. In order to clarify the thermal-hydraulic phenomena in the accidents, the flow pattern and the interface in multi-phase flow must be investigated. The JNC cooperative research scheme on the nuclear fuel cycle with the University of Tokyo has been carried to develop a simultaneous measurement system of concentration and velocity profiles and to evaluate influence of chemical reaction on mixing phenomena. In the experiments, aqueous liquor of acetic acid and ammonium hydroxide are selected as a simulant fluid instead of liquid sodium and water vapor. The following conclusions are obtained in this research. Laser Induced Fluorescence (LIF) technique was adopted to measure reacting zone and pH distribution in chemically reacting liquid round free jet. As a result, it was found that the chemical reaction, which took place at the interface between the jet and outer flow, suppressed the mixing phenomenon (in 2001 research). Dynamic Particle Image Velocimetry (PIV) method was developed to measure instantaneous velocity profile with high temporal resolution. In the Dynamic PIV, a high-speed video camera coupled with a high-speed laser pulse generator was implemented. A time-line trend of interfacial area in the free jet was investigated with the Dynamic PIV. This technique was also applied to a complicated geometry (in 2002 research). A new algorithms for image analysis was developed to evaluated the Dynamic PIV data in detail. The characteristics of the mixing phenomenon with reacting jet such as the turbulent kinetic energy and the Reynolds stress were estimated in a spatial and temporal spectrum (in 2003 research). (author)

Ionic Liquids: An Environmentally Friendly Media for Nucleophilic Substitution Reactions

International Nuclear Information System (INIS)

Jorapur, Yogesh R.; Chi, Dae Yoon

2006-01-01

Ionic liquids are alternative reaction media of increasing interest and are regarded as an eco-friendly alternatives, of potential use in place of the volatile organic solvents typically used in current chemical processing methods. They are emerging as the smart and excellent solvents, which are made of positive and negative ions that they are liquids near room temperature. The nucleophilic substitution reaction is one of the important method for inserting functional groups into a carbon skeleton. Many nucleophilic substitution reactions have been found with enhanced reactivity and selectivity in ionic liquid. In this review, some recent interesting results of nucleophilic substitution reactions such as hydroxylations, ether cleavages, carbon-X (X = carbon, oxygen, nitrogen, fluorine) bond forming reactions, and ring opening of epoxides in ionic liquids are discussed

Contribution to the liquid-vapour equilibrium of potassium and sodium mixtures

International Nuclear Information System (INIS)

Schreinlechner, I.; Schwarz, N.

1975-10-01

In this paper the phase diagram of the binary system potassium-sodium in the liquid-vapour range was calculated for different pressures and temperatures, assuming the two metals acting as ideal solution. The assumption was verified by experimental results. It is thus possible to calculate the separation factor for the rectification of potassium and to estimate the content of sodium in the vapour phase during experiments with vapourized potassium from the data of the vapour pressures of the pure metals. (author)

Prediction of liquid-liquid equilibria for polyethylene glycol based aqueous two-phase system by ASOG and UNIFAC method

Directory of Open Access Journals (Sweden)

M. Perumalsamy

2009-03-01

Full Text Available Liquid-Liquid equilibrium data were obtained for the polyethylene glycol2000(PEG2000-sodium citrate-water system at 298.15, 308.15 and 318.15 K. The effect of temperature on binodal and tie line data was studied and published in a previous article (Murugesan and Perumalsamy, 2005. The interaction parameters of ASOG and UNIFAC models were estimated using the LLE data of PEG2000-sodium citrate-water system and are used to predict the LLE data for PEG6000-sodium citrate-water system at 298.15, 308.15 and 318.15 K (literature data. The predicted LLE data by both ASOG and UNIFAC models showed good agreement with the experimental and literature data.

Liquid sodium testing of in-house phased array EMAT transducer for L-wave applications

Energy Technology Data Exchange (ETDEWEB)

Le Bourdais, F.; Le Polles, T. [Non Destructive Testing Department at the French Atomic Energy Commission (CEA), Saclay, 91191 Gif sur Yvette CEDEX, (France); Baque, F. [Department of Sodium Technology at the French Atomic Energy Commission (CEA), Cadarache, 13108 St Paul lez Durance CEDEX, (France)

2015-07-01

This paper describes the development of an in-house phased array EMAT transducer for longitudinal wave inspection in liquid sodium. The work presented herein is part of an undergoing project aimed at improving in-service inspection techniques for the ASTRID reactor project. The design process of the phased array EMAT probe is briefly explained and followed by a review of experimental test results. We first present test results obtained in the laboratory while the last part of the paper describes the liquid sodium testing and the produced ultrasound images. (authors)

Liquid sodium testing of in-house phased array EMAT transducer for L-wave applications

International Nuclear Information System (INIS)

Le Bourdais, F.; Le Polles, T.; Baque, F.

2015-01-01

This paper describes the development of an in-house phased array EMAT transducer for longitudinal wave inspection in liquid sodium. The work presented herein is part of an undergoing project aimed at improving in-service inspection techniques for the ASTRID reactor project. The design process of the phased array EMAT probe is briefly explained and followed by a review of experimental test results. We first present test results obtained in the laboratory while the last part of the paper describes the liquid sodium testing and the produced ultrasound images. (authors)

Exothermic potential of sodium nitrate salt cake

International Nuclear Information System (INIS)

Beitel, G.A.

