WorldWideScience

Sample records for sodium coolant flow

  1. The sodium coolant

    International Nuclear Information System (INIS)

    Rodriguez, G.

    2004-01-01

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

  2. Sodium as a reactor coolant

    International Nuclear Information System (INIS)

    Cesar, S.B.G.

    1989-01-01

    This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

  3. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  4. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

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

  5. Formation and hydraulic effects of deposits in high temperature sodium coolant systems

    International Nuclear Information System (INIS)

    Yunker, W.

    1976-01-01

    Deposition of sodium impurities in the high temperature (600 0 C), high flow (Reynolds Number approximately equal to 8 x 10 4 ) regions of a sodium coolant circuit is being studied to determine its possible hydraulic effects. Increases in flow impedance (pressure drop/volume flow 2 ) of up to 30 percent have been detected in an annular flow sensor. The apparatus and preliminary results of these tests are presented. Continuing tests are to specifically identify the materials involved and the system conditions under which the formations occur

  6. Calorimetric and reactor coolant system flow uncertainty

    International Nuclear Information System (INIS)

    Bates, L.; McLean, T.

    1991-01-01

    This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

  7. A new inexpensive electrochemical meter for oxygen in sodium coolant

    International Nuclear Information System (INIS)

    Periaswami, G.; Rajan Babu, S.S.; Mathews, C.K.

    1987-01-01

    This report describes the development of an inexpensive oxygen meter for sodium coolant and gives the results of the test experiments. Calcia stabilized zirconia has been found to have necessary domain boundary characteristics at low temperatures for use as oxygen sensor in liquid sodium system. It is possible to obtain acceptable sensor cell resistance at temperatures as low as 230 C if K, K 2 O or Na, Na 2 O is used as reference electrode. The performance of these cells has been tested in bench top sodium loops over long periods. Their performance in terms of cell-out put variation with change in oxygen concentration in sodium has been found to be satisfactory. They also have sufficiently long life times since the kinetics of sodium attack on the electrolyte is slow at low temperatures. (author). 17 refs., 6 figs

  8. Reactor coolant flow measurements at Point Lepreau

    International Nuclear Information System (INIS)

    Brenciaglia, G.; Gurevich, Y.; Liu, G.

    1996-01-01

    The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

  9. Fast reactors bulk sodium coolant disposal NOAH process application

    International Nuclear Information System (INIS)

    Magny, E. de; Berte, M.

    1997-01-01

    Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

  10. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  11. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  12. Heat transfer and fluid flow aspects of fuel--coolant interactions

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

  13. Automatic coolant flow control device for a nuclear reactor assembly

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  14. New method for NPP sodium coolant pipeline austenization

    International Nuclear Information System (INIS)

    Malashonok, V.A.; Rotshtejn, A.V.; Gotshalk, A.L.; Miryushchenko, E.F.

    1980-01-01

    Heat treatment technology is considered for pipelines intended for the NPP cooling systems employing sodium coolant. Various techniques are discussed which are used for protecting the pipeline internal surfaces against oxidation in the process of heat treatment. It is noted that the austenite formation of welded joints of steel 12Kh18N9 and steel Kh16N11M3 at temperatures of 1050 and 1100 deg C releases welding-induced stresses and reduces a possibility of local damages. Evacuation down to 1 mm Hg appears to be the most rational protective technique. The considered procedure of the pipeline heat treatment has been utilized for mounting the equipment of the BN-600 reactor at the Beloyarskaya NPP. The economic gain resulting from the use of the procedure, owing to decrease in argon consumption and reduction of labour input, makes up 150 000 roubles

  15. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

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

  16. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  18. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  19. Numerical experimentation on convective coolant flow in Ghana ...

    African Journals Online (AJOL)

    Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

  20. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  1. Liquid metal coolant flow rate regulation

    International Nuclear Information System (INIS)

    Vitkovskij, I.V.; Glukhikh, V.A.; Kirillov, I.R.; Smirnov, A.M.

    1981-01-01

    Some aspects of fast reactor and experimental bench operation related to liquid metal flow rate regulation are considered. Requirements to the devices for the flow rate regulation are formulated. A new type of these devices namely magnetohydrodynamic (MHD) throttles is described. Structural peculiarities of MHD throttles of different types are described as well. It is noted that the MHD throttles with a screw channel have the best energy mass indices. On the basis of the comparison of the MHD throttles with mechanical valves it is concluded that the MHD throttles described are useful for regulating the flow rates of any working media. Smoothness and accuracy of the flow rate regulation by the throttles are determined by the electric control circuit and may be practically anyone. The total coefficient of hydraulic losses in the throttle channel in the absence of a magnetic field is ten and more times lesser than in completely open mechanical valve. Electromagnetic time constant of the MHD throttles does not exceed several tenths of a second [ru

  2. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  3. A study on fission product retention capability in a sodium coolant system

    International Nuclear Information System (INIS)

    Satoh, K.; Kubo, S.; Hashiguchi, Y.; Itooka, S.; Akatsu, Y.; Miyagi, K.; Wakamatsu, M.; Endo, H.; Tachino, T.

    1992-01-01

    Three kinds of separate model tests have been performed using water and air, focusing on the transport behavior of FP gas bubbles from subassembly outlets into a cover gas region, to study the dominant processes regarding the retention for volatiles ejected with inert gas into sodium after fuel failures. In the case that whole fuel pin failures occurring coherently in a subassembly were assumed, a periodic formation of globules was observed at the subassembly outlet. The globules rapidly broke up into small bubbles of less than 10 mm in mean diameter. The small bubbles at the top region had a tendency to be coalesced during rising through the upper plenum. As the coolant flow rate increased, bubble deformation and breakup were accelerated, but the bubble transport time did not vary remarkably. It is expected that bubbles in sodium would play in a similar way as in the water test, and the importance of the bubble behavior for the retention capability of volatiles has been confirmed. (author)

  4. Quantitative determination of a hydrogen impurity in a sodium coolant by hydride thermal dissociation

    Science.gov (United States)

    Ivanovskiy, M. N.; Pavlova, G. D.; Shmatko, B. A.; Milovidova, A. V.; Konovalov, E. YE.; Arnoldov, M. N.; Pleshivtsev, A. D.

    1988-01-01

    A molten sodium coolant containing hydrogen was heated in a vacuum at 450 C, and the gases generated pumped through a liquid nitrogen trap, and the H2 was then oxidized on a copper oxide substrate heated to 400 C. The accuracy of the method is 1.5 percent and the sensitivity is 1x10 to the -5 wt percent hydrogen.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  6. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  7. Labelling Of Coolant Flow Anomaly Using Fractal Structure

    International Nuclear Information System (INIS)

    Djainal, Djen Djen

    1996-01-01

    This research deals with the instrumentation of the detection and characterization of vertical two-phase flow coolant. This type of work is particularly intended to find alternative method for the detection and identification of noise in vertical two-phase flow in a nuclear reactor environment. Various new methods have been introduced in the past few years, an attempt to developed an objective indicator off low patterns. One of new method is Fractal analysis which can complement conventional methods in the description of highly irregular fluctuations. In the present work, Fractal analysis was applied to analyze simulated boiling coolant signal. This simulated signals were built by sum random elements in small subchannels of the coolant channel. Two modes are defined and both are characterized by their void fractions. In the case of uni modal -PDF signals, the difference between these modes is relatively small. On other hand, bimodal -PDF signals have relative large range. In this research, Fractal dimension can indicate the characters of that signals simulation

  8. Boundary layer attenuation in turbulent sodium flows

    International Nuclear Information System (INIS)

    Tenchine, D.

    1994-01-01

    Temperature fluctuations are produced in the sodium coolant of Liquid Metal Reactors when flows at different temperatures are mixing. That occurs in various areas of the reactor plant, in the primary and the secondary circuits. This paper deals with secondary circuit pipings, specifically the Superphenix steam generator outlet. The possibility of thermal striping in this area is studied because of the mixing of a main 'hot' flow surrounded by a smaller 'cold' flow in the vertical pipe located below the steam generator. This work was developed in the frame of a collaboration between CEA, EDF and FRAMATOME. The purpose of our study is to measure temperature fluctuations in the fluid and on the structures, on a sodium reduced scale model of the outlet region of the steam generator. We want to evidence the boundary layer attenuation by comparing wall and fluid measurements. From these experimental data, we shall propose a methodology to predict the boundary layer attenuation and the temperature fluctuations at the surface of the structure, for pipe flow configurations

  9. 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

  10. Regulation of liquid metal coolant flow rate in experimental loops

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Laptev, G.I.

    1987-01-01

    The possibility to use the VRT-2, RPA-T and R 133 analog temperature regulators for the automated regulation of liquid metal flow rate in the experimental loops for investigations on sodium and sodium-potassium alloy technology is considered. The RPA-T device is shown to be the most convenient one; it is characterized by the following parameters: measuring modulus transfer coefficient is 500; the range of regulating modulus proportionality factor variation - 0.3 - 50; the range of the regulating modulus intergrating time constant variation - 5 - 500 s

  11. A Review of Fragmentation Models Relative to Molten UO2 Breakup when Quenched in Sodium Coolant

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Grolmes, M.A.

    1976-01-01

    An important aspect of the fuel-coolant interaction problem relative to liquid metal fast breeder reactor (LMFBR) safety analysis is the fragmentation of molten oxide fuel during contact with liquid sodium coolant. A proper description of the kinetics of such an event requires an understanding of the breakup process and an estimate of the size and dispersion of such finely divided fuel in coolant. In recent years, considerable interest has centered on the problem of determining the nature of such fragmentation. In this paper, both analytic and experimental studies pertaining to such breakup are reviewed in light of recent developments in the understanding of heat transfer and solidification phenomena during quenching of UO 2 in sodium. A more extensive review of this subject can be found in Ref. 1. In conclusion: As discussed, a number of models have been proposed in an attempt to understand the nature of the UO 2 fragmentation process. The four principle mechanisms considered likely to cause such fragmentation (impact forces, boiling, violent gas release, and shell solidification) have been developed to the point where comparative analysis is possible. In addition, recent developments in the understanding of the physics of oxide fuel behavior in sodium coolant (boiling regime criteria, vapor nucleation theories, and prediction of solidification kinetics enable us to asses whether or not the various model assumptions are realistic. In view of this knowledge the following conclusions are made. For the case of hydrodynamic influence on fragmentation, it can be said that although the disruptive forces of impact and viscous drag may contribute to breakup, their effects are not controlling with respect to high temperature materials, including UO 2 -sodium. With respect to the vapor bubble growth and collapse mechanism it was shown that for sodium quenching, where coolant contact may, be expected (as opposed to water), the thermodynamic work potential of the bubble is

  12. BWR fuel assembly bottom nozzle with one-way coolant flow valve

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1987-01-01

    In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

  13. Numerical simulation on coolant flow and heat transfer in core

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

  14. Metallic sodium as a coolant of high speed nuclear reactors, (2)

    International Nuclear Information System (INIS)

    Atsumo, Hideo

    1975-01-01

    Tables are given on all the sodium loops in Japan and most of the sodium loops all over the world. Name and purpose of the loops, time of establishment, highest temperature, amount of sodium, flow rate, the materials used for the construction of the loops, and the diameter of the main pipings are given. Also, the problems related with these loops are discussed. For example, the high temperature sodium facility at HEDL-WADCO was made for the FFTF component test and instrument test, and uses 50,000 gallons of metallic sodium. The highest temperature is 590 0 C. The sodium flows at the rate of 60 g/m. The body is made of Type 304 stainless steel. Main data of existing sodium-cooled reactors in the world are also tabulated. The data include thermal output, electric output, the structure of the reactor cores, the dimensions of the cores, fuel used, the highest temperature in the reactors, the temperature of sodium at the inlet and outlet, the rate of multiplication, the amount of sodium used, number of control rods, number of heat exchangers, and the pressure of steam. The Monju type nuclear reactor in Japan uses 1,800 ton of sodium. (Fukutomi, T.)

  15. Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

  16. Eddy-current flow rate meter for measuring sodium flow rates

    International Nuclear Information System (INIS)

    Knaak, J.

    1976-01-01

    For safety reasons flow rate meters for monitoring coolant flow rates are inserted in the core of sodium-cooled fast breeder reactors. These are so-called eddy-current flow rate meters which can be mounted directly above the fuel elements. In the present contribution the principle of measurement, the mechanical construction and the circuit design of the flow rate measuring device are described. Special problems and their solution on developing the measuring system are pointed out. Finally, results of measurement and experience with the apparatus in several experiments are reported, where also further possibilities of application were tested. (orig./TK) [de

  17. Analysis of Sodium-23 Data Cross-Sections for Coolant on Generation IV Reactor - SFR

    International Nuclear Information System (INIS)

    Suwoto; Zuhair

    2009-01-01

    The integral tests of sodium-23 neutron cross-sections for coolant contained in JENDL-3.3, ENDF/B-VII.0, BROND-2.2 and JEFF-3.1 files have been performed. Cross-sections analysis of sodium-23 such as total cross-sections, elastic scattering, in-elastic scattering and radiative capture cross-sections for several temperature i.e. 300K, 800K and 1500K. The sodium-23 total cross-sections analysis based on MAEKER, R.E. experimental result through Broomstick experiment calculation. Differences between among other evaluated nuclear data file for elastic scattering, in-elastic scattering and radiative capture cross-sections were done analyzed and compared to ENDF/B-VII.0 as standard reference. Analysis of total cross-sections sodium-23 through broomstick calculation show JENDL-3.3 file give the best result on C/E statistical average value is 1.1043 compared to another nuclear data files. Differences sodium-23 total cross-sections on JEFF-3.1 file for all temperature work specially for energy 40keV and 1MeV-2MeV is about 0.2%. Meanwhile, relative small differences on in-elastic total scattering cross-sections are shown for all temperatures are about ±0.1% in JENDL-3.3. On the other hand, BROND-2.2 file give ±6% higher on sodium-23 in-elastic total scattering cross sections for energy range 450keV-550keV. Clearly significant differences on sodium-23 radiative capture cross sections for BROND-2.2 file especially in energy 109.659keV is somewhat higher than 446%, otherwise JENDL-3.3 and JEFF-3.1 give 16% higher than ENDF/B-VII.0 file. Overall result show that JENDL-3.3, ENDFB-VII.0, BROND-2.2 and JEFF-3.1 have little bit differences in total, elastic scattering, in-elastic scattering total cross sections, except BROND-2.2 file due to radiative capture cross-sections with larger discrepancies. (author)

  18. Cavitation erosion in sodium flow, sodium cavitation tunnel testing

    International Nuclear Information System (INIS)

    Courbiere, Pierre.

    1981-04-01

    The high-volume sodium flows present in fast neutron reactors are liable to induce cavitation phenomena in various portion of the sodium lines and pumps. The absence of sufficient data in this area led the C.E.A. to undertake an erosion research program in cavitating sodium flow. This paper discusses the considerations leading to the definition and execution of sodium cavitation erosion tests, and reviews the tests run with 400 0 C sodium on various steel grades: 316, 316 L, 316 Ti (Z8CNDT17-12), Poral (Z3CND18-12), 304 L and LN2 - clad 316 L (Ni coating-clad 316 L). Acoustic detection and signal processing methods were used with an instrument package designed and implemented at the Cadarache Nuclear Research Center

  19. Sodium aerosol formation in an argon flow over hot sodium

    International Nuclear Information System (INIS)

    Clement, C.F.; Dolias, M.J.; UKAEA Atomic Energy Research Establishment, Harwell. Thermal Hydraulics Div.)

    1987-01-01

    Vapour evaporation, which partly forms aerosol, occurs when a cold gas flows over a hot liquid. A previous well-mixed model is extended to predict the final vapour plus aerosol content of such a flow in terms of its initial and final temperatures. The predictions are compared to results of the Copacabana II experiment in which argon passed over a sodium pool. Agreement is obtained for the final sodium density at moderate flow rates, and physical reasons are given as to why deviations occur at low and high flow rates. (author)

  20. Coolant flow monitoring in a PWR core using noise analysis

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1992-01-01

    Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)

  1. Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2017-02-15

    When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

  2. Correlation measurements of sodium flow rate with magnetic sensors

    International Nuclear Information System (INIS)

    Kebadze, B.V.; Krasnoyarov, N.V.; Adamovskij, L.A.; Golushko, V.V.; Sroelov, V.S.

    1978-01-01

    The results of bench-mark experiments and those carried out at the BOR-60 reactor to measure the sodium coolant flow rate by a correlation method are presented. The method is based on detecting the eddy type flow hydraulic nonuniformities using magnetic flowmeters. The measurements were fulfilled in a broad range of flow rates (G=10-10 4 m 3 /h, Re=2x10 5 -2x10 7 ). The measured and calculated mutual correlation functions are presented with parallel and perpendicular orientations of the flowmeters magnetic fields. A good accord is stated. Prerequirements to the arrangement of the measuring systems are formulated. As an important advantage of the correlation method a possibility of the flowmeter calibration in situ is hydhlighted

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

    International Nuclear Information System (INIS)

    2002-06-01

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

  4. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    International Nuclear Information System (INIS)

    Peng, Wei; Sun, Xiaokai; Jiang, Peixue; Wang, Jie

    2017-01-01

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  5. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  6. A system for cooling electronic elements with an EHD coolant flow

    International Nuclear Information System (INIS)

    Tanski, M; Kocik, M; Barbucha, R; Garasz, K; Mizeraczyk, J; Kraśniewski, J; Oleksy, M; Hapka, A; Janke, W

    2014-01-01

    A system for cooling electronic components where the liquid coolant flow is forced with ion-drag type EHD micropumps was tested. For tests we used isopropyl alcohol as the coolant and CSD02060 diodes in TO-220 packages as cooled electronic elements. We have studied thermal characteristics of diodes cooled with EHD flow in the function of a coolant flow rate. The transient thermal impedance of the CSD02060 diode cooled with 1.5 ml/min EHD flow was 7.8°C/W. Similar transient thermal impedance can be achieved by applying to the diode a large RAD-A6405A/150 heat sink. We found out that EHD pumps can be successfully applied for cooling electronic elements.

  7. Sodium evaporation into a forced argon flow

    International Nuclear Information System (INIS)

    Kumada, Toshiaki; Kasahara, Fumio; Ishiguro, Ryoji

    1975-01-01

    Evaporation from a rectangular sodium free surface into an argon flow was measured. Tests were carried out with varying sodium temperature, argon velocity and argon temperature respectively under conditions of fog formation being possible. In order to clarify the enhancement of evaporation by fog formation, convection heat transfer from a plate of the same geometry into an air flow was also measured. The evaporation rate and Sherwood number were compared with those predicted by both the heat transfer experiment and the theory proposed by Hill and Szekely, and also a comparison was run with the previously reported experimental results of sodium evaporation. As a result it was shown that the sodium evaporation rate in this experiment is at least four times as large as that predicted by the heat transfer experiment and varies almost linearly with the heat transfer rate and the sodium vapour pressure. (auth.)

  8. Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

    International Nuclear Information System (INIS)

    Curtis, Franklin G.; Ekici, Kivanc; Freels, James D.

    2011-01-01

    The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

  9. System and method for determining coolant level and flow velocity in a nuclear reactor

    Science.gov (United States)

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  10. Flow in Rotating Serpentine Coolant Passages With Skewed Trip Strips

    Science.gov (United States)

    Tse, David G.N.; Steuber, Gary

    1996-01-01

    Laser velocimetry was utilized to map the velocity field in serpentine turbine blade cooling passages with skewed trip strips. The measurements were obtained at Reynolds and Rotation numbers of 25,000 and 0.24 to assess the influence of trips, passage curvature and Coriolis force on the flow field. The interaction of the secondary flows induced by skewed trips with the passage rotation produces a swirling vortex and a corner recirculation zone. With trips skewed at +45 deg, the secondary flows remain unaltered as the cross-flow proceeds from the passage to the turn. However, the flow characteristics at these locations differ when trips are skewed at -45 deg. Changes in the flow structure are expected to augment heat transfer, in agreement with the heat transfer measurements of Johnson, et al. The present results show that trips are skewed at -45 deg in the outward flow passage and trips are skewed at +45 deg in the inward flow passage maximize heat transfer. Details of the present measurements were related to the heat transfer measurements of Johnson, et al. to relate fluid flow and heat transfer measurements.

  11. Optimization of mass flow rate in RGTT200K coolant purification for Carbon Monoxide conversion process

    International Nuclear Information System (INIS)

    Sumijanto; Sriyono

    2016-01-01

    Carbon monoxide is a species that is difficult to be separated from the reactor coolant helium because it has a relatively small molecular size. So it needs a process of conversion from carbon monoxide to carbondioxide. The rate of conversion of carbon monoxide in the purification system is influenced by several parameters including concentration, temperature and mass flow rate. In this research, optimization of the mass flow rate in coolant purification of RGTT200K for carbon monoxide conversion process was done. Optimization is carried out by using software Super Pro Designer. The rate of reduction of reactant species, the growth rate between the species and the species products in the conversion reactions equilibrium were analyzed to derive the mass flow rate optimization of purification for carbon monoxide conversion process. The purpose of this study is to find the mass flow rate of purification for the preparation of the basic design of the RGTT200K coolant helium purification system. The analysis showed that the helium mass flow rate of 0.6 kg/second resulted in an un optimal conversion process. The optimal conversion process was reached at a mass flow rate of 1.2 kg/second. A flow rate of 3.6 kg/second – 12 kg/second resulted in an ineffective process. For supporting the basic design of the RGTT200K helium purification system, the mass flow rate for carbon monoxide conversion process is suggested to be 1.2 kg/second. (author)

  12. Analytical evaluation of local fault in sodium cooled small fast reactor (4S). Preliminary evaluation of partial blockage in coolant channel

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Ueda, Nobuyuki

    2007-01-01

    Local faults are fuel failures that result from heat removal imbalance within a single subassembly especially in FBRs. Although the occurrence frequency of local faults is quite low, the licensing body required local faults evaluations in previous FBR plants to confirm the potential for the occurrence of severe fuel subassembly failure and its propagation. A conceptual design of 4S (Super-Safe, Small and Simple) is a sodium cooled fast reactor, which aims at an application to dispersed energy source and long core lifetime. It has a dense arrangement of fuel pins to achieve a long lifetime. Therefore, from the viewpoint of thermal hydraulics, the 4S reactor is considered to have more potential for coolant boiling and fuel pin failure caused by formation of local blockage, comparing these potential in the conventional FBRs. The objective of the present study is to evaluate the effect of local blockage on the coolant flow pattern and temperature rise in the 4S-type fuel subassembly under the normal operation condition. A series of three-dimensional thermal-hydraulic analysis in a single subassembly with local blockage was conducted by the commercialized CFD code 'PHOENICS'. Analytical results show that the peak coolant temperature behind the blockage rises with increasing the blockage area, however, the coolant boiling does not occur under the present analytical conditions. On the other hand, it is found that the liquid phase formation caused by eutectic reactions will occur between the metallic fuel and the cladding under the local blockage condition. However, the penetration rate of liquid phase at fuel-cladding interface is quit low. Therefore, it is expected that rapid fuel pin failure and its propagation to surrounding pins due to liquid phase formation will not occur. (author)

  13. Ultrasonic flow-meter test in sodium

    International Nuclear Information System (INIS)

    Ishii, Y.; Uno, O.; Kamei, M.

    1978-01-01

    As a part of the R and D programme for the prototype fast breeder reactor MONJU, an ultrasonic flow-meter (USFM) test is being carried out in sodium in the O-Arai Engineering Center of PNC. Prior to the present test, an in-water test was done at the manufacturer's as a preliminary investigation. The results reported here are the results up to the present. Calibration tests using the actual fluid were conducted on a 12-inch ultrasonic flow-meter with guide rods fabricated for sodium flow measurement. The test conditions in sodium were a temperature of 200 approximately 400 0 C and flow-rates of 0 approximately 6m/s. The main results are: (1) The linearity of output signal was good and accuracy was within 1%; (2) The alternating type of the USFM was much better than the fixed type in temperature change; (3) 2MHz of transducer frequency was better than 3MHz in sodium; (4) The S/N ratio of the ultrasonic signal and the length/diameter effect in a wide range in sodium surpassed the in-water test. (author)

  14. Effect of coolant flow rate on the power at onset of nucleate boiling in a swimming pool type research reactor

    International Nuclear Information System (INIS)

    Khan, L.A.; Ahmad, N.; Ahmad, S.

    1998-01-01

    The effect of flow rate of coolant on power of Onset Nucleate Boiling (ONB) in a reference core of a swimming pool type research reactor has been studied using a as standard computer code PARET. It has been found that the decrease in the coolant flow rate results in a corresponding decrease in power at ONB. (author)

  15. Recent developments in the analysis of coolant sodium for fast breeder reactors

    International Nuclear Information System (INIS)

    Keough, R.F.; McCown, J.J.

    1976-01-01

    The measurement of impurities in sodium is an important part of the sodium technology for the fast breeder reactor program. A knowledge of impurity levels in sodium provides an important source of information on such things as corrosion rates, air and water leak location and detection, and the effectiveness of the sodium purification systems. A discussion is presented of some of the analytical techniques for sodium characterization in reactor heat transfer systems. Emphasis is placed on some recently developed alternatives to laboratory analysis

  16. FBR structural material test facility in flowing sodium environment

    International Nuclear Information System (INIS)

    Shanmugasundaram, M.; Kumar, Hemant; Ravi, S.

    2016-01-01

    In Fast Breeder Reactor (FBR), components such as Control and Safety Rod Drive Mechanism (CSRDM), Diverse Safety Rod Drive Mechanism (DSRDM), Transfer arm and primary sodium pumps etc., are experiencing friction and wear between the moving parts in contact with liquid sodium at high temperature. Hence, it is essential to evaluate the friction and wear behaviour to validate the design of components. In addition, the above core structural reactor components such as core cover plate, control plugs etc., undergoes thermal striping which is random thermal cycling induced by flow stream resulting from the mixing of non isothermal jets near that component. This leads to development of surface cracks and assist in crack growth which in turn may lead to failure of the structural component. Further, high temperature components are often subjected to low cycle fatigue due to temperature gradient induced cyclic thermal stresses caused by start-ups, shutdowns and transients. Also steady state operation at elevated temperature introduces creep and the combination of creep and fatigue leads to creep-fatigue interactions. Therefore, resistance to low cycle fatigue, creep and creep-fatigue are important considerations in the design of FBR components. Liquid sodium is used as coolant and hence the study of the above properties in dynamic sodium are equally important. In view of the above, facility for materials testing in sodium (INSOT) has been constructed and in operation for conducting the experiments such as tribology, thermal stripping, low cycle fatigue, creep and creep-fatigue interaction etc. The salient features of the operation and maintenance of creep and fatigue loops of INSOT facility are discussed in detail. (author)

  17. Identification of flow patterns by neutron noise analysis during actual coolant boiling in thin rectangular channels

    International Nuclear Information System (INIS)

    Kozma, R.; van Dam, H.; Hoogenboom, J.E.

    1992-01-01

    The primary objective of this paper is to introduce results of coolant boiling experiments in a simulated materials test reactor-type fuel assembly with plate fuel in an actual reactor environment. The experiments have been performed in the Hoger Onderwijs Reactor (HOR) research reactor at the Interfaculty Reactor Institute, Delft, The Netherlands. In the analysis, noise signals of self-powered neutron detectors located in the neighborhood of the boiling region and thermocouple in the channel wall and in the coolant are used. Flow patterns in the boiling coolant have been identified by means of analysis of probability density functions and power spectral densities of neutron noise. It is shown that boiling has an oscillating character due to partial channel blockage caused by steam slugs generated periodically between the plates. The observed phenomenon can serve as a basis for a boiling detection method in reactors with plate-type fuels

  18. Evaluation of effective coolant flow rate in advanced design of the small scale VHTR core

    International Nuclear Information System (INIS)

    Fumizawa, Motoo; Suzuki, Kunihiko; Murakami, Tomoyuki.

    1988-02-01

    This report describes the evaluation of effective coolant flow rate in the advanced design of the small scale VHTR core. The analytical design study was carried out after the 2nd stage of detailed design in order to reduce the cost of construction. The summary of the analytical results are as follows: (1) Crossflow loss coefficient of flange type fuel block having 0.1 mm of sealing gap is about 100 times higher than that of dowel type block adopted in the 2nd stage of detailed design. (2) In case that coolant channel outer diameter is 52 mm and hydraulic diameter is 6 mm, the effective coolant flow rates using flange and dowel type fuel blocks are 80 % and 70 % respectively. Because the crossflow loss coefficients of dowel type are lower than that of flange type. (3) The effective coolant flow rate, when crossflow loss coefficients are distributed along with the axial direction, agrees well with that using mean value of crossflow loss coefficient i.e. 5 x 10 11 m -4 . (author)

  19. Dryout heat flux in a debris bed with forced coolant flow from below

    International Nuclear Information System (INIS)

    Bang, Kwang-Hyun; Kim, Jong-Myung

    2004-01-01

    The objective of the present study is to experimentally investigate the enhancement of dryout heat flux in debris beds with coolant flow from below. The experimental facility consists mainly of an induction heater (40 kW, 35 kHz), a double-wall quartz-tube test section containing steel-particle bed and coolant injection and recovery condensing loop. A fairly uniform heating of particle bed was achieved by induction heating. This paper reports the experimental data for 5 mm particle bed and 300 mm bed height. The dryout heat rate data were obtained of both top-flooding case and forced coolant injection from below with the injection mass flux up to 1.5 kg/m 2 s. For the top-flooded case, the volumetric dryout heat rate was about 4 MW/m 3 and it increased as the rate of coolant injection from below was increased. At the coolant injection mass flux of 1.5 kg/m 2 s, the volumetric dryout heat rate was about 10 MW/m 3 , the enhancement factor was more than two. (author)

  20. The Experimental Characterization of the Magnetic Field Effect on a Liquid Sodium Flow

    International Nuclear Information System (INIS)

    Kim, Hee Reyoung; Kim, Jong Man; Cha, Jae Eun; Choi, Jong Hyun; Nam, Ho Yoon

    2006-01-01

    A liquid sodium coolant is used for a LMR such as KALIMER and a magnetic field is generated in the electromagnetic pump or flowmeter. The magnetic field has an effect on the electrically conducting metal flow by a generation of an electromagnetic pressure drop. Therefore, in the present study, a theoretical calculation is carried out for the effect of an external magnetic field and the magnetic field is measured over the electromagnet system manufactured for the magnetohydrodynamic experiments

  1. The Theoretical Investigation of the Magnetic Field Effect on a Liquid Sodium Flow

    International Nuclear Information System (INIS)

    Kim, Hee Reyoung; Kim, Jong Man; Cha, Jae Eun; Choi, Jong Hyun; Nam, Ho Yoon

    2005-01-01

    The liquid sodium coolant is used for LMR such as KALIMER and magnetic field is generated in the electromagnetic pump or flowmeter. The magnetic field takes an effect on the electrically conducting metal flow by the generation of the electromagnetic pressure drop. Therefore, in the present study, the theoretical calculation is carried out for an effect from the external magnetic field and the magnetic field is firstly measured over the electromagnet system manufactured for the magnetohydrodynamic experiments

  2. Determination of primary flow by correlation of temperatures of the coolant

    International Nuclear Information System (INIS)

    Villanueva, Jose

    2003-01-01

    Correlation techniques are often used to assess primary coolant flow in nuclear reactors. Observable fluctuations of some physical or chemical coolant properties are suitable for this purpose. This work describes a development carried out at the National Atomic Energy Commission of Argentina (CNEA) to apply this technique to correlate temperature fluctuations. A laboratory test was performed. Two thermocouples were installed on a hydraulic loop. A stationary flow of water circulated by the mentioned loop, where a mechanical turbine type flowmeter was installed. Transit times given by the correlation flowmeter, for different flow values measured with the mechanical flowmeter, were registered and a calibration between them was done. A very good linear behavior was obtained in all the measured range. It was necessary to increase the fluctuation level by adding water at different temperatures at the measuring system input. (author)

  3. Analysis of coolant flow in central tube of WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Zsiros, G.; Toth, S.; Attila Aszodi, A.

    2011-01-01

    Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

  4. Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

    International Nuclear Information System (INIS)

    Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

    2013-01-01

    The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

  5. Problems of technology and corrosion in sodium coolant and cover gas

    International Nuclear Information System (INIS)

    Kuenstler, K.; Ullmann, H.

    1977-07-01

    The meeting encloses the following themes: (i) Reactions in the system sodium-steel-cover gas (ii) Corrosion behaviour of structural and cladding materials (iii) Determination of impurities in sodium and cover gas (iv) Technology of sodium and cover gas (v) Testing equipments (vi) Safety problems

  6. 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.

  7. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    Science.gov (United States)

    Xie, Huaqing; Li, Yang; Yu, Wei

    2010-05-01

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2O 3, ZnO, TiO 2, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2O 3, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  8. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    International Nuclear Information System (INIS)

    Xie Huaqing; Li Yang; Yu Wei

    2010-01-01

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2 O 3 , ZnO, TiO 2 , and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2 O 3 , and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  9. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    Energy Technology Data Exchange (ETDEWEB)

    Xie Huaqing, E-mail: hqxie@eed.sspu.c [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China); Li Yang; Yu Wei [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China)

    2010-05-31

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al{sub 2}O{sub 3}, ZnO, TiO{sub 2}, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al{sub 2}O{sub 3}, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  10. Full sized tests on a french coolant pump under two-phase flow

    International Nuclear Information System (INIS)

    Huchard, J.C.; Bore, C.; Dueymes, E.

    1997-01-01

    The French Safety Authorities required EDF to demonstrate the ability of the new N4 main coolant pump to withstand two-phase flow conditions without damage. Therefore three full sized tests, simulating a bleeding flow on the primary system, were performed on a laboratory test loop under real operating conditions (temperature = 290 deg. C, pressure = 155 b, flowrate = 7 m 3 /s; electrical power = 7 MW). The maximum value of the mean void fraction reached 75 %. The outcome of the tests is very positive: the mechanical behaviour of the main coolant pump is good, even at high void fraction. The maximum vibration levels were below the limits fixed by the manufacturer. Correlations between the mechanical behaviour of the pump and the pressure pulsation in the test loop have been found. (authors)

  11. Numerical Simulation of Non-Rotating and Rotating Coolant Channel Flow Fields. Part 1

    Science.gov (United States)

    Rigby, David L.

    2000-01-01

    Future generations of ultra high bypass-ratio jet engines will require far higher pressure ratios and operating temperatures than those of current engines. For the foreseeable future, engine materials will not be able to withstand the high temperatures without some form of cooling. In particular the turbine blades, which are under high thermal as well as mechanical loads, must be cooled. Cooling of turbine blades is achieved by bleeding air from the compressor stage of the engine through complicated internal passages in the turbine blades (internal cooling, including jet-impingement cooling) and by bleeding small amounts of air into the boundary layer of the external flow through small discrete holes on the surface of the blade (film cooling and transpiration cooling). The cooling must be done using a minimum amount of air or any increases in efficiency gained through higher operating temperature will be lost due to added load on the compressor stage. Turbine cooling schemes have traditionally been based on extensive empirical data bases, quasi-one-dimensional computational fluid dynamics (CFD) analysis, and trial and error. With improved capabilities of CFD, these traditional methods can be augmented by full three-dimensional simulations of the coolant flow to predict in detail the heat transfer and metal temperatures. Several aspects of turbine coolant flows make such application of CFD difficult, thus a highly effective CFD methodology must be used. First, high resolution of the flow field is required to attain the needed accuracy for heat transfer predictions, making highly efficient flow solvers essential for such computations. Second, the geometries of the flow passages are complicated but must be modeled accurately in order to capture all important details of the flow. This makes grid generation and grid quality important issues. Finally, since coolant flows are turbulent and separated the effects of turbulence must be modeled with a low Reynolds number

  12. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

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

    International Nuclear Information System (INIS)

    Hunsbedt, A.; Boardman, C.E.

    1993-01-01

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

  14. Corrosion phenomena in sodium-potassium coolant resulting from solute interaction in multicomponent solution

    Science.gov (United States)

    Krasin, V. P.; Soyustova, S. I.

    2018-03-01

    The solubility of Fe, Cr, Ni, V, Mn and Mo in sodium-potassium melt has been calculated using the mathematical framework of pseudo-regular solution model. The calculation results are compared with available published experimental data on mass transfer of components of austenitic stainless steel in sodium-potassium loop under non-isothermal conditions. It is shown that the parameters of pair interaction of oxygen with transition metal can be used to predict the corrosion behavior of structural materials in sodium-potassium melt in the presence of oxygen impurity. The results of calculation of threshold concentration of oxygen of ternary oxide formation of sodium with transitional metals (Fe, Cr, Ni, V, Mn, Mo) are given in conditions when pure solid metal comes in contact with sodium-potassium melt.

  15. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  16. Q-factor of coolant flow in the primary circuit of NPP with pressurised water reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Belikov, S.O.; Novikov, K.S.

    2011-01-01

    Systems of preoperational vibration dynamic monitoring in of WWER are presented. The results of measurements during commission of NPP with WWER are presented. The paper provides the result of the research, that estimation of coolant fluctuations caused by pulse perturbation of pressure in the primary circuit NPP. It is shown that results could be received at known value of a Q - factor of acoustical oscillatory system only. The research demonstrates the results of dependence of the sound speed from the mass steam content in the coolant flow thru reactor core. The worked out results can be used for identification of the reasons of abnormal growth of level of vibrations of fuel assembly, fuel rod, equipment and internals, and for forecasting the operation conditions which provide of vibration - acoustical resonances in the primary loop equipment. (author)

  17. Analytical study on coolant temperature of several leak flows in the experimental VHTr core

    International Nuclear Information System (INIS)

    Fumizawa, Motoh; Arai, Taketoshi; Miyamoto, Yoshiaki

    1982-08-01

    This report describes heat transfer analysis of several leak flows which bypass main coolant flow path in the experimental VHTR core. The analysis contains the leak flow at permanent reflectors, replaceable reflectors and gaps between fuel columns. The summary of the results are as follows: (1) the temperature of the leak flow gas increases up to the surface temperature of permanent reflectors, (2) the gas temperature at replaceable reflectors increases at least 40 0 C in case of the worst analytical condition, (3) the gas temperature increases remarkably with decreasing equivalent diameter which is changed by the angle of bevel edge of the reflector, (4) while the gas temperature is low at the upper part of the fuel element, the temperature increases rapidly when it flow down along the gap of the fuel columns. (author)

  18. Flow in sodium loop surge tank

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1977-01-01

    The alternate liquid flow, the condition of vortex formation, gas entrainment in the discharge and the liquid level characteristics are studied using the models of the vertical and horizontal surge tanks of a sodium circuit with pump and heat exchangers. The conditions for vortex formation are more favourable in the vertical cylindrical tank than in the horizontal tank. The size of the vortex produced in the tank is affected by the initial speed circulation, due as a rule to an unsuitable inlet design. The proposed design considers an inlet below the sodium level using capped perforated pipes. Vortex formation, gas transport to the discharge pipe and turbulences of the liquid in the tank may be prevented by dividing the tank to the discharge and the inlet areas using perforated partitions, and by inserting the discharge cylinder above the discharge pipe outflow. The liquid level in the tank may be calmed by screens or by perforated plates. The adaptation of the surge tank of the sodium circuit will probably eliminate vortex formation and the entrainment of cover gas into the discharge piping and the sodium circuit under nominal conditions. (J.B.)

  19. Technology for sodium purity control

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Ji Young; Kim, B. H.; Kim, T. J. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    When sodium is used as heat transfer fluid, the plugging in coolant flow, the corrosion of structure material and the transfer of radioactive material caused by the impurities in sodium are worth considerable. Accordingly, these impurities must be monitored and controlled continuously by sodium purification devices in the heat transfer system which sodium is used as coolant. Sodium purification loop was constructed for the purpose of accumulating the technology for purity control of the coolant, developing and verifying further efficient instruments for sodium purification. The plugging meter and the cold trap is used as the implement for measuring and controlling the oxygen and the hydrogen, the main impurities in sodium coolant. They are capable of excellent performance as the implements which could detect and monitor the impurities to the concentration limit required for nuclear reactor. Sodium purification loop could be used variably according to the experimental purpose. 18 refs., 34 figs., 8 tabs. (Author)

  20. Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

    2013-06-01

    Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

  1. A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

    International Nuclear Information System (INIS)

    Markfort, D.; Kaiser, A.; Dohmen, A.

    1975-01-01

    In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

  2. A numerical approach to the simulation of one-phase and two phase reactor coolant flow around nuclear fuel spacers

    International Nuclear Information System (INIS)

    Stosic, Z.V.; Stevanovic, V.D.

    2001-01-01

    A methodology for the simulation and analysis of one-phase and two-phase coolant flows around one or a row of spacers is presented. It is based on the multidimensional two-fluid mass, momentum and energy balance equations and application of adequate turbulence models. Necessary closure laws for interfacial transfer processes are presented. The stated general approach enables simulation and analyses of reactor coolant flow around spacers on different scale levels of the rod bundle geometry: detailed modelling of coolant flow around spacers and investigation of the influence of spacer's geometry on the coolant thermal-hydraulics, as well as prediction of global thermal-hydraulic parameters within the whole rod bundle with the investigation of the influence of rows of spacers on the bulk thermal-hydraulic processes. Sample problems are included illustrating these different modelling approaches. (author)

  3. JOYO coolant sodium and cover gas purity control database (MK-II core)

    International Nuclear Information System (INIS)

    Ito, Kazuhiro; Nemoto, Masaaki

    2000-03-01

    The experimental fast reactor 'JOYO' served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon dioxide, methane and helium in argon gas with the reactor condition. (author)

  4. A dynamic model of the reactor coolant system flow for KMRR plant simulation

    International Nuclear Information System (INIS)

    Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

    1990-01-01

    To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

  5. Development of two-phase Flow Model, 'SOBOIL', for Sodium

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Chang, Won Pyo; Kim, In Chul; Kwon, Young Min; Lee, Yong Bum

    2000-03-01

    The objective of this research is to develop a sodium two-phase flow analysis model, 'SOBOIL', for the assessment of the initial stage of the KALIMER HCDA (Hypotherical Core Disruptive Accident). The 'SOBOIL' is basically similar to the multi-bubble slug ejection model used in SAS2A[1]. When a bubble is formed within the liquid slug, the bubble fills the whole cross section of the coolant channel except for a film left on the cladding or on the structure. Up to nine bubbles, separated by the liquid slugs, are allowed in the channel at any time. Each liquid slug flow rate in the model is performed in 2 steps. In the first step, the preliminary flow rate in the liquid slug is calculated neglecting the effect of changes in the vapor bubble pressures over the time step. The temperature and pressure distributions, and interface velocity at the interface between the liquid slug and vapor bubble are also calculated during this process. The new vapor temperature and pressure are then determined from the balance between the net energy transferred into the vapor and the change of the vapor energy. The liquid flow is finally calculated considering the change of the vapor pressure over a time step and the calculation is repeated until specified elapsed time is met. Continuous effort, therefore, must be made on the examination and improvement for the model to become reliable. To this end, much interest must be concentrated in the relevant international collaborations for access to a reference model or test data for the verification

  6. Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

    International Nuclear Information System (INIS)

    In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

    2008-01-01

    Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

  7. A thermal analysis computer programme package for the estimation of KANUPP coolant channel flows and outlet header temperature distribution

    International Nuclear Information System (INIS)

    Siddiqui, M.S.

    1992-06-01

    COFTAN is a computer code for actual estimation of flows and temperatures in the coolant channels of a pressure tube heavy water reactor. The code is being used for Candu type reactor with coolant flowing 208 channels. The simulation model first performs the detailed calculation of flux and power distribution based on two groups diffusion theory treatment on a three dimensional mesh and then channel powers, resulting from the summation of eleven bundle powers in each of the 208 channels, are employed to make actual estimation of coolant flows using channel powers and channel outlet temperature monitored by digital computers. The code by using the design flows in individual channels and applying a correction factor based on control room monitored flows in eight selected channels, can also provide a reserve computational tool of estimating individual channel outlet temperatures, thus providing an alternate arrangements for checking Rads performance. 42 figs. (Orig./A.B.)

  8. Study of coolant flow distribution within the PWR type reactor vessel

    International Nuclear Information System (INIS)

    Eberle, L.M.M.

    1983-01-01

    The thermohydraulic design of a pressurized water reactor requires the determination of the coolant flow distributions within the reactor vessel, particulary at the core inlet. In this work it is proposed the study of this flow, using potencial flow theory governed by Laplace's equation, nabla 2 φ = O. The solution of the potential field is obtained by the finite element method, which simplifies considerably the treatment of complex geometrical configurations. The equation is solved by the finite element computer code ANSYS, developed and licensed for structural and thermal analysis by using the analogy between steady state heat transfer equation without heat generation, nabla 2 T=O, and Laplace's equation of the velocity potential. The proposed method has been applied to a commercial reactor, and the results are consistent with the available experimental data. (author) [pt

  9. Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

    2017-07-01

    Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

  10. Effect of cross-flow direction of coolant on film cooling effectiveness with one inlet and double outlet hole injection

    Directory of Open Access Journals (Sweden)

    Guangchao Li

    2012-12-01

    Full Text Available In order to study the effect of cross-flow directions of an internal coolant on film cooling performance, the discharge coefficients and film cooling effectiveness with one inlet and double outlet hole injections were simulated. The numerical results show that two different cross-flow directions of the coolant cause the same decrease in the discharge coefficients as that in the case of supplying coolant by a plenum. The different proportion of the mass flow out of the two outlets of the film hole results in different values of the film cooling effectiveness for three different cases of coolant supplies. The film cooling effectiveness is the highest for the case of supplying coolant by the plenum. At a lower blowing ratio of 1.0, the film cooling effectiveness with coolant injection from the right entrance of the passage is higher than that from the left entrance of the passage. At a higher blowing ratio of 2.0, the opposite result is found.

  11. Using the coolant temperature noise for measuring the flow rate in the RBMK technological channels

    International Nuclear Information System (INIS)

    Selivanov, V.M.; Karlov, N.P.; Martynov, A.D.; Prostyakov, V.V.; Lysikov, B.V.; Kuznetsov, B.A.; Pallagi, D.; Khorani, Sh.; Khargitai, T.; Tezher, Sh.

    1983-01-01

    The problems are considered connected with the possibility of using thermometric correlation method to measure the coolant flow rate in the RBMK reactor technological channels. The main attention is paid to the study of the physical nature of the coolant temperature pulsations and to estimation of the effect of parameters of the primary thermaelectrical converter (TEC) on the results of measurements. In the process of reactor inspections made using the thermometric correlation flowmeter of a special design, the temperature noise distribution in the points of flow rate measurement is studied, the noise intensity and physical nature are determined, as well as the effect of different TEC parameters (TEC inertia and base distance between them) on the measurement accuracy. On the basis of the analysis of the effect on the results of the TEC thermal inertia measured value divergence, tausub(α) and transport time, tau sub(T), a conclusion is made on the necessity of choosing the base distance between TEC with tausub(T)>tausub(d)

  12. Combustion suppressing device for leaked sodium

    International Nuclear Information System (INIS)

    Ooto, Akihiro.

    1985-01-01

    Purpose: To suppress the atmospheric temperature to secure the building safety and shorten the recovery time after the leakage in a chamber for containing sodium leaked from coolant circuit equipments or pipeways of LMFBR type rector by suppressing the combustion of sodium contained in the chamber. Constitution: To the inner wall of a chamber for containing sodium handling equipments, are vertically disposed a panel having a coolant supply port at the upper portion and a coolant discharge port at the lower portion thereof and defined with a coolant flowing channel and a panel for sucking the coolant discharged from the abovementioned panel and exhausting the same externally. Further, a corrugated combustion suppressing plate having apertures for draining the condensated leaked sodium is disposed near the sodium handling equipments. If ruptures are resulted to the sodium handling equipments or pipeway, leaked sodium is passed through the drain apertures in the suppressing plate and stored at the bottom of the containing chamber. (Horiuchi, T.)

  13. Experimental and numerical study of hydrodynamics of flow-accelerated corrosion in CANDU primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Supa-Amornkul, S

    2006-07-01

    In CANDU-6 reactors, the pressurised high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310{sup o}C with up to 0.30 steam voidage, turns through 90{sup o} as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations reported excessive corrosion of their outlet feeder pipes, especially over the first metre, which consists of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion (FAC). Local shear stress, which is believed to be one of the important factors contributing to FAC, was approximated in the studies with standard empirical correlations. In order to understand the hydrodynamics of the coolant in the outlet feeders, flow-visualisation studies were done at AECL and UNB. At AECL, the observations were confined to a transparent simulation of an outlet feeder bend but at UNB a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting was fabricated. The feeder consisted of a 54 mm (inside diameter) acrylic pipe with a 73{sup o} bend, connected to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outside diameter, and an outer pipe, 150 mm inside diameter, both 1.907 m long. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 0.019 m{sup 3}/s and the volume fraction of air varied from 0.05 to 0.56. In characterizing the flow in the UNB study, particular attention was paid to the patterns at the inside of the bend, where a CFD (computational fluid dynamics) code

  14. Experimental and numerical study of hydrodynamics of flow-accelerated corrosion in CANDU primary coolant

    International Nuclear Information System (INIS)

    Supa-Amornkul, S.

    2006-01-01

    In CANDU-6 reactors, the pressurised high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310 o C with up to 0.30 steam voidage, turns through 90 o as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations reported excessive corrosion of their outlet feeder pipes, especially over the first metre, which consists of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion (FAC). Local shear stress, which is believed to be one of the important factors contributing to FAC, was approximated in the studies with standard empirical correlations. In order to understand the hydrodynamics of the coolant in the outlet feeders, flow-visualisation studies were done at AECL and UNB. At AECL, the observations were confined to a transparent simulation of an outlet feeder bend but at UNB a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting was fabricated. The feeder consisted of a 54 mm (inside diameter) acrylic pipe with a 73 o bend, connected to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outside diameter, and an outer pipe, 150 mm inside diameter, both 1.907 m long. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 0.019 m 3 /s and the volume fraction of air varied from 0.05 to 0.56. In characterizing the flow in the UNB study, particular attention was paid to the patterns at the inside of the bend, where a CFD (computational fluid dynamics) code - Fluent 6.1- had

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  16. The problems of using a high-temperature sodium coolant in nuclear power plants for the production of hydrogen and other innovative applications

    Science.gov (United States)

    Sorokin, A. P.; Alexeev, V. V.; Kuzina, Ju. A.; Konovalov, M. A.

    2017-11-01

    The intensity of the hydrogen sources arriving from the third contour of installation in second in comparison with the hydrogen sources on NPP BN-600 increases by two - three order at using of high-temperature nuclear power plants with the sodium coolant (HT-NPP) for drawing of hydrogen and other innovative applications (gasification and a liquefaction of coal, profound oil refining, transformation of biomass to liquid fuel, in the chemical industry, metallurgy, the food-processing industry etc.). For these conditions basic new technological solutions are offered. The main condition of their implementation is raise of hydrogen concentration in the sodium coolant on two - three order in comparison with the modern NPP, in a combination to hydrogen removal from sodium and its pumping out through membranes from vanadium or niobium. The researches with use diffusive model have shown possibility to expel a casium inflow in sodium through a leakproof shell of fuel rods if vary such parameters as a material of fuel rods shell, its thickness and maintenance time at design of fuel rods for high-temperature NPP. However maintenance of high-temperature NPP in the presence of casium in sodium is inevitable at loss of leakproof of a fuel rods shell. In these conditions for minimisation of casium diffusion in structural materials it is necessary to provide deep clearing of sodium from cesium.

  17. Design of the solid target structure and the study on the coolant flow distribution in the solid target using the 2-dimensional flow analysis

    International Nuclear Information System (INIS)

    Haga, Katsuhiro; Terada, Atsuhiko; Ishikura, Shuichi; Teshigawara, Makoto; Kinoshita, Hidetaka; Kobayashi, Kaoru; Kaminaga, Masaki; Hino, Ryutaro; Susuki, Akira

    1999-11-01

    A solid target cooled by heavy water is presently under development under the Neutron Science Research Project of the Japan Atomic Energy Research Institute (JAERI). Target plates of several millimeters thickness made of heavy metal are used as the spallation target material and they are put face to face in a row with one to two millimeters gaps in between though which heavy water flows, as the coolant. Based on the design criteria regarding the target plate cooling, the volume percentage of the coolant, and the thermal stress produced in the target plates, we conducted thermal and hydraulic analysis with a one dimensional target plate model. We choosed tungsten as the target material, and decided on various target plate thicknesses. We then calculated the temperature and the thermal stress in the target plates using a two dimensional model, and confirmed the validity of the target plate thicknesses. Based on these analytical results, we proposed a target structure in which forty target plates are divided into six groups and each group is cooled using a single pass of coolant. In order to investigate the relationship between the distribution of the coolant flow, the pressure drop, and the coolant velocity, we conducted a hydraulic analysis using the general purpose hydraulic analysis code. As a result, we realized that an uniform coolant flow distribution can be achieved under a wide range of flow velocity conditions in the target plate cooling channels from 1 m/s to 10 m/s. The pressure drop along the coolant path was 0.09 MPa and 0.17 MPa when the coolant flow velocity was 5 m/s and 7 m/s respectively, which is required to cool the 1.5 MW and 2.5 MW solid targets. (author)

  18. Low cycle fatigue testing in flowing sodium at elevated temperatures

    International Nuclear Information System (INIS)

    Flagella, P.N.; Kahrs, J.R.

    1976-01-01

    The paper describes equipment developed to obtain low cycle strain-controlled fatigue data in flowing sodium at elevated temperatures. Operation and interaction of the major components of the system are discussed, including the calibration technique using remote strain measurement and control. Confirmation of in-air results using the special technique is demonstrated, with data presented for Type 316 stainless steel tested in high purity flowing sodium at 593 0 C. The fatigue life of the material in sodium is essentially the same as that obtained in air for delta epsilon/sub t/= 1 percent. On the other hand, sodium pre-exposure at 650 0 C for 5000 hours increased the fatigue life in-sodium by a factor of two, and sodium pre-exposure at 718 0 C for 5000 hours increased the fatigue life in-sodium by a factor of three

  19. PBDOWN: A computer code for simulation of core material discharge and expansion in the upper coolant plenum in a hypothetical unprotected loss of flow accident in a LMFBR

    International Nuclear Information System (INIS)

    Royl, P.

    1985-01-01

    The report gives a description of the code PBDOWN (Pool Blow Down), its equations, input specifications and subroutines and it lists the input and output for some samples. Besides that some analysis results for the SNR-300 are discussed, that were obtained with this code. PBDOWN is an integral blow-down and expansion code, which simulates core material discharge and expansion into a sodium filled upper coolant plenum after build-up of vapour pressures in an unprotected loss of flow accident. The model includes the effect of sodium entrainment into an expending bubble of fuel or steel vapour with various assumptions for the heat transfer and vaporization of the entrained sodium droplets. The expanding vapour bubble is connected to the discharging pool via an orifice of a given size through which a time dependent ejection is simulated using quasi-stationary blow down correlations. The model allows bounding analysis of the possible influence of sodium vapour as a secondary working fluid, that is activated outside the pool on the overall expansion energy and discharge

  20. Sodium flow rate measurement method of annular linear induction pump

    International Nuclear Information System (INIS)

    Araseki, Hideo

    2011-01-01

    This report describes a method for measuring sodium flow rate of annular linear induction pumps arranged in parallel and its verification result obtained through an experiment and a numerical analysis. In the method, the leaked magnetic field is measured with measuring coils at the stator end on the outlet side and is correlated with the sodium flow rate. The experimental data and the numerical result indicate that the leaked magnetic field at the stator edge keeps almost constant when the sodium flow rate changes and that the leaked magnetic field change arising from the flow rate change is small compared with the overall leaked magnetic field. It is shown that the correlation between the leaked magnetic field and the sodium flow rate is almost linear due to this feature of the leaked magnetic field, which indicates the applicability of the method to small-scale annular linear induction pumps. (author)

  1. The relationship between dynamic and average flow rates of the coolant in the channels of complex shape

    Science.gov (United States)

    Fedoseev, V. N.; Pisarevsky, M. I.; Balberkina, Y. N.

    2018-01-01

    This paper presents interconnection of dynamic and average flow rates of the coolant in a channel of complex geometry that is a basis for a generalization model of experimental data on heat transfer in various porous structures. Formulas for calculation of heat transfer of fuel rods in transversal fluid flow are acquired with the use of the abovementioned model. It is shown that the model describes a marginal case of separated flows in twisting channels where coolant constantly changes its flow direction and mixes in the communicating channels with large intensity. Dynamic speed is suggested to be identified by power for pumping. The coefficient of proportionality in general case depends on the geometry of the channel and the Reynolds number (Re). A calculation formula of the coefficient of proportionality for the narrow line rod packages is provided. The paper presents a comparison of experimental data and calculated values, which shows usability of the suggested models and calculation formulas.

  2. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  3. Sodium flow distribution test of the air cooler tubes

    International Nuclear Information System (INIS)

    Uchida, Hiroyuki; Ohta, Hidehisa; Shimazu, Hisashi

    1980-01-01

    In the heat transfer tubes of the air cooler which is installed in the auxiliary core cooling system of the fast breeder prototype plant reactor ''Monju'', sodium freezing may be caused by undercooling the sodium induced by an extremely unbalanced sodium flow in the tubes. Thus, the sodium flow distribution test of the air cooler tubes was performed to examine the flow distribution of the tubes and to estimate the possibility of sodium freezing in the tubes. This test was performed by using a one fourth air cooler model installed in the water flow test facility. As the test results show, the flow distribution from the inlet header to each tube is almost equal at any operating condition, that is, the velocity deviation from normalized mean velocity is less than 6% and sodium freezing does not occur up to 250% air velocity deviation at stand-by condition. It was clear that the proposed air cooler design for the ''Monju'' will have a good sodium flow distribution at any operating condition. (author)

  4. Sodium flow measurement in large pipelines of sodium cooled fast breeder reactors with bypass type flow meters

    International Nuclear Information System (INIS)

    Rajan, K.K.; Jayakumar, T.; Aggarwal, P.K.; Vinod, V.

    2016-01-01

    Highlights: • Bypass type permanent magnet flow meters are more suitable for sodium flow measurement. • A higher sodium velocity through the PMFM sensor will increase its sensitivity and resolution. • By modifying the geometry of bypass line, higher sodium velocity through sensor is achieved. • With optimized geometry the sensitivity of bypass flow meter system was increased by 70%. - Abstract: Liquid sodium flow through the pipelines of sodium cooled fast breeder reactor circuits are measured using electromagnetic flow meters. Bypass type flow meter with a permanent magnet flow meter as sensor in the bypass line is selected for the flow measurement in the 800 NB main secondary pipe line of 500 MWe Prototype Fast Breeder Reactor (PFBR), which is at the advanced stage of construction at Kalpakkam. For increasing the sensitivity of bypass flow meters in future SFRs, alternative bypass geometry was considered. The performance enhancement of the proposed geometry was evaluated by experimental and numerical methods using scaled down models. From the studies it is observed that the new configuration increases the sensitivity of bypass flow meter system by around 70%. Using experimentally validated numerical tools the volumetric flow ratio for the bypass configurations is established for the operating range of Reynolds numbers.

  5. Evaluation of the corrosion, reactivity and chemistry control aspects for the selection of an alternative coolant in the secondary circuit of sodium fast reactors

    International Nuclear Information System (INIS)

    Brissonneau, L.; Simon, N.; Balbaud-Celerier, F.; Courouau, J.L.; Martinelli, L.; Grabon, V.; Capitaine, A.; Conocar, O.; Blat, M.

    2009-01-01

    Full text of publication follows: Sodium Fast Reactors are promising fourth generation reactors as they can contribute to reduce resource demand in uranium and considerably reduce waste level due to their fast spectrum. However, progress can be obtained for these reactors on the investment cost and on safety improvement. To achieve these goals, one of the innovative solutions consists in eliminating the reaction of sodium with water in the steam generators, by replacing the sodium in the secondary circuit by another coolant. A work group composed of experts from CEA, Areva NP and EdF was in charge to evaluate several alternative coolants as Heavy Liquid Metals (HLM), nitrate salts and hydroxide mixtures, through a multi-criteria analysis. Three important criteria for the selection of one coolant are its 'Interactions with the structures', and its 'chemistry control', and 'Reactivity with fluids' which are strongly correlated. The assessment, mainly based on the state-of-art from published literature on these points, is detailed in this paper. The mechanisms of corrosion of steels by the HLM depend on the oxygen content. For Pb-Bi, it has been modelled for oxidation and release domains. The corrosion of steels by nitrate salts presents similarity with the oxidation induced by HLM. The highly corrosive hydroxide mixture requires the use of nickel base alloys, for which oxidation and mass transfer are nevertheless significant. The HLM requires a fine regulation of oxygen content, through measurements and control systems, both to prevent lead oxide precipitation at high level and release corrosion at low level. Nitrate salts decompose into nitrites at sufficiently high temperature, which might induce pressure build-up in the circuit. The hydroxides must be kept under reducing atmosphere to lower the corrosion rate. Though these coolants are relatively inert to air and water, one of the main drawbacks of HLM and nitrate salts are their reactivity with sodium. Bismuth

  6. 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.

  7. Particle image velocimetry measurement of complex flow structures in the diffuser and spherical casing of a reactor coolant pump

    Directory of Open Access Journals (Sweden)

    Yongchao Zhang

    2018-04-01

    Full Text Available Understanding of turbulent flow in the reactor coolant pump (RCP is a premise of the optimal design of the RCP. Flow structures in the RCP, in view of the specially devised spherical casing, are more complicated than those associated with conventional pumps. Hitherto, knowledge of the flow characteristics of the RCP has been far from sufficient. Research into the nonintrusive measurement of the internal flow of the RCP has rarely been reported. In the present study, flow measurement using particle image velocimetry is implemented to reveal flow features of the RCP model. Velocity and vorticity distributions in the diffuser and spherical casing are obtained. The results illuminate the complexity of the flows in the RCP. Near the lower end of the discharge nozzle, three-dimensional swirling flows and flow separation are evident. In the diffuser, the imparity of the velocity profile with respect to different axial cross sections is verified, and the velocity increases gradually from the shroud to the hub. In the casing, velocity distribution is nonuniform over the circumferential direction. Vortices shed consistently from the diffuser blade trailing edge. The experimental results lend sound support for the optimal design of the RCP and provide validation of relevant numerical algorithms. Keywords: Diffuser, Flow Structures, Particle Image Velocimetry, Reactor Coolant Pump, Spherical Casing, Velocity Distribution

  8. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2015-01-01

    Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

  9. Three-dimensional analysis of the coolant flow characteristics in the fuel assemblies of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Dinh Van Thin; Tran Thi Nhan

    2015-01-01

    Computational Fluid Dynamics (CFD) is a widely used method around the world for complex flow and heat industrial problems. In this paper, the coolant flow parameters were investigated in subchannels of VVER-1000 reactor’s fuel assemblies by ANSYS V14.5 programme. The different mesh solutions and turbulence models were carried out to deal with the water flow problems such as velocity distribution, streamline, temperature and pressure change as well as the hydraulic resistances of the spacer grids. The obtained results are good agreement with the measured values and the published reports from other authors. (author)

  10. CFD analysis of flow distribution of reactor core and temperature rise of coolant in fuel assembly for VVER reactor

    International Nuclear Information System (INIS)

    Du Daiquan; Zeng Xiaokang; Xiong Wanyu; Yang Xiaoqiang

    2015-01-01

    Flow field of VVER-1000 reactor core was investigated by using computational fluid dynamics code CFX, and the temperature rise of coolant in hot assembly was calculated. The results show that the maximum value of flow distribution factor is 1.12 and the minimum value is 0.92. The average value of flow distribution factor in hot assembly is 0.97. The temperature rise in hot assembly is higher than current warning limit value ΔT t under the deviated operation condition. The results can provide reference for setting ΔT t during the operation of nuclear power plant. (authors)

  11. Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

    Directory of Open Access Journals (Sweden)

    Istvan Farkas

    2016-08-01

    Full Text Available The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively with experimental results.

  12. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

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

  13. Large Eddy Simulation of turbulent flow in wire wrapped fuel pin bundles cooled by sodium

    International Nuclear Information System (INIS)

    Saxena, Aakanksha; Cadiou, Thierry; Bieder, Ulrich; Viazzo, Stephane

    2013-06-01

    The objective of the study is to understand the thermal hydraulics in a core sub-assembly with liquid sodium as coolant by performing detailed numerical simulations. The passage for the coolant flow between the fuel rods is maintained by thin wires wrapped around the rods. The contact point between the fuel pin and the spacer wire is the region of creation of hot spots and a cyclic variation of temperature in hot spots can adversely affect the mechanical properties of the clad due to the phenomena like thermal stripping. The current status quo provides two different models to perform the numerical simulations, namely Reynolds Averaged Navier-Stokes (RANS) and Large Eddy Simulation (LES). The two models differ in the extent of modelling used to close the Navier-Stokes equations. LES is a filtered approach where the large scale of motions are explicitly resolved while the small scale motions are modelled whereas RANS is a time averaging approach where all scale of motions are modelled. Thus LES involves less modelling as compared to RANS and so the results are comparatively more accurate. An attempt has been made to use the LES model. The simulations have been performed using the code Trio-U (developed by CEA). The turbulent statistics of the flow and thermal quantities are calculated. Finally the goal is to obtain the frequency of temperature oscillations at the region of hot spots near the spacer wire. (authors)

  14. Heat and momentum transfer in a gas coolant flow through a circular pipe in a high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Ogawa, Masuro

    1989-07-01

    In Japan Atomic Energy Research Institute (JAERI), a very high temperature gas cooled reactor (VHTR) has been researched and developed with a purpose of attaining a coolant temperature of around 1000degC at the reactor outlet. In order to design VHTR, comprehensive knowledge is required on thermo-hydraulic characteristics of laminar-turbulent transition, of coolant flow with large thermal property variation due to temperature difference, and of heat transfer deterioration. In the present investigation, experimental and analytical studies are made on a gas flow in a circular tube to elucidate the thermo-hydraulic characteristics. Friction factors and heat transfer coefficients in transitional flows are obtained. Influence of thermal property variation on the friction factor is qualitatively determined. Heat transfer deterioration in the turbulent flow subjected to intense heating is experimentally found to be caused by flow laminarization. The analysis based on a k-kL two-equation model of turbulence predicts well the experimental results on friction factors and heat transfer coefficients in flows with thermal property variation and in laminarizing flows. (author)

  15. 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.

  16. 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.)

  17. A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

    Energy Technology Data Exchange (ETDEWEB)

    Lister, D. [University of New Brunswick, Fredericton, NB (Canada). Dept. of Chemical Engineering; Lang, L.C. [Atomic Energy of Canada Ltd., Chalk River Lab., ON (Canada)

    2002-07-01

    Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

  18. A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

    International Nuclear Information System (INIS)

    Lister, D.

    2002-01-01

    Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

  19. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

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

  20. On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

    International Nuclear Information System (INIS)

    Melesed'Hospital, G.

    1983-01-01

    The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

  1. Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

    International Nuclear Information System (INIS)

    Carter, G.W.

    1979-01-01

    Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

  2. Effect of removal of a central thimble on coolant flow distribution in a research reactor fuel element

    International Nuclear Information System (INIS)

    Green, W.J.

    1977-01-01

    Using two twice full-size models of a HIFAR research reactor fuel element, experiments have been performed to determine how the flow distribution of coolant gas through the element in a transfer flask is affected by removal of the central instrumentation thimble. With the thimble present, experimental flow results agree with theoretical predictions. Over the range of total flowrates considered, mass flow apportioning among the five annular channels was independent of annular channel Reynolds number (in the range 3500 to 10,500) and ranged between 13% and 27% of the total flowrate. For the case with the thimble removed, interesting experimental flow characteristics were obtained which could not have been predicted. Flow apportioning among the annular channels was found to be uniquely dependent upon total flowrate and ranged between 3% and 8% for the experimental conditions investigated (annular channel Reynolds numbers in the range 800 to 4000). (Author)

  3. On possibility of application of the parallel-mixed type coolant flow scheme to NPP steam generators linked with superheaters

    International Nuclear Information System (INIS)

    Malkis, V.A.; Lokshin, V.A.