1977-06-01

High-Level radioactive liquid waste is being reduced to a liquid slurry by an evaporation and crystallization process and stored in the existing single-shell tanks. Continuous pumping of the waste storage tank will reduce the present 30 to 50% moisture to the minimum possible. The reduced waste is a relatively immobile salt cake consisting predominantly of sodium nitrate (NaNO 3 ) with lesser amounts of sodium nitrite (NaNO 2 ), sodium metaaluminate (NaAlO 2 ), and sodium hydroxide (NaOH). Trace amounts of fission products, transuranics, and a broad spectrum of organic materials in small but unknown amounts are also present. A program was initiated in 1973 to determine whether or not conditions exist which could lead to an exothermic reaction in the salt cake. Results of the latest series of tests conducted to determine the effects of mass and pressure are summarized. Hanford salt cake, as stored, cannot support combustion, and does not ignite when covered with a burning volatile hydrocarbon

First-principles quantum mechanical investigations: Catalytic reactions of furfural on Pd(111) and at the water/Pd(111) interface

Science.gov (United States)

Xue, Wenhua

Bio-oils have drawn more and more attention from scientists as a promising new clean, cheap energy source. One of the most interesting relevant issues is the effect of catalysts on the catalytic reactions that are used for producing bio-oils. Furfural, as a very important intermediate during these reactions, has attracted significant studies. However, the effect of catalysts, including particularly the liquid/solid interface formed by a metal catalyst and liquid water, in the catalytic reactions involving furfural still remains elusive. In this research, we performed ab initio molecular dynamics simulations and first-principles density-functional theory calculations to investigate the atomic-scale mechanisms of catalytic hydrogenation of furfural on the palladium surface and at the liquid/state interface formed by the palladium surface and liquid water. We studied all the possible mechanisms that lead to formation of furfuryl alcohol (FOL), formation of tetrahydrofurfural (THFAL), and formation of tetrahydrofurfurfuryl alcohol (THFOL). We found that liquid water plays a significant role in the hydrogenation reactions. During the reaction in the presence of water and the palladium catalyst, in particular, water directly participates in the hydrogenation of the aldehyde group of furfural and facilitates the formation of FOL by reducing the activation energy. Our calculations show that water provides hydrogen for the hydrogenation of the aldehyde group, and at the same time, a pre-existing hydrogen atom, which is resulted from dissociation of molecular hydrogen (experimentally, molecular hydrogen is always supplied for hydrogenation) on the palladium surface, is bonded to water, making the water molecule intact in structure. In the absence of water, on the other hand, formation of FOL and THFAL on the palladium surface involves almost the same energy barriers, suggesting a comparable selectivity. Overall, as water reduces the activation energy for the formation of FOL

Experimental studies of water hammer in propellant feed system of reaction control system

Directory of Open Access Journals (Sweden)

Avanish Kumar

2018-03-01

Full Text Available Water hammer pressure transient produces large dynamic forces which can damage the pipes and other assemblies in the feed line of a reaction control system (RCS. It has led to the failure of pressure transducers monitoring the manifold pressure in the feed line of RCS. Therefore, water hammer studies have been carried out to understand its effect in feed line. Feedline system has been simplified to develop a mathematical model and experiments have been carried out at extensive levels. The mathematical model was developed considering pipe of uniform c/s and moving liquid-gas interface. The experimental studies have been done using water as working medium instead of actual propellant. The studies showed that rate of pressurization have a very critical role on the water hammer amplitude. Sensitivity studies have been also carried out to study the effect of density, friction and initial liquid column length on water hammer amplitude. Keywords: Water hammer, Reaction control system (RCS, Propellant feed system, Experimental study, Testing

The corrosion of steels by hot sodium melts

International Nuclear Information System (INIS)

Currie, R.

1996-01-01

Considerable research has been performed by AEA Technology on the corrosion of steels by hot sodium melts containing sodium hydroxide and sodium oxide. This research has principally been in support of understanding the effects of sodium-water reactions on the internals of fast reactor steam generators. The results however have relevance to sodium fires. It has been determined that the rate of corrosion of steels by melts of pure NaOH can be significantly increased by the addition of Na 2 O. In the case of a sodium-water reaction jet created by a leak of steam into sodium, the composition of the jet varies from 100% sodium through to 100% steam, with a full range of concentrations of NaOH and Na 2 O, depending on axial and radial position. The temperature in the jet also varies with position, ranging from bulk sodium temperature on one boundary to expanded steam temperature on the other boundary, with internal temperatures ranging up to 1300 deg. C, depending on the local pre-reaction mole ratio of steam to sodium. In the case of sodium-water reaction jets, it has been possible to develop a model which predicts the composition of the reaction jet and then, using the data generated on the corrosivity of sodium melts, predict the rate of corrosion of a steel target in the path of the jet. In the case of a spray sodium fire, the sodium will initially contain a concentration of NaOH and the combustion process will generate Na 2 O. If there is sufficient humidity, conversion of some of the Na 2 O to NaOH will also occur. There is therefore the potential for aggressive mixtures of NaOH and Na 2 O to exist on the surface of the sodium droplets. It is therefore possible that the rate of corrosion of steels in the path of the spray may be higher than expected on the basis of assuming that only Na and Na 2 O were present. In the case of a pool sodium fire, potentially corrosive mixtures of NaOH and Na 2 O may be formed at some locations on the surface. This could lead to

Corrosion and cleaning aspects of sodium side crevices in components of LMFBR's

International Nuclear Information System (INIS)

Chirer, E.G.