    1983-01-01

    Optimum distribution of the coolant straight-through flow between the superheater, evaporator and economizer is determined and the parallel-mixed type flow scheme is compared with other schemes. The calculations are performed for the 250 MW(e) steam generator for the WWER-1000 reactor unit the inlet and outlet primary coolant temperature of which is 324 and 290 deg C, respectively, while the feed water and saturation temperatures are 220 and 278.5 deg C, respectively. The rated superheating temperature is 300 deg C. The comparison of different schemes has been performed according to the average temperature head value at the steam-generator under the condition of equality as well as essential difference in the heat transfer coefficients in certain steam-generator sections. The calculations have shown that the use of parallel-mixed type flow permits to essentially increase the temperature head of the steam generator. At a constant heat transfer coefficient in all steam generator sections the highest temperature head is reached. At relative flow rates in the steam generator, economizer and evaporator equal to 6, 8 and 86%, respectively. The superheated steam generator temperature head in this case by 12% exceeds the temperature head of the WWER-1000 reactor unit wet steam generator. In case of heat transfer coefficient reduction in the superheater by a factor of three, the choice of the primary coolant, optimum distribution permits to maintain the steam generator temperature head at the level of the WWER-1000 reactor unit wet-steam steam generator. The use of the parallel-mixed type flow scheme permits to design a steam generator of slightly superheated steam for the parameters of the WWER-1000 unit

  4. Sodium flow rate measurement method of annular linear induction pumps

    International Nuclear Information System (INIS)

    Araseki, Hideo; Kirillov, Igor R.; Preslitsky, Gennady V.

    2012-01-01

    Highlights: ► We found a new method of flow rate monitoring of electromagnetic pump. ► The method is very simple and does not require a large space. ► The method was verified with an experiment and a numerical analysis. ► The experimental data and the numerical results are in good agreement. - Abstract: The present paper proposes a method for measuring sodium flow rate of annular linear induction pumps. The feature of the method lies in measuring the leaked magnetic field with measuring coils near the stator end on the outlet side and in correlating it with the sodium flow rate. This method is verified through an experiment and a numerical analysis. The data obtained in the experiment reveals that the correlation between the leaked magnetic field and the sodium flow rate is almost linear. The result of the numerical analysis agrees with the experimental data. The present method will be particularly effective to sodium flow rate monitoring of each one of plural annular linear induction pumps arranged in parallel in a vessel which forms a large-scale pump unit.

  5. APPLICATION FEATURES OF SPATIAL CONDUCTOMETRY SENSORS IN MODELLING OF COOLANT FLOW MIXING IN NUCLEAR POWER UNIT EQUIPMENT

    Directory of Open Access Journals (Sweden)

    A. A. Barinov

    2016-01-01

    Full Text Available Coolant flow mixing processes with different temperatures and concentrations of diluted additives widely known in nuclear power units operation. In some cases these processes make essential impact on the resource and behavior of the nuclear unit during transient and emergency situations. The aim of the study was creation of measurement system and test facility to carry out basic tests and to embed spatial conductometry method in investigation practice of turbulent coolant flows. In the course of investigation measurement system with sensors and experimental facility was designed, several first tests were carried out. A special attention was dedicated to calibration and clarification of conductometry sensor application methodologies in studies of turbulent flow characteristics. Investigations involved method of electrically contrast tracer jet with concurrent flow in closed channel of round crosssection. The measurements include both averaged and unsteady realizations of measurement signal. Experimental data processing shows good agreement with other tests acquired from another measurement systems based on different physical principles. Calibration functions were acquired, methodical basis of spatial conductometry measurement system application was created. Gathered experience of spatial sensor application made it possible to formulate the principles of further investigation that involve large-scale models of nuclear unit equipment. Spatial wire-mesh sensors proved to be a perspective type of eddy resolving measurement devices.

  6. Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event

    International Nuclear Information System (INIS)

    Costa, A. L.; Cherubini, M.; D'Auria, F.; Giannotti, W.; Moskalev, A.

    2007-01-01

    One of the transients that have received considerable attention in the safety evaluation of RBMK reactors is the partial break of a group distribution header (GDH). The coolant flow rate blockage in one GDH might lead to excessive heat-up of the pressure tubes and can result in multiple fuel channels (FC) ruptures. In this work, the GDH flow blockage transient has been studied considering the Smolensk-3 RBMK NPP (nuclear power plant). In the RBMK, each GDH distributes coolant to 40-43 FC. To investigate the behavior of each FC belonging to the damaged GDH and to have a more realistic trend, one (affected) GDH has been schematised with its forty-two FC, one by one. The calculations were performed using the 0-D NK (neutron kinetic) model of the RELAP5-3.3 stand-alone code. The results show that, during the event, the mass flow rate is disturbed differently according to the power distribution established for each FC in the schematization. The start time of the oscillations in mass flow rate depends strongly on the attributed power to each FC. It was also observed that, during the event, the fuel channels at higher thermal power values tend to undergo first cladding rupture leaving the reactor to scram and safeguarding all the other FCs connected to the affected GDH.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  8. CFD simulation of a coolant flow and a heat transfer in a pebble bed reactor - HTR2008-58334

    International Nuclear Information System (INIS)

    In, W. K.; Lee, W. J.; Hassan, Y. A.

    2008-01-01

    This CFD study is to simulate a coolant(gas) flow and heat transfer in a PBR core during a normal operation. This study used a pebble array with direct area contacts among the pebbles which is one of the pebbles arrangements for a detailed simulation of PBR core CFD studies. A CFD model is developed to more adequately represent the pebbles randomly stacked in the PBR core. The CFD predictions showed a large variation of the temperature on the pebble surface as well as in the pebble core. The temperature drop in the outer graphite layer is smaller than that in the pebble-core region. This is because the thermal conductivity of graphite is higher than the fuel (UO, mixture) conductivity in the pebble core. Higher pebble surface temperature is predicted downstream of the pebble contact due to a reverse flow. Multiple vortices are predicted to occur downstream of the spherical pebbles due to a flow separation. The coolant flow structure and fuel temperature in the PBR core appears to largely depend on the in-core distribution of the pebbles. (authors)

  9. Measurement and analysis of vibrational behavior of an SNR-fuel element in sodium flow

    International Nuclear Information System (INIS)

    Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

    1975-01-01

    Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 deg C system temperature in the AKB sodium loop at Interatom, Bensberg. It is known that the coolant flow through a subassembly can induce flutter or vibrations of structural parts such as single pins, the wrapper and the total pin bundle all of which have been of interest during this test. To detect these vibrations of different structural parts simultaneously with a minimum of instrumentation only 3 weldable high temperature strain gauges were employed. These strain gauges were especially prepared and bent in such a way as to form a bridge between the inner wrapper and a fuel pin top and spot-welded to both the wrapper and the fuel pin. Although this arrangement seems to be a rather unusual one, the simultaneous-measurement of bundle, wrapper and pin vibrations was possible and periodic flow fluctuations were also detected. The presented results are only relative due to calibration difficulties with these deformed strain gauges which were first used during this test. It is, however, believed that this arrangement, in connection with the proposed anlytical approach, leads to a simple and technical representation of the vibrational behavior of core elements during sodium tests. Detailed information needed for check and calibration of computer codes are however displayed by the respective power spectral density functions

  10. Sodium flow distribution in test fuel assembly P-23B

    International Nuclear Information System (INIS)

    Taylor, J.P.S.

    1978-08-01

    Relatively large cladding diametral increases in the exterior fuel pins of HEDL's test fuel subassembly P-23B were successfully explained by a thermal-hydraulic/solid mechanics analysis. This analysis indicates that while at power, the subassembly flow was less than planned and that the fuel pins were considerably displaced and bowed from their nominal position. In accomplishing this analysis, a method was developed to estimate the sodium flow distribution and pin distortions in a fuel subassembly at power

  11. Chaotic Dynamos Generated by a Turbulent Flow of Liquid Sodium

    International Nuclear Information System (INIS)

    Ravelet, F.; Monchaux, R.; Aumaitre, S.; Chiffaudel, A.; Daviaud, F.; Dubrulle, B.; Berhanu, M.; Fauve, S.; Mordant, N.; Petrelis, F.; Bourgoin, M.; Odier, Ph.; Plihon, N.; Pinton, J.-F.; Volk, R.

    2008-01-01

    We report the observation of several dynamical regimes of the magnetic field generated by a turbulent flow of liquid sodium (VKS experiment). Stationary dynamos, transitions to relaxation cycles or to intermittent bursts, and random field reversals occur in a fairly small range of parameters. Large scale dynamics of the magnetic field result from the interactions of a few modes. The low dimensional nature of these dynamics is not smeared out by the very strong turbulent fluctuations of the flow

  12. 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

  13. Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-01-01

    This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

  14. Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1994-07-01

    This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

  15. Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

    International Nuclear Information System (INIS)

    Shieh, D.J.; Upadhyaya, B.R.

    1986-01-01

    In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

  16. Sodium evaporation into a forced argon flow, (1)

    International Nuclear Information System (INIS)

    Kumada, Toshiaki; Kasahara, Fumio; Ishiguro, Ryoji

    1976-01-01

    Measurements were made on the rate of evaporation from a rectangular-shaped free surface of liquid sodium into argon flow. Tests were carried out at various levels of sodium temperature, of argon velocity and of argon temperature, under conditions where fog formation could be expected. To gain information on the enhancement of evaporation occasioned by fog formation, a supplementary experiment was performed on convection heat transfer into flowing air from a heated plate of the same geometry as the free surface of the sodium in the preceding measurements. The values obtained for the rate of evaporation and Sherwood number were compared with those predicted by the heat transfer experiment and by the theory by Hill and Szekely. The overall results revealed that the rate of sodium evaporation can amount to as much as three times that predicted by the heat transfer experiment, and that it varies roughly linearly with the heat transfer rate and with the sodium vapor pressure prevailing at the free surface. (auth.)

  17. Device for preventing coolant outflow in a reactor

    International Nuclear Information System (INIS)

    Nemoto, Kiyomitsu; Mochizuki, Keiichi.

    1975-01-01

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

  18. Optimisation of the flow path in a conceptual pool type reactor under natural circulation with lead coolant

    International Nuclear Information System (INIS)

    Thiele, R.; Anglart, H.

    2014-01-01

    This contribution investigates the effects of a bypass flow blocking bottom plate and the influence of the heat transfer between the hot and cold leg in a small pool type reactor cooled through natural convection with lead coolant. The computations are carried out using 3D computational fluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled using a porous media approach. The introduction of full conjugate heat transfer shows that the heat transfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculation zones. These zones become even more pronounced with the introduction of a bottom plate, which on the other hand also increases the flow through the core and lowers the maximum temperature in the core by approximately 150 K. Based on the results, redesign suggestions for the bottom plate and the internal wall are made. (author)

  19. Development of Sodium Two Phase Flow Model for Kalimer Core Analysis

    International Nuclear Information System (INIS)

    Chang, W.P.; Hahn, Dohee

    2002-01-01

    An algorithm for sodium boiling is developed in order to extend the applicability of SSC-K, which is a main system analysis code for the KALIMER (Korea Advanced LIquid Metal Reactor) conceptual design. As the capability of the current SSC-K version is limited to simulation of only a single-phase sodium flow, its applicable range should not be enough to assess the fuel integrity under some of HCDA (Hypothetical Core Disruptive Accident) initiating events where sodium boiling is anticipated. The two-phase flow model similar to that used for the light water system is known to be no more effective directly to liquid metal reactors, because the phenomena observed between two reactor coolant systems are definitely different. The developing algorithm is based on a multiple-bubble slug ejection model, which allows a finite number of bubbles in a channel at any time. The present work is a continuous effort following the former study to confirm a qualitative acceptance on the model. Since the model has been applied only to the active fuel region in the former study, a part of its qualification seems to have already been demonstrated. For its application to the whole KALIMER core channel, however, the model needs to be examined the applicability to the fuel regions other than the active fuel. The present study primarily focuses on that point. In a result, although the model may be improved in a sense through the present study over the previous modeling, a clear limitation is also confirmed with the validity of the model. The further development, therefore, is required for this model to achieve its goal by resolving such limitations. (authors)

  20. The development of a model to study the thermal behaviour of the coolant in the blind elements of a fast sodium-cooled breeder in the case of a severe hypothetical accident during the initial phase

    International Nuclear Information System (INIS)

    Genter, G.

    1981-03-01

    The enthalpy level of the coolant is studied in the interior of gaps and special elements of a fast sodium coded breeder reactor during the initial and the final stages of a hypothetical accident. For this purpose numerical models are presented to calculate the heat transport in the special element on the basis of heat conduction and axial convection. (orig./RW) [de

  1. Multidimensional analysis of fluid flow in the loft cold leg blowdown pipe during a loss-of-coolant experiment

    International Nuclear Information System (INIS)

    Demmie, P.N.; Hofmann, K.R.

    1979-03-01

    A computer analysis of fluid flow in the Loss-of-Fluid Test (LOFT) cold leg blowdown pipe during a loss-of-coolant experiment (LOCE) was performed using the computer program K-FIX/MOD1. The purpose of this analysis was to evaluate the capability of K-FIX/MOD1 to calculate theoretical fluid quantity distributions in the blowdown pipe during a LOCE for possible application to the analysis of LOFT experimental data, the determination of mass flow, or the development of data reduction models. A rectangular section of a portion of the LOFT blowdown pipe containing measurement Station BL-1 was modeled using time-dependent boundary conditions. Fluid quantities were calculated during a simulation of the first 26 s of LOFT LOCE L1-4. Sensitivity studies were made to determine changes in void fractions and velocities resulting from specific changes in the inflow boundary conditions used for this simulation

  2. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

  3. A review of flow analysis methods for determination of radionuclides in nuclear wastes and nuclear reactor coolants.

    Science.gov (United States)

    Trojanowicz, Marek; Kołacińska, Kamila; Grate, Jay W

    2018-06-01

    The safety and security of nuclear power plant operations depend on the application of the most appropriate techniques and methods of chemical analysis, where modern flow analysis methods prevail. Nevertheless, the current status of the development of these methods is more limited than it might be expected based on their genuine advantages. The main aim of this paper is to review the automated flow analysis procedures developed with various detection methods for the nuclear energy industry. The flow analysis methods for the determination of radionuclides, that have been reported to date, are primarily focused on their environmental applications. The benefits of the application of flow methods in both monitoring of the nuclear wastes and process analysis of the primary circuit coolants of light water nuclear reactors will also be discussed. The application of either continuous flow methods (CFA) or injection methods (FIA, SIA) of the flow analysis with the β-radiometric detection shortens the analysis time and improves the precision of determination due to mechanization of certain time-consuming operations of the sample processing. Compared to the radiometric detection, the mass spectrometry (MS) detection enables one to perform multicomponent analyses as well as the determination of transuranic isotopes with much better limits of detection. Copyright © 2018 Elsevier B.V. All rights reserved.

  4. Numerical Analysis of Flow Distribution in a Sodium Chamber of a Finned-tube Sodium-to-Air Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Youngchul; Son, Seokkwon; Kim, Hyungmo; Eoh, Jaehyuk; Jeong, Jiyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    DHR systems consist of two diverse heat removal loops such as passive and active DHR systems, and the heat load imposed on the primary sodium pool is safely rejected into the environment through different kinds of sodium-to-air heat exchangers, e.g. M-shape and helical-coil type air-coolers. The former is called as an FHX(Forced-draft sodium-to-air Heat Exchanger) and the latter is simply called as an AHX(natural-draft sodium-to-Air Heat Exchanger). In a general sodium-to-air heat exchanger design, convection resistance in a shell-side air flow path becomes dominant factor affecting the mechanism of conjugate heat transfer from the sodium flow inside the tube to the air path across the sodium tube wall. Hence verification of the flow and heat transfer characteristics is one of the most important tasks to demonstrate decay heat removal performance. To confirm a kind of ultimate heat sink heat exchanger, a medium-scale Sodium thermal-hydraulic Experiment Loop for Finned-tube sodium-to-Air Heat exchanger (here after called the SELFA) has been designed and is recently being constructed at KAERI site. The introduction of the flow baffle inside the upper sodium chamber of the model FHX unit in the SELFA facility is briefly proposed and discussed as well. The present study aims at introducing a flow baffle design inside the upper sodium chamber to make more equalized flowrates flowing into each heat transfer tube of the model FHX unit. In the cases without the flow baffle geometry, it was observed lager discrepancies in flowrates at the heat transfer tubes. However it was also found that those kinds of discrepancies could be definitely decreased at around 1/10 by employing a flow baffle.

  5. Numerical Simulation of Three-Dimensional Flow Through Full Passage and Performance Prediction of Nuclear Reactor Coolant Pump

    International Nuclear Information System (INIS)

    Li Ying; Zhou Wenxia; Zhang Jige; Wang Dezhong

    2009-01-01

    In order to achieve the level of self-design and domestic manufacture of the reactor coolant pump (nuclear main pump), the software FLUENT was used to simulate the three-dimensional flow through full passage of one nuclear main pump basing on RNG κ-ε turbulence model and SIMPLE algorithm. The distribution of pressure and velocity of the flow in the impeller's surface was analyzed in different working conditions. Moreover, the performance of the pump was predicted based on the simulation results. The results show that the distributions of pressure and velocity are reasonable in both the working and back face of the blade in the steady working condition. The pressure of the flow is increased from the inlet to the outlet of the pump, and shows the maximal value in the impeller region. Comparatively satisfactory efficiency and head value were obtained in the condition of the pump design. The shaft power of the nuclear main pump is gradually increased with the increase of the flow flux. These results are helpful in understanding the change of the internal flow field in the nuclear main pump, which is of some importance for the pre-exploration and theoretical research on the domestic manufacture of the nuclear main pump. (authors)

  6. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

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

  7. Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Behafarid, F.; Shaver, D. R. [Rensselaer Polytechnic Inst., Troy, NY (United States); Bolotnov, I. A. [North Carolina State Univ., Raleigh, NC (United States); Jansen, K. E. [Univ. of Colorado, Boulder, CO (United States); Antal, S. P.; Podowski, M. Z. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2012-07-01

    The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approach to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)

  8. Investigation of straitified and countercurrent flows in horizontal piping during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bourteele, J.P.

    1980-06-01

    The ECTHOR program consists in a loop having as objective to study the flow regimes in horizontal pipings (stratification, countercurrent flows) in conditions representative of small break transients within commercial PWR. The ECTHOR tests are in process. Experimental results are already available and are presented in this paper: scaling problem, U tube experiments, hot leg experiments, high pressure tests

  9. BWR core response to fluctuations in coolant flow and pressure, with implications on noise diagnosis and stability monitoring

    International Nuclear Information System (INIS)

    Blomstrand, J.H.; Andersson, S.A.

    1982-01-01

    Reactor dynamic tests, utilizing sinuosidal oscillations in pressure and recirculation flow, have been conducted in operating BWRs in Sweden and Finland. Test data recorded, as well as recordings of process noise, have been analyzed in terms of dynamic core properties. The results obtained show good qualitative agreement with model predictions of BWR core dynamics. Model studies can often support interpretation of dynamic information obtained from operating plants. Comparisons between model studies, dynamic tests and process noise may also provide improved understanding of test results and noise patterns; in this way it can be demonstrated that some neutron flux noise is caused by noise in coolant flow and steam flow. From reactor test data nd noise recordings, core stability parameters have been evaluated by a number of methods. These have been found to provide essentially the same results. The cores investigated were found to be very stable under normal operating conditions. In special operating points, outside the normal operating range, higher decay ratios may occur. The experience indicates that for BWR cores, operated at decay ratios above quarter damping, the stability parameters may be identified from the oscillatory behavior of the autocorrelation in the time domain of the neutron flux noise

  10. Thermal experiments with LMFBR subassembly models in sodium flow

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.

    1982-01-01

    Within the framework of the Fast Breeder Project research work has been undertaken at the Karlsruhe Nuclear Research Center on the thermal and fluid dynamics of nominal and distorted core subassemblies. In 19-rod bundle models (P/D=1.30, W/R=1.38) three-dimensional temperature distributions were measured in the cladding tubes exposed to sodium flow. Results of measurements of the azimuthal temperature profiles of rotated rods in the duct wall zone are indicated for different operating conditions 80 2 , evenly distributed load and oblique load; different axial positions of the spacer grids; and different positions of one bowed rod

  11. The study of two-phase critical flow characteristics in nuclear reactor coolant system

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Chang, Seok Kyu

    1993-01-01

    This report presents the physical characteristics of two-phase critical flow whcih can be occured in a light water nuclear power plant during LOCA and also reviews the critical flow models and their applications in detail. The existing experimental data base are reviewed and classified. The typical critical flow models which have been applied to the computer code for the accident are also reviewed. Some suggestions are presented for the development of advanced analytical models and the extension of useful experimental database. (Author)

  12. Development of two-phase Flow Model, 'SOBOIL', for Sodium

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Chang, Won Pyo; Kim, In Chul; Kwon, Young Min; Lee, Yong Bum

    2000-03-01

    The objective of this research is to develop a sodium two-phase flow analysis model, 'SOBOIL', for the assessment of the initial stage of the KALIMER HCDA (Hypotherical Core Disruptive Accident). The 'SOBOIL' is basically similar to the multi-bubble slug ejection model used in SAS2A[1]. When a bubble is formed within the liquid slug, the bubble fills the whole cross section of the coolant channel except for a film left on the cladding or on the structure. Up to nine bubbles, separated by the liquid slugs, are allowed in the channel at any time. Each liquid slug flow rate in the model is performed in 2 steps. In the first step, the preliminary flow rate in the liquid slug is calculated neglecting the effect of changes in the vapor bubble pressures over the time step. The temperature and pressure distributions, and interface velocity at the interface between the liquid slug and vapor bubble are also calculated during this process. The new vapor temperature and pressure are then determined from the balance between the net energy transferred into the vapor and the change of the vapor energy. The liquid flow is finally calculated considering the change of the vapor pressure over a time step and the calculation is repeated until specified elapsed time is met. Continuous effort, therefore, must be made on the examination and improvement for the model to become reliable. To this end, much interest must be concentrated in the relevant international collaborations for access to a reference model or test data for the verification.

  13. The study of flow resistances in nuclear reactor Maria under coolant boiling conditions

    International Nuclear Information System (INIS)

    Czerski, P.

    1998-01-01

    The report presents hydrodynamic phenomena recorded in experimental work done on WIW-300 installation. In experiments in which critical heat flux was obtained, were observed such phenomena as : flow pattern in two-phase flow, Ledinegg instability and pressure oscillations. The installation WIW-300 and the course of experiments were presented in detail. The observations were the basis for formulation the steam pillow hypothesis. The pressure drop oscillations were presented on graphs in new way. They were interpolated with polynominals. (author)

  14. Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor

    International Nuclear Information System (INIS)

    Zeuch, W.R.; A-Moneim, M.T.

    1979-01-01

    A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the reference accident postulated for the test. Based on a thermal-hydraulics analysis and assumptions for fuel-coolant interaction in the test section, a pressure pulse of 144 MPa maximum pressure and pulse width of 1.32 ms has been calculated as the reference accident. The response of the test loop to the pressure transient was obtained with the ICEPEL and STRAW codes. Modelling of the test section was completed with STRAW and the remainder of the loop was modelled by ICEPEL

  15. Application of TEMPPC code to the IEA-R1 nuclear reactor core hydrothermal calculations operating at 2 MW for determining the minimal coolant flow

    International Nuclear Information System (INIS)

    Frajndlich, R.; Sousa, J.A. de.

    1985-01-01

    A thermohydraulic study of the IEA-R1 nuclear reactor core on steady-state operating condition and forced convection, is presented. The objective of this calculation is to obtain the minimal flow rate of coolant necessary at the reactor core, limited by the temperature associated to the beginning of nucleate boiling over the fuel plates at a normal operating power (2MW) for a certain inlet coolant temperature. The coolant system safety level is also calculated in this paper, which is divided in three steps: thermohydraulic calculation, without using the uncertainty factors and, after that, considering these factor by two methods: the statistical and the conventional ones. Whichever the method accepted, the results obtained by the program TEMPPC show a great safety margin with respect to the termohydraulic parameters from the IEA-R1 nuclear reactor. (Author) [pt

  16. Study on parameters of self-oscillations of the coolant flow rate in an evaporating channel of a boiling-type reactor

    International Nuclear Information System (INIS)

    Proshutinskij, A.P.; Lobachev, A.G.

    1979-01-01

    The experimental data on the oscillation frequencies and amplitudes of the coolant flow rate at the limit of the thermohydraulic stability of the boiling type reactor evaporating channel are presented. The experiments have been carried out on the channel simulators of three modifications -smooth-tube, with intensifiers of a transverse crimp type and of an inner spiral ribbing type. The range of the investigated regime parameters is as follows: the pressure - 2.5-14MPa; the heat flux density is 0.015-0.8MV/m 2 , mass velocity is 252-2520 kg/(m 2 xs), the temperature at the channel entrance is from 50 deg C up to (tsub(s) -5)deg C. The experimental data analysis is carried out on the assumption that the period of parameter oscillations in the steam generating channel equals the time of the coolant transfer through the channel. The formular is obtained which provides 25% accuracy of the oscillation frequency calculation in the range of underheating parameter variation B=0.5-3.0. As a result the following conclusions have been made: the oscillation frequency of the coolant flow rate is connected with the time of its transfer through the channel and does not practically depend on the type of the heat exchange intensifiers and the degree of the flux throttling at the channel entrance; the self-oscillation amplitude of the coolant flow rate depends on the regime and structural parameters as well

  17. Erosion of sodium bentonite by flow and colloid diffusion

    International Nuclear Information System (INIS)

    Moreno, L.; Liu, L.; Neretnieks, I.

    2011-01-01

    Smectite gel formed at the outer part of a bentonite buffer in granitic rock could expand into rock fractures with seeping water. Such a gel can release colloids into low ionic strength waters. In addition the gel/sol can itself slowly flow downstream when it has reached a low particle concentration sufficient to decrease the viscosity to allow flow. The erosion due to the combined effects of particle diffusion and gel/sol flow is modelled for a thin fracture into which the gel expands influenced by various forces between and on particles. Some of the forces such as the electrical double layer force and viscous force are strongly influenced by the ionic strength of the pore water. Changes in the ionic strength due to diffusion and dilution of ions in the expanding clay are modelled simultaneously with the gel expansion, flow of gel and colloid release to the seeping water. The model includes description of flow of the seeping fluid, which gradually turns from pure water to sol to more dense gel as the smectite source is approached. The model also describes expansion of the gel/sol and colloid release and flow and diffusion of ions in the system. The coupled models are solved using a numerical code. The results show that the gel will flow with a non-negligible flowrate when its volume fraction is below 1%, but that the erosion and loss of smectite is not much influenced by the concentration of sodium in the clay or in the approaching seeping water, if they are kept below the Critical Coagulation Concentration, CCC. (authors)

  18. Fast Flux Test Facility primary sodium valves

    International Nuclear Information System (INIS)

    Rabe, G.B.; Ezra, B.C.

    1977-01-01

    The design and development of the valves used in the primary sodium coolant loop of the Fast Flux Test Facility is described. One tilting-disk check valve is used in the cold leg of the coolant loop. It is designed to limit flow reversal in the loop while maintaining a low pressure drop during forward flow. Two isolation valves are used in each coolant loop--one in the cold leg and one in the hot leg. They are of the motor-operated swinging-gate type. The design, analysis, and testing programs undertaken to develop and qualify these valves are described

  19. A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gaballah, I.

    1978-09-01

    A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident. A theory was developed for the calculation of a dispersed two phase flow with heat addition in a channel with general area change. The theory was used to study different thermodynamic and gasdynamic processes, which may occur during the emergency cooling after a LOCA of a pressurized water reactor. The basic equations were formulated and solved numerically. The heat transfer mechanism was examined. Calculations have indicated that the radiative heat flux component is small compared to the convective component. A drop size spectrum was used in the calculations. Its effect on the heat transfer was investigated. It was found that the calculation with a mean drop diameter gives good results. Significant thermal non-equilibrium has been evaluated. The effect of different operating parameters on the degree of thermal non-equilibrium was studied. The flow and heat transfer in a channel with cross-sectional area change were calculated. It was shown that the channel deformation affects the state properties and the heat transfer along the channel very strongly. (orig.) 891 GL [de

  20. Assessment of RANS at low Prandtl number and simulation of sodium boiling flows with a CMFD code

    Energy Technology Data Exchange (ETDEWEB)

    Mimouni, S., E-mail: stephane.mimouni@edf.fr; Guingo, M.; Lavieville, J.

    2017-02-15

    Highlights: • Modelling of boiling sodium flows in a multiphase flow solver. • Rod heated with a constant heat flux in a pipe liquid metal flow. • Sodium boiling flow around a rod heated with a constant heat. • Computations in progress in an assembly constituted of 19 pins equipped with a wrapped wire. - Abstract: In France, Sodium-cooled Fast Reactors (SFR) have recently received a renewed interest. In 2006, the decision was taken by the French Government to initiate research in order to build a first Generation IV prototype (called ASTRID) by 2020. The improvement in the safety of SFR is one of the key points in their conception. Accidental sequences may lead to a significant increase of reactivity. This is for instance the case when the sodium coolant is boiling within the fissile zone. As a consequence, incipient boiling superheat of sodium is an important parameter, as it can influence boiling process which may appear during some postulated accidents as the unexpected loss of flow (ULOF). The problem is that despite the reduction in core power, when boiling conditions are reached, the flow decreases progressively and vapour expands into the heating zone. A crucial investigating way is to optimize the design of the fissile assemblies of the core in order to lead to stable boiling during a ULOF accident, without voiding of the fissile zone. Moreover, in order to evaluate nuclear plant design and safety, a CFD tool has been developed at EDF in the framework of the nuclear industry. Advanced models dedicated to boiling flows have been implemented and validated against experimental data for ten years now including a wall law for boiling flows, wall transfer for nucleate boiling, turbulence and polydispersion model. This paper aims at evaluating the generalization of these models to SFR. At least two main issues are encountered. Firstly, at low Prandtl numbers such as those of liquid metal, classical approaches derived for unity or close to unity fail to

  1. Bypass valve and coolant flow controls for optimum temperatures in waste heat recovery systems

    Science.gov (United States)

    Meisner, Gregory P

    2013-10-08

    Implementing an optimized waste heat recovery system includes calculating a temperature and a rate of change in temperature of a heat exchanger of a waste heat recovery system, and predicting a temperature and a rate of change in temperature of a material flowing through a channel of the waste heat recovery system. Upon determining the rate of change in the temperature of the material is predicted to be higher than the rate of change in the temperature of the heat exchanger, the optimized waste heat recovery system calculates a valve position and timing for the channel that is configurable for achieving a rate of material flow that is determined to produce and maintain a defined threshold temperature of the heat exchanger, and actuates the valve according to the calculated valve position and calculated timing.

  2. Regulation of the flow rate of liquid-metal coolants on experimental stands

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Laptev, G.I.

    1988-01-01

    Systems for automatic regulation of the flow rate of alkali metals, based on the series ENIV, VIN, and TsLIN three-phase electromagnetic pumps with a pumping rate of 0.5-200 m 3 per hour, were evaluated. The stability of each system was investigated by the method of undamped oscillations. The possibility of employing the analog temperature regulators VRT-2, RPA-T, and R113 was assessed. The functions performed by the most suitable automatic regulation unit, the RPA-T, were described. The limiting period of flow rate oscillations with a maximum gain of the RPA-T in alkali metal regulation systems equaled about 0.5 sec and the minimum integration time of the RPA-T was an order of magnitude longer than the optimal interval. Use of the systems on experimental stands enabled raising the quality of the studies and expanding the zone of servicing of the facilities by the same personnel

  3. The study of flow resistance in nuclear reactor Maria under coolant boiling condition

    International Nuclear Information System (INIS)

    Czerski, P.

    1999-01-01

    This study describes an analysis of experiments carried out in the WIW-300 installation located in the Institute of Atomic Energy (Swierk, Poland). The flow, simulated in the annular gap of test section, was similar to the flow in Maria reactor fuel channel. Experimental character of the work lead to the conclusions related to the physical nature of the hydrodynamic phenomena investigated as well as to the practical aspects of future research. A hypothesis defining a cause of pressure changes was formulated and specific problems related to the mathematical model were defined. The analysis shows that hydrodynamic phenomena studies are of basic significance for the prediction of burnout effects and that heat exchange is very often determined by local phenomena. All described observations are the base for further research on thermodynamic aspects of investigated phenomena. (author)

  4. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  5. Determination of primary flow by correlation of temperatures of the coolant; Medicion de caudal primario por correlacion de temperaturas del refrigerante

    Energy Technology Data Exchange (ETDEWEB)

    Villanueva, Jose [Comision Nacional de Energia Atomica, Ezeiza (Argentina). Centro Atomico Ezeiza

    2003-07-01

    Correlation techniques are often used to assess primary coolant flow in nuclear reactors. Observable fluctuations of some physical or chemical coolant properties are suitable for this purpose. This work describes a development carried out at the National Atomic Energy Commission of Argentina (CNEA) to apply this technique to correlate temperature fluctuations. A laboratory test was performed. Two thermocouples were installed on a hydraulic loop. A stationary flow of water circulated by the mentioned loop, where a mechanical turbine type flowmeter was installed. Transit times given by the correlation flowmeter, for different flow values measured with the mechanical flowmeter, were registered and a calibration between them was done. A very good linear behavior was obtained in all the measured range. It was necessary to increase the fluctuation level by adding water at different temperatures at the measuring system input. (author)

  6. Influence of flowing sodium on the creep rupture behavior of type 304 SS at 550 degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Huthmann, H; Grosser, E D [INTERATOM, Bergisch Gladbach (Germany); Tas, H [SCK/CEN, Mol (Belgium); Borgstedt, H U [GfK, Karlsruhe (Germany)

    1977-07-01

    Substantial components of Liquid Metal Fast Breeder Reactors operate in a temperature range, where the knowledge of time dependent properties is necessary for design purposes. Current high temperature design rules for this temperature range are based on material properties data, which have been obtained in air tests. During the exposure to the flowing sodium environment at elevated temperatures the structural materials are subjected to compositional and microstructural changes, which could alter the mechanical behavior as compared to that in air tests. Within the SNR 300 project a cooperative program between INTERATOM (Federal Republic of Germany), the Nuclear Research Center, GfK, of Karlsruhe and the Studiecentrum voor Kernenergie, SCK/CEN, in Belgium Is conducted to determine quantitatively the influence of sodium environment on the creep-rupture behavior of Type 304 stainless steel. This program is aimed to identify design properties reflecting the possible influence of the coolant. In the present paper, preliminary data are presented on the effect of dynamic un-isothermal sodium on the uniaxial creep-rupture behavior of Type 304 stainless steel at 550 deg. C.