1978-01-01

Although the presence of crevices is excluded in critical areas of sodium components by design, their occurrence in other areas cannot be eliminated completely. During the lifetime of a component high concentrations of sodium compounds, such as caustics, may be formed in crevices. These compounds can remain within the crevices for some time. In this respect the following situations are recognized: - Reaction products from initial contaminants such as oxide scales. The component with crevices containing oxide scale either from the manufacturing process or insufficient cleaning after water pressure testing is exposed to sodium during actual operation. - Reaction products formed during or after cleaning. Sodium in the crevices of a drained component reacts with water vapour or water during cleaning or during subsequent storing or handling under non-perfect conditions. Before refilling with sodium the component is heated to preheat temperature. Same situation as above, however the component is exposed to sodium at operating temperature. These cycles can be repeated several times. - Products from a small sodium-water reaction. Caustic products from a small sodium-water reaction may be present in crevices or dead ends of a component which is exposed to high temperature during sodium operation or during vacuum distillation. The aims of the investigations are: determination of the corrosive aspects of high concentration of caustic reaction products of sodium in crevices on the structural materials of the component; comparison of the effectiveness of different cleaning procedures in respect to removal of sodium from crevices, e.g. water, steam, alcohol cleaning, vacuum distillation. Concerning the first item, in particular the possibility of the occurrence of intercrystalline corrosion and stress corrosion cracking is investigated. Materials investigated are the Cr-Mo steels 2 1/4Cr1Mo stabilized with Nb, 9Cr1Mo, 12Cr1Mo and the austenitic stainless steal AISI 304. The

Supported ionic liquids: versatile reaction and separation media

DEFF Research Database (Denmark)

Riisager, Anders; Fehrmann, Rasmus; Haumann, Marco

2006-01-01

The latest developments in supported ionic liquid phase (SILP) systems for catalysis and separation technology are surveyed. The SILP concept combines the advantages of homogeneous catalysis with heterogeneous process technology, and a variety of reactions have been studied where supported ionic ...... liquid catalysts proved to be more active and selective than common systems. In separation applications the use of supported ionic liquids can facilitate selective transport of substrates across membranes.......The latest developments in supported ionic liquid phase (SILP) systems for catalysis and separation technology are surveyed. The SILP concept combines the advantages of homogeneous catalysis with heterogeneous process technology, and a variety of reactions have been studied where supported ionic...

Fragmentation of molten copper drop caused by entrapment of liquid sodium

International Nuclear Information System (INIS)

Abe, N.; Sugiyama, K.; Nishimura, S.; Kinoshita, I.

2001-01-01

In core meltdown accidents, it is possible to occur thermal interactions between molten fuel and coolant. Analysis of the steam explosion, which is one of the most severe phenomena in such thermal interactions, is important for the safety evaluation. The steam explosion is a phenomenon that intensive pressure waves are caused by the explosive thermal interaction between high and low temperature liquids, and is considered to be one of the phenomena that can cause a serious failure of the nuclear reactor structures. In a large-scale steam explosion, the fragmentation of hot molten material causes a rapid increase of heat transfer area, and it is achieved to transmit instantaneously a large amount of heat to coolant. Two ideas are chiefly considered as the mechanism of the fragmentation. The one is the hypothesis that hydrodynamic effect causes fragmentation of hot liquid. According to this hypothesis, the high temperature drops flake off from the surface. The other is that fragmentation is caused by the interface instability accompanied by collapse of the steam bubble formed around a hot liquid. In this research, the possibility of the internal fragmentation caused by the coolant jet is focused in. Experiments were conducted on the condition that the surface of melt drops solidify at the moment drops contact the coolant. The possibility of the fragmentation of hot liquid from its surface was eliminated in this condition. To satisfy this condition, molten copper was chosen as hot liquid, and liquid sodium was used as coolant to verify the effect of the driving force of the sodium jet. (author)

Recent developments in biocatalysis in multiphasic ionic liquid reaction systems.

Science.gov (United States)

Meyer, Lars-Erik; von Langermann, Jan; Kragl, Udo

2018-06-01

Ionic liquids are well known and frequently used 'designer solvents' for biocatalytic reactions. This review highlights recent achievements in the field of multiphasic ionic liquid-based reaction concepts. It covers classical biphasic systems including supported ionic liquid phases, thermo-regulated multi-component solvent systems (TMS) and polymerized ionic liquids. These powerful concepts combine unique reaction conditions with a high potential for future applications on a laboratory and industrial scale. The presence of a multiphasic system simplifies downstream processing due to the distribution of the catalyst and reactants in different phases.

Sodium immersible high temperature microphone design description

International Nuclear Information System (INIS)

Gavin, A.P.; Anderson, T.T.; Janicek, J.J.

1975-02-01

Argonne National Laboratory has developed a rugged high-temperature (HT) microphone for use as a sodium-immersed acoustic monitor in Liquid Metal Fast Breeder Reactors (LMFBRs). Microphones of this design have been extensively tested in room temperature water, in air up to 1200 0 F, and in sodium up to 1200 0 F. They have been successfully installed and employed as acoustic monitors in several operating liquid metal systems. The design, construction sequence, calibration, and testing of these microphones are described. 6 references. (U.S.)

Sodium-water clusters and their role in radiation chemistry

International Nuclear Information System (INIS)

Dhar, S.; Kestner, N.R.

1988-01-01

Studies of sodium-water clusters are presented which could serve as models for the recently suggested intermediate species in the radiation chemistry of water. The ionization potentials and the lower excited states of sodium with n-water molecules are calculated by ab initio quantum chemistry methods. The ionization potential calculated at the SCF level for the water monomer is 4.10 eV, which becomes 4.34 at the MP2 correlation level. The experimental value is 4.379 ± 0.002 eV. Structural data is presented for the lower members of the sodium with n-water clusters. In addition the Hartree-Fock calculations indicate that there should be some strong charge transfer to solvent transitions at higher energies. (author)

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

Laboratory-scale sodium-carbonate aggregate concrete interactions

International Nuclear Information System (INIS)

Westrich, H.R.; Stockman, H.W.; Suo-Anttila, A.