  7. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  10. Calculation of local characteristics of velocity field in turbulent coolant flow in fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Muehlbauer, P.

    1981-08-01

    Experience is described gained with the application of computer code VELASCO in calculating the velocity field in fast reactor fuel assemblies taking into account configuration disturbances due to fuel pin displacement. Theoretical results are compared with the results of experiments conducted by UJV on aerodynamic models HEM-1 (model of the fuel assembly central part) and HEM-2 (model of the fuel assembly peripheral part). The results are reported of calculating the distribution of shear stress in wetted rod surfaces and in the assembly wall (model HEM-2) and the corresponding experimental results are shown. The shear stress distribution in wetted surfaces obtained using the VELASCO code allowed forming an opinion on the code capability of comprising local parameters of turbulent flow through a fuel rod bundle. The applicability was also tested of the code for calculating mean velocities in the individual zones, eg., in elementary cells. (B.S.)

  11. Development of a model to predict flow oscillations in low-flow sodium boiling

    International Nuclear Information System (INIS)

    Levin, A.E.; Griffith, P.

    1980-04-01

    Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed

  12. MIXING LOSSES INVESTIGATION DOWNSTREAM OF TURBINE BLADE CASCADE WITH COOLANT FLOW BLOWING

    Directory of Open Access Journals (Sweden)

    ASSIM HAMEED YOUSIF

    2011-04-01

    Full Text Available A major cause of noise and vibration characteristics of turbomachinery has caused by wakes. The characteristics of the wake, the wake decay, the path that it follows, and the mechanisms of mixing losses generated due to the mixing of blade trailing edge cold jet issued into the hot cross flow are important to find adequate solution to the problem. At the present work the wake characteristic was observed by introducing experimental work inside a cascade test rig to investigate the wake domain downstream of blade cascade with the aid of five-hole probe. The case studies were done with cold jets blowing ratios 1.58, 1.667 and 1.935 with jet stream wise angle and jet lateral injection angle 37.5° and 35 º respectively. The measurement showed that there is a certain harmonization in the region of high reverse pressure loss coefficient which reflects the concentration of wake region. Also it was observed three distinct wake regions located in the centre of the passage vortex region. The wake characteristics measurements of the movement path, the growth of wake width, and the physical awareness of the wake propagating may help to explain the mechanisms of mixing losses.

  13. Renal blood flow, early distal sodium, and plasma renin concentrations during osmotic diuresis

    DEFF Research Database (Denmark)

    Leyssac, P P; Holstein-Rathlou, N H; Skøtt, O

    2000-01-01

    .6 mmHg. Urine flow increased 10-fold, and sodium excretion increased by 177%. Plasma renin concentration (PRC) increased by 58%. Renal blood flow and glomerular filtration rate decreased, however end-proximal flow remained unchanged. After a similar volume of hypotonic glucose (152 mM), ED......(NaCl) increased by 3.6 mM, (P renal hemodynamics, urine flow, sodium excretion rate, or PRC. Infusion of 300 micromol NaCl in a smaller volume caused ED(NaCl) to increase by 6.4 mM without significant changes in PRC. Urine flow and sodium excretion increased significantly...

  14. The development of NRTM-turbine flow meter and measurement of the coolant flow rate in-core of 5 MW heating reactor

    International Nuclear Information System (INIS)

    Zha Meisheng; Wang Xiuqin; Ni Mengchen

    1995-01-01

    In order to measure the coolant flow rate in-core of 5 MW Heating Reactor the special turbine flowmeter of the type of NRTM has been developed. It consists of a body, a turbine with long screw blade and six pieces of Alnico magnets, and a coil mounted on the body. The advantage of this turbine flowmeter is of low resistance and long working-life. Another advantage is that when the turbine is working or not working its factor of resistance is about the same. It is very important for a natural circulation heating reactor. Because the cable, which is welded to the coil assembly, is long enough to extend out of the reactor vessel to the control room, the signal of flow rate is easy to be disturbed by noise in the case. The traditional method of counting the frequency of the A-C voltage which is induced in the coil has a poor ability for resisting noise. The method of the frequency-spectrum analysis of the frequency of the A-C voltage is used to make sure the accuracy of the measurement of the turbine flow meter. Compared with the method of the count it has a good ability for resisting noise. After three years operation a lot of valuable data were obtained

  15. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

  16. Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

    International Nuclear Information System (INIS)

    Komen, E.M.J.; Koning, H.

    1995-05-01

    This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

  17. Thermal response of core and central-cavity components of a high-temperature gas-cooled reactor in the absence of forced convection coolant flow

    International Nuclear Information System (INIS)

    Whaley, R.L.; Sanders, J.P.

    1976-09-01

    A means of determining the thermal responses of the core and the components of a high-temperature gas-cooled reactor after loss of forced coolant flow is discussed. A computer program, using a finite-difference technique, is presented together with a solution of the confined natural convection. The results obtained are reasonable and demonstrate that the computer program adequately represents the confined natural convection

  18. Evaluation of conservatism in analysis of fuel-coolant interaction

    International Nuclear Information System (INIS)

    Reynolds, A.B.; Erdman, C.A.; Garner, P.L.; Haas, P.M.; Allen, C.L.

    Using the ANL parametric model developed by Cho e.a. the following mechanisms and parameters involved in fuel-coolant interaction were examined: coherence of fuel-sodium mixing; two-phase heat transfer; sodium-to-fuel mass ratio; fuel particle size; heat transfer to plenum and core cladding; constraint geometry. Both overpower and loss-of-flow transients were studied. Main attention is given to the maximum mechanical work to be expected. As a general conclusion, it can be stated that more realistic models will result in a reduction of the estimated mechanical work

  19. Experimental study on combustion characteristics of sodium fire in a columnar flow

    International Nuclear Information System (INIS)

    Zhang Zhigang; Peng Kangwei; Guo Ming; Huo Yan

    2014-01-01

    In the operation of the sodium-cooled fast reactor, the accident caused by the leakage and combustion of liquid sodium is common and frequent in sodium-related facilities. This paper is based on an experimental study of sodium fire in a columnar flow, which was carried out to focus on the burning characteristics by analyzing the temperature fields in the burner. The injection of 200°C liquid sodium with the flux of 0.5 m 3 /h was poured into a 7.9 m 3 volume stainless steel cylindrical burner to shape a sodium fire, and the data of temperature fields in the burner have been collected by dozens of thermocouples which are laid in the combustion space and sodium collection plate. These results show that the sodium fire in a columnar flow is composed of the foregoing centered columnar fire, the subsequent spray fire caused by atomization and the pool fire on the collection plate. The temperature close to the burning sodium flow maximally reaches up to 950°C. The radial temperatures apart from the sodium flow are relatively low and generally about 200°C, and maximally just 300°C even when close to the sodium collection plate. The maximum temperature of the burning sodium dropping on the collection plate rises in the center of plate, about 528°C. This study is helpful to evaluate the combustion characteristics, formation process and composing forms of the sodium fire in the sodium-related facilities. (author)

  20. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

    2017-08-15

    The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  1. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

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

    International Nuclear Information System (INIS)

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

    1976-04-01

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

  3. Experiment studies of fuel rod vibration in coolant flow for substantiation of vibration stability of fuel rods with no fretting-wear

    International Nuclear Information System (INIS)

    Egorov, Yu. V.; Afanasiev, A. V.; Makarov, V. V.; Matvienko, I. V.

    2013-01-01

    For substantiation of vibration stability it is necessary to determine the ultimate permissible vibration levels which do not cause fretting, to compare them with the level of fuel rod vibration caused by coolant flow. Another approach is feasible if there is experience of successful operation of FA-prototypes. In this case in order to justify vibration stability it may be sufficient to demonstrate that the new element does not cause increased vibration of the fuel rod. It can be done by comparing the levels of hydro-dynamic fuel rod vibration and FA new designs. Program of vibration tests of TVS-2M model included studies of forced oscillations of 12 fuel rods in the coolant flow in the spans containing intensifiers, in the reference span without intensifiers, in the lower spans with assembled ADF and after its disassembly. The experimental results for TVS-2M show that in the spans with intensifier «Sector run» the level of movements is 6% higher on the average than in the span without intensifiers, in the spans with intensifier «Eddy» it is 2% higher. The level of fuel rod vibration movements in the spans with set ADF is 2 % higher on the average than without ADF. During the studies of TVS-KVADRAT fuel rod vibration, the following tasks were solved: determination of acceleration of the middle of fuel rod spans at vibration excited due to hydrodynamics; determination of influence of coolant thermal- hydraulic parameters (temperature, flowrate, dynamic pressure) on fuel rod vibration response; determination of influence of span lengths on the vibration level. Conclusions: 1) The vibration tests of the full-scale model of TVS-2M in the coolant flow showed that the new elements of TVS-2M design (intensifiers of heat exchange and ADF) are not the source of fuel rod increased vibration. Considering successful operation of similar fuel rod spans in the existing TVS-2M design, vibration stability of TVS-2M fuel rods with new elements is ensured on the mechanism of

  4. Sodium confluent rates of flow values, on 0,5 mixer, of a sodium italian SS-050 circuit component

    International Nuclear Information System (INIS)

    Walsh, L.M.

    1987-01-01

    Sodium lines on different temperatures, during an emergency drainage on 0,5 mixer was found. To future valuation by DIMEC of tensions that occurs on that component of SS-050, the confluent rates of flow values were calculated. (L.M.J.) [pt

  5. Sodium

    Science.gov (United States)

    Table salt is a combination of two minerals - sodium and chloride Your body needs some sodium to work properly. It helps with the function ... in your body. Your kidneys control how much sodium is in your body. If you have too ...

  6. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  7. Research of the fluid flow in a radially orientated coolant channel of a turbine blade; Untersuchung der Stroemung in einem radial gerichteten Kuehlkanal eines Turbinenlaufrades

    Energy Technology Data Exchange (ETDEWEB)

    Hein, O.

    1999-07-01

    Due to rotation (Coriolis forces) in a coolant channel a secondary flow is superimposed to the basic flow. This leads to a change in the local heat transfer over the surface of the coolant channel as well as a change in the overall value of the heat transfer. Also the pressure loss over the channel length will change by rotation. By means of computational fluid dynamics (Finite Element Method) it was achieved to figure out the interaction between changing fluid flow and heat transfer. To validate the results obtained by a numerical flow simulation, a new measurement technique was developed. A laser-two-focus velocimeter has been combined with a rotation prism which allows continued measurements in a rotating scaled up channel. (orig.) [German] Bedingt durch die Rotationsbewegung eines Kuehlkanals wird die Grundstroemung von einem Sekundaerwirbel ueberlagert (Corioliskraefte). Durch diese Einfluesse aendert sich sowohl der lokale Waermeuebergang ueber der Kanaloberflaeche als auch die globalen Waermeuebertragungsraten ueber dem gesamten Kanal. Ebenfalls aendert sich durch die Rotation der Druckverlust ueber der Kanallaenge. Durch eine numerische Stroemungssimulation (Finite-Element-Methode) war es moeglich, einen detaillierten Zusammenhang zwischen dem veraenderten Stroemungsverhalten und dem Waermeuebertragungsverhalten darzustellen. Um die numerisch gewonnenen Ergebnisse experimentell abzusichern, wurde eine neuartige Messtechnik entwickelt. Ein Laser-2-Fokus-Velozimeter wurde mit einem Bilddrehprisma kombiniert, und dies erlaubte eine kontinuierliche Messung in einem rotierenden vergroesserten Modellkanal. (orig.)

  8. Method of preventing sodium from flowing when pipes of a fast breeder reactor are injured

    International Nuclear Information System (INIS)

    Nakai, Yasushi; Yamagishi, Yoshiaki; Koga, Tomonari.

    1975-01-01

    Object: To inject high pressure sodium into an inlet nozzle portion when fluid pressure in the inlet nozzle portion of a core cooling pipe on the inlet side is in an abnormal condition, to thereby quickly and positively prevent the flow of sodium in a high pressure chamber in a reactor vessel, when pipes are injured. Structure: When the core cooling pipe on the inlet side is injured and as a consequence the pressure gage detects an abnormal condition of fluid pressure in the inlet nozzle, the valve is opened to allow high pressure sodium to inject into the inlet nozzle through a high pressure sodium supply pipe, thereby blocking a back-flow of sodium in the high pressure chamber into the core cooling pipe. (Kamimura, M.)

  9. Optimal estimate of the coolant flow in the assemblies of a BWR of natural circulation in real time; Estimacion optima del flujo de refrigerante en los ensambles de un BWR de circulacion natural en tiempo real

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Division de Estudios de Posgrado, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Espinosa P, G., E-mail: julfi_jg@yahoo.com.mx [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2012-10-15

    The present work exposes the design and the implementation of an advanced controller that allows estimating the coolant flow in the fuel assemblies of a BWR reactor of natural circulation in real time. To be able to reduce the penalizations that are established in the calculations of the operation limits due to the magnitude of the uncertainties in the coolant flows of a natural circulation reactor, is the objective of this research. In this work the construction of the optimal controller that allows estimating the coolant flows in a fuel channels group of the reactor is shown, as well as the operation of this applied to a reduced order model that simulates the dynamics of a natural circulation reactor. The controller design required of an estimator of the valuation variables not directly of the plant and of the estimates use of the local distributions of the coolant flow. The controller construction of the estimator was based mathematically in the filter Kalman whose algorithm allows to be carried out an advanced control of the system. To prove the estimator operation was development a simplified model that reproduces the basic dynamics of the flowing coolant in the reactor, which works as observer of the system, this model is coupled by means of the estimator controller to a detail model of the plant. The results are presented by means of graphics of the interest variables and the estimate flow, and they are documented in the chart that is presented at the end of this article. (Author)

  10. Cavitation erosion in a 400 deg. C sodium flow

    International Nuclear Information System (INIS)

    Courbiere, P.

    2002-01-01

    This paper presents the results of cavitation erosion tests conducted in the Cavitation Tunnel at the Cadarache Nuclear Research Center. The CANASTA system was used for acoustic monitoring of cavitation noise during the experiments. Comparative results are also presented for sodium and water tests. (author)

  11. Generation of a Magnetic Field by Dynamo Action in a Turbulent Flow of Liquid Sodium

    International Nuclear Information System (INIS)

    Monchaux, R.; Chiffaudel, A.; Daviaud, F.; Dubrulle, B.; Gasquet, C.; Marie, L.; Ravelet, F.; Berhanu, M.; Fauve, S.; Mordant, N.; Petrelis, F.; Bourgoin, M.; Moulin, M.; Odier, Ph.; Pinton, J.-F.; Volk, R.

    2007-01-01

    We report the observation of dynamo action in the von Karman sodium experiment, i.e., the generation of a magnetic field by a strongly turbulent swirling flow of liquid sodium. Both mean and fluctuating parts of the field are studied. The dynamo threshold corresponds to a magnetic Reynolds number R m ∼30. A mean magnetic field of the order of 40 G is observed 30% above threshold at the flow lateral boundary. The rms fluctuations are larger than the corresponding mean value for two of the components. The scaling of the mean square magnetic field is compared to a prediction previously made for high Reynolds number flows

  12. Experimental study on combustion and suppression characteristics of sodium fire in a columnar flow using extinguishing powder

    International Nuclear Information System (INIS)

    Huo Yan; Zhang Zhigang; Li Jinke; Liu Zhongkun; Ma Yaolong

    2017-01-01

    In the operation of the sodium-cooled fast reactor, the leakage and fire accident of liquid sodium is common and it is frequent in sodium-related facilities. This study focuses on the combustion and suppression characteristics of sodium fire in a columnar flow. Liquid sodium (250°C) is injected into a 7.9 m"3 cylindrical chamber at a flow rate of about 1.0 m"3/h to create a columnar sodium fire, and 18.4 kg class D extinguishing powder is sprayed after the liquid sodium injection. The temperature in the chamber space and sodium collection plate and the heat release rate from sodium fire are measured and analyzed. Based on the temperature data the sodium fire under suppression could be divided into four phases of dropping sharply, continuously remaining lower, rising and declining mildly, and depressing. The sodium fire in the space could be suppressed and cooled down if the extinguishing agent could spray in the early period of the liquid sodium injection. The extinguishing agent could suppress the combustion and spreading of liquid sodium dropping on the collection plate, limit the pool combustion area and postpone the commencement of sodium pool burning in spite of its later re-ignition happening. This study promises to evaluate the combustion and suppression characteristics of sodium fire in the sodium-related facilities. (author)

  13. 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)

  14. Sodium evaporation into a forced argon flow, (2)

    International Nuclear Information System (INIS)

    Kumada, Toshiaki; Ishiguro, Ryoji; Kasahara, Fumio.

    1978-01-01

    The critical supersaturation model (CSM) was examined by comparing its predictions with the experimental data of the forced evaporation of sodium reported in Part (I) of this series of papers. The results revealed that the CSM predicted a fairly lower Sherwood number than that predicted in the experiment. This discrepancy may be due mainly to the oversimplified assumptions in the CSM. A calculation of the fog movement in a boundary layer was also attempted by taking into account the force on fog particles by thermophoresis. The calculation showed that the particles were submerged deeply in the boundary layer and that they absorbed sodium vapor in accordance with their size. It was concluded that the fog movement in the boundary layer was essentially important for the exact estimation of the evaporation rate. (author)

  15. Coolant Passage

    Directory of Open Access Journals (Sweden)

    Tom I.-P. Shih

    2001-01-01

    Full Text Available Computations were performed to study the three-dimensional flow and heat transfer in a U-shaped duct of square cross section with inclined ribs on two opposite walls under rotating and non-rotating conditions. Two extreme limits in the Reynolds number (25,000 and 350,000 were investigated. The rotation numbers investigated are 0, 0.24, and 0.039. Results show rotation and the bend to reinforce secondary flows that align with it and to retard those that do not. Rotation was found to affect significantly the flow and heat transfer in the bend even at a very high Reynolds number of 350,000 and a very low Rotation number of 0:039. When there is no rotation, the flow and heat transfer in the bend were dominated by rib-induced secondary flows at the high Reynolds number limit and by bend-induced pressure-gradients at the low Reynolds number limit. Long streaks of reduced surface heat transfer occur in the bend at locations where streamlines from two contiguous secondary flows merge and then flow away from the surface. The location and size of these streaks varied markedly with Reynolds and rotation numbers.

  16. Numerical investigation on turbulence mixing characteristics under thermal striping flows. Investigations on fluid temperature fluctuation phenomena in air and sodium

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, Satoshi [Customer System Co. Ltd., Tokai, Ibaraki (Japan); Muramatsu, Toshiharu

    1999-05-01

    A three-dimensional thermal striping analysis was carried out using a direct numerical simulation code DINUS-3, for a coaxial jet configuration using air and sodium as a working fluid, within the framework of the EJCC thermo-hydraulic division. From the analysis, the following results have been obtained: (1) Calculated potential core length in air and sodium turbulence flows agreed with a theoretical value (5d - 7d ; d : diameter of jet nozzle) in the two-dimensional free jet theory. (2) Hydraulic characteristics in sodium flows as the potential core length can be estimated by the use of that of air flow characteristics. (3) Shorter thermally potential core length defined by spatial temperature distribution was evaluated in sodium flows, compared with that in air flows. This is due to the higher thermal conductivity of sodium. (4) Thermal characteristics in sodium flows as the thermally potential core length can not be evaluated, based on that air thermal characteristics. (author)

  17. Study on flow-induced vibration of large-diameter pipings in a sodium-cooled fast reactor. Influence of elbow curvature on velocity fluctuation field

    International Nuclear Information System (INIS)

    Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira

    2010-02-01

    The main cooling system of Japan Sodium-cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9m/s in the primary hot leg pipe with large-diameter (1.3m). The maximum Reynolds number in the piping reaches 4.2x10 7 . The hot leg pipe having a 90 degree elbow with curvature ratio of r/D=1.0, so-called 'short elbow', which enables a compact reactor vessel. In sodium cooled fast reactors, the system pressure is so low that thickness of pipings in the cooling system is thinner than that in LWRs. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned at the short elbow. The evaluation of the structural integrity of pipings in JSFR should be conducted based on a mechanistic approach of FIV at the elbow. It is significant to obtain the knowledge of the fluctuation intensity and spectra of velocity and pressure fluctuations in order to grasp the mechanism of the FIV. In this study, water experiments were conducted. Two types of 1/8 scaled elbows with different curvature ratio, r/D=1.0, 1.5, were used to investigate the influence of curvature on velocity fluctuation at the elbow. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow at the elbow outlet and separation flow at the inner wall of elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow. (author)

  18. 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

  19. THE CONTROL ALGORITHM OF THE DRYING PROCESS PARTICULATE MATERIALS IN THE APPARATUS WITH THE SWIRLING FLOW OF COOLANT AND MICROWAVE ENERGY SUPPLY

    Directory of Open Access Journals (Sweden)

    S. T. Antipov

    2015-01-01

    Full Text Available The technical task of the process is to improve the drying quality of the final product, increasing the precision and reliability of control, the reduction of specific energy consumption. One of the ways to improve the process is complex and i ts local automation. This paper deals with the problems of development and creation of a new control algorithm drying process of the particulate material. Identified a number of shortcomings of the existing methods of automatic control of the process. As a result, the authors proposed a method for drying particulate materials in the device with swirling flow and the microwave energy supply and its automatic control algorithm. The description of the operating principle of the drying apparatus consists in that the particulate material is wet by using a tangential flow of coolant supplied to the cylinder-drying apparatus which also serves the axial coolant flow, whereby the heat transfer fluid with the particulate material begins to undergo a complex circular movement along the circumference apparatus, thereby increasing its speed and its operation control algorithm. The work of this scheme is carried out at three levels of regulation on the basis of determining the coefficient of efficiency of the dryer, which makes it possible to determine the optimal value of the power equipment and to forecast the cost of electricity. All of the above allows you to get ready for a high quality product while minimizing thermal energy and material costs by optimizing the operating parameters of the drying of the particulate material in the dryer with a combined microwave energy supply and ensure the rational use of heat energy by varying their quantity depending on the characteristics to be dried particulate material and the course of the process.

  20. Analysis of clad motion during a loss of flow (LOF) accident in a fast sodium cooled reactor

    International Nuclear Information System (INIS)

    Henkel, P.

    1985-10-01

    A new model describing clad motion during a Loss of Flow (LOF) accident in a Liquid Metal Cooled Fast (Breeder) Reactor (LMFBR) is presented. Its special features are Clad motion is treated within a fuel pin bundle. The bundle geometry is represented by an equivalent annular geometry which serves as the descriptional basis for the clad motion analysis; Several flow regimes are considered. These include a wave or film flow along the fuel pin surfaces as well as a drop flow within the coolant channels. A new entrainment criterion is successfully applied to describe the entrainment of molten cladding and the coolant flow is modelled as a two-dimensional, monstationary flow. Therefore, radial cross flows in a pin bundle can be calculated. Especially, thermal incoherency effects can be treated consistently. The analysis of clad motion in the two experiments STAR1 and STAR2 using the subsequently presented SANDCMOT model gives good agreement with the experimental data. (orig.) [de

  1. BACCHUS-3D/SP. A computer programme for the three-dimensional description of sodium single-phase flow in bundle geometry

    International Nuclear Information System (INIS)

    Bottoni, M.; Dorr, B.; Homann, C.; Struwe, D.

    1983-07-01

    The computer programme BACCHUS implemented at KfK includes a steady-state version, a two-dimensional and a three-dimensional transient single-phase flow version describing the thermal-hydraulic behaviour of the coolant (sodium or water) in bundle geometry under nominal or accident conditions. All versions are coupled with a pin model describing the temperature distribution in fuel (or electrical heaters) and cladding. The report describes the programme from the viewpoints of the geometrical model, the mathematical foundations and the numerical treatment of the basic equations. Although emphasis is put on the three-dimensional version, the two-dimensional and the steady state versions are also documented in self-consistent sections. (orig.) [de

  2. Coolant-fuel interaction in Sodium-cooled Fast Reactors: Structural investigations of The Na-An-O (An = U, Np, Pu) systems

    International Nuclear Information System (INIS)

    Smith, A.L.; Raison, P.E.; Bykov, D.M.; Konings, R.J.; Caciuffo, R.; Cheetham, A.K.

    2014-01-01

    Nuclear energy has the potential to provide Europe with a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gases emissions. The interest for Sodium-cooled-Fast-spectrum Reactors (SFRs), when compared to Pressurized Water Reactors (PWRs), lies in their more efficient management of plutonium and other actinides as well as their ability to use almost all of the energy in the natural uranium versus 1% utilized in thermal spectrum systems. The high fuel efficiency of fast reactors could greatly dampen concerns about fuel supply. But these reactors have also several drawbacks when compared to PWRs (i.e sodium fire, Na reaction with O2 and H2O, interaction of sodium with oxide fuels). Their development at an industrial scale needs therefore an exhaustive safety assessment that comprises both experimental work and development of sophisticated modelling tools able to describe the reactor behaviour in normal or incidental conditions

  3. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

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

  4. Decreased cerebral blood flow after administration of sodium bicarbonate in the distressed newborn infant

    DEFF Research Database (Denmark)

    Lou, H C; Lassen, N A; Fris-Hansen, B

    1978-01-01

    In the course of our studies on cerebral blood flow in newborn infants, we have observed a striking depressing effect of sodium bicarbonate infusion on cerebral blood flow which in some cases may severely aggravate cerebral ischemia. We measured cerebral blood flow before and after the treatment...... with 1 to 8 meqs of sodium bicarbonate in seven distressed newborn infants. The 133 Xe clearance technique was used. The results showed in six of the seven cases a decrease in cerebral blood flow, which in most cases was reduced to 14 to 22 ml/100 g/min, which is about half the value prior...... to the bicarbonate infusion. In one case an extreme reduction occurred: cerebral blood flow was reduced to 3 ml/100 g/min, well below the level compatible with tissue survival. The results are discussed with regard to the optimal treatment of the acidotic newborn....

  5. Thermophysical properties of sodium nitrate and sodium chloride solutions and their effects on fluid flow in unsaturated media

    International Nuclear Information System (INIS)

    Xu, Tianfu; Pruess, Karsten

    2001-01-01

    Understanding movement of saline sodium nitrate (NaNO 3 ) waste solutions is important for assessing the contaminant migration near leaking waste storage tanks in the unsaturated zone at the Hanford site (Washington, USA). The purpose of this study is to contribute a basic understanding of effects of the thermophysical behavior of NaNO 3 solutions on fluid flow in unsaturated media. We first present mathematical expressions for the dependence of density, viscosity, solubility and vapor pressure of NaNO 3 solutions on both salt concentration and temperature, which were determined by fitting from published measured data. Because the previous studies of thermophysical behavior of sodium chloride (NaCl) solutions can provide a basis for those of NaNO 3 solutions, we also present a comparison of thermophysical properties of both salt solutions. We have implemented the functional thermophysical properties of NaNO 3 solutions into a new TOUGH2 equation-of-state module EWASG-NaNO 3 , which is modified from a previous TOUGH2 equation-of-state module EWASG for NaCl. Using the simulation tool, we have investigated effects of the thermophysical properties on fluid flow in unsaturated media. The effect of density and viscosity of saline solutions has been long recognized. Here we focus our attention on the effect of vapor pressure lowering due to salinity. We present simulations of a one-dimensional problem to study this salinity-driven fluid flow. A number of simulations were performed using different values of thermal conductivity, permeability, and temperature, to illustrate conditions and parameters controlling these processes. Results indicate that heat conduction plays a very important role in this salinity-driven vapor diffusion by maintaining a nearly constant temperature. The smaller the permeability, the more water is transferred into the saline environment. Effects of permeability on water flow are also complicated by effects of capillary pressure and tortuosity. The

  6. The state-of-the-art report for flow and pressure measurement techniques in the piping system

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Lee, Jeong Han; Kim, Bong Soo; Kim, Jong Man; Choi, Jong Hyun

    2002-12-01

    The sodium is widely used as one of the coolants in the liquid metal reactor system since it has important safety features such as a long thermal response time, a large margin to coolant boiling, and operating in near atmospheric pressure, etc. The state-of-the-art on the flow and pressure measurement techniques in the piping system worldwide is investigated and reviewed to utilize it as a basis for developing a new technique applying for the high temperature sodium flow environment.

  7. Sodium boiling and mixed oxide fuel thermal behavior in FBR undercooling transients; W-1 SLSF experiment results

    International Nuclear Information System (INIS)

    Henderson, J.M.; Wood, S.A.; Knight, D.D.

    1981-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) Experiment was conducted to study fuel pin heat release characteristics during a series of LMFBR Loss-of-Piping Integrity (LOPI) transients and to investigate a regime of coolant boiling during a second series of transients at low, medium and high bundle power levels. The LOPI transients produced no coolant boiling and showed only small changes in coolant temperatures as the test fuel microstructure changed from a fresh, unrestructured to a low burnup, restructured condition. During the last of seven boiling transients, intense coolant boiling produced inlet flow reversal, cladding dryout and moderate cladding melting

  8. 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

  9. Brief description of out-of-pile test facilities for study in corrosion and fission product behaviour in flowing sodium

    International Nuclear Information System (INIS)

    Iizawa, K.; Sekiguchi, N.; Atsumo, H.

    1976-01-01

    The experimental methods to perform tests for study in corrosion and fission products behaviour in flowing sodium are outlined. Flow diagrams for the activated materials and fission products behaviour test loop are given

  10. Flow induced vibration of the large-sized sodium valve for MONJU

    International Nuclear Information System (INIS)

    Sato, K.

    1977-01-01

    Measurements have been made on the hydraulic characteristics of the large-sized sodium valves in the hydraulic simulation test loop with water as fluid. The following three prototype sodium valves were tested; (1) 22-inch wedge gate type isolation valve, (2) 22-inch butterfly type isolation valve, and (3) 16-inch butterfly type control valve. In the test, accelerations of flow induced vibrations were measured as a function of flow velocity and disk position. The excitation mechanism of the vibrations is not fully interpreted in these tests due to the complexity of the phenomena, but the experimental results suggest that it closely depends on random pressure fluctuations near the valve disk and flow separation at the contracted cross section between the valve seat and the disk. The intensity of flow induced vibrations suddenly increases at a certain critical condition, which depends on the type of valve and is proportional to fluid velocity. (author)

  11. Flow induced vibration of the large-sized sodium valve for MONJU

    Energy Technology Data Exchange (ETDEWEB)

    Sato, K [Sodium Engineering Division, O-arai Engineering Centre, Power Reactor and Nuclear Fuel Development Corporation, Nariata-cho, O-arai Machi, Ibaraki-ken (Japan)

    1977-12-01

    Measurements have been made on the hydraulic characteristics of the large-sized sodium valves in the hydraulic simulation test loop with water as fluid. The following three prototype sodium valves were tested; (1) 22-inch wedge gate type isolation valve, (2) 22-inch butterfly type isolation valve, and (3) 16-inch butterfly type control valve. In the test, accelerations of flow induced vibrations were measured as a function of flow velocity and disk position. The excitation mechanism of the vibrations is not fully interpreted in these tests due to the complexity of the phenomena, but the experimental results suggest that it closely depends on random pressure fluctuations near the valve disk and flow separation at the contracted cross section between the valve seat and the disk. The intensity of flow induced vibrations suddenly increases at a certain critical condition, which depends on the type of valve and is proportional to fluid velocity. (author)

  12. Sodium fires and its extinguishment

    International Nuclear Information System (INIS)

    Mikhedov, V.G.

    1979-01-01

    The fire safety problems of NPP with sodium coolants in USSR are presented. The design of sodium reactors is made with premises with sodium coolants being hermetic and filled with nitrogen. Some engineering solutions of fire safety including design, elaboration and choice of construction and protection materials are presented. Some theoretical aspects of sodium burning are presented as well as methods of sodium fire extinguishing methods including the use of powder

  13. sodium

    International Development Research Centre (IDRC) Digital Library (Canada)

    Les initiatives de réduction de la consommation de sel qui visent l'ensemble de la population et qui ciblent la teneur en sodium des aliments et sensibilisent les consommateurs sont susceptibles de réduire la consommation de sel dans toutes les couches de la population et d'améliorer la santé cardiovasculaire. Ce projet a ...

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Decreased cerebral blood flow after administration of sodium bicarbonate in the distressed newborn infant

    DEFF Research Database (Denmark)

    Lou, H C; Lassen, N A; Fris-Hansen, B

    1978-01-01

    with 1 to 8 meqs of sodium bicarbonate in seven distressed newborn infants. The 133 Xe clearance technique was used. The results showed in six of the seven cases a decrease in cerebral blood flow, which in most cases was reduced to 14 to 22 ml/100 g/min, which is about half the value prior...

  16. BLOW-3A. A theoretical model to describe transient two-phase flow conditions in LMFBR coolant channels

    International Nuclear Information System (INIS)

    Bottoni, M.; Struwe, D.

    1982-12-01

    The computer programme BLOW-3A describes sodium boiling phenomena in subassemblies of fast breeder reactors as well as in in-pile or out-of-pile experiments simulating different failure conditions. This report presents a complete documentation of the code from three main viewpoints: the theoretical foundations of the programme are first described with particular reference to the most recent developments; the structure of the programme is then explained in all details necessary for the user to get a rapid acquaintance with it; eventually several examples of the programme validation are discussed thus enabling the reader to acquire a full picture of the possible applications of the code and at the same time to know its validity range. (orig.) [de

  17. Development of a model to predict flow oscillations in low-flow sodium boiling. [Loss-of-Piping Integrity accidents

    Energy Technology Data Exchange (ETDEWEB)

    Levin, A.E.; Griffith, P.

    1980-04-01

    Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed.

  18. Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chenu, A.