1983-09-01

A series of laboratory-scale experiments was made at 600 0 C to identify the important heat-producing chemical reactions between sodium and carbonate aggregate concretes. Reactions between sodium and carbonate aggregate were found to be responsible for the bulk of heat production in sodium-concrete tests. Exothermic reactions were initiated at 580+-30 0 C for limestone and dolostone aggregates as well as for hydrated limestone concrete, and at 540+-10 0 C for dehydrated limestone concrete, but were ill-defined for dolostone concrete. Major reaction products included CaO, MgO, Na 2 CO 3 , Na 2 O, NaOH, and elemental carbon. Sodium hydroxide, which forms when water is released from cement phases, causes slow erosion of the concrete with little heat production. The time-temperature profiles of these experiments have been modeled with a simplified version of the SLAM computer code, which has allowed derivation of chemical reaction rate coefficients

Application of a gradient diffusion and dissipation time scale ratio model for prediction of mean and fluctuating temperature fields in liquid sodium downstream of a multi-bore jet block

International Nuclear Information System (INIS)

Bremhorst, K.; Listijono, J.B.H.; Krebs, L.; Mueller, U.

1989-01-01

A previously developed diffusivity based based model, for the prediction of mean and fluctuating temperatures in water flow downstream of a multi-bore jet block in which one jet is heated, is applied to a flow of sodium in apparatus of similar geometry. Some measurements not readily possible in sodium or water flows for this geometry are made using air in order to verify assumptions used in the model. The earlier derived mathematical model is modified to remove assumptions relating to turbulence. Reynolds number and turbulence Peclet number in the relationship between velocity and temperature microscales. Spalding's model, relating fluctuating velocity and temperature dissipation rates, is tested. A significant effect on this relationship due to the low Prandtl number of liquid sodium is identified. Measurements performed behind a multi-bore jet block with air as the working fluid have verified the non-isotropic nature of the large-scale flow. Results clearly show that measurements performed in water can be transferred to liquid sodium provided that molecular diffusion is included in the mean temperature equation, allowance is made for the Prandtl number effect on the dissipation time scale ratio and the coefficient of gradient diffusion of mean square temperature fluctuations is assumed equal to the eddy diffusivity of heat. (author)

Experimental investigations of different steel resistances in the sodium-steam reaction zone

International Nuclear Information System (INIS)

Mazanov, A.S.; Kulpin, B.V.; Petukhov, V.P.; Ledovskikh, N.M.

1975-01-01

The results are stated on the experimental investigations of IX2M, IXI8HIOT, OXI2H2M and Sanikro-31 steel resistance in the sodium-steam reaction zone. A target in the form of a pipe was used in the experiments within which the excessive pressure of 95 atm. was produced with an inert gas. Steam was supplied through the nozzle to the sodium tank in which there was a movable target unit. The dependence of time-to failure and failure rate on the distance of the nozzle and target wall thickness was estimated for these steels. It was shown that the resistance of Sanikro-31 in the sodium steam reaction zone was 2.5, 3.5 and 6 times that of IXI8HIOT, OXI2H2M, IX2M steels, respectively. The failure curves were obtained on thin targets for two steels (author)

Pulse radiolysis study on solvated electrons in ionic liquid with controlling water content

International Nuclear Information System (INIS)

Yang Jinfeng; Kondoh, T.; Yoshida, Y.; Nagaishi, R.

2006-01-01

Room-temperature ionic liquids, which are nonvolatile and nonflammable, have been proposed as 'green solvents' for new applications in chemical synthesis, separation chemistry, electrochemistry and other areas. In the separation chemistry, the hydrophobic ionic liquids have been practically expected to be alternative to traditional organic solvents for solvent extraction of 4f and 5f elements from the viewpoints of the immiscibility in water, especially in the spent nuclear fuel reprocessing. However, the chemical reaction or kinetics studies are important to apply the ionic liquids for various processes. To understand the effects of ionic liquids on chemical reactions, pulse radiolysis studies of ionic liquid have been carried out on nanosecond scale by using a 27 MeV electron beam and an analyzing light source of xenon lamp. In the experiment, a hydrophobic ionic liquid of diethylmethyl(2-methoxy)ammonium-bis(trifluoromethanesulfonyl)imide (DEMMA-TFSI) salt was used. The ionic liquid of DEMMA-TFSI was prepared by reacting equimolar amounts of diethylmethyl(2-methoxy)ammonium chloride (C 10 H 20 F 6 N 2 O 5 S 2 Cl, >98%, Nisshinbo) with lithium bis(trifluoromethanesulfonyl)imide (LiN(SO 2 CF 3 ) 2 , SynQuest Labs., Inc.) in aqueous solutions at room temperature. The ionic liquid was separated from the aqueous phase, purified by repeated extractions with water to LiCl and excess reagent, and finally dried at 110 degree C under vacuum. The transient absorptions of the ionic liquid were measured at wavelengths from 350 to 1400 nm, in which two photodiodes of silicon ( 1000 nm) were used. The spectrum of solvated electrons in the ionic liquid of DEMMA-TFSI was obtained with an absorption peak of 1060 nm and a wide bandwidth of about 600 nm (FWHM). The decay constant of the solvated electrons in the ionic liquid was 1.54 x 10 7 s -1 , which is independent on the wavelength. The absorption peak of the spectrum was blue-shifted from 1060 to 780 nm with increasing water

The Dounreay PFR Liquid-Metal Disposal Project

International Nuclear Information System (INIS)

Sherwood, D.V.; Comline, A.; Small, J.; Blyth, J.