    2011-10-01

    Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

  19. Heat transfer characteristics of rectangular coolant channels with various aspect ratios in the plasma-facing components under fully developed MHD laminar flow

    International Nuclear Information System (INIS)

    Takase, K.; Hasan, M.Z.

    1995-01-01

    Convective heat transfer in MHD laminar flow through rectangular channels in the plasma-facing components of a fusion reactor has been analyzed numerically to investigate the effects of channel aspect ratio, defined as the ratio of the lengths of the plasma-facing side to the other side. The adverse effect of the nonuniformity of surface heat flus on Nusselt number (Nu) at the plasma-facing side can be alleviated by increasing the aspect ratio of a rectangular duct. At the center and corner of the plasma-facing side of a square duct, the Nu of non-MHD flow are 6.8 and 2.2, respectively, for uniform surface heat flux. In the presence of a strong magnetic field, Nu at the center and corner increases to 22 and 3.6, respectively. However, when the heat flux is highly nonuniform, as in the plasma-facing components, Nu decreases from 22 to 3.1 at the center and from 3.6 to 3.1 at the corner. When the aspect ratio is increased to 4, Nu at the center and corner increase to 5 and 4.7. Along the circumference of a rectangular channel, there are locations where the wall temperature is equal to or less than the bulk coolant temperature, thus making the Nu with conventional definition infinity or negative. The ratio between Nu of MHD flow and Nu of non-MHD flow for various aspect ratios is constant in the region of Hartmann number of more than 200 at least. On the other hand, its ratio increases monotonously with increasing the aspect ratio

  20. Measurement and analysis of vibrational behaviour of an SNR-fuel element in sodium flow

    International Nuclear Information System (INIS)

    Hess, B.F.H.; Ruppert, E.; Schmidt, H.; Vinzens, K.

    1975-01-01

    Within the framework of SNR-300 fuel element development programme a complete full size fuel element dummy has been tested thoroughly for nearly 3000 hours at 650 0 C system temperature in the AKB sodium loop at Interatom, Bensberg. Investigations of the hydraulic characteristics by measurements of specific pressure losses, flow velocities, leakage flow through the piston rings and investigations of its vibrational behaviour were part of this endurance test at elevated temperatures. The pressure drop versus flow and the leakage measurement are mentioned briefly to confirm the correctness of the test hydraulics. The vibrational behaviour of the element and the approach to analysis is the main object of this report. (Auth.)

  1. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

  2. USSR position paper for the specialists' meeting on thermal stratification in sodium

    International Nuclear Information System (INIS)

    Rinejskij, A.A.; Shcherbakov, S.I.

    1983-01-01

    The problem of adequate coolant mixing and flow pattern optimisation becomes more significant while considering pool type fast reactors. Within a vessel of this reactor type there are considerable bulks of coolant at different temperatures. Because of the complex design requirements certain difficulties exist in the problem of ensuring sufficient mixing of the coolant at the outlet of the core, radial blanket, in-pile fuel storage and in-pile neutron shielding. n the framework of the R and D activities carried out to justify fast reactor design solutions a proper attention is given to investigations of the coolant flow pattern within plena and in those sections of pool type reactor, where the coolant now takes place. In this connection It should be noted that, up to now, we did not obtain any clear evidence of the presence of a pronounced thermal stratification of sodium within reactor mixing chambers under normal operating conditions. The conclusion may be drawn on the basis of our experiments and design experience that there are certain gradients of temperature and Inefficient mixing of coolant flows with different temperatures within the chambers (both 'hot' and 'cold' ones). All these factors to a certain degree complicate a reactor operation. However, from our point of view, the main point in the problem of arranging an 'optimum' coolant flow pattern within the chambers is to insure the stability of coolant motion in the chambers (Including flows with different temperatures) by means ol special design solutions based on results of experimental and calculation studies. It is considered that the instability of coolant flow could be the most important factor influencing temperature distribution within structures and components. That is why coolant flow in all reactor chambers and cavities should be stable, including natural convection flows

  3. 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

  4. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

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

  5. High energy resolution and high count rate gamma spectrometry measurement of primary coolant of generation 4 sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Coulon, R.

    2010-01-01

    Sodium-cooled Fast Reactors are under development for the fourth generation of nuclear reactor. Breeders reactors could gives solutions for the need of energy and the preservation of uranium resources. An other purpose is the radioactive wastes production reduction by transmutation and the control of non-proliferation using a closed-cycle. These thesis shows safety and profit advantages that could be obtained by a new generation of gamma spectrometry system for SFR. Now, the high count rate abilities, allow us to study new methods of accurate power measurement and fast clad failure detection. Simulations have been done and an experimental test has been performed at the French Phenix SFR of the CEA Marcoule showing promising results for these new measurements. (author) [fr

  6. Single-phase coolant flow CFD simulations inside the CANDU channel for the 37 and the 43 elements bundles

    International Nuclear Information System (INIS)

    Pauna, E.; Olteanu, G.; Catana, A.

    2013-01-01

    In this paper, a Computation Fluid Dynamics (CFD) simulation was performed in order to find the flow conditions in the CANDU Channel for the standard (37 elements) and the new designed bundle (43 elements) using the CFD Code S aturne software. Due to the fact that the code is a single-phase one it was considered an inlet temperature of 250 O C, a flow rate of 24.17 kg/s, an outlet pressure of 10.3 MPa and a linear power of 800 kW/m. The flow conditions were achieved by using a CFD typical chain of steps which was performed starting from preprocessing (geometry, mesh and boundary conditions), through solver and post-processing. Open Source platform (Salome-Meca geometry and mesh modules, the Code S aturne solver, Paraview and Visit for post-processing) were used as computational tool kit and an unsteady state was considered. Some simplifications were considered: the tube creep was not taken into account and all the bundles were considered aligned. The three dimensional thermal-hydraulic distributions of the temperature, pressure and velocity parameters offered information for the geometry comparison and the results were in agreement with some experimental data. CFD analysis results provided valuable data regarding the thermal-hydraulic operating conditions inside the CANDU reactor channel. (authors)

  7. Local velocity measurements in lead-bismuth and sodium flows using the ultrasound doppler velocimetry

    International Nuclear Information System (INIS)

    Eckert, S.; Gerbeth, G.

    2003-01-01

    We will present measurements of the velocity profiles in liquid sodium and eutectic lead-bismuth by means of the Ultrasonic Doppler Velocimetry (UDV). A sodium flow in a rectangular duct exposed to an external, transverse magnetic field has been examined. To demonstrate the capability of UDV the transformation of the well-known turbulent, piston-like profile to an M-shaped velocity profile for growing magnetic field strength was observed. The significance of artifacts such as caused by the existence of reflecting interfaces in the measuring domain will be discussed. In the sodium case, the measurements were performed through the channel wall. An integrated ultrasonic sensor with acoustic wave-guide has been developed to overcome the limitation of ultrasonic transducers to temperatures lower than 200 .deg. C. This sensor can presently be applied at maximum temperatures up to 800 .deg. C. Stable and robust measurements have been performed in various PbBi flows in our laboratory at FZR as well as at the THESYS loop of the KALLA laboratory of the ForschungsZentrum Karlsruhe (FZK). We will also present experimental results obtained in a PbBi bubbly flow at 250...300 .deg. C. Argon bubbles were injected through a single orifice in a cylindrical container filled with stagnant PbBi. Velocity profiles were measured in the bubble plume. Mean values of the liquid as well as the bubble velocity were extracted from the data and will be presented as function of the gas flow rate

  8. Large eddy simulation of mixing between hot and cold sodium flows - comparison with experiments

    Energy Technology Data Exchange (ETDEWEB)

    Simoneau, J.P.; Noe, H.; Menant, B.

    1995-09-01

    The large eddy simulation is becoming a potential powerful tool for the calculation of turbulent flows. In nuclear liquid metal cooled fast reactors, the knowledge of the turbulence characteristics is of great interest for the prediction and the analysis of thermal stripping phenomena. The objective of this paper is to give a contribution in the evaluation of the large eddy simulation technique is an individual case. The problem chosen is the case of the mixing between hot and cold sodium flows. The computations are compared with available sodium tests. This study shows acceptable qualitative results but the simple model used is not able to predict the turbulence characteristics. More complex models including larger domains around the fluctuating zone and fluctuating boundary conditions could be necessary. Validation works are continuing.

  9. Characteristics of liquid and boiling sodium flows in heating pin bundles

    International Nuclear Information System (INIS)

    Menant, Bernard

    1976-01-01

    This study is related to cooling accidents which could occur in sodium cooled fast reactors. Thermo-hydraulic aspects of boiling experiments in pin bundles with helical wire-wrap spacer systems, in the case of undamaged geometries, are analyzed. Differences and analogies in the behavior of multi-rod bundle flows and one-dimensional channel flows are studied. A boiling model is developed for bundle geometries, and predictions obtained with the FLICA code using this models are presented. These predictions are compared with experimental results obtained in a water 19-rod bundle. Then, results of sodium boiling experiments through a 19-rod bundle are interpreted. Both cases of high power and reduced power are envisaged. (author) [fr

  10. LOFA [loss of flow accident] and LOCA [loss of coolant accident] in the TIBER-II engineering test reactor: Appendix A-4

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510 0 C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs

  11. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Steady state and transients computations

    International Nuclear Information System (INIS)

    Martin, A.; Alvarez, D.; Cases, F.

    1996-03-01

    The paper explains the chronological account and the first results obtained in the R and D program on the mixing in the 900 MW PWR vessels. After the presentation of the plant type simulated, we define the numerical tool, the (Finite Element Modelling) FEM N3S code. Two results are presented with a comparison with the experiment results issued of the BORA BORA mock up. The first case is dealing with the isothermal steady state mixing in the vessel with the three loops mass flow rate balanced. This case identified as a validation of our numerical tool shows a good agreement. The second case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. We compare the numerical and experiment results giving the mean boron concentration at the core inlet for several clear water plugs. The results show again a good agreement. (authors). 12 refs., 10 figs., 1 tab

  12. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. Secondary coolant purification system

    International Nuclear Information System (INIS)

    Stiteler, F.Z.; Donohue, J.P.

    1978-01-01

    The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

  14. Comparison of thermohydraulic characteristics in the use of various coolants

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Suda, Kazunori; Yamaguchi, Akira

    2000-11-01

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

  15. [Influence of non-sodium restricted diet with diuretics on plasma rennin, renal blood flow and in patients with cirrhotic ascites].

    Science.gov (United States)

    Zhu, Yin-fang; Gu, Xi-bing; Zhu, Hong-ying; Yang, Xiao-juan; Wang, Dong; Yu, Ping

    2013-02-01

    To explore influence of sodium restricted diet and non-sodium restricted diet on plasma rennin (PRA), angiotensin II (All), ALD, renal blood flow (RBF) and subside of ascites in patients with cirrhotic ascites. Eighty cases of hepatitis B with cirrhotic ascites were randomly divided into sodium restricted diet group and non-sodium restricted diet group. 39 cases were in non-sodium restricted diet group, taking sodium chloride 6500-8000 mg daily; 41 cases were in sodium restricted diet group, taking sodium chloride 5000 mg daily. Both groups received diuretics furosemide and spironolactone. Blood sodium, urine sodium, PRA, AII, ALD, RBF ascites subsiding were compared after treatment. In non-sodium restricted diet group, blood sodium and urine sodium increased 10 days after treatment compared with those before treatment, and compared with those of sodium restricted diet group 10 days after treatment, P Renal damage induced by low blood sodium after treatment was less in non-sodium restricted diet group than that in sodium restricted diet group, P blood sodium, thus increasing excretion of urine sodium and diuretic effect. It can also decrease levels of PRA, AII and ALD, increase renal blood flow and prevent renal damage induced by low blood sodium and facilitate subsiding of ascites.

  16. Coolant channel module CCM

    International Nuclear Information System (INIS)

    Hoeld, Alois

    2007-01-01

    A complete and detailed description of the theoretical background of an '(1D) thermal-hydraulic drift-flux based mixture-fluid' coolant channel model and its resulting module CCM will be presented. The objective of this module is to simulate as universally as possible the steady state and transient behaviour of the key characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. Due to the possibility that different flow regimes can appear along any channel, such a 'basic (BC)' 1D channel is assumed to be subdivided into a number of corresponding sub-channels (SC-s). Each SC can belong to only two types of flow regime, an SC with just a single-phase fluid, containing exclusively either sub-cooled water or superheated steam, or an SC with a two-phase mixture flow. After an appropriate nodalisation of such a BC (and therefore also its SC-s) a 'modified finite volume method' has been applied for the spatial discretisation of the partial differential equations (PDE-s) which represent the basic conservation equations of thermal-hydraulics. Special attention had to be given to the possibility of variable SC entrance or outlet positions (which describe boiling boundaries or mixture levels) and thus the fact that an SC can even disappear or be created anew. The procedure yields for each SC type (and thus the entire BC), a set of non-linear ordinary 1st order differential equations (ODE-s). To link the resulting mean nodal with the nodal boundary function values, both of which are present in the discretised differential equations, a special quadratic polygon approximation procedure (PAX) had to be constructed. Together with the very thoroughly tested packages for drift-flux, heat transfer and single- and two-phase friction factors this procedure represents the central part of the here presented 'Separate-Region' approach, a theoretical model which provides the basis to the very effective working code package CCM

  17. Temperature fluctuation of sodium in annular flow channel heated by single-pin with blockage

    International Nuclear Information System (INIS)

    Miyazaki, Keiji; Kimura, Jiro; Ogawa, Masuro; Okada, Toshio

    1978-01-01

    Root mean square (RMS) value and power spectral density (PSD) of temperature fluctuation were measured with use of forced-circulating sodium in an annular channel (6.5 mm I.D., 20mm O.D.) with concentric disk to simulate blockage (about 80%) of sodium flow. The experimental range of the heat flux was 40 -- 150 W/cm 2 and the bulk flow velocity 0.14--0.41m/sec (Re=7.7x10 3 --2.3x10 4 ) under a temperature of 500--800 0 C. The RMS value measured at the exit of heating section (150mm downstream from the blockage) is larger by a factor of 2 -- 3 than that in the wake (10 -- 20mm downstream from the blockage), marking a few deg.C for a heat flux of 105W/cm 2 and a flow velocity of 0.27m/sec. The RMS value is proportional to the wall-to-bulk-fluid temperature difference in heat transfer, presenting the similar dependence on the heat flux and flow velocity. The fluctuations of temperature are greatly attenuated in the upper unheated section where the radial temperature gradient is absent, and consequently it is suggested that the fluctuations of temperature should be caused by the local turbulence of flow, such as a vortex street due to blockage in the present experiment, under the presence of large gradient of temperature near the heating surface. (auth.)

  18. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

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

  19. Optimization of magnetic amplification by flow constraints in turbulent liquid sodium

    International Nuclear Information System (INIS)

    Nornberg, M. D.; Taylor, N. Z.; Forest, C. B.; Rahbarnia, K.; Kaplan, E.

    2014-01-01

    Direct measurements of the vector turbulent emf in a driven two-vortex flow of liquid sodium were performed in the Madison Dynamo Experiment [K. Rahbarnia et al., Astrophys. J. 759, 80 (2012)]. The measured turbulent emf is anti-parallel with the mean current and is almost entirely described by an enhanced resistivity, which increases the threshold for a kinematic dynamo. We have demonstrated that this enhanced resistivity can be mitigated by eliminating the largest-scale eddies through the introduction of baffles. By tailoring the flow to reduce large-scale components and control the helical pitch, we have reduced the power required to drive the impellers, doubled the magnetic flux generated by differential rotation, and increased the decay time of externally applied magnetic fields. Despite these improvements, the flows remain sub-critical to the dynamo instability due to the reemergence of turbulent fluctuations at high flow speeds

  20. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

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

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

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

  3. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Numerical simulation of the accurate RCP start-up flow rate

    International Nuclear Information System (INIS)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S.

    1997-06-01

    This report explains the last results about the mixing in the 900 MW PWR vessels. The accurate fluid flow transient, induced by the RCP starting-up, is represented. In a first time, we present the Thermalhydraulic Finite Element Code N3S used for the 3D numerical computations. After that, results obtained for one reactor operation case are given. This case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. A comparison made between two injection modes; a steady state fluid flow conditions or the accurate RCP transient fluid flow conditions. The results giving the local minimum of concentration and the time response of the mean concentration at the core inlet are compared. The results show the real importance of the unsteadiness characteristics of the fluid flow transport of the clear water plug. (author)

  4. Self-welding evaluation of stellite 6 and stellite 156 in flowing sodium

    International Nuclear Information System (INIS)

    Chang, J.Y.; Schrock, S.L.; Johnson, R.N.

    1977-01-01

    The self-welding behavior of two similar materials, Stellite 6 and Stellite 156, in sodium are discussed. The materials were tested at temperatures from 850 to 1140 0 F for time periods up to six-months while immersed in flowing sodium. Contact stresses ranged from 6000 psi to 16,000 psi on contact areas from 0.35 to 0.47 square inches. All separation tests to determine the extent of self-welding were conducted in a tensile mode. The surface morphologies of the samples before and after each test were measured. At temperatures of 1115 0 F and above, almost all the Stellite 6 specimens indicated a significant tendency toward self-welding within a relatively short period of time (one week). Stellite 156 couples also developed a strong self-weld bond at 1060 0 F after six-month exposure

  5. Design of creep machine and creep specimen chamber for carrying out creep tests in flowing liquid sodium

    Energy Technology Data Exchange (ETDEWEB)

    Ravi, S., E-mail: sravi@igcar.gov.in; Laha, K.; Sakthy, S.; Mathew, M.D.; Jayakumar, T.

    2014-02-15

    Highlights: • Design of a lever type creep machine for carrying out creep test in flowing sodium. • Leveling of lever during creep was achieved by automated movement of fulcrum. • Design of creep chamber for providing constant sodium flow rate across creep specimen. • Minimum use of bellow in chamber for sodium containment and mechanical isolation. • Mini-lever mechanism to counter balance load reduction on specimen due to bellow stiffness. - Abstract: A creep testing system has been designed, fabricated, installed and validated for carrying out creep tests in flowing liquid sodium. The testing system consists of two sections namely creep testing machine and an environmental chamber. The testing system has the ability of (i) applying tensile load to the test specimen through a lever, (ii) monitoring continuously the creep elongation and (iii) allowing sodium to flow around the creep specimen at constant velocity. The annular space between the creep specimen and the environmental chamber has been suitably designed to maintain constant sodium flow velocity. Primary and secondary bellows are employed in the environmental chamber to (i) mechanically isolate the creep specimen, (ii) prevent the flowing sodium in contact with air and (iii) maintain an argon gas cover to the leaking sodium if any from primary bellow, with a provision to an alarm get activated by a spark plug. The lever-horizontality during creep test has been maintained by automatically lifting up the fulcrum instead of lowering down the pull rod as conventionally used. A mini lever mechanism has been incorporated in the load train to counter balance the load reduction on specimen from the changing stiffness of the bellows. The validation of the testing system has been established by carrying out creep tests on 316L(N) stainless steel at 873 K over a wide stress range and comparing the results with those obtained in air by employing the developed and conventional creep testing machines.

  6. Influence of flowing sodium on creep deformation and rupture behaviour of 316L(N) austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Ravi, S., E-mail: sravi@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Laha, K.; Mathew, M.D. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Vijayaraghavan, S.; Shanmugavel, M.; Rajan, K.K. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102 (India)

    2012-08-15

    The influence of flowing sodium on creep deformation and rupture behaviour of AISI 316L(N) austenitic stainless steel has been investigated at 873 K over a stress range of 235-305 MPa. The results were compared with those obtained from testing in air environment. The steady state creep rates of the material were not influenced appreciably by the testing environments. The time to onset of tertiary stage of creep deformation was delayed in sodium environment. The creep-rupture lives of the material increased in sodium environment, which became more pronounced at lower applied stresses. The increase in rupture life of the material in flowing sodium was accompanied by an increase in rupture ductility. The creep damage on specimen surface as well as inside the specimen was less in specimen tested in sodium. SEM fractographic investigation revealed predominantly transgranular dimple failure for the specimen tested in sodium, whereas predominantly intergranular creep failure was observed in the air tested specimens. Almost no oxidation was observed in the specimens creep tested in the sodium environment. Absence of oxidation and less creep damage cavitation extended the secondary state in liquid sodium tests and lead to increase in creep rupture life and ductility of the material as compared to in air.

  7. The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

    2007-01-01

    CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application in dispersed energy supply and multipurpose use, in conjunction with Toshiba Corporation. The 4S is sodium cooled fast reactor and their electrical output has two options of 10MWe and 50MWe. In this paper, 10MWe 4S (4S-10M) was proposed. 4S-10M has some unique features. It employs a burn-up control system with annular reflector in place of the control rod that requires the frequent maintenance service. The core life time of the 4S-10M is 30 years and the fuel transport is not required during core life time. All temperature feedback coefficients are negative during core life time. In the latest design for 4S-10M, a pool and tall type reactor design was selected to reduce the construction cost. Two types of decay heat removal system (Reactor Vessel Auxiliary Cooling System; RVACS, Intermediate Reactor Auxiliary Cooling System; IRACS) using natural convection power were adopted. It is necessary to confirm that these two heat removal system can operate appropriately. The transition analyses were executed by the CERES code to evaluate the design feasibility and the thermal hydraulic characteristics of the 4S-10M. CERES is a multi-dimensional plant dynamics simulation code for liquid metal reactors developed by the CRIEPI. CERES can perform simulations ranging from forced circulation (full/partial power operation) to natural circulation. Components (pumps, IHXs, SGs, pipings, etc.) of the reactor are modeled as one-dimensional. Multi-dimensional plena are connected to such components. Two loss-of-flow accident sequences are considered. In the first case, it is assumed that the primary and the secondary pump were stopped by the total station black out. The reactor shut down system was assumed to be success. This sequence is referred to as the protected loss-of-flow accident (PLOF). In the second case, it is assumed that the reactor shut down systems fail to operate and the

  8. Evaluation of tests for coastdown of reactor coolant flow and measure of primary circuit flow of Angra-1 nuclear power plant

    International Nuclear Information System (INIS)

    Galetti, M.R.S.; Camargo, C.T.M.; Pontedeiro, A.C.

    1987-05-01

    The Angra 1 Nuclear Power Plant first reload license was issued after several technical discussions among CNEN, FURNAS and KWU. During the license process CNEN has established that the plant could return to anormal operation if the requirements described in the letter CNEN-DExL-C 06/86 were satisfied. The requirements according to the CNEN Transient and Thermohydraulic Group Analysis were to do again the following tests: 'Primary Flow Measurement' to check if the excess flow measured in the first cycle was held; and Pump Coastdown' to check if the Westinghouse and KWU fuel elements are thermo-hydraulicaly compatibles during transients. The mixed core must keep at least the same safety margin presented on Angra 1 FSAR for the original core. The tests and the analysis of results are described. (Author) [pt

  9. Chemistry of liquid metal coolants and sensors

    International Nuclear Information System (INIS)

    Gnanasekaran, T.

    2015-01-01

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

  10. An evaluation of the fluid-elastic instability for Intermediate Heat Exchanger of Prototype Sodium-cooled fast Reactor

    International Nuclear Information System (INIS)

    Cho, Jaehun; Kim, Sungkyun; Koo, Gyeonghoi

    2014-01-01

    The sodium-cooled fast reactor (SFR) module consists of the vessel, containment vessel, head, rotating plug (RP), upper internal structure (UIS), intermediate heat exchanger (IHX), decay heat exchanger (DHX), primary pump, internal structure, internal components and reactor core. The IHXs transfer heat from the radioactive sodium coolant (primary sodium) in the primary heat transport system to the nonradioactive sodium coolant (secondary sodium) in the intermediate heat transport system. Each sodium flows like Fig. 1. Primary sodium flows inside of tube and secondary sodium flows outside. During transferring heat two sodium to sodium, the fluid-elastic instability is occurred among tube bundle by cross flow. Large amplitude vibration occurred by the fluid-elastic instability is caused such as crack and wear of tube. Thus it is important to decrease the fluid-elastic instability in terms of a safety. The purpose of this paper is to evaluate the fluid-elastic instability for tube bundle in the IHX following ASME code. This paper evaluated the fluid-elastic instability of tube bundle in the SFR IHX. According evaluation results, the fluid-elastic instability of IHX tube bundle is occurred. A installing an additional TSP under the upper tubesheet can decrease a probability of fluid-elastic instability. If a location of an additional TSP does not exceed tube length to become a 750 mm, tube bundle of IHX is safety from the fluid-elastic instability

  11. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

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

  12. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

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

  13. High Magnetic Shear Gain in a Liquid Sodium Stable Couette Flow Experiment: A Prelude to an α-Ω Dynamo

    International Nuclear Information System (INIS)

    Colgate, Stirling A.; Beckley, Howard; Si, Jiahe; Martinic, Joe; Westpfahl, David; Slutz, James; Westrom, Cebastian; Klein, Brianna; Schendel, Paul; Scharle, Cletus; McKinney, Travis; Ginanni, Rocky; Bentley, Ian; Mickey, Timothy; Ferrel, Regnar; Li, Hui; Pariev, Vladimir; Finn, John

    2011-01-01

    The Ω phase of the liquid sodium α-Ω dynamo experiment at New Mexico Institute of Mining and Technology in cooperation with Los Alamos National Laboratory has demonstrated a high toroidal field B φ that is ≅8xB r , where B r is the radial component of an applied poloidal magnetic field. This enhanced toroidal field is produced by the rotational shear in stable Couette flow within liquid sodium at a magnetic Reynolds number Rm≅120. Small turbulence in stable Taylor-Couette flow is caused by Ekman flow at the end walls, which causes an estimated turbulence energy fraction of (δv/v) 2 ∼10 -3 .

  14. Experimental characterization of MHD pressure drop of liquid sodium flow under uniform magnetic field

    International Nuclear Information System (INIS)

    Kim, Hee Reyoung; Park, Jon Ho; Kim, Jong Man; Nam, Ho Yoon; Choi, Jong Hyun

    2001-01-01

    Magnetic field has many effects on the hydraulic pressure drop of fluids with high electrical conductivity. The theoretical solution about MHD pressure drop is sought for the uniform current density model with simplified physical geometry. Using the MHD equation in the rectangular duct of the sodium liquid flow under a transverse magnetic field, the electrical potential is sought in terms of the duct geometry and the electrical parameters of the liquid metal and duct material. By the product of the induced current inside the liquid metal and transverse magnetic field, the pressure gradients is found as a function of the duct size and the electrical conductivity of the liquid metal. The theoretically predicted pressure drop is compared with experimental results on the change of flow velocity and magnetic flux density

  15. Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

    International Nuclear Information System (INIS)

    Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

    1978-12-01

    The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

  16. Effect of flowing sodium on corrosion and tensile properties of AISI type 316LN stainless steel at 823 K

    Science.gov (United States)

    Sivai Bharasi, N.; Thyagarajan, K.; Shaikh, H.; Balamurugan, A. K.; Bera, Santanu; Kalavathy, S.; Gurumurthy, K.; Tyagi, A. K.; Dayal, R. K.; Rajan, K. K.; Khatak, H. S.

    2008-07-01

    AISI type 316LN stainless steel was exposed to flowing sodium in mass transfer loop (MTL) at 823 K for 16 000 h and then examined for changes in the tensile properties due to the mass transfer and corrosion effects. Comparisons in microstructural and mechanical properties were made between annealed, thermally aged and sodium exposed materials. Microstructural examination of thermally aged and sodium exposed materials revealed precipitation of carbides at the grain boundaries. The sodium exposed samples contained a degraded layer at the surface up to a depth of around 10 μm and a surface carburized layer of about 30 μm. There was about 15% increase in yield strength and a decrease of about 20% in ductility for the sodium exposed material vis-a-vis thermally aged material and this was attributed to carburization effects and microstructural changes.

  17. Effect of flowing sodium on corrosion and tensile properties of AISI type 316LN stainless steel at 823 K

    International Nuclear Information System (INIS)

    Sivai Bharasi, N.; Thyagarajan, K.; Shaikh, H.; Balamurugan, A.K.; Bera, Santanu; Kalavathy, S.; Gurumurthy, K.; Tyagi, A.K.; Dayal, R.K.; Rajan, K.K.; Khatak, H.S.

    2008-01-01

    AISI type 316LN stainless steel was exposed to flowing sodium in mass transfer loop (MTL) at 823 K for 16 000 h and then examined for changes in the tensile properties due to the mass transfer and corrosion effects. Comparisons in microstructural and mechanical properties were made between annealed, thermally aged and sodium exposed materials. Microstructural examination of thermally aged and sodium exposed materials revealed precipitation of carbides at the grain boundaries. The sodium exposed samples contained a degraded layer at the surface up to a depth of around 10 μm and a surface carburized layer of about 30 μm. There was about 15% increase in yield strength and a decrease of about 20% in ductility for the sodium exposed material vis-a-vis thermally aged material and this was attributed to carburization effects and microstructural changes

  18. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

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

  19. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

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

  20. Experimental analysis of upward vertical two-phase flow in four-cusp channels simulating the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident

    International Nuclear Information System (INIS)

    Assad, A.C.A.

    1984-01-01

    The present work deals with an experimental analysis of upward vertical two-phase flow in channels with circular and four-cusp cross-sections. The latter simulates the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident. Simultaneous flow of air and water has been employed to simulate adiabatic steam-water flow. The installation of air-water separators helped eliminate instabilities during pressure-drop measurements. The gamma ray attenuation was utilized for the void fraction determination. For the four-cusp geommetry, new criteria for two-phase flow regime transitions have been determined, as well as new correlatins for pressure drop and void fraction, as function of the Lockhart-Martinelli factor and vapour mass-fraction, respectively. (Author) [pt

  1. Research on coolant radiochemistry

    International Nuclear Information System (INIS)

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

    2004-04-01

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

  2. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

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

    International Nuclear Information System (INIS)

    Pion, Agnes

    1983-01-01

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

  4. 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

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

    International Nuclear Information System (INIS)

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

    2004-11-01

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

  6. Early Jurassic Carbon and Sodium Sequestration in a CAMP basalt flow

    Science.gov (United States)

    Block, K. A.; Puffer, J. H.

    2017-12-01

    The initial HTQ-type CAMP Orange Mountain Basalt flow, as well as related pillowed flows and the overlying Preakness flows, locally underwent substantial and well documented albitization, chloritization, and sulphate, carbonate, and zeolite mineralization. Layers representing at least 25 vol % of the Orange Mountain Basalt have undergone a major net increase in sodium and carbon content and a major redistribution of magnesium and calcium. Most alteration occurred during the development of a widespread early Jurassic geothermal system similar to the active system of Iceland. In both cases alteration was controlled by active circulation of basin brines through vesicular layers during rapid burial at temperatures that were kept elevated by recurring magmatism. Whole rock Na2O levels typically increased from 2.2 wt. % in unaltered layers to 3.2 wt. % in vesicular layers, and commonly reached levels exceeding 5 wt. %. The environmental implications of the removal of such massive amounts of sodium from the geothermal system on the chlorine budget and the salt content of Early Jurassic lakes are currently being evaluated. Massive amounts of carbon sequestration from the geothermal system may have mitigated an increased burden on the early Jurassic atmosphere where geothermal CO2 may have otherwise been vented at hot springs or solfataras. Calcite amygdules typically account for 5 to 10 vol. % of the vesiculated layers amounting to 66 to 132 kg of CO2 per m3 of basalt. If 25 vol. % of the 160 thick Orange Mountain Basalt is vesiculated that would equate to about 2640 to 5280 kg of CO2 per m2 of basalt. The full extent of calcite enrichment across the entire CAMP province, however, has not yet been determined.

  7. Electricity generation by nuclear fission reactor and closed cycle gas turbines, with core automatically shut down by coolant flow failure and dropped out of plant for sealing if temperature is excessive

    International Nuclear Information System (INIS)

    Pedrick, A.P.

    1976-01-01

    A reactor system is described in which if there is a failure of coolant flow the core automatically drops down to its control rods, so that criticality is reduced, but if the temperature of the core still stays dangerously high the core is allowed to drop down a deep shaft. Concrete blocks automatically come together after the ejected reactor core has moved past them to prevent the escape of radiation or radioactive material, until such time that the core temperature has dropped to a level that it can, with safety, be returned to its normal position in the plant. (U.K.)

  8. Effect of composition on the corrosion behavior of 316 stainless steel in flowing sodium

    International Nuclear Information System (INIS)

    Bates, J.F.; Brehm, W.F.

    1976-03-01

    Type 316 stainless steel specimens irradiated and non-irradiated, with minor variations in C, N, Cr, Ni, Si, and Mn content, were exposed to flowing sodium in the Source Term Control Loop-1 facility. Test conditions of 604 0 C, a sodium velocity of 6.7 m/s, and an oxygen content of 0.5 and 2.5 ppM were used to ascertain the effect of these compositional variations on the corrosion rate and on the 54 Mn release rate. Variations in C, N, and Mn had no significant effect on the corrosion rate. The corrosion rate increased somewhat as the Ni and Si content of the steel was increased, and decreased with an increase in Cr content. Microprobe examination showed preferential leaching of Mn, Si, Cr, and Ni to depths of up to 10 μm (0.4 mil) after exposure times ranging from 5000 to 8000 hours. Variations in natural ( 55 Mn) content did not affect release rates of 54 Mn generated by 54 Fe(n,p) 54 Mn. 10 fig, 4 tables

  9. Coolant leakage detecting device

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  10. Erosion protection by calcium lactate/sodium fluoride rinses under different salivary flows in vitro.

    Science.gov (United States)

    Borges, Alessandra B; Scaramucci, Taís; Lippert, Frank; Zero, Domenick T; Hara, Anderson T

    2014-01-01

    This study investigated the effect of a calcium lactate prerinse on sodium fluoride protection in an in vitro erosion-remineralization model simulating two different salivary flow rates. Enamel and dentin specimens were randomly assigned to 6 groups (n = 8), according to the combination between rinse treatments - deionized water (DIW), 12 mM NaF (NaF) or 150 mM calcium lactate followed by NaF (CaL + NaF) - and unstimulated salivary flow rates - 0.5 or 0.05 ml/min - simulating normal and low salivary flow rates, respectively. The specimens were placed into custom-made devices, creating a sealed chamber on the specimen surface connected to a peristaltic pump. Citric acid was injected into the chamber for 2 min, followed by artificial saliva (0.5 or 0.05 ml/min) for 60 min. This cycle was repeated 4×/day for 3 days. Rinse treatments were performed daily 30 min after the 1st and 4th erosive challenges, for 1 min each time. Surface loss was determined by optical profilometry. KOH-soluble fluoride and structurally bound fluoride were determined in specimens at the end of the experiment. Data were analyzed by 2-way ANOVA and Tukey tests (α = 0.05). NaF and CaL + NaF exhibited significantly lower enamel and dentin loss than DIW, with no difference between them for normal flow conditions. The low salivary flow rate increased enamel and dentin loss, except for CaL + NaF, which presented overall higher KOH-soluble and structurally bound fluoride levels. The results suggest that the NaF rinse was able to reduce erosion progression. Although the CaL prerinse considerably increased F availability, it enhanced NaF protection against dentin erosion only under hyposalivatory conditions.

  11. Studies to single subassembly flow monitoring with a complete 7 element array under sodium

    International Nuclear Information System (INIS)

    Hess, B.; Ruppert, E.; Stehle, H.; Vinzens, K.

    1975-01-01

    As part of the SNR-300 R and D programme a complete clamped array, consisting of 4 full size fuel elements and 3 blanket elements was tested for more than 4000 hours at 600 deg C in the AKB sodium loop at Interatom, Bensberg. The test was split into two phases and the total cluster was prestrained in the second phase to simulate 15 mm subassembly displacement at the level of the upper bearing pads. Although this test was mainly considered as an endurance test to demonstrate the integrity of prestrained core elements, efforts were made to study the feasibility of single subassembly flow monitoring with this full size model of a core section. The results of these investigations are presented and discussed in this paper

  12. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  13. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  14. Development of industrial utilization of metallic sodium

    International Nuclear Information System (INIS)

    Yuhara, Shunichi

    1995-01-01

    Sodium exists in large quantity, being ranked to 6th in the existence proportion of elements, and takes 2.83% of the matters composing earth crust. Sodium is an alkali metal which is light weight, chemically very active and a strong reducing substance. It is excellent in the compatibility with iron and steel materials, and it possesses good heat conduction and flow characteristics and stable nuclear characteristics. Since the industrial production of sodium became practical, its utilization was developed as the reducing agent and catalyst in chemical industry, the core coolant and heat transport medium for nuclear reactors, the material composing the secondary batteries for storing electric power, and the auxiliaries for metal refining and so on. The industrial production of metallic sodium is carried out by the electrolysis of melted salt, namely Downs process. The production of metallic sodium in Japan is 3000-6000 t yearly, and its import is 300-350 t. Its main use is for organic chemical industry including dye production. The grades of metallic sodium products and their uses are shown. The utilization of sodium for large fast reactors, the utilization of NaK as the heat transport and cooling medium for space use nuclear reactors and deep sea fast reactor system, and the utilization of sodium as the catalyst in dye production, for silicon carbide fiber production and for agricultural and medical chemical production are reported. (K.I.)

  15. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

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

  16. Evaluation of organic coolants for the transportation of LMFBR spent fuel rods

    International Nuclear Information System (INIS)

    Arnold, C. Jr.

    1978-05-01

    The physical and chemical processes that are likely to occur when sodium coated LMFBR spent fuel rods are submerged in various aromatic organic coolants was defined by means of immersion experiments carried out with sodium coated 304 stainless steel coupons. Upon immersion of sodium coated coupons at 220 0 C in hydrocarbon type coolants such as Therminol 88, a mixture of terphenyls, not only was the metallic sodium retained on the coupon, but a carbonaceous coating formed on the surface of the sodium. In contrast, coolants that contained aromatic ether bonds, such as Dowtherm A, reacted with sodium at 220 0 C to form phenolate and other salts, which precipitated from the coolant in the form of a dark sludge. With Dowtherm A, removal of metallic sodium from the coupon was essentially complete in a matter of hours at temperatures of 160--220 0 C. Data on the rate and efficiency of sodium removal upon immersion in Dowtherm A at elevated temperatures were obtained. In addition the kinetics and chemistry of the sodium/Dowtherm A reaction were defined. Because sodium sludges are potentially incompatible with the containing structural materials and the fuel elements, it is recommended that sodium be removed prior to immersion in the coolant via reaction with benzoic acid; this method should be adaptable to the facilities at reactor sites. In aging studies Dowtherm A was found to be thermally stable up to 400 0 C and radiatively stable at ambient conditions. The combined effect of heat and radiation was not defined

  17. Development of a Program for Predicting Flow Instability in a Once-through Sodium-Heated Steam Generator (III)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eui Kwang; Yoon, Jung; Kim, Jong Bum; Jeong, Jiyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Two-phase flow systems can be subjected to several types of instability problems. Density-wave oscillation is the most common and important type of instability in boiling channels. Such instability gives difficulties in predictions of system performance and system control, and component failure due to thermal fatigue. A computer program developed for predicting two-phase flow instability in a steam generator heated by liquid sodium was presented in the previous works. Limit cycle was predicted even in a fixed node system. The amplitude of inlet flow rate is larger than that of outlet flow rate. The amplitude of phase change location oscillation within boiling-to-vapor boundary node is larger than that of liquid-to-boiling boundary node. Sodium and steam temperature are invariant at tube exit.

  18. Comparison of fractional flow reserve measurements using intracoronary adenosine versus intracoronary sodium nitroprusside infusions in moderately stenotic coronary artery lesions

    Energy Technology Data Exchange (ETDEWEB)

    Safi, Morteza; Namazi, Mohammad Hasan; Fooladi, Esfandiar; Vakili, Hossein; Parsa, Saeed Alipour; Khaheshi, Isa [Cardiovascular Research Center, Modarres hospital, Shahid Beheshti University of Medical Sciences, Tehran (Iran, Islamic Republic of); Abbasi, Mohammad Amin [Department of Internal Medicine, Shahid Beheshti University of Medical Sciences, Tehran (Iran, Islamic Republic of); Movahed, Mohammad Reza, E-mail: rmova@aol.com [CareMore, Arizona, Tucson, AZ (United States); University of Arizona, Sarver Heart Center, Tucson, AZ (United States)

    2016-10-15

    Introduction: The aim of this study was to investigate the efficacy and safety of intracoronary (IC) sodium nitroprusside infusion in comparison to IC adenosine for fractional flow reserve (FFR) measurement in moderately diseased coronary artery lesions for functional assessment. Methods: During a nine month period, a consecutive of 98 patients with suspected or known coronary artery disease with moderate stenosis found during angiography (40% to 70% stenosis), were enrolled in this study. Hyperemia was induced by bolus doses of IC adenosine followed by sodium nitroprusside for FFR measurement. Results: Both IC adenosine and IC sodium nitroprusside induced similar and significant reduction in FFR. There was no statistically difference in FFR values between adenosine vs sodium nitroprusside infusions (mean FFR 84.3 ± 6.3 vs 85.7 ± 6.2, p = 0.1) respectively. Furthermore, comparing different FFR cut-off points between the groups (FFR < 0.75, 0.75–0.8 and > 0.8) showed no significant differences (p value = 0.7). Conclusion: An IC bolus of sodium nitroprusside (0.6 μg/kg) infusion induces a similar degree of hyperemia to IC bolus of 100–300 μg of adenosine. Therefore, IC sodium nitroprusside could be considered as an alternative drug to adenosine for FFR measurement with lower side effect profile. - Highlights: • Intracoronary (IC) sodium nitroprusside was compared with IC adenosine for FFR test. • IC adenosine and IC sodium nitroprusside induced similar reduction in FFR. • Different FFR cut-off points between the groups showed no significant differences. • IC sodium nitroprusside could be considered as an alternative to adenosine for FFR.

  19. Studies to single subassembly flow monitoring with a complete 7 element array under sodium

    International Nuclear Information System (INIS)

    Hess, B.; Ruppert, E.; Stehle, H.; Vinzens, K.

    1975-01-01

    A core restraint system in a fast reactor serves to limit fuel element movement leading to reactivity changes and misalignment of control rod drives and instrumentation. To guarantee proper control rod function the upper ring of the passive restraint system for the SNR-300 should keep subassembly displacement below 20 mm, whereas a free bowing up to 25 mm does not impair subassembly handling. With respect to single subassembly instrumentation the influences of subassembly displacement on temperature and flow monitoring were not exactly known. As part of the SNR-300 R and D programme a complete clamped array, consisting of 4 full size fuel elements and 3 blanket elements was tested for more than 4000 hours at 600 0 C in the AKB sodium loop at Interatom, Bensberg. The test was split into two phases and the total cluster was prestrained in the second phase to simulate 15 mm subassembly displacement at the level of the upper pads. Although this test was mainly considered as an endurance test to demonstrate the integrity of prestrained core elements, effort were made to study the feasibility of single subassembly flow monitoring with this full size model of a core section. (Auth.)

  20. Coolant system decontamination

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Mizutani, Akihiko; Sekimoto, Hiroshi

    1998-01-01

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

  2. Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm2 in the cold leg of primary loop using RELAP5 code

    International Nuclear Information System (INIS)

    Borges, Eduardo M.; Sabundjian, Gaiane

    2017-01-01

    The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm 2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

  3. Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm{sup 2} in the cold leg of primary loop using RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: borges.em@hotmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm{sup 2} of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

  4. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

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

    2007-06-01

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

  5. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  6. An automatic sodium-loop for testing the lon-term behaviour of sintered bodies flowed through by gas

    International Nuclear Information System (INIS)

    Barkleit, G.; George, G.; Haase, I.; Kiessling, W.

    1980-08-01

    An automatic sodium loop NAKOS for testing the long-term behaviour of porous stainless steel bodies which are flowed through by gas is described. The loop using a special safety protection system is capable of working without control up to 1000 h. During a 500 h-experiment the safety system and the gas permeability measuring method for testing the porous bodies were tested. Both first results of the behaviour of sintered bodies in liquid sodium of high purity and temperatures of about 850 K and some details of the production of these bodies are given. (author)

  7. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

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

  8. Health physics aspects of processing EBR-I coolant

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  9. Investigation and prevention of droplet splashing during operation of a sodium free jet flow

    International Nuclear Information System (INIS)

    Stoppel, L.; Gordeev, S.; Wetzel, T.; Fellmoser, F.; Daubner, M.

    2010-01-01

    Many accelerator application concepts consider liquid metal as a windowless target, at which the particle beam does directly hit the liquid. One of such concepts is studied in the European project ''DIRAC-Secondary beams - Design Study''. This project is focused on the preliminary research work for construction of a new international particle accelerator - Facility for Antiproton and Ion Research (FAIR) in Darmstadt. The planned accelerator is aimed to work with high-energy heavy-ions, such as U 238 . One of the key elements of the FAIR facility is a liquid-metal-target, made in the form of a rectangular shaped Lithium jet aligned with the gravity vector. In the course of preliminary investigations the theoretical and practical conditions for a stable liquid-metal-jet conforming to the FAIR-requirements have been studied in the Karlsruhe Liquid Metal Laboratory (KALLA) sodium facility. The acquired scientific and technological results can be transferred to liquid-metal targets in nuclear applications, for example, the IFMIF-Target for the study of fusion reactor materials and the Myrrah/XT-ADS target. The main goals of the KALLA-part of the project were to design and build a facility for experimental research on hydrodynamic phenomena of the free surface liquid metal flow as well as to look at technological problems influencing the hydrodynamic stability of such flows. One of such problems emerged already during the startup of the facility: Splashing of liquid metal drops in the vacuum volume of the target box. As a result of such splashing process, liquid metal droplets are accumulated on various internal constructional elements of the target box, for example, on the inspection windows. This effect prohibits long term operation with the facility. The present paper describes the methods used to reduce the splashing to a minimum. (orig.)

  10. Measurements of local temperature distributions in rod bundles with sodium flow

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1984-12-01

    In an electrically heated 19-rod bundle (P/D = 1.30, W/R = 1.40) with sodium flow the three-dimensional temperature fields in the rod clads were measured. The main characteristics of the test section are three adjacent heater rods in the duct wall zone instrumented on four measuring planes and rotatable by 360 0 under full power conditions; furthermore spacer grids which are axially movable, and a system allowing to bow one heater rod over the last third of its heated length. The results of measurements of the azimuthal temperature variations of the rotatable rods are presented for different operating conditions (80 2 ), different spacer grid positions relative to the measuring planes and different bowing positions of one rod. For better understanding of the experimental results cross sections of the 19-rod bundle were prepared. It became evident, that a well-known bundle geometry is very important for the interpretation of the experimental results. (orig.) [de

  11. Results of studying of turbulent heat transfer deterioration and their application for development of engineering methods of calculation of heat transfer and pressure drop in supercritical-pressure coolant flow

    International Nuclear Information System (INIS)

    Vladimir A Kurganov; Yuri A Zeigarnik

    2005-01-01

    Full text of publication follows: Using of the supercritical-pressure (SCP) water as a working medium is an apparent way to increase specific capacity and economic efficiency of nuclear power installations. Nevertheless, to provide safe operation of SCP nuclear power units, it is necessary to considerably improve reliability and accuracy of calculations of pressure drop and heat transfer in the SCP working media and coolants flows and the methods of forecasting such a dangerous phenomenon as deterioration of the turbulent heat transfer at a certain level of heat flux density. A value of the latter changes within a very large range depending on the specific conditions of the process under consideration. In the paper, the main results of the experimental study of heat transfer, pressure drop, and velocity and temperature fields in both upward and downward flows of the SCP CO 2 in tubes are considered. This study was conducted at OIVT RAN under conditions of heat input and embraced the regimes of normal and deteriorated heat transfer as well. On the basis of this data, the concept regarding to physical mechanism of incipience of the regimes of deteriorated heat transfer was developed. Classification of different modes of heat transfer deterioration in vertical channels is proposed. A degree of a danger of certain regimes is assessed. It is shown that the above phenomenon is caused by transformation of the structure of nonisothermal flow of SCP fluid due to changes in proportions between the forces acting upon a flow, specifically, because of an increase in the inertia forces due to thermal acceleration of a flow and/or in Archimedes' (buoyancy) forces up to the level comparable or higher than that of friction forces. The efficiency of the most thorough correlations for calculating normal and deteriorated heat transfer in flows of SCP water and CO 2 is analyzed. Reliability of existed recommendations to determine boundaries of normal heat transfer regimes is considered

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  13. Transient behaviour of main coolant pump in nuclear power plants

    International Nuclear Information System (INIS)

    Delja, A.

    1986-01-01

    A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

  14. Saline-induced natriuresis and renal blood flow in conscious dogs: effects of sodium infusion rate and concentration

    DEFF Research Database (Denmark)

    Sandgaard, N C F; Andersen, J L; Holstein-Rathlou, N-H

    2005-01-01

    AIM: This study focused on static and dynamic changes in total renal blood flow (RBF) during volume expansion and tested whether a change in RBF characteristics is a necessary effector mechanism in saline-induced natriuresis. METHODS: The aortic flow subtraction technique was used to measure RBF...... continuously. Identical amounts of NaCl (2.4 mmol kg(-1)) were given as slow isotonic (Iso, 120 min), slow hypertonic (Hyper, 120 min), and rapid isotonic loads (IsoRapid, 30 min). RESULTS: During Iso and IsoRapid, arterial blood pressure increased slightly (6-7 mmHg), and during Hyper it remained unchanged...... saline loading simulating daily sodium intake, the rate of sodium excretion may increase 10-20-fold without any change in mean arterial blood pressure or in RBF. Regulatory responses to changes in total body NaCl levels appears, therefore, to be mediated primarily by neurohumoral mechanisms and may occur...

  15. Measurement of renal blood flow by 131I-labelled sodium iodohippurate imaging in a rat model of herpes encephalitis

    International Nuclear Information System (INIS)

    Cleator, G.M.; Klapper, P.E.; Lewis, A.G.; Sharma, H.L.; Smith, A.M.; Manchester Univ.

    1990-01-01

    Renal blood flow was assessed qualitatively over a 30 min period in a rat model of herpes encephalitis by extra-renal scintigraphic imaging following the injection of 131 I-labelled sodium iodohippurate. No significant differences were observed for renal blood flow in either kidney between (or within) infected and control groups. Our data suggest that kidney function is not compromised in this animal model of encephalitis. The results are discussed in the context of the development of a non-invasive protocol for the in vivo diagnosis of herpes encephalitis. (orig.)

  16. Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 3. Behaviour of high pressure coolant injection system (HPCI) based on thermodynamic model

    International Nuclear Information System (INIS)

    Maruyama, Shigenao

    2014-01-01

    In order to clarify the process of Accident of Fukushima Nuclear Plants, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 3 is analyzed from the data open to the public. Phase equilibrium process model was introduced in which the vapor and water are at saturation point in the vessels. The present accident scenario assumes that the high pressure coolant injection system (HPCI) did not worked properly, but the steam in the reactor pressure vessel (RPV) leaked through the turbine of HPCI to the suppression chamber since 12/3/2011 12:35. It is assumed that the Tsunami flooded the torus room where the suppression chamber was placed. Proposed accident scenario agrees with the data of the plant parameters obtained just after the accident. It is estimated that the water injection by HPIC was stopped since around at 13/3 19:00 and the water level in RPV decreased since then. It is estimated that the RPV broke at 14/3 8:55 and water could injected from fire engines due to the depression due to the rupture of RPV. There was little water left in RPV at the time of the rupture. If the present scenario is correct, the behavior that operators in the plant stopped HPCI at 13/3 2:42 did not affect seriously on the RPV rupture. If HPCI was working properly until the operators stopped it, the plant parameters obtained in the accident cannot be explained. (author)

  17. Effect of inorganic additive sodium pyrophosphate tetrabasic on positive electrolytes for a vanadium redox flow battery

    International Nuclear Information System (INIS)

    Park, Se-Kook; Shim, Joonmok; Yang, Jung Hoon; Jin, Chang-Soo; Lee, Bum Suk; Lee, Young-Seak; Shin, Kyoung-Hee; Jeon, Jae-Deok

    2014-01-01

    Sodium pyrophosphate tetrabasic (SPT) is employed as an inorganic additive in the positive electrolyte of a vanadium redox flow battery (VRFB) to improve its long-term stability and electrochemical performance. The results of precipitation tests show that the long-term stability of positive electrolytes (2 MV(V) solution in 4 M total sulfates with 0.05 M SPT additive) is improved compared to the blank one. UV-vis and cyclic voltammetry (CV) measurements also suggest that the addition of SPT can effectively delay the formation of precipitation in positive electrolytes, and no new substances are formed in V(V) electrolytes with SPT. The calcined precipitates extracted from the electrolytes with and without a SPT additive are identified as V 2 O 5 by X-ray diffraction (XRD) analysis. A VRFB single-unit cell employing positive electrolytes with an additive exhibits the high energy efficiency of 74.6% at a current density of 40 mA cm 2 at the 500 th cycle at 20°C, compared to 71.8% for the cell employing the electrolyte without an additive. Moreover, the cell employing the electrolyte with an additive exhibits less discharge capacity fading during cycling in comparison with the pristine one. The disassembled cell without an additive shows a large number of V 2 O 5 precipitation particles on the felt electrode after 500 cycles. Meanwhile, the felt electrode of the cell with an additive has little precipitation. That precipitation gives rise to an imbalance between the positive and negative half-cell electrolytes, which results in a significant capacity loss. The additive has shown positive results under limited laboratory short-term and small-scale conditions

  18. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon

    2015-01-01

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable

  19. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

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

  20. Fuel-coolant interaction-phenomena under prompt burst conditions

    International Nuclear Information System (INIS)

    Jacobs, H.; Young, M.F.; Reil, K.O.

    1979-01-01

    The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2

  1. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

  2. Compatibility Behavior of the Ferritic-Martensitic Steel Cladding under the Liquid Sodium Environment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Baek, Jong Hyuk; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Fuel cladding is a component which confines uranium fuel to transport energy into the coolant as well as protect radioactive species from releasing outside. Sodium-cooled Fast Reactor (SFR) has been considered as one of the most probable next generation reactors in Korea because it can maximize uranium resource as well as reduce the amount of PWR spent fuel in conjunction with pyroprocessing. Sodium has been selected as the coolant of the SFR because of its superior fast neutron efficiency as well as thermal conductivity, which enables high power core design. However, it is reported that the fuel cladding materials like austenitic and ferritic stainless steel react sodium coolant so that the loss of the thickness, intergranular attack, and carburization or decarburization process may happen to induce the change of the mechanical property of the cladding. This study aimed to evaluate material property of the cladding material under the liquid sodium environment. Ferritic-martensitic steel (FMS) coupon and cladding tube were exposed at the flowing sodium then the microstructural and mechanical property were evaluated. mechanical property of the cladding was evaluated using the ring tension test

  3. SNR coolant system components

    International Nuclear Information System (INIS)

    De Haas Van Dorsser, A.H.; Mausbeck, H.

    1976-01-01

    The DEBENELUX prototype fast reactor power plant SNR 300 at Kalkar has a loop-type heat transfer system similar to that of the prototype LMFBR plants in the USA and Japan. There exist three 257 MW/sub th/ primary sodium loops, each with a hot leg centrifugal pump and three 85.6 MW/sub th/ intermediate heat exchangers in parallel. From there the heat is transferred to the steam generators via three secondary sodium loops with one cold leg sodium circulating pump in each. At a nominal reactor outlet temperature of 819 0 K and a turbine inlet power of 771 MW/sub th/ super heated steam of 166 bar and 733 0 K is produced, giving rise to a plant rating of 327 MW/sub e/ gross. The primary and secondary loops are described in detail

  4. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    International Nuclear Information System (INIS)

    Siamidis, John; Mason, Lee

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS O pt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids

  5. Reactor coolant pump seal leakage monitoring

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

    1986-01-01

    Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

  6. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 37-Rod Bundle for a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Hyungmo; Bae, Hwang; Chang, Seok-Kyu; Choi, Sun Rock; Lee, Dong Won; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Lee, Hyeong-Yeon

    2014-01-01

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are very important. Wrapped wires make a cross flow in a around the fuel rod) of the fuel rod, and this effect lets flow be mixed. Experimental results of flow mixing can be meaningful for verification and validation of thermal mixing correlation in a reactor core thermo-hydraulic design code. A wire mesh sensing technique can be useful method for measuring of flow mixing characteristics. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, it has been recently reported that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. This can be powerfully adapted to recognize flow mixing characteristics by wrapped wires in SFR core thermal design. In this work, we conducted the flow mixing experiments using a custom designed wire mesh sensor. To verify and validate computer codes for the SFR core thermal design, mixing experiments were conducted at a hexagonally arrayed 37-pin wire-wrapped fuel rod bundle test section. The well-designed wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable. In addition, by uncertainty analysis, the system errors and the random error were estimated in experiments. Therefore, the present results and methods can be used for design code verification and validation

  7. Modelling of the steam-water-countercurrent flow in the rewetting and flooding phase after loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Curca-Tivig, F.

    1990-01-01

    A new interphase momentum exchange model has been developed to simulate the Refill- Reflood Phase after LOCAs. Special phenomena of steam/water- countercurrent flow - like limitation or onset of downward-watee penetration - have been modelled and integrated into a flooding model. The interphase momentum exchange model interconnected with the flooding model has been implemented into the advanced system code RELAP5/MOD1. The new version of this code can now be utilized to predict the hot leg emergency-core-cooling (ECC) injection for German PWRs. The interfacial momentum transfer model developed includes the interphase frictional drag, the force due to virtual mass and the momenta due to interphase mass transfer. The modelling of the interfacial shear or drag accounts for the effects of phase and velocity profiles. The flooding model predicts countercurrent-flow limitation, onset of water penetration and partial delivery. The flooding correlation specifies the maximum down flow liquid velocity in case of countercurrent flow through flow restrictions for a given vapor velocity. (orig./HP) [de

  8. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

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

  9. CFD analyses of coolant channel flowfields

    Science.gov (United States)

    Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

    1993-01-01

    The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

  10. Study on two-phase flow in a coolant channel of a plate-type fuel with use of neutron radiography technique

    International Nuclear Information System (INIS)

    Mishima, K.; Hibiki, T.; Nishihara, H.

    1992-01-01

    Two-phase flow in a narrow rectangular duct is important related to abnormal cooling conditions of a MTR type research reactor. In view of this, flow regime, void fraction, slug bubble velocity and pressure loss were measured for rectangular ducts with a narrow gap. The neutron radiography technique was used to visualize the flow and the void fraction was obtained by image processing. The void fraction was correlated well by the drift flux model with existing correlation for the distribution parameter which was about 1.35. Similar results were obtained for slug bubble velocity, however the distribution parameter was in the range from 1.0 to 1.2. The frictional pressure loss was correlated well by the Chisholm-Laird correlation. In collaboration with previously obtained data, it was found that the Chisholm's parameter C, however, changed from 21 to zero as the gap decreased. (author)

  11. Sodium fires and nuclear power station safety

    International Nuclear Information System (INIS)

    Ivanenko, V.N.; Zubin, A.; Drobyshev, A.V.

    1986-01-01

    The danger of sodium aerosol release at a design basis accident (DBA) of a sodium-cooled fast reactor that involves coolant leakage and burning, is being analyzed. It has been shown that radioactive and toxic releases at DBA do not exceed permissible values. Some means of sodium fire fighting are described. (author)

  12. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

    2000-01-01

    There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  13. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

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

  14. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

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

  15. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

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

  16. 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

  17. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. Investigations of flow and temperature field development in bare and wire-wrapped reactor fuel pin bundles cooled by sodium

    International Nuclear Information System (INIS)

    Govindha Rasu, N.; Velusamy, K.; Sundararajan, T.; Chellapandi, P.

    2013-01-01

    Highlights: ► We study sodium flow and temperature development in fuel pin bundles. ► Pin diameter, number of pins, wire wrap and ligament gap are varied as parameters. ► Flow development is achieved within ∼30–40 hydraulic diameters. ► Thermal development is attained only for small pin diameter and less number of pins. ► Wire wrap and ligament gap strongly influence Nusselt number. - Abstract: Simultaneous development of liquid sodium flow and temperature fields in the heat generating pin bundles of reactor has been investigated. Development characteristics are seen to be strongly influenced by pin diameter, number of pins, helical wire-wrap, ligament gap between the last row of pins and hexcan wall and Reynolds number. Flow development is achieved within an axial length of ∼125 hydraulic diameters, for all the pin bundle configurations considered. But temperature development is attained only if the pin diameter is small or the number of pins is less. In the case of large pin diameter with more pins, temperature development could not be achieved even after a length of ∼1000 hydraulic diameters. The reason for this behavior is traced to be the weak communication among sub-channels in tightly packed bundles. It is seen that the pin Nusselt number decreases from center to periphery in a bundle. Also, if the ligament gap is narrow, the Nusselt number is large and more uniform. Flow development length is short if the Reynolds number is large and the converse is true for thermal development length. Helical wire-wrap shortens the thermal entry length and significantly enhances the global Nusselt number. But, its influence on hydrodynamic entry length is not significant

  19. 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)

  20. On the method of heat exchange calculation for critical and postcritical regimes in sodium-water steam generators

    International Nuclear Information System (INIS)

    Khudasko, V.V.; Kardash, D.Yu.; Grachev, N.S.

    1986-01-01

    Technique for calculating heat exchange in sodium-water steam generators with provisions for steam-water flow non-equilibrium character and moisture additional evaporation in pipes is suggested. Zone of heat exchange crisis representing the zone of transition from developed boiling to postcritical zone is considered. Comparison of estimated and experimental data performed for the following ranges of steame generator parameters: pressure p=7.8-14.0 MPa, coolant flow rate ρw=350-1000 kg/(m 2 xs), inlet sodium temperature T=590-825 K shows their good agreement

  1. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Anand, Nk [Texas A & M Univ., College Station, TX (United States)

    2016-03-30

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  2. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Hassan, Yassin; Anand, Nk

    2016-01-01

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  3. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. Assertion based verification methodology for HDL designs of primary sodium pump speed and eddy current flow measurement systems of PFBR

    International Nuclear Information System (INIS)

    Misra, M.K.; Menon, Saritha P.; Thirugnana Murthy, D.

    2013-01-01

    With the growing complexity and size of digital designs, functional verification has become a huge challenge. The validation and testing process accounts for a significant percentage of the overall development effort and cost for electronic systems. Many studies have shown that up to 70% of the design development time and resources are spent on functional verification. Functional errors manifest themselves very early in the design flow, and unless they are detected upfront, they can result in severe consequences - both financially and from a safety viewpoint. This paper covers the various types of verification methodologies and focuses on Assertion Based Verification Methodology for HDL designs, taking as case studies, the Primary Sodium Pump Speed and Eddy Current Flow Measurement Systems of PFBR. (author)

  6. Transformation of sodium from the Rapsodie fast breeder reactor into sodium hydroxide

    International Nuclear Information System (INIS)

    Roger, J.; Latge, C.; Rodriguez, G.

    1994-01-01

    One of the major problems raised by decommissioning a fast breeder reactor (FBR) concerns the disposal of the sodium coolant. The Desora operation was undertaken to eliminate the Rapsodie primary sodium as part of the partial decommissioning program, and to develop an operational sodium treatment unit for other needs. The process involves reacting small quantities of sodium in water inside a closed vessel, producing aqueous sodium hydroxide and hydrogen gas. It is described in this work. (O.L.). 4 figs

  7. Numerical experiment on different validation cases of water coolant flow in supercritical pressure test sections assisted by discriminated dimensional analysis part I: the dimensional analysis

    International Nuclear Information System (INIS)

    Kiss, A.; Aszodi, A.

    2011-01-01

    As recent studies prove in contrast to 'classical' dimensional analysis, whose application is widely described in heat transfer textbooks despite its poor results, the less well known and used discriminated dimensional analysis approach can provide a deeper insight into the physical problems involved and much better results in all cases where it is applied. As a first step of this ongoing research discriminated dimensional analysis has been performed on supercritical pressure water pipe flow heated through the pipe solid wall to identify the independent dimensionless groups (which play an independent role in the above mentioned thermal hydraulic phenomena) in order to serve a theoretical base to comparison between well known supercritical pressure water pipe heat transfer experiments and results of their validated CFD simulations. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  9. Methodology for Extraction of Remaining Sodium of Used Sodium Containers

    International Nuclear Information System (INIS)

    Jung, Minhwan; Kim, Jongman; Cho, Youngil; Jeong, Jiyoung

    2014-01-01

    Sodium used as a coolant in the SFR (Sodium-cooled Fast Reactor) reacts easily with most elements due to its high reactivity. If sodium at high temperature leaks outside of a system boundary and makes contact with oxygen, it starts to burn and toxic aerosols are produced. In addition, it generates flammable hydrogen gas through a reaction with water. Hydrogen gas can be explosive within the range of 4.75 vol%. Therefore, the sodium should be handled carefully in accordance with standard procedures even though there is a small amount of target sodium remainings inside the containers and drums used for experiment. After the experiment, all sodium experimental apparatuses should be dismantled carefully through a series of draining, residual sodium extraction, and cleaning if they are no longer reused. In this work, a system for the extraction of the remaining sodium of used sodium drums has been developed and an operation procedure for the system has been established. In this work, a methodology for the extraction of remaining sodium out of the used sodium container has been developed as one of the sodium facility maintenance works. The sodium extraction system for remaining sodium of the used drums was designed and tested successfully. This work will contribute to an establishment of sodium handling technology for PGSFR. (Prototype Gen-IV Sodium-cooled Fast Reactor)

  10. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

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

  11. 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

  12. Decontamination of main coolant pumps

    International Nuclear Information System (INIS)

    Roofthooft, R.

    1988-01-01

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

  13. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

  14. The dismantling of fast reactors: sodium processing

    International Nuclear Information System (INIS)

    Rodriguez, G.; Berte, M.; Serpante, J.P.

    1999-01-01

    Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

  15. Study thermal characteristics of millet grain, dried in the machine with the swirling flow of the coolant and the microwave energy supply

    Directory of Open Access Journals (Sweden)

    S. T. Antipov

    2016-01-01

    Full Text Available The article discusses the problems of determining the thermal characteristics of millet. The choice of the research object. The paper presents the principle of operation of the plant and the parameters of the standard, organic glass for measurements. Method was to study millet grains and organic glass, which are brought into contact on a common plane. The heater is brought into contact with the product and passed the constant heat flow, which passed through a layer of millet grain at different speeds. As a result, the temperature in the contact plane of the changed and recorded on the chart of the potentiometer in the form of the curve, by which you can determine the time and temperature change. The thermal diffusivity and thermal conductivity determined by empirical formulas obtained by solving a system of differential equations, made up for the system of two bodies, one of which includes the unknown thermal characteristics. Test two bodies in contact on a common plane, resulting in mathematical physics principles constitute two differential equations with uniform initial and boundary conditions of the first kind, due to the parameters of ongoing experience. It is a plot of thermal performance of the temperature and humidity. Revealed linear dependence of the physical thermal characteristics, showing that with increasing temperature the thermal diffusivity value decreases, and the thermal conductivity and specific heat capacity are increasing character. Revealed the equations describing the thermal characteristics of millet grain with a humidity in the range of 13.6–35.1% and the temperature range 293–373 K.

  16. Sodium fires and its extinguishment; Gorenie, sredstva i sposoby tusheniya natriya

    Energy Technology Data Exchange (ETDEWEB)

    Mikhedov, V G

    1979-03-01

    The fire safety problems of NPP with sodium coolants in USSR are presented. The design of sodium reactors is made with premises with sodium coolants being hermetic and filled with nitrogen. Some engineering solutions of fire safety including design, elaboration and choice of construction and protection materials are presented. Some theoretical aspects of sodium burning are presented as well as methods of sodium fire extinguishing methods including the use of powder.

  17. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  18. Effect of Flowing Water on Sr Sorption Changes of Hydrous Sodium Titanate

    OpenAIRE

    Youko Takahatake; Atsuhiro Shibata; Kazunori Nomura; Tsutomu Sato

    2017-01-01

    Radioactive contaminated water has been generated at the Fukushima Daiichi Nuclear Power station (F1NPS). Hydrous sodium titanate (SrTreat®) is able to remove radioactive Strontium (Sr) from this water. Knowing the amount of radioactive nuclides in the used as-received SrTreat® is important for effective disposal and deposition of the F1NPS waste. This study investigated changes in the ability of SrTreat® to sorb Sr, and to understand the causes of changes in the sorbing. An investigation of ...