2005-01-01

The UKAEA Prototype Fast Reactor at Dounreay had a liquid sodium-cooled core. Following its shutdown in 1994, the liquid metal is being removed from the reactor and other vessels by means of specialized equipment and reacted with an aqueous solution of sodium hydroxide in a special vessel. The reaction products are neutralized with hydrochloric acid to produce a saline solution.The reactor sodium delivery and processing equipment is all of novel design. As sodium has been withdrawn from the vessel, it has been necessary to switch off the primary sodium pumps (used to heat the sodium), and the reactor is now kept at temperature by a purpose-designed electric heater and a NaK loop heater.A primary sodium extract pump has currently removed [approximately]450 tonnes of primary sodium from the reactor. As the level falls special equipment will be used to punch a hole in the primary circuit pipe work and to drill the strongback to allow trapped sodium to drain for extraction

Measurement and Estimation of Effective Thermal Conductivity for Sodium based Nanofluid using 3-Omega Method

Energy Technology Data Exchange (ETDEWEB)

Oh, Sun Ryung; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Kim, Moo Hwan [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The sodium-cooled fast reactor (SFR) is one of generation IV type reactors and has been extensively researched since 1950s. A strong advantage of the SFR is its liquid sodium coolant which is well-known for its superior thermal properties. However, in terms of possible pipe leakage or rupture, a liquid sodium coolant possesses a critical issue due to its high chemical reactivity which leads to fire or explosion. Due to its safety concerns, dispersion of nanoparticles in liquid sodium has been proposed to reduce the chemical reactivity of sodium. In case of sodium based titanium nanofluid (NaTiNF), the chemical reactivity suppression effect when interacting with water has been proved both experimentally and theoretically [1,2]. Suppression of chemical reactivity is critical without much loss of high heat transfer characteristic of sodium. As there is no research conducted for applying 3-omega sensor in liquid metal as well as high temperature liquid, the sensor development is performed for using in NaTiNF as well as effective thermal conductivity model validation. Based on the acquired effective thermal conductivity of NaTiNF, existing effective thermal conductivity models are evaluated. Thermal conductivity measurement is performed for liquid sodium based titanium nanofluid (NaTiNF) through 3-Omega method. The experiment is conducted at three temperature points of 120, 150, and 180 .deg. C for both pure liquid sodium and NaTiNF. By using 3- omega sensor, thermal conductivity measurement of liquid metal can be more conveniently conducted in labscale. Also, its possibility to measure the thermal conductivity of high temperature liquid metal with metallic nanoparticles being dispersed is shown. Unlike other water or oil-based nanofluids, NaTiNF exhibits reduction of thermal conductivity compare with liquid sodium. Various nanofluid models are plotted, and it is concluded that the MSBM which considers interfacial resistance and Brownian motion can be used in predicting

Measurement and Estimation of Effective Thermal Conductivity for Sodium based Nanofluid using 3-Omega Method

International Nuclear Information System (INIS)

Oh, Sun Ryung; Park, Hyun Sun; Kim, Moo Hwan

2016-01-01

The sodium-cooled fast reactor (SFR) is one of generation IV type reactors and has been extensively researched since 1950s. A strong advantage of the SFR is its liquid sodium coolant which is well-known for its superior thermal properties. However, in terms of possible pipe leakage or rupture, a liquid sodium coolant possesses a critical issue due to its high chemical reactivity which leads to fire or explosion. Due to its safety concerns, dispersion of nanoparticles in liquid sodium has been proposed to reduce the chemical reactivity of sodium. In case of sodium based titanium nanofluid (NaTiNF), the chemical reactivity suppression effect when interacting with water has been proved both experimentally and theoretically [1,2]. Suppression of chemical reactivity is critical without much loss of high heat transfer characteristic of sodium. As there is no research conducted for applying 3-omega sensor in liquid metal as well as high temperature liquid, the sensor development is performed for using in NaTiNF as well as effective thermal conductivity model validation. Based on the acquired effective thermal conductivity of NaTiNF, existing effective thermal conductivity models are evaluated. Thermal conductivity measurement is performed for liquid sodium based titanium nanofluid (NaTiNF) through 3-Omega method. The experiment is conducted at three temperature points of 120, 150, and 180 .deg. C for both pure liquid sodium and NaTiNF. By using 3- omega sensor, thermal conductivity measurement of liquid metal can be more conveniently conducted in labscale. Also, its possibility to measure the thermal conductivity of high temperature liquid metal with metallic nanoparticles being dispersed is shown. Unlike other water or oil-based nanofluids, NaTiNF exhibits reduction of thermal conductivity compare with liquid sodium. Various nanofluid models are plotted, and it is concluded that the MSBM which considers interfacial resistance and Brownian motion can be used in predicting

Method of removing bulk sodium from metallic surfaces

International Nuclear Information System (INIS)

Maffei, H.P.; Borisch, R.R.

1975-01-01

A process of removing sodium from an article, particularly one made of stainless steel, by treating it with a mixture of water vapor and a gas which is inert to sodium is described. By selecting combinations of temperature and water vapor-to-gas ratio, the reaction temperature is controlled to prevent damage to the articles

Metastable liquid-liquid transition in a molecular model of water

Science.gov (United States)

Palmer, Jeremy C.; Martelli, Fausto; Liu, Yang; Car, Roberto; Panagiotopoulos, Athanassios Z.; Debenedetti, Pablo G.