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

    Science.gov (United States)

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

    2017-11-01

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

  20. Liquid metal reactor development -Studies on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author)

  1. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  2. Flowing and freezing of molten core materials during unprotected loss of flow accidents in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Maschek, W.; Royl, P.

    1988-09-01

    Flowing and freezing of mobile core materials change the fissile material distribution and core-inventory under hypothetical accident conditions and determine the path to permanent shutdown of the neutronic events and the energetic potentials. The report classifies the bondary conditions for such flowing and freezing processes by going through the different situations under which these processes can occur in the scenario of the unprotected loss of flow (ULOF) accident. The classification is based on ULOF-accident simulations for a homogeneous reactor core concept of a 300 MWe LMFBR (e. g. SNR-300), but many boundary conditions are also characteristic for other core designs. A review of the relevant experiments is then made to correlate the available experimental information with these classified boundary conditions and to look at the resulting flowing and freezing processes. Boundary conditions that have been experimentally shown to be important are assigned high priorities. The data are specifically valued in relation to these boundary conditions of high priorities. The review includes the major experimental programs with published results. The discussion shows that the results from most clean condition tests for melt relocations are valuable for a better understanding of basic phenomena and analytical model development, but are not directly applicable to real accident conditions. The database for relevant boundary conditions from the ULOF scenario is limited and largely included in integral sequence tests from which quantitative information for modelling is difficult to obtain. Needs for additional investigations are identified. The suggestions are mainly restricted to investigations of the early phase of fuel removal. They are given with reference to candidate facilities and include relocations in the subassemblies and in the inter-subassembly gaps. Particular emphasis is put on the leading edge properties and possible driving forces to which more attention

  3. Steady-state, local temperature fields with turbulent liquid sodium flow in nominal and disturbed bundle geometries with spacer grids

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.

    1980-01-01

    The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). The essential results are: - Outside the spacer grids, the azimuthal temperature variations of the side and corner rods are approximately 10-fold those of rods in the central bundle zone. - The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. - Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to adjacent cladding tubes. (orig.)

  4. Effect of anionic polyelectrolytes on the flow of activated sodium bentonite drilling mud

    Directory of Open Access Journals (Sweden)

    Chalah Kaci

    2018-01-01

    Full Text Available Bentonite is often used in water-based drilling fluids. The xanthan gum is widely used as to increase the viscosity of the bentonite suspension. For the stabilization of the drilled layers, we use filtrate reducers: sodium carboxymethylcellulose low viscosity and cellulose polyanionic low viscosity. The objective of this work is to explain the effect of the polymers on the rheological behavior of the 5% bentonite suspensions. These results will provide practical recommendations for the rational use of different types of additives in water-based drilling muds. Our work is based on rheological trials on a viscometer. The results obtained on the bentonite 5%-xanthane suspension show a rheofluidifying behavior with yield stress conform to the Herschel-bulckly modal. While increasing the concentration of filtrate reducer decreases the yield stress and reduces the viscosity. The effect of CMC LV is more pronounced than PAC L.

  5. Estimation of flows cavitation by similarity test: application to the couple water/liquid sodium

    International Nuclear Information System (INIS)

    Espanet, Laurent

    2000-01-01

    In many industrial flows, it is necessary to perform a similarity test to assess the cavitation intensity. The similarity laws must take into account the physical properties of the two fluids and especially the surface tension. To obtain such laws, a continuous description of the two-phase flow has been used. The non linear model of Cahn and Hilliard allows to obtain the equations of motion for these fluids, and non-dimensional conditions yield to the similarity laws. This method has been checked experimentally by achieving two different flows through a diaphragm. When respecting the similarity laws, the comparison of these flows allows to validate the transposition laws and to determine their associated tolerance. At last, the number of dusts and particles is an important parameter for cavitation inception threshold since it leads to the number of bubbles. Using the Laplace theory, it is shown that for each value of flow rate, it exists an upper bound value for the bubble density, which allows to establish a similarity law concerning this number of bubbles. (author) [fr

  6. Reactor coolant pump transportation incident

    International Nuclear Information System (INIS)

    Noce, D.

    1992-01-01

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

  7. Experimental Investigation of Heat Transfer Characteristics of Automobile Radiator using TiO2-Nanofluid Coolant

    Science.gov (United States)

    Salamon, V.; Senthil kumar, D.; Thirumalini, S.

    2017-08-01

    The use of nanoparticle dispersed coolants in automobile radiators improves the heat transfer rate and facilitates overall reduction in size of the radiators. In this study, the heat transfer characteristics of water/propylene glycol based TiO2 nanofluid was analyzed experimentally and compared with pure water and water/propylene glycol mixture. Two different concentrations of nanofluids were prepared by adding 0.1 vol. % and 0.3 vol. % of TiO2 nanoparticles into water/propylene glycol mixture (70:30). The experiments were conducted by varying the coolant flow rate between 3 to 6 lit/min for various coolant temperatures (50°C, 60°C, 70°C, and 80°C) to understand the effect of coolant flow rate on heat transfer. The results showed that the Nusselt number of the nanofluid coolant increases with increase in flow rate. At low inlet coolant temperature the water/propylene glycol mixture showed higher heat transfer rate when compared with nanofluid coolant. However at higher operating temperature and higher coolant flow rate, 0.3 vol. % of TiO2 nanofluid enhances the heat transfer rate by 8.5% when compared to base fluids.

  8. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  9. Influence of sodium chloride on shear flow induced starch-gluten separation from Soissons wheat dough

    NARCIS (Netherlands)

    Zalm, van der E.E.J.; Goot, van der A.J.; Boom, R.M.

    2010-01-01

    Wheat dough can be separated into a starch-rich and a gluten-rich fraction by subjecting the dough to curvilinear shear flow. This paper presents the effect of salt (NaCl) addition on the shear-induced separation process. The separation (defined as the changes in protein concentration in the various

  10. TEGENA: Detailed experimental investigations of temperature and velocity distributions in rod bundle geometries with turbulent sodium flow

    International Nuclear Information System (INIS)

    Moeller, R.

    1989-02-01

    Precise knowledge of the velocity and temperature distributions is necessary in fuel element design (rod bundles with longitudinal flow). The detail codes required in the fine analysis of non-uniformly cooled bundle zones are presently at the stage of development. In order to verify these computer codes, the mean fluid temperatures and the related RMS values of the temperature fluctuations were measured in a heated bundle TEGENA, containing 4 rods arranged in one row (P/D = W/D = 1.147) with sodium cooling (Pr ≅ 0.005). The temperature distribution in the structures was determined as the necessary boundary condition for the temperature profiles in the fluid. The experiments were carried out with different types of heating (uniform load and load tilting) and the flow conditions were varied in the range from 4000 ≤ Re ≤ 76.000, 20 ≤ Pe ≤ 400. The essential process of thermal development took place under uniform load within a heated bundle length of about 100 hydraulic diameters. In the main measuring plane at the end of the heated zone, after 200 hydraulic diameters, the flow can be termed largely developed thermally. There, the temperature profiles measured in the fluid exhibit pronounced maxima in the narrowest gaps of the subchannels as well as pronounced minima in the centers of the subchannels at the unheated wall. In the zones of maximum temperature gradients the temperature fluctuations attain maximum and minimum values, respectively, at the points of disappearance of the temperature gradients. In all cases of load tilting investigated the flow at the end of the heated zone had not yet developed thermally. By inspection of all thermocouples in isothermal experiments performed at regular intervals, by redundant arrangement of the mobile probe thermocouples and by demonstration of the reproducibility of results of measurement the experiments have been validated satisfactorily. (orig./GL) [de

  11. TEGENA: Detailed experimental investigations of temperature and velocity distributions in rod bundle geometries with turbulent sodium flow

    International Nuclear Information System (INIS)

    Moeller, R.

    1989-12-01

    Precise knowlege of the velocity and temperature distributions is necessary in fuel element design (rod bundles with longitudinal flow). The detail codes required in the fine analysis of non-uniformly cooled bundle zones are presently at the stage of development. In order to verify these computer codes, the mean fluid temperatures and the related RMS values of the temperature fluctuations were measured in a heated bundle, TEGENA, containing four rods arranged in one row (P/D = W/D = 1.147) with sodium cooling (Pr≅0.005). The temperature distribution in the structures was determined as the necessary boundary condition for the temperature profiles in the fluid. The experiments were carried out with different types of heating (uniform load and flux tilting) and the flow conditions were varied in the ranges 4000≤Re≤76,000; 20≤Pe≤400. The essential processes of thermal development took place under uniform load within a heated bundle length of about 100 hydraulic diameters. In the main measuring plane at the end of the heated zone, after 200 hydraulic diameters, the flow can be termed largely developed thermally. There, the temperature profiles measured in the fluid exhibit pronounced maxima in the narrowest gaps of the subchannels as well as pronounced minima in the centers of the subchannels at the unheated wall. In the zones of maximum temperature gradients the temperature fluctuations attain maximum and minimum values, respectively, at the points of disappearance of the temperature gradients. In all cases of flux tilting investigated the flow at the end of the heated zone had not yet developed thermally. (orig.) [de

  12. APPLICATION OF MULTIHOLE PRESSURE PROBE FOR RESEARCH OF COOLANT VELOCITY PROFILE IN NUCLEAR REACTOR FUEL ASSEMBLIES

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2015-01-01

    Full Text Available Development of heat and mass transfer intensifiers is a major engineering task in the design of new and modernization of existing fuel assemblies. These devices create lateral mass flow of coolant. Design of intensifiers affects both the coolant mixing and the hydraulic resistance. The aim of this work is to develop a methodology of measuring coolant local velocity in the fuel assembly models with different mixing grids. To solve the problems was manufactured and calibrated multihole pressure probe. The air flow velocity measuring method with multihole pressure probe was used in the experimental studies on the coolant local hydrodynamics in fuel assemblies with mixing grids. Analysis of the coolant lateral velocity vector fields allowed to study the formation of the secondary vortex flows behind the mixing grids, and to determine the basic laws of coolant flow in experimental models. Quantitative data on the coolant flow velocity distribution obtained with a multihole pressure probe make possible to determine the magnitude of the flow lateral velocities in fuel rod gaps, as well as to determine the distance at which damping occurs during mixing. 

  13. Effect of Flowing Water on Sr Sorption Changes of Hydrous Sodium Titanate

    Directory of Open Access Journals (Sweden)

    Youko Takahatake

    2017-12-01

    Full Text Available Radioactive contaminated water has been generated at the Fukushima Daiichi Nuclear Power station (F1NPS. Hydrous sodium titanate (SrTreat® is able to remove radioactive Strontium (Sr from this water. Knowing the amount of radioactive nuclides in the used as-received SrTreat® is important for effective disposal and deposition of the F1NPS waste. This study investigated changes in the ability of SrTreat® to sorb Sr, and to understand the causes of changes in the sorbing. An investigation of the Sr sorption ability of SrTreat® is important for calculating the initial radioactive inventory of used SrTreat®. This study carries out Sr sorption studies with acid-base titrations and X-ray photoelectron spectroscopy (XPS to characterize the properties. After exposure to simulated treated water for 99 h, the surface structure of the SrTreat® was changed, and the percentage of sorbed Sr and the buffer capacity for protons decreased. When the amount of radioactive nuclides contained in the used SrTreat® is calculated from the sorption data of the as-received SrTreat®, the radioactive Sr content will be overestimated with a concomitant increase in the deposition and disposal costs of the used SrTreat®.

  14. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Choi, Y D; Choi, J H; Kim, T J; Jeong, K C; Kwon, S W; Kim, B H; Jeong, J Y; Park, J H; Kim, K R; Jo, B R

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

  15. A study on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H 2 gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs

  16. Tests of an experimental steam generator heated by sodium-potassium flow

    International Nuclear Information System (INIS)

    Robin, M.G.; Duchatelle, L.; Nucheze, L. de

    1974-01-01

    The first boiler heated by liquid alkaline metal flow, installed at the C.E.A. Heat Test Centre at Grand-Quevilly, near Rouen, was a double-walled 5MW model fed with water at 220 deg C and producing superheated steam at 545 deg C and at an effective pressure of 125 bar. From 1965 to 1967, in the course of more than 8,000 hours of operation under varied conditions (specific water flow rates between 110 and 1,200kg/sq.m./sec., steam pressures of 85 and 125 bar), some 200 results of stationary operating conditions were obtained. These were compared with calculated predictions. After specifying the laws of exchange utilized, the authors discuss the heat transfer results and pressure loss values [fr

  17. Renal blood flow, fractional excretion of sodium and acute kidney injury: time for a new paradigm?

    Science.gov (United States)

    Prowle, John; Bagshaw, Sean M; Bellomo, Rinaldo

    2012-12-01

    Global renal blood flow is considered pivotal to renal function. Decreased global renal blood flow (decreased perfusion) is further considered the major mechanism of reduced glomerular filtration rate responsible for the development of acute kidney injury (AKI) in critically ill patients. Additionally, urinary biochemical tests are widely taught to allow the differential diagnosis of prerenal (functional) AKI and intrinsic [structural AKI (so-called acute tubular necrosis)]. In this review we will examine recent evidence regarding these two key clinical paradigms. Recent animal experiments and clinical studies in humans using cine-phase contrast magnetic resonance technology are not consistent with the decreased perfusion paradigm. They suggest instead that changes in the intra-renal circulation including modification in efferent arteriolar function and intra-renal shunting are much more likely to be responsible for AKI, especially in sepsis. Similarly, recent human studies indicate the urinary biochemistry has limited diagnostic or prognostic ability and is dissociated form biomarker and microscopic evidence of tubular injury. Intra-renal microcirculatory changes are likely more important than changes in global blood flow in the development of AKI. Urinary biochemistry is not a clinically useful diagnostic or prognostic tool in critically ill patients at risk of or with AKI.

  18. Vertical reactor coolant pump instabilities

    International Nuclear Information System (INIS)

    Jones, R.M.

    1985-01-01

    The investigation conducted at the Tennessee Valley Authority's Sequoyah Nuclear Power Plant to determine and correct increasing vibrations in the vertical reactor coolant pumps is described. Diagnostic procedures to determine the vibration causes and evaluate the corractive measures taken are also described

  19. Temperature noise analysis and sodium boiling detection in the fuel failure mockup

    International Nuclear Information System (INIS)

    Sides, W.H. Jr.; Fry, D.N.; Leavell, W.H.; Mathis, M.V.; Saxe, R.F.

    1976-01-01

    Sodium temperature noise was measured at the exit of simulated, fast-reactor fuel subassemblies in the Fuel Failure Mockup (FFM) to determine the feasibility of using temperature noise monitors to detect flow blockages in fast reactors. Also, acoustic noise was measured to determine whether sodium boiling in the FFM could be detected acoustically and whether noncondensable gas entrained in the sodium coolant would affect the sensitivity of the acoustic noise detection system. Information from these studies would be applied to the design of safety systems for operating liquid-metal fast breeder reactors (LMFBRs). It was determined that the statistical properties of temperature noise are dependent on the shape of temperature profiles across the subassemblies, and that a blockage upstream of a thermocouple that increases the gradient of the profile near the blockage will also increase the temperature noise at the thermocouple. Amplitude probability analysis of temperature noise shows a skewed amplitude density function about the mean temperature that varies with the location of the thermocouple with respect to the blockage location. It was concluded that sodium boiling in the FFM could be detected acoustically. However, entrained noncondensable gas in the sodium coolant at void fractions greater than 0.4 percent attenuated the acoustic signals sufficiently that boiling was not detected. At a void fraction of 0.1 percent, boiling was indicated only by the two acoustic detectors closest to the boiling site

  20. Coolant Chemistry Control: Oxygen Mass Transport in Lead Bismuth Eutectic

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  1. Vascular Type 1A Angiotensin II Receptors Control BP by Regulating Renal Blood Flow and Urinary Sodium Excretion.

    Science.gov (United States)

    Sparks, Matthew A; Stegbauer, Johannes; Chen, Daian; Gomez, Jose A; Griffiths, Robert C; Azad, Hooman A; Herrera, Marcela; Gurley, Susan B; Coffman, Thomas M

    2015-12-01

    Inappropriate activation of the type 1A angiotensin (AT1A) receptor contributes to the pathogenesis of hypertension and its associated complications. To define the role for actions of vascular AT1A receptors in BP regulation and hypertension pathogenesis, we generated mice with cell-specific deletion of AT1A receptors in smooth muscle cells (SMKO mice) using Loxp technology and Cre transgenes with robust expression in both conductance and resistance arteries. We found that elimination of AT1A receptors from vascular smooth muscle cells (VSMCs) caused a modest (approximately 7 mmHg) yet significant reduction in baseline BP and exaggerated sodium sensitivity in mice. Additionally, the severity of angiotensin II (Ang II)-dependent hypertension was dramatically attenuated in SMKO mice, and this protection against hypertension was associated with enhanced urinary excretion of sodium. Despite the lower BP, acute vasoconstrictor responses to Ang II in the systemic vasculature were largely preserved (approximately 80% of control levels) in SMKO mice because of exaggerated activity of the sympathetic nervous system rather than residual actions of AT1B receptors. In contrast, Ang II-dependent responses in the renal circulation were almost completely eliminated in SMKO mice (approximately 5%-10% of control levels). These findings suggest that direct actions of AT1A receptors in VSMCs are essential for regulation of renal blood flow by Ang II and highlight the capacity of Ang II-dependent vascular responses in the kidney to effect natriuresis and BP control. Copyright © 2015 by the American Society of Nephrology.

  2. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  3. Room Temperature, Hybrid Sodium-Based Flow Batteries with Multi-Electron Transfer Redox Reactions

    Science.gov (United States)

    Shamie, Jack S.; Liu, Caihong; Shaw, Leon L.; Sprenkle, Vincent L.

    2015-01-01

    We introduce a new concept of hybrid Na-based flow batteries (HNFBs) with a molten Na alloy anode in conjunction with a flowing catholyte separated by a solid Na-ion exchange membrane for grid-scale energy storage. Such HNFBs can operate at ambient temperature, allow catholytes to have multiple electron transfer redox reactions per active ion, offer wide selection of catholyte chemistries with multiple active ions to couple with the highly negative Na alloy anode, and enable the use of both aqueous and non-aqueous catholytes. Further, the molten Na alloy anode permits the decoupled design of power and energy since a large volume of the molten Na alloy can be used with a limited ion-exchange membrane size. In this proof-of-concept study, the feasibility of multi-electron transfer redox reactions per active ion and multiple active ions for catholytes has been demonstrated. The critical barriers to mature this new HNFBs have also been explored. PMID:26063629

  4. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  5. Coolant voiding analysis following SGTR for an HLMC reactor

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

    2000-01-01

    Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

  6. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  7. Development of a transfer model for design of sodium purification systems for Fast Breeder Reactors

    International Nuclear Information System (INIS)

    Khatcheressian, N.

    2013-01-01

    Operating a Sodium Fast Reactor (SFR) in reliable and safe conditions requires to master the quality of the sodium fluid coolant, regarding oxygen and hydrogen impurities contents. A cold trap is a purification unit in SFR, designed for maintaining oxygen and hydrogen contents within acceptable limits. The purification of these impurities is based on crystallization of sodium hydride on cold walls and sodium oxide or hydride on wire mesh packing. Indeed, as oxygen and hydrogen solubilities are nearly nil at temperatures close to the sodium fusion point, i.e. 97.8 C, on line sodium purification can be performed by crystallization of sodium oxide and hydride from liquid sodium flows. However, the management of cold trap performances is necessary to prevent from unforeseen maintenance operations, which could induce shut-down of the reactor. It is thus essential to understand how a cold trap fills up with impurities crystallization in order to optimize the design of this system and to overcome any problems during nominal operation. The objective is to develop a design and simulation tool for cold traps able to predict the location and the amount of the impurities deposited. Crystallization model involve phenomena coupling in a porous medium with hydrodynamics, heat and mass transfer, distinguishing nucleation and growth phases for each impurity. It enables to understand how thermo hydraulic conditions and growing impurities interact on each other. This analysis will adapt operating and management conditions in order to optimize purification requirements. (author) [fr

  8. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Measurement of gas-liquid two-phase flow around horizontal tube bundle using SF6-water. Simulating high-pressure high-temperature gas-liquid two-phase flow of PWR/SG secondary coolant side at normal pressure

    International Nuclear Information System (INIS)

    Ishikawa, Atsushi; Imai, Ryoj; Tanaka, Takahiro

    2014-01-01

    In order to improve prediction accuracy of analysis code used for design and development of industrial products, technology had been developed to create and evaluate constitutive equation incorporated in analysis code. The experimental facility for PWR/SG U tubes part was manufactured to measure local void fraction and gas-liquid interfacial velocity with forming gas-liquid upward two-phase flow simulating high-pressure high-temperature secondary coolant (water-steam) rising vertically around horizontal tube bundle. The experimental facility could reproduce flow field having gas-liquid density ratio equivalent to real system with no heating using SF6 (Sulfur Hexafluoride) gas at normal temperature and pressure less than 1 MPa, because gas-liquid density ratio, surface tension and gas-liquid viscosity ratio were important parameters to determine state of gas-liquid two-phase flow and gas-liquid density ratio was most influential. Void fraction was measured by two different methods of bi-optical probe and conductivity type probe. Test results of gas-liquid interfacial velocity vs. apparent velocity were in good agreement with existing empirical equation within 10% error, which could confirm integrity of experimental facility and appropriateness of measuring method so as to set up original constitutive equation in the future. (T. Tanaka)

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

    International Nuclear Information System (INIS)

    Monta, Kazuo.

    1974-01-01

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

  11. Cooling Characteristics of the V-1650-7 Engine. II - Effect of Coolant Conditions on Cylinder Temperatures and Heat Rejection at Several Engine Powers

    Science.gov (United States)

    Povolny, John H.; Bogdan, Louis J.; Chelko, Louis J.

    1947-01-01

    An investigation has been conducted on a V-1650-7 engine to determine the cylinder temperatures and the coolant and oil heat rejections over a range of coolant flows (50 to 200 gal/min) and oil inlet temperatures (160 to 2150 F) for two values of coolant outlet temperature (250 deg and 275 F) at each of four power conditions ranging from approximately 1100 to 2000 brake horsepower. Data were obtained for several values of block-outlet pressure at each of the two coolant outlet temperatures. A mixture of 30 percent by volume of ethylene glycol and 70-percent water was used as the coolant. The effect of varying coolant flow, coolant outlet temperature, and coolant outlet pressure over the ranges investigated on cylinder-head temperatures was small (0 deg to 25 F) whereas the effect of increasing the engine power condition from ll00 to 2000 brake horsepower was large (maximum head-temperature increase, 110 F).

  12. Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Navid Bozorgan

    2015-07-01

    Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

  13. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse

    2012-01-01

    the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The threedimensional governing equations for the fluid flow and the heat......The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find...... heat sink configurations reduces the coolant pumping power in the system....

  14. Trace organics in AGR coolants

    International Nuclear Information System (INIS)

    Smith, R.; Green, L.O.; Johnson, P.A.V.

    1980-01-01

    Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

  15. Sodium bicarbonate causes dose-dependent increases in cerebral blood flow in infants and children with single ventricle physiology

    Science.gov (United States)

    Buckley, Erin M.; Naim, Maryam Y.; Lynch, Jennifer M.; Goff, Donna A.; Schwab, Peter J.; Diaz, Laura K.; Nicolson, Susan C.; Montenegro, Lisa M.; Lavin, Natasha A.; Durduran, Turgut; Spray, Thomas L.; Gaynor, J. William; Putt, Mary E.; Yodh, A.G.; Fogel, Mark A.; Licht, Daniel J.

    2013-01-01

    Background Sodium bicarbonate (NaHCO3) is a common treatment for metabolic acidemia, however little definitive information exists regarding its treatment efficacy and cerebral hemodynamic effects. This pilot observational study quantifies relative changes in cerebral blood flow (rCBF) and oxy and deoxy-hemoglobin concentrations (ΔHbO2 and ΔHb) due to bolus administration of NaHCO3 in patients with mild base deficits. Methods Infants and children with hypoplastic left heart syndrome (HLHS) were recruited prior to cardiac surgery. NaHCO3 was given as needed for treatment of base deficit. Diffuse optical spectroscopies were employed for 15 minutes post-injection to non-invasively monitor ΔHb, ΔHbO2 and rCBF relative to baseline prior to NaHCO3 administration. Results Twenty-two anesthetized and mechanically ventilated HLHS patients (1 day to 4 years old) received a median (interquartile range) dose of 1.1 (0.8, 1.8) mEq/kg NaHCO3 administered intravenously over 10–20 seconds to treat a base deficit of −4 (−6, −3) mEq/l. NaHCO3 caused significant dose-dependent increases in rCBF, however population averaged ΔHb or Δ4HbO2 compared to controls were not significant. Conclusions Dose-dependent increases in cerebral blood flow (CBF) caused by bolus NaHCO3 are an important consideration in vulnerable populations wherein risk of rapid CBF fluctuations does not outweigh the benefit of treating a base deficit. PMID:23403802

  16. Steady-state, local temperature fields with turbulent sodium flow in nominal and disturbed bundle geometries with spacer grids

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1980-12-01

    The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). These are the essential results obtained: Outside the spacer grids the azimuthal temperature variations of the side and corner rods are greater by approximately the factor 10 in the bundle geometry under consideration as compared to rods in the central bundle zone. The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to the adjacent cladding tube zones. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  18. Low and high cycle fatigue behaviour of steel-X6CRNI1811 (Type 304 SS) in air and flowing sodium

    International Nuclear Information System (INIS)

    Huthmann, H.; Jenner, G.

    Strain controlled LCF-tests were performed on X6CrNi1811 steel (Type 304 SS) in air and flowing sodium in a non-isothermal sodium loop. The results measured at 550 deg. C in an upstream position of the sodium loop show an increase in fatigue life in the strain range from about 1.5% to 0.4% for the base material in the as-received condition while the welded joints remain unaffected by the environment. The LCF-behaviour of base material specimens tested at 500 deg. C in a downstream position after a prior pre-exposure to sodium (4058 h, 500 deg. C) is also improved in comparison to parallel specimens tested in air with comparable thermal aging. Load controlled HCF-tests performed on X6CrNi1811 steel at 550 deg. C show a significant increase in fatigue life in the low cycle region (Nsub(f) 6 cycles) and a higher endurance limit for the base material in sodium but no influence of the environment is indicated for the welded joints. (author)

  19. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

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

  20. A miniature inductive temperature sensor to monitor temperature noise in the coolant of an LMFBR

    International Nuclear Information System (INIS)

    Dean, S.A.; Sandham, C.W.

    1980-01-01

    A description is given of the design and performance of miniature inductive sensors developed to monitor fast temperature fluctuations in the sodium coolant above the core of a LMFBR. These instruments, designed to be installed within existing thermocouple containment thimbles, also provide a steady-state temperature indication for reactor control purposes. (author)

  1. Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Brachet, Alain.

    1981-01-01

    This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

  2. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, EC-JRC, Westerduinweg 3, P.O. Box 2, NL-0 1755 ZG Petten (Netherlands)

    2006-07-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  3. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    International Nuclear Information System (INIS)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-01-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  4. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  5. Experiments on simulation of coolant mixing in fuel assembly head and core exit channel of WWER-440 reactor

    International Nuclear Information System (INIS)

    Kobzar, L.L; Oleksyuk, D.A.

    2006-01-01

    RRC 'Kurchatov Institute' has performed coolant mixing investigation in a head of a full-size simulator of WWER-440 fuel assembly. The experiments were focused on obtaining the data important for investigating the trends in temperature difference between the value registered by a ICIS thermocouple and the value of average temperature. The completed experiments ensure representative of configuration simulation by reproducing every construction peculiar feature of flow part of fuel assembly in the domain between the lower spacing grid and thermocouple location, and also by slightly modified fuel assembly regular elements (or analogues thereof). For the purpose of effectiveness of coolant mixing assessment within the head cross section of FA simulator, we measured coolant temperature distribution both in the place where coolant flow leaves the rod bundle simulator (in 39 data points along the cross section) and in the cross section location of regular ICIS thermocouple simulator (30 data points). The testing was conducted with pressure of (90 - 95) bar, mass coolant flow rates up to 2000 kg/(m 2 .s), temperature of coolant heating in 'hot' parts of the bundle up to 35.. and differences between coolant temperature extremes measured in rod bundle simulator outlet up to 20... Temperature fields were registered in 63 conditions that differ in coolant flow and inlet coolant temperature, electrical heating rate of FA simulator, and radial coolant distribution. In certain registered conditions we simulated coolant leakage to the space between the fuel assemblies. The received test data may be important both for investigation of dependencies between the coolant temperature in regular thermocouple location or average outlet temperature in assembly head, and for validation of CFD codes or subchannel codes (Authors)

  6. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  7. TRANSENERGY S: computer codes for coolant temperature prediction in LMFBR cores during transient events

    International Nuclear Information System (INIS)

    Glazer, S.; Todreas, N.; Rohsenow, W.; Sonin, A.

    1981-02-01

    This document is intended as a user/programmer manual for the TRANSENERGY-S computer code. The code represents an extension of the steady state ENERGY model, originally developed by E. Khan, to predict coolant and fuel pin temperatures in a single LMFBR core assembly during transient events. Effects which may be modelled in the analysis include temporal variation in gamma heating in the coolant and duct wall, rod power production, coolant inlet temperature, coolant flow rate, and thermal boundary conditions around the single assembly. Numerical formulations of energy equations in the fuel and coolant are presented, and the solution schemes and stability criteria are discussed. A detailed description of the input deck preparation is presented, as well as code logic flowcharts, and a complete program listing. TRANSENERGY-S code predictions are compared with those of two different versions of COBRA, and partial results of a 61 pin bundle test case are presented

  8. Carbon in sodium: a status review of the USA R and D work

    International Nuclear Information System (INIS)

    McCown, J.J.; Bagnall, C.

    1979-01-01

    A review is presented concerning R and D work on carbon in sodium with reference to LMFBR primary coolant circuits. The chemistry of carbon in sodium, analysis and monitoring of carbon in sodium, carbon meters, and problems of carbon in sodium are described. 31 references

  9. Simultaneous release of diclofenac sodium and papaverine hydrochloride from tablets and pellets using the flow-through cell apparatus described by dimensionless equations.

    Science.gov (United States)

    Kasperek, Regina

    2011-01-01

    The release of diclofenac sodium and papaverine hydrochloride from tablets and pellets using the flow-through cell apparatus was studied. The influence of excipients and of a size of the solid dosage forms on the amount of the released substances at the intervals of time using the different rates of flow of the dissolution medium was investigated. Physical parameters corresponding to the dissolution process as the mass transfer coefficient, the thickness of the boundary diffusion layer and the concentration of the saturated solution at this layer were calculated. The results of release were described by dimensionless equations.

  10. Coolant controls of a PEM fuel cell system

    Science.gov (United States)

    Ahn, Jong-Woo; Choe, Song-Yul

    When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

  11. Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

    International Nuclear Information System (INIS)

    Chvetsov, I.; Kouznetsov, I.; Volkov, A.

    2000-01-01

    GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

  12. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  13. Feasibility study on the type of KALIMER coolant circulation pump

    International Nuclear Information System (INIS)

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

    1997-07-01

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

  14. Friction and wear in sodium

    International Nuclear Information System (INIS)

    Hoffman, N.J.; Droher, J.J.

    1973-01-01

    In the design of a safe and reliable sodium-cooled reactor one of the more important problem areas is that of friction and wear of components immersed in liquid sodium or exposed to sodium vapor. Sodium coolant at elevated temperatures may severely affect most oxide-bearing surface layers which provide corrosion resistance and, to some extent, lubrication and surface hardness. Consequently, accelerated deterioration may be experienced on engaged-motion contact surfaces, which could result in unexpected reactor shutdown from component malfunction or failure due to galling and seizure. An overall view of the friction and wear phenomena encountered during oscillatory rubbing of surfaces in high-temperature, liquid-sodium environments is presented. Specific data generated at the Liquid Metal Engineering Center (LMEC) on this subject is also presented. (U.S.)

  15. Toxicology of plutonium-sodium

    International Nuclear Information System (INIS)

    Hackett, P.L.

    1982-01-01

    Scenarios for liquid-metal fast breeder reactor (LMFBR) accidents predict the loss of sodium coolant, with subsequent core melt-down and release of mixed sodium-fuel aerosols [Na-(PuU)O 2 ] into the environment. Studies in other laboratories demonstrated that mixed aerosols of Na 2 O-PuO 2 were more readily transported from the lung than PuO 2 aerosols. We therefore devised a continuous aerosol-generating system for animal exposures in which laser-generated fuel aerosols were swept through sodium vapor to form sodium-fuel aerosols. These fuel and sodium-fuel aerosols were compared with regard to their physicochemical properties and their biological behavior following inhalation studies in rats and dogs

  16. Coolant inlet device for nuclear reactors

    International Nuclear Information System (INIS)

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

    1969-01-01

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

  17. Effect of sodium nitroprusside-induced hypotension on the blood flow in subcutaneous and intramuscular BT4An tumors and normal tissues in rats

    International Nuclear Information System (INIS)

    Krossnes, Baard Kronen; Mella, Olav; Tyssebotn, Ingvald

    1996-01-01

    Purpose: To examine the effect of infusion of the vasodilator sodium nitroprusside (SNP) on the blood flow in normal tissues and BT 4 An tumors growing subcutaneously or intramusculary in BD IX rats. Methods and Materials: Sodium nitroprusside was given as a continuous intravenous infusion to keep the mean arterial pressure stable at 60 mmHg. The cardiac output, organ blood flow, and perfusion of the BT 4 An tumors were measured by injection of radiolabelled microspheres at control conditions and after 20 min SNP infusion in each animal. Two series of experiments were performed with two anesthetics with different mechanisms of action, Inactin and the midazolam-fentanyl-fluanisone combination (MFF), to secure reliable conclusions. Results: Cardiac output, heart rate, and blood flow to the skeletal muscles, heart, and liver increased during SNP infusion in either anesthetic group. In the kidneys and particularly in the skin, decreased blood flow by SNP was observed. When located subcutaneously on the foot, the blood flow in the tumor fell to 23.4% and 21.4% of the control values in the MFF- and Inactin-anesthetized animals, respectively. This was accompanied by a similar fall in the blood flow in the foot (tumor bed) itself. In the intramuscular tumor the blood flow fell to 24.8% of the control value in the MFF group, whereas the corresponding figure was 36.2% in the Inactin group. In the surrounding muscle (tumor bed) the blood flow increased significantly, most pronounced in the MFF experiment, where it was tripled. Conclusion: The fall in the tumor perfusion by SNP may be exploited therapeutically to increase the tumor temperature during hyperthermia. Predominant heating of the tumor compared to the tumor bed can be expected if the tumor is growing in or near skeletal muscles

  18. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

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

    1991-04-01

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

  19. Physical properties of organic coolants

    International Nuclear Information System (INIS)

    Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

    1963-03-01

    Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

  20. Flow injection determination of diclofenac sodium based on its sensitizing effect on the chemiluminescent reaction of acidic potassium permanganate-formaldehyde.