2014-06-01

Liquid water's isothermal compressibility and isobaric heat capacity, and the magnitude of its thermal expansion coefficient, increase sharply on cooling below the equilibrium freezing point. Many experimental, theoretical and computational studies have sought to understand the molecular origin and implications of this anomalous behaviour. Of the different theoretical scenarios put forward, one posits the existence of a first-order phase transition that involves two forms of liquid water and terminates at a critical point located at deeply supercooled conditions. Some experimental evidence is consistent with this hypothesis, but no definitive proof of a liquid-liquid transition in water has been obtained to date: rapid ice crystallization has so far prevented decisive measurements on deeply supercooled water, although this challenge has been overcome recently. Computer simulations are therefore crucial for exploring water's structure and behaviour in this regime, and have shown that some water models exhibit liquid-liquid transitions and others do not. However, recent work has argued that the liquid-liquid transition has been mistakenly interpreted, and is in fact a liquid-crystal transition in all atomistic models of water. Here we show, by studying the liquid-liquid transition in the ST2 model of water with the use of six advanced sampling methods to compute the free-energy surface, that two metastable liquid phases and a stable crystal phase exist at the same deeply supercooled thermodynamic condition, and that the transition between the two liquids satisfies the thermodynamic criteria of a first-order transition. We follow the rearrangement of water's coordination shell and topological ring structure along a thermodynamically reversible path from the low-density liquid to cubic ice. We also show that the system fluctuates freely between the two liquid phases rather than crystallizing. These findings provide unambiguous evidence for a liquid-liquid transition in

Deuterium exchange reaction between hydrogen and water in a trickle-bed column packed with novel catalysts

International Nuclear Information System (INIS)

Ahn, D. H.; Baek, S. W.; Lee, H. S.; Kim, K. R.; Kang, H. S.; Lee, S. H.; Jeong, H. S.

1998-01-01

The activity of a novel catalyst (Pt/SDBC) for deuterium exchange reaction between water and hydrogen streams in a trickle bed was measured. The performance of the catalyst was compared with a commercial catalyst with same metal content. The catalytic activity for the bed of wet-proofed catalyst diluted with hydrophillic packing material also measured. The Pt/SDBC catalyst shows higher activity in the liquid phase reaction than the commercial catalyst as measured in the vapor phase reaction. The performance for 50% dilution of the Pt/SDBC catalyst bed with hydrophillic packing material is better than that of the 100% bed due to more liquid holdup and better water distribution

Influence of sodium dodecyl sulfate on the reaction between Nile Blue A and hydrogen peroxide

Directory of Open Access Journals (Sweden)

IVANA A. JANKOVIC

1999-05-01

Full Text Available The influence of the anionic surfactant sodium dodecyl sulfate on the rate of the reaction between the cationic form of Nile Blue A and hydrogen peroxide was investigated in the pH range from 5 to 8.5. A retardation of the oxidation of Nile Blue A with hydrogen peroxide of three orders of magnitude was observed at pH 8.5 in the presence of anionic micelles compared to the kinetic data in water. The retardation effect was less pronounced at lower pH values. These effects were explained by the electrostatic interaction of the species involved in the reaction with the negatively charged micellar surface and their effective separation in the vicinity of the micellar surface.

The liquid metal embrittlement of iron and ferritic steels in sodium

International Nuclear Information System (INIS)

Hilditch, J.P.; Hurley, J.R.; Tice, D.R.; Skeldon, P.

1995-01-01

The liquid metal embrittlement of iron and A508 III, 21/4Cr-1Mo and 15Mo3 steels in sodium at 200-400 o C has been studied, using dynamic straining at 10 -6 s -1 , in order to investigate the roles of microstructure and composition. The steels comprised bainitic, martensitic, tempered martensitic and ferritic/pearlitic microstructures. All materials were embrittled by sodium, the embrittlement being associated generally with quasicleavage on fracture surfaces. Intergranular cracking was also found with martensitic and ferritic/pearlitic microstructures. The susceptibility to embrittlement was greater in higher strength materials and at higher temperatures. The embrittlement was similar to that encountered previously in 9Cr steel, which depends upon the presence of non-metallic impurities in the sodium. (author)

Operational techniques employed for the liquid sodium source term control loops

International Nuclear Information System (INIS)

Chulos, L.E.

1976-01-01

Four Source Term Control Loops (STCLs) have been designed, constructed, and placed into operation at the Hanford Engineering Development Laboratory (HEDL) as part of the Radioactivity Control Technology program. The data obtained are used to determine the corrosion and deposition of LMFBR materials, including corrosion product radionuclides, in a non-isothermal flowing sodium system. The paper discusses operation of the STCL Facilities and, in particular, the methods used for controlling the oxygen content of the liquid sodium. These methods include cold trapping techniques, hot trapping, seeding the cold traps with sodium oxide, and precipitating the oxygen in the cold trap in a controlled manner. Operational problems encountered with the STCL Facilities and the techniques for correcting these problems are also discussed

Kinetics and mechanism of the furan peroxide formation in the reaction of furfural with hydrogen peroxide in the presence and absence of sodium molybdate

International Nuclear Information System (INIS)

Grunskaya, E.P.; Badovskaya, L.A.; Kaklyugina, T.Ya.; Poskonin, V.V.

2000-01-01

Kinetics of the initial stage of the reaction of furfural with hydrogen peroxide are studied in the presence of Na 2 MoO 4 in water and without catalytic additions in n-butanol. Organic peroxide having in its disposal Mo(6), which is the only product on the initial stage of the reaction, is formed since the first minutes of oxidation of furfural by hydrogen peroxide with the presence of Na 2 MoO 4 . The mechanisms of conversion of furfural in the Na 2 MoO 4 - H 2 O system and its oxidation by peroxide without sodium molybdate are discussed. Schemes of formation of furfural complexes based on the results of kinetic studies are suggested. Comparison of obtained data demonstrates that presence of the sodium molybdates in the reaction medium trends to change of reaction procedure in the hydrogen peroxide [ru

Speciation of mercury in water samples by dispersive liquid-liquid microextraction combined with high performance liquid chromatography-inductively coupled plasma mass spectrometry

Energy Technology Data Exchange (ETDEWEB)

Jia Xiaoyu; Han Yi; Liu Xinli [State Key Laboratory of Electroanalytical Chemistry, Changchun Institute of Applied Chemistry, Chinese Academy of Science, Changchun 130022 (China); Graduate School of Chinese Academy of Sciences, Beijing 100039 (China); Duan Taicheng, E-mail: tcduan@ciac.jl.cn [State Key Laboratory of Electroanalytical Chemistry, Changchun Institute of Applied Chemistry, Chinese Academy of Science, Changchun 130022 (China); Chen Hangting, E-mail: htchen@ciac.jl.cn [State Key Laboratory of Electroanalytical Chemistry, Changchun Institute of Applied Chemistry, Chinese Academy of Science, Changchun 130022 (China)

2011-01-15

The dispersive liquid-liquid microextraction (DLLME) combined with high performance liquid chromatography-inductively coupled plasma mass spectrometry for the speciation of mercury in water samples was described. Firstly methylmercury (MeHg{sup +}) and mercury (Hg{sup 2+}) were complexed with sodium diethyldithiocarbamate, and then the complexes were extracted into carbon tetrachloride by using DLLME. Under the optimized conditions, the enrichment factors of 138 and 350 for MeHg{sup +} and Hg{sup 2+} were obtained from only 5.00 mL sample solution. The detection limits of the analytes (as Hg) were 0.0076 ng mL{sup -1} for MeHg{sup +} and 0.0014 ng mL{sup -1} for Hg{sup 2+}, respectively. The relative standard deviations for ten replicate measurements of 0.5 ng mL{sup -1} MeHg{sup +} and Hg{sup 2+} were 6.9% and 4.4%, respectively. Standard reference material of seawater (GBW(E)080042) was analyzed to verify the accuracy of the method and the results were in good agreement with the certified values. Finally, the developed method was successfully applied for the speciation of mercury in three environmental water samples.

Development of sodium droplet combustion analysis methodology using direct numerical simulation in 3-dimensional coordinate (COMET)

International Nuclear Information System (INIS)

Okano, Yasushi; Ohira, Hiroaki

1998-08-01

In the early stage of sodium leak event of liquid metal fast breeder reactor, LMFBR, liquid sodium flows out from a piping, and ignition and combustion of liquid sodium droplet might occur under certain environmental condition. Compressible forced air flow, diffusion of chemical species, liquid sodium droplet behavior, chemical reactions and thermodynamic properties should be evaluated with considering physical dependence and numerical connection among them for analyzing combustion of sodium liquid droplet. A direct numerical simulation code was developed for numerical analysis of sodium liquid droplet in forced convection air flow. The numerical code named COMET, 'Sodium Droplet COmbustion Analysis METhodology using Direct Numerical Simulation in 3-Dimensional Coordinate'. The extended MAC method was used to calculate compressible forced air flow. Counter diffusion among chemical species is also calculated. Transport models of mass and energy between droplet and surrounding atmospheric air were developed. Equation-solving methods were used for computing multiphase equilibrium between sodium and air. Thermodynamic properties of chemical species were evaluated using dynamic theory of gases. Combustion of single sphere liquid sodium droplet in forced convection, constant velocity, uniform air flow was numerically simulated using COMET. Change of droplet diameter with time was closely agree with d 2 -law of droplet combustion theory. Spatial distributions of combustion rate and heat generation and formation, decomposition and movement of chemical species were analyzed. Quantitative calculations of heat generation and chemical species formation in spray combustion are enabled for various kinds of environmental condition by simulating liquid sodium droplet combustion using COMET. (author)

Disposition of nonflammable low-level radioactive wastes using supercritical water with ruthenium(IV) oxide catalyst

International Nuclear Information System (INIS)

Sugiyama, Wataru

2013-01-01

This paper presents the distribution behavior of iron, cobalt, cesium, iodine and strontium attached to nonflammable organic materials, in solid, liquid and gas phases during the decomposition of these materials using supercritical water with ruthenium(IV) oxide (RuO 2 ) catalyst. The distributions of these elements under various conditions (initial amounts, with/without precipitation reagent) were determined by using their radioisotopes as simulated low-level radioactive wastes (LLW) in order to ease the detection of trace amounts of elements even in solid and gas phases. Iron and cobalt were found only in the solid phase when iron hydroxide was added as a precipitation reagent before the supercritical water reaction. Cesium, iodine and strontium were found in the liquid phase after the reaction. Therefore, by adding precipitation reagents such as sodium tetraphenylborate, and sodium carbonate (Na 2 CO 3 ) (or sodium hydrogen carbonate (NaHCO 3 )) and silver nitrate (AgNO 3 ) aqueous solutions to each resultant liquid phase containing cesium, strontium and iodine, respectively, these elements can be successfully recovered only in the solid phase. The gases produced during the decomposition of the organic material contain no radioactivity under all conditions in this study. These results indicate that all of the elements investigated in this study (iron, cobalt, cesium, iodine and strontium) can be recovered successfully by this supercritical water process using RuO 2 Consequently, this process is suggested as a predominant candidate for the treatment of nonflammable organic materials in LLW. (author)

Dissolution of manganese and cobalt and their deposition on Type 304 stainless steel in liquid sodium

International Nuclear Information System (INIS)