    Science.gov (United States)

    Song, Jingjing; Sun, Pulv; Ji, Zhongling; Li, Jianguo

    2015-02-01

    A sensitive and simple chemiluminescent (CL) method for the determination of diclofenac sodium has been developed by combining the flow injection technique and its sensitizing effect on the weak CL reaction between formaldehyde and acidic potassium permanganate. A calibration curve is constructed for diclofenac sodium under optimized experimental parameters over the range 0.040-5.0 µg/mL and the limit of detection is 0.020 µg/mL (3σ). The inter-assay relative standard deviation for 0.040 µg/mL diclofenac sodium (n = 11) is 2.0%. This method is rapid, sensitive, simple, and shows good selectivity and reproducibility. The proposed method has been successfully applied to the determination of the studied diclofenac sodium in pharmaceutical preparations with satisfactory results. Furthermore, the possible mechanism for the CL reaction has been discussed in detail on the basis of UV and CL spectra. Copyright © 2014 John Wiley & Sons, Ltd.

  1. Effects of sodium bicarbonate on the end-tidal CO2, PaCO2, HCO3-, PH and cerebral blood flow

    International Nuclear Information System (INIS)

    Komatani, Akio; Akutsu, Tooru; Yoshida, Michihiko; Yamaguchi, Koichi; Seo, Hiroshi

    1992-01-01

    To estimate the quantitative reactivity of cerebral blood flow (CBF), the effects of sodium bicarbonate on the end-tidal CO 2 , arterial partial pressure of CO 2 (PaCO 2 ), HCO 3 - , pH and CBF were examined. The CBF was measured by 133 Xe inhalation method with ring type SPECT (HEADTOME). Activation study with sodium bicarbonate administration was performed after 30 minutes of resting study, and the reactivity of each parameters was investigated. The arterial HCO 3 - and pH increased with similar reactivity, but PaCO 2 , end-tidal CO 2 and CBF in the non-injured hemisphere changed with irregular reactivity. The excellent correlation between PaCO 2 and end-tidal CO 2 was vanished by the administration of sodium bicarbonate. The reactivity of CBF did not correlate with reactivity of PaCO 2 and end-tidal CO 2 , but correlated with arterial HCO 3 - and pH. Thus the measurement of arterial HCO 3 - and pH may be indispensable to estimate the CBF reactivity with the administration of sodium bicarbonate. (author)

  2. Effects of sodium bicarbonate on the end-tidal CO[sub 2], PaCO[sub 2], HCO[sub 3][sup -], PH and cerebral blood flow

    Energy Technology Data Exchange (ETDEWEB)

    Komatani, Akio; Akutsu, Tooru; Yoshida, Michihiko; Yamaguchi, Koichi; Seo, Hiroshi (Yamagata Univ. (Japan). School of Medicine)

    1992-09-01

    To estimate the quantitative reactivity of cerebral blood flow (CBF), the effects of sodium bicarbonate on the end-tidal CO[sub 2], arterial partial pressure of CO[sub 2] (PaCO[sub 2]), HCO[sub 3][sup -], pH and CBF were examined. The CBF was measured by [sup 133]Xe inhalation method with ring type SPECT (HEADTOME). Activation study with sodium bicarbonate administration was performed after 30 minutes of resting study, and the reactivity of each parameters was investigated. The arterial HCO[sub 3][sup -] and pH increased with similar reactivity, but PaCO[sub 2], end-tidal CO[sub 2] and CBF in the non-injured hemisphere changed with irregular reactivity. The excellent correlation between PaCO[sub 2] and end-tidal CO[sub 2] was vanished by the administration of sodium bicarbonate. The reactivity of CBF did not correlate with reactivity of PaCO[sub 2] and end-tidal CO[sub 2], but correlated with arterial HCO[sub 3][sup -] and pH. Thus the measurement of arterial HCO[sub 3][sup -] and pH may be indispensable to estimate the CBF reactivity with the administration of sodium bicarbonate. (author).

  3. In-pile TREAT Test L04: simulating a lead sub-assembly in an unprotected LMFBR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Tylka, J.P.; Bauer, T.H.; Wright, A.E.; Davies, A.L.; Herbert, R.; Woods, W.J.

    1983-01-01

    Test L04 in the PFR/TREAT series is the first multi-pin, in-pile simulation of a LMFBR transient undercooling/overpower (TUCOP) accident using full length prototypic fuel irradiated in a fast reactor. L04 is a gridded 7-pin bundle test performed in the ANL Mk-III integral loop in a flowing sodium environment and uses prototypic, bottom plenum, UK reactor fuel, preirradiated in the PFR to an axial peak burn-up of 4.2 a/o. The objective of L04 was the study, by simulation, of coolant voiding and fuel motion during the initiating phase of a hypothetical TUCOP accident in a large LMFBR. Test L04 is intended to study the behavior of a centrally located, lead subassembly with the highest power-to-flow ratio

  4. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

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

  5. Fuel-coolant interaction-phenomena under prompt burst conditions. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, H.; Young, M.F.; Reil, K.O.

    1979-01-01

    The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2 (190 MPa peak).

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

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

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

  7. 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)

  8. Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor. Influence of flow collector of backup CR guide tube

    International Nuclear Information System (INIS)

    Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

    2016-01-01

    Design study of an advanced large-scale sodium-cooled fast reactor (SFR) has been conducted in JAEA. In the region between the bottom of the Upper Internal Structure (UIS) and the core outlet, the hot sodium from the fuel subassembly mixes with the cold sodium from the neighbor control rod (CR) channel. Therefore, temperature fluctuation due to mixing fluids at different temperatures may cause high cycle thermal fatigue at the bottom of the UIS. In the advanced design, installation of a flow guide structure named Flow-Collector (FC) to the backup control rod (BCR) guide tube is considered to enhance reliable operation of self-actuated shutdown system (SASS) and to ensure reactor shutdown operation. Previously, water experiments without the FC model had been examined in JAEA to investigate effective countermeasures to the significant temperature fluctuation generation at the bottom of the UIS. Since the FC may affect the thermal mixing behavior at the bottom of the UIS, influence of the FC on characteristics of the temperature fluctuation around the BCR channels was investigated using a water experimental facility with structure model of the FC. Through the experiment, small influence of the FC on the temperature fluctuation distribution at the bottom of the UIS was indicated. (author)

  9. Experimental studies of flow induced vibrations of the fuel assembly for the PEC reactor

    International Nuclear Information System (INIS)

    Pitimada, D.; Presaghi, M.; Tampone, O.; Cesari, F.

    1977-01-01

    The vibration behaviour of an assembly of seven mock-up fuel bundles of PEC reactor has been investigated. The assembly was excited by a parallel flow of water simulating sodium. The motion of the group (or of a single bundle in the group) has been measured in transverse sections detecting two orthogonal components of displacement. During the experiences the following parameters were varied: bundle foot and pads restraints, flow rate condition, coolant flow outlet conditions at the head of fuel bundles. Experimental data were processed in order to obtain: trajectories of three points of fuel bundle axis, power density spectra of measured vibration amplitudes, correlations between coolant flow rate and vibration amplitude R.M.S. (author)

  10. Theoretical studying the stability of steady-state regime of a channel with a coolant condensation

    International Nuclear Information System (INIS)

    Savikhin, O.G.

    1987-01-01

    Based on the boiling channel stability theory, the channel steady-state stability with the coolant condensation is studied. Condensable coolants are used in the NPP steam-separator superheaters as well as in cryogenic technique. Under certain conditions the coolant flow rate and temperature fluctuations may be excited in the parallel channel system with coolant condensation, which produce a sufficient effect on the heat exchange equipment operation reliability. To describe unsteady processes of heat and mass transfer in the channel, a homogeneous two-phase flow one dimensional model is used. The results obtained allow one to make a conclusion concerning the effect of some parameters on condensing channel steady-state regime stability: reduction of inlet and outlet unheated communication length, pressure drop increase at the outlet plate and its reduction at the inlet one lead to the increase of stability margin

  11. Sound velocity in the coolant of boiling nuclear reactors

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  12. Contact condensation effects in the main coolant pipe

    International Nuclear Information System (INIS)

    Haefner, W.; Fischer, K.

    1990-01-01

    Contact condensation effects may occur in a pressurized water reactor (PWR) after a loss of coolant accident (LOCA) when emergency core cooling (ECC) water is injected contact with escaping steam which is generated within the core. The condensation which takes place may cause a sudden depressurization leading to the formation of water slugs. The interaction between the transient condensation and the inertia of the flow may also result in large amplitude flow and pressure oscillations. These contact condensation effects are of great importance for the mass flow distribution and the coolant water supply to the reactor core. To examine those complex processes, large computer codes are necessary. The development and verification of analytical models requires greatly simplified flow boundary conditions from experiments and a sufficiently large base of experimental data. Separate models have been developed for interfacial exchange of mass, momentum and energy with respect to the associated flow regime. Therefore, an adequate description of the condensation process requires the modeling of two different topics: the prediction of the flow regime and the calculation of the interfacial exchange. (author)

  13. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

    The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

  14. PUMP: analog-hybrid reactor coolant hydraulic transient model

    International Nuclear Information System (INIS)

    Grandia, M.R.

    1976-03-01

    The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

  15. Organic coolant in Winnipeg riverbed sediments

    International Nuclear Information System (INIS)

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

    1979-03-01

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

  16. New Configurations of Micro Plate-Fin Heat Sink to Reduce Coolant Pumping Power

    Science.gov (United States)

    Rezania, A.; Rosendahl, L. A.

    2012-06-01

    The thermal resistance of heat exchangers has a strong influence on the electric power produced by a thermoelectric generator (TEG). In this work, a real TEG device is applied to three configurations of micro plate-fin heat sink. The distance between certain microchannels is varied to find the optimum heat sink configuration. The particular focus of this study is to reduce the coolant mass flow rate by considering the thermal resistances of the heat sinks and, thereby, to reduce the coolant pumping power in the system. The three-dimensional governing equations for the fluid flow and the heat transfer are solved using the finite-volume method for a wide range of pressure drop laminar flows along the heat sink. The temperature and the mass flow rate distribution in the heat sink are discussed. The results, which are in good agreement with previous computational studies, show that using suggested heat sink configurations reduces the coolant pumping power in the system.

  17. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  18. Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

    Directory of Open Access Journals (Sweden)

    Zhi Li

    2017-09-01

    Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

  19. Experimental determination of local temperature field variations due to spacer grids in the cladding tubes of a rod cluster flowed through by sodium

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.

    1978-01-01

    If spacer grids are used to keep the fuel rods in their places - as in the fuel elements of the SNR series, exact tests are necessary to find out whether and to what extent temperature peaks near the supporting points affect cladding tube design. To clarify this special problem, experimental investigations have been carried out for the first time in a rod cluster model of the SNR-300 fuel element cross-flowed with sodium. The investigations and findings so far are reported on. (orig./RW) [de

  20. 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)

  1. 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)

  2. Numerical model simulation of atmospheric coolant plumes

    International Nuclear Information System (INIS)

    Gaillard, P.

    1980-01-01

    The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr

  3. Speed control device for coolant recycling pump

    International Nuclear Information System (INIS)

    Kageyama, Takao.

    1992-01-01

    The present invention intends to increase a margin relative of the oscillations of neutron fluxes when the temperature of feedwater is lowered in a compulsory recycling type BWR reactor. That is, when the operation point represented by a reactor thermal power and a reactor core inlet flow rate is in a state approximate to an oscillation limit of the reactor power, the device of the present invention controls the recycling pump speed in the increasing direction depending on the lowering range of the feedwater temperature from a stationary state. With such a constitution, even if the reactor power is in the operation region near the oscillation limit in the BWR type reactor and a feedwater heating loss is caused, the speed of the coolant recycling pump is increased by 10% at the maximum depending on the extent of the reduction of the feedwater temperature, so that the oscillation of the reactor power can be prevented from lasting for a long period of time even if a reactivity external disturbance should occur in the reactor. (I.S.)

  4. Method of decontaminating primary coolant circuits

    International Nuclear Information System (INIS)

    Ishibashi, Masaru; Sumi, Masao.

    1981-01-01

    Purpose: To eliminate hard contaminated layers as well as soft contaminated layers without injuring substrate materials, upon decontamination of radiation contaminated portions in equipments and pipeways constituting primary coolant circuits. Constitution: High pressure water from a high pressure pump is jetted out from the nozzle of a spray gun to the radiation contaminated portions in equipments, for example, to the surface of water chamber in a vapor evaporator. High pressure pure water or aqueous boric acid is jetted out from the periphery and boric oxide particles (of about 1 - 100 μ particle size) are jetted out from the center of the nozzle of the spray gun. The particles (blasting material) jetted out together with the high pressure water impinge on the contaminated surfaces to remove the contaminated layers. Upon impingement, the high pressure water acts as the shock absorber for the blasting material and, after the impingement, it flows down to the bottom of the water chamber, and the blasting material is dissolved in the high pressure water. (Horiuchi, T.)

  5. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    International Nuclear Information System (INIS)

    DeVan, J.H.

    1979-01-01

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

  6. Coolant cleanup method in a nuclear reactor

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  7. Application of heat-resistant non invasive acoustic transducers for coolant control in the NPP pipelines

    International Nuclear Information System (INIS)

    Melnikov, V.; Nigmatulin, B.

    1997-01-01

    The use of ultrasonic waves enables remote testing of the coolant flow, detection of solid and gaseous occlusions and measuring of the water velocity and level. Analysis of the acoustic noise makes it possible to detect coolant leaks and diagnose the state and operation of the rotating mechanisms and bearings. Results are given of the research in the development of highly reliable waveguide-type non-invasive acoustic transducers with a long service life. Examples are given of the use of transducers in various fields of nuclear technology: detection of gas in coolant, indication of the coolant level, control of pipe filling and drainage, measurement of liquid film velocity at the pipe inner surface. (M.D.)

  8. TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

    Science.gov (United States)

    Gaugler, R. E.

    1994-01-01

    As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

  9. Cerebral blood flow and metabolism during controlled hypotension with sodium-nitroprusside and general anaesthesia for total hip replacement a.m. Charnley

    International Nuclear Information System (INIS)

    Buenemann, L.; Jensen, K.; Thomsen, L.; Riisager, S.

    1987-01-01

    Cerebral blood flow (CBF) and cerebral metabolic rate of oxygen (CMRo 2 ) were studied during hypotension induced with sodium nitroprusside (SNP) in 10 patients undergoing total hip replacement a.m. Charnley. Cerebral blood flow was measured using an injection of xenon-133 into an arm vein. The decay curve was detected by five scintillation counters placed over each hemisphere and analysed with the Novo 10a cerebrograph. Blood samples were drawn from the radial artery and the jugular venous bulb to calculate the CMRo 2 . In the gropu as a whole, there were significant decreases in mean arterial pressure and in cerebrovascular resistance. There were no significant changes, in either CBF or CMRo 2 in the gropu as a whole, but there were substantial individual differences. In conclusion, the use of SNP-induced hypotension for extracranial surgery should be used only in patients monitored closely. (author)

  10. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

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

  11. EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1977-09-01

    In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

  12. 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)

  13. Creep in sodium

    International Nuclear Information System (INIS)

    Charnock, W.; Cordwell, J.E.

    1978-03-01

    Available information on the creep of austenitic, ferritic and Alloy-800 type steels in liquid sodium is critically reviewed. Creep properties of stainless steels can be affected by element transfer and corrosion. At reactor structural component temperatures environmental effects are likely to be less important than changes due to thermal ageing. At high clad temperatures (700 0 C) decarburisation may cause the loss of strength and ductility in unstabilised steels while cavity formation may cause embrittlement in stabilised steels. The properties of Alloy 800 are, in some experiments, found to deteriorate while in others they are enhanced. This may be a consequence of the metallurgical complexity of the material or arise from the nature of the various techniques employed. Low alloy ferritic steels tend to decarburise in sodium at temperatures greater than 500 0 C and this leads to loss of strength and an increase in ductility. High alloy ferritics are immune to this effect and appear to be able to tolerate a degree of carburisation. Although intergranular cracking may be enhanced in liquid sodium the mechanical consequences are not significant and evidence for the existence of an embrittlement effect not associated with element transfer or corrosion is weak. Stress and strain may enhance element transfer at crack tips. However in real cracks the gettering or supply action of the crack faces conditions the chemistry of the cracks in sodium and protects the crack tip from element transfer. Thus creep crack extension rates should be independent of changes in bulk coolant chemistry. (author)

  14. Method of injecting iron ion into reactor coolant

    International Nuclear Information System (INIS)

    Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

    1988-01-01

    Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

  15. Corrosion performance of advanced structural materials in sodium.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L. (Nuclear Engineering Division)

    2012-05-16

    This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux

  16. Corrosion performance of advanced structural materials in sodium

    International Nuclear Information System (INIS)

    Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L.

    2012-01-01

    This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and

  17. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  18. Coolant processing device for nuclear reactor

    International Nuclear Information System (INIS)

    Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

    1981-01-01

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

  19. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

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

  20. Selection of nuclear reactor coolant materials

    International Nuclear Information System (INIS)

    Shi Lisheng; Wang Bairong

    2012-01-01

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

  1. Standardized sampling system for reactor coolants

    International Nuclear Information System (INIS)

    Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

    1982-09-01

    A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

  2. Liquid metal cooled nuclear power plant with direct heat transfer from the primary coolant to the working medium

    International Nuclear Information System (INIS)

    Hahn, G.

    1974-01-01

    The cooling systems of the sodium-cooled reactor are entirely inside a containment. The heat transfer from the primary to the secondary coolant - i.e. water - is done in heat exchangers with three-layer tubes. As there is no component cooling heat exchanger, it is advantageous that the layers that are in touch with the primary coolant form part of the wall of the containment. An emergency cooling system inside the containment is also made of three-layer tubes. The tubes of the primary loops have the shape of loops, helices, and spirals surrounding the reactor tank or a biological shield. Between the tubes and the safety wall there are maintenance areas which are accessible from the outside. The three-layer construction prevents a reaction of leaked-out or evaporated sodium with the secondary coolant. (DG) [de

  3. A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui, E-mail: rhu@anl.gov; Yu, Yiqi

    2016-11-15

    Highlights: • Developed a computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The accuracy and efficiency of the method is confirmed with a 7-assembly test problem. • The effects of different spatial discretization schemes are investigated and compared to the RANS-based CFD simulations. - Abstract: For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

  4. Single-phase sodium pump model for LMFBR thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Agrawal, A.K.

    1979-01-01

    A single-phase, homologous pump model has been developed for simulation of safety-related transients in LMFBR systems. Pump characteristics are modeled by homologous head and torque relations encompassing all regimes of operation. These relations were derived from independent model test results with a centrifugal pump of specific speed equal to 35 (SI units) or 1800 (gpm units), and are used to analyze the steady-state and transient behavior of sodium pumps in a number of LMFBR plants. Characteristic coefficients for the polynomials in all operational regimes are provided in a tabular form. The speed and flow dependence of head is included through solutions of the impeller and coolant dynamic equations. Results show the model to yield excellent agreement with experimental data in sodium for the FFTF prototype pump, and with vendor calculations for the CRBR pump. A sample pipe rupture calculation is also performed to demonstrate the necessity for modeling the complete pump characteristics

  5. Coolant void effect investigation - case of a na-cooled fast reactor

    International Nuclear Information System (INIS)

    Glinatsis, G.; Gugiu, D.

    2013-01-01

    In the frame of the last EURATOM-FP7 Program, a large sized Sodium-cooled FR (SFR) has been studied. Mixed carbides fuel (U, Pu)C has been adopted for the backup core solution and important work has been also performed in order to obtain an ''optimised'' backup configuration ''close'' to the reference one, which is fueled by mixed oxides fuel (U, Pu)Ox. The peculiarity of both core designs (the reference configuration and the optimised backup configuration) is the adoption of a 60 cm Plenum zone in the upper part of each fuel assembly (FA), that is filled by coolant, in order to mitigate (when emptied) the core positive coolant void effect. This paper presents some results of a detailed study of the coolant void effect for the above SFR with mixed carbides core. Many aspects, like geometric heterogeneity, the burnup state, the operating conditions, etc., have been taken into consideration in order to obtain information about the ''propagation'' and the behaviour of the coolant void effect itself. The performed study investigates also the coolant void effect consequences on some reactivity coefficients, which are important for a safe behaviour of the reactor. The investigation consisted in the steady state simulations of the reactor on different operating conditions in Monte Carlo approach. (authors)

  6. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  7. Continuous method for refining sodium. [for use in LMFBR type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Batoux, B; Laurent-Atthalin, A; Salmon, M

    1973-11-16

    The invention relates to a refining method according to which commercial sodium provides a high purity sodium with, in particular, a very small calcium content. The method consists in continuously feeding a predetermined amount of sodium peroxide into a sodium stream, mixing and causing said sodium peroxide to reach with sodium at an appropriate temperature, and, finally, separating the reaction products from sodium by decanting and filtering same. The thus obtained high purity sodium meets the requirements of atomic industries in particular, in view of its possible use as coolant in nuclear reactors of the ''breeder'' type.

  8. Experimental investigation of thermoelectric power generation versus coolant pumping power in a microchannel heat sink

    DEFF Research Database (Denmark)

    Kolaei, Alireza Rezania; Rosendahl, Lasse; Andreasen, Søren Juhl

    2012-01-01

    The coolant heat sinks in thermoelectric generators (TEG) play an important role in order to power generation in the energy systems. This paper explores the effective pumping power required for the TEGs cooling at five temperature difference of the hot and cold sides of the TEG. In addition......, the temperature distribution and the pressure drop in sample microchannels are considered at four sample coolant flow rates. The heat sink contains twenty plate-fin microchannels with hydraulic diameter equal to 0.93 mm. The experimental results show that there is a unique flow rate that gives maximum net-power...

  9. Coolant and ambient temperature control for chillerless liquid cooled data centers

    Science.gov (United States)

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

    2016-02-02

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

  10. Fuel-coolant interactions in a jet contact mode

    International Nuclear Information System (INIS)

    Konishi, K.; Isozaki, M.; Imahori, S.; Kondo, S.; Furutani, A.; Brear, D.J.

    1994-01-01

    Molten fuel-coolant interactions in a jet contact mode was studied with respect to the safety of liquid-metal-cooled fast reactors (LMFRs). From a series of molten Wood's metal (melting point: 79 deg. C, density: -8400 kg/m 3 ) jet-water interaction experiments, several distinct modes of interaction behaviors were observed for various combinations of initial temperature conditions of the two fluids. A semi-empirical model for a minimum film boiling temperature criterion was developed and used to reasonably explain the different interaction modes. It was concluded that energetic jet-water interactions are only possible under relatively narrow initial thermal conditions. Preliminary extrapolation of the present results in an oxide fuel-sodium system suggests that mild interactions with short breakup length and coolable debris formation should be most likely in LMFRs. (author)

  11. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  12. Experimental and numerical investigation of the coolant mixing during fast deboration transients

    International Nuclear Information System (INIS)

    Hoehne, T.; Rohde, U.; Weiss, F.P.

    1999-01-01

    For the analysis of boron dilution transients and main steam line break scenarios the modeling of the coolant mixing inside the reactor vessel is important, because the reactivity insertion strongly depends on boron acid concentration or the coolant temperature distribution. Calculations for steady state flow conditions for the VVER-440 were performed with a CFD code (CFX-4). The comparison with experimental data and an analytical mixing model which is implemented in the neutron-kinetic code DYN3D showed a good agreement for near-nominal conditions. First experiments at the Rossendorf Mixing Test Facility ROCOM were performed simulating the start-up of the first main coolant pump. The reference reactor for the geometrically 1:5 scaled Plexiglas model is the German Konvoi type PWR. After demonstrating the capability of the CFD code to simulate these complicated flow transients, calculations were performed for the start-up of the first pump in a VVER-440 type reactor. These calculations are a first step of understanding the coolant mixing in the RPV of a VVER-440 type reactor under transient conditions. The results of the calculation show a very complex flow in the downcomer. A high downcomer of VVER-440 and the existence of the lower control rod chamber support coolant mixing is concluded. (author)

  13. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  14. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

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

    International Nuclear Information System (INIS)

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

    1984-07-01

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

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

    International Nuclear Information System (INIS)

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

    1977-04-01

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

  17. Conceptural design of multipurpose sodium test loop

    International Nuclear Information System (INIS)

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

    1982-01-01

    This report describes the conceptural design of the multipurpose sodium test loop (MSTL). This MSTL consists mainly of impurity control and measurement system, corrosion and masstransfer system and heat transfer system. Problems associated with liquid sodium coolant will be studied and operating experiences will be obtained by the use of this facility. This technology will be used to evaluate safety and reliability of large sodium facility in the future. The total cost excluding the cost of building construction is estimated to 175 thousand dollars. (Author)

  18. RETRAN analysis of inter-system LOCA within the primary coolant pump

    International Nuclear Information System (INIS)

    Gangadharan, A.; Pratt, G.F.

    1992-01-01

    One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  20. 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

  1. 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

  2. Generation IV SFR Nuclear Reactors: Under Sodium Robotics for ASTRID

    International Nuclear Information System (INIS)

    Jouan-de-Kervenoael, T.; Rey, F.; Baque, F.

    2013-06-01

    For non-removable components of the future ASTRID prototype, repair operations will be performed in a gas environment. If the faulty area is located under the sodium free level, the gas-tight system will have to contain the inspection and repair tools and to protect them from the surrounding liquid sodium. Concerning repair tools, the unique laser tool has been selected for future SFRs: the repair scenario for in-sodium structures will first involve removing the sodium (after bulk draining), then machining and finally welding. Concerning conventional tools (brush or gas blower for sodium removal, milling machine for machining and TIG for welding for which its feasibility was demonstrated in the 1990's) are still considered as a back-up solution. The maintenance of future ASTRID nuclear reactor prototype (inspection, repair) will be performed during shut down periods with some robotic carriers which have to be introduced within the main vessel, in primary 200 deg. C sodium coolant with argon gas cover. Inspection campaigns will be 20 days long. These robots (or carriers) will allow bringing inspection and repairing tools up to concerned components and structures. The needed degrees of freedom associated to these operations will be assumed either directly by the carrier itself or by specifics lower end carrier device for accurate local positioning. Several carriers will be designed, well adapted to specific needs: type of maintenance operation and location of inspection and repair sites. Feedback experience was gained during Superphenix SFR operation with the MIR robot which allowed to successfully make the Non Destructive Examination of main vessel welding joints, the carrier being outside bulk sodium. Operating conditions for the ASTRID robots will be harder from those of the MIR robot: temperature ranging from 180 deg. C to 200 deg. C, radiation dose ranging from 105 to 106 Gy, but mainly direct immersion within liquid sodium coolant. At the design phase of

  3. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  4. HTGR-GT primary coolant transient resulting from postulated turbine deblading

    International Nuclear Information System (INIS)

    Cadwallader, G.J.; Deremer, R.K.

    1980-11-01

    The turbomachine is located within the primary coolant system of a nuclear closed cycle gas turbine plant (HTGR-GT). The deblading of the turbine can cause a rapid pressure equilibration transient that generates significant loads on other components in the system. Prediction of and design for this transient are important aspects of assuring the safety of the HTGR-GT. This paper describes the adaptation and use of the RATSAM program to analyze the rapid fluid transient throughout the primary coolant system during a spectrum of turbine deblading events. Included are discussions of (1) specific modifications and improvements to the basic RATSAM program, which is also briefly described; (2) typical results showing the expansion wave moving upstream from the debladed turbine through the primary coolant system; and (3) the effect on the transient results of different plenum volumes, flow resistances, times to deblade, and geometries that can choke the flow

  5. Under-Sodium-Viewing as one technique for periodic inspections in sodium-cooled fast reactors-- possibilities and limits

    International Nuclear Information System (INIS)

    Weiss, H.

    1979-07-01

    Periodic inspections are gaining increasingly technical importance for fast sodium cooled reactors. Among others the reactor tank and its internals have to be inspected, whereby licensing experts partly are requesting the standards of Light Water Reactors. This leads to difficulties in sodium cooled reactors because of the non-transparent coolant sodium and their compact structure. In order to avoid the complete dumping of the sodium, the under sodium viewing shall be applied besides other inspection methods. Since this is a new method, which is still in its development phase, this report presents and discusses the technical and physical basis and outlines possibilities and limits [de

  6. Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

    International Nuclear Information System (INIS)

    Luo Bangqi

    2012-01-01

    The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

  7. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

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

  8. Numerical investigation on critical heat flux and coolant volume required for transpiration cooling with phase change

    International Nuclear Information System (INIS)

    He, Fei; Wang, Jianhua

    2014-01-01

    Highlights: • Five states during the transpiration cooling are discussed. • A suit of applicable program is developed. • The variations of the thickness of two-phase region and the pressure are analyzed. • The relationship between heat flux and coolant mass flow rate is presented. • An approach is given to define the desired case of transpiration cooling. - Abstract: The mechanism of transpiration cooling with liquid phase change is numerically investigated to protect the thermal structure exposed to extremely high heat flux. According to the results of theoretical analysis, there is a lower critical and an upper critical external heat flux corresponding a certain coolant mass flow rate, between the two critical values, the phase change of liquid coolant occurs within porous structure. A strongly applicable self-edit program is developed to solve the states of fluid flow and heat transfer probably occurring during the phase change procedure. The distributions of temperature and saturation in these states are presented. The variations of the thickness of two-phase region and the pressure including capillary are analyzed, and capillary pressure is found to be the main factor causing pressure change. From the relationships between the external heat flux and coolant mass flow rate obtained at different cooling cases, an approach is given to estimate the maximal heat flux afforded and the minimal coolant consumption required by the desired case of transpiration cooling. Thus the pressure and coolant consumption required in a certain thermal circumstance can be determined, which are important in the practical application of transpiration cooling

  9. Numerical simulation of complex multi-dimensional two-phase flows in nuclear power plant coolant circuits by means of the best-estimate thermal-hydraulic code BAGIRA

    International Nuclear Information System (INIS)

    Kalinichenko, S.D.; Kroshilin, A.E.; Kroshilin, V.E.; Smirnov, A.V.

    2009-01-01

    Recent results are exposed, obtained by applying the best-estimate thermal hydraulic code BAGIRA for three-dimensional modeling complex two-phase flow dynamics inside the vessel of the horizontal steam generator PGV-1000 used in reactor units with VVER-1000. Spatial volumetric void fraction and velocity distributions are calculated and compared with available experimental data. (author)

  10. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  11. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  12. Iron crud supply device to reactor coolant

    International Nuclear Information System (INIS)

    Baba, Takao.

    1993-01-01

    In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

  13. Limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  14. Determination of temperature distributions in fast reactor core coolants

    International Nuclear Information System (INIS)

    Tillman, M.

    1975-04-01

    An analytical method of determination of a temperature distribution in the coolant medium in a fuel assembly of a liquid-metal-fast-breeder-reactor (LMFBR) is presented. The temperature field obtained is applied for a constant velocity (slug flow) fluid flowing, parallel to the fuel pins of a square and hexagonal array assembly. The coolant subchannels contain irregular boundaries. The geometry of the channel due to the rod adjacent to the wall (edge rod) differs from the geometry of the other channels. The governing energy equation is solved analytically, assuming series solutions for the Poisson and diffusion equations, and the total solution is superposed by the two. The boundary conditions are specified by symmetry considerations, assembly wall insulation and a continuity of the temperature field and heat fluxes. The initial condition is arbitrary. The method satisfies the boundary conditions on the irregular boundaries and the initial condition by a least squares technique. Computed results are presented for various geometrical forms, with ratio of rod pitch-to-diameter typical for LMFBR cores. These results are applicable for various fast-reactors, and thus the influence of the transient solution (which solves the diffusion equation) on the total depends on the core parameters. (author)

  15. Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

    International Nuclear Information System (INIS)

    Gonzalez, Jose; Gimenez, Marcelo

    2000-01-01

    We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

  16. Measuring device for the coolant flowrate in a reactor core

    International Nuclear Information System (INIS)

    Sawa, Toshihiko.

    1983-01-01

    Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

  17. Study on the quench behavior of molten fuel material jet into coolant

    International Nuclear Information System (INIS)

    Abe, Yutaka; Kizu, Tetsuya; Arai, Takahiro; Nariai, Hideki; Chitose, Keiko; Koyama, Kazuya

    2004-01-01

    In a core disruptive accident (CDA) of a Fast Breeder Reactor, the post accident heat removal (PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. In the present experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The distributions of the fragmented droplet diameter from the molten material jet are evaluated by correcting the solidified particles. The experimental results of the mean fragmented droplet diameter are compared with the existing theories. Consequently, the fragmented droplet diameter is close to the value estimated based on the Kelvin-Helmholtz instability. Once the particle diameter of the fragmented molten material could be known from a hydrodynamic model, it becomes possible to estimate the mass ratio of the molten particle to the total injected mass by combining an appropriate heat transfer model. The heat transfer model used in the present study is composed of the fragmentation model based on the Kelvin-Helmholtz instability. The mass ratio of the molten fragment to total mass of the melted mixed oxide fuel in sodium coolant estimated in the present study is very small. The result means that most of the molten mixed oxide fuel material injected into the sodium coolant can be cooled down under the solidified temperature, that is so called quenched, if the amount of the coolant is sufficient. (author)

  18. Heat exchanger with oscillating flow

    Science