Yokota, Norikatsu; Shimoyashiki, Shigehiro

1989-01-01

Dissolution of manganese and cobalt and their deposition on Type 304 stainless steel in liquid sodium at 833 K for 3.6 x 10 3 ks were examined using a liquid sodium pot. Manganese was easily dissolved in sodium from the iron-manganese alloy specimen and deposited on the steel to form two kind of deposition particles, α-phase (body-centered cubic) composed of iron and γ-phase (face-centered cubic) composed of iron and manganese, respectively. Cobalt which was less easily dissolved than manganese also deposited on the Type 304 stainless steel, giving an iron-cobalt alloy. These three deposition particles corresponded to the precipitation lines of iron-manganese and iron-cobalt phase diagrams at 833 K, respectively. Therefore, the deposition process of manganese or cobalt in sodium was explained as a precipitation process of iron-manganese or iron-cobalt in the solid region of the binary phase diagram. A sodium chromite (NaCrO 2 ) layer was formed on the steel surface. (author)

Renal function, sodium and water homeostasis in patients with ...

African Journals Online (AJOL)

At baseline there were no differences in inulin clearance, PAH clearance, fractional excretion of sodium and free water excretion. During and after the saline infusion both groups showed a significant increase in sodium excretion with a reduction in water excretion, while the PAH and inulin clearances remained unchanged.

Acoustic emissions of a boiling liquid - an experimental survey in water and extrapolation to SFRs

International Nuclear Information System (INIS)

Vanderhaegen, M.; Paumel, K.; Tourin, A.

2013-06-01

The acoustic detection of sodium boiling is seen as a promising and innovative surveillance technique for sodium-cooled fast reactors (SFRs). It could be especially useful to detect in-core boiling that are the consequence of initiating accidents or whilst the mean subassembly temperature is very close to the nominal value. This latter is a consequence of the fuel assembly design of SFRs. Furthermore, it is a technique that has been proven to be successful in the past to follow the boiling behavior during SFR experiments that were aimed at simulating accidental conditions. However its effectiveness as in-core instrumentation still has to be demonstrated. In that aim, the acoustic emissions of sodium boiling in subassemblies are studied. Experimental studies are however limited to the boiling of common coolants due to the complications that arise when boiling liquid metals. As such, simple water experiments are performed. And although the results of these experiments are not completely representative for sodium boiling due to the incomplete thermo-hydraulic similarities between sodium and water, they can provide an interesting knowledge of the many influences that control the acoustic pressure field. In this article we'll specifically show how the condensation of vapor in subcooled liquid, the principal contribution to the acoustic emissions during boiling and hence the acoustic spectrum, is influenced by a pin-bundle geometry. We study this influence by comparing pool boiling experimental acoustic recordings with those of a simple pin-bundle geometry boiling experiment. The qualitative link, between this relatively simple pin-bundle experiment and the condensation phenomena that take place during sodium boiling inside SFR subassemblies, is used as a basis for this analysis. This simple experimental evidence, together with other theoretical arguments based on a thorough analysis of the sodium material properties, enables us to deduce that simple sodium

Modified ionic liquid cold-induced aggregation dispersive liquid-liquid microextraction followed by atomic absorption spectrometry for trace determination of zinc in water and food samples

International Nuclear Information System (INIS)

Zeeb, M.; Sadeghi, M.

2011-01-01

We report on a new method for the microextraction and determination of zinc (II). The ion is accumulated via ionic-liquid cold-induced aggregation dispersive liquid-liquid microextraction (IL-CIA-DLLME) followed by flame atomic absorption spectrometry (FAAS). The ionic liquid (IL) 1-hexyl-3-methylimidazolium hexafluorophosphate is dispersed into a heated sample solution containing sodium hexafluorophosphate as a common ion source. The solution is then placed in an ice-water bath upon which a cloudy solution forms due to the decrease of the solubility of the IL. Zinc is complexed with 8-hydroxyquinoline and extracted into the IL. The enriched phase is dissolved in a diluting agent and introduced to the FAAS. The method is not influenced by variations in the ionic strength of the sample solution. Factors affecting the performance were evaluated and optimized. At optimum conditions, the limit of detection is 0.18 μg L -1 , and the relative standard deviation is 3.0% (at n=5). The method was validated by recovery experiments and by analyzing a certified reference material and successfully applied to the determination of Zn (II) in water and food samples. (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.

Thermodynamic properties by Equation of state of liquid sodium under pressure

Science.gov (United States)

Li, Huaming; Sun, Yongli; Zhang, Xiaoxiao; Li, Mo

Isothermal bulk modulus, molar volume and speed of sound of molten sodium are calculated through an equation of state of a power law form within good precision as compared with the experimental data. The calculated internal energy data show the minimum along the isothermal lines as the previous result but with slightly larger values. The calculated values of isobaric heat capacity show the unexpected minimum in the isothermal compression. The temperature and pressure derivative of various thermodynamic quantities in liquid Sodium are derived. It is discussed about the contribution from entropy to the temperature and pressure derivative of isothermal bulk modulus. The expressions for acoustical parameter and nonlinearity parameter are obtained based on thermodynamic relations from the equation of state. Both parameters for liquid Sodium are calculated under high pressure along the isothermal lines by using the available thermodynamic data and numeric derivations. By comparison with the results from experimental measurements and quasi-thermodynamic theory, the calculated values are found to be very close at melting point at ambient condition. Furthermore, several other thermodynamic quantities are also presented. Scientific Research Starting Foundation from Taiyuan university of Technology, Shanxi Provincial government (``100-talents program''), China Scholarship Council and National Natural Science Foundation of China (NSFC) under Grant No. 11204200.

Analogy of water as compared to sodium in cavitation

International Nuclear Information System (INIS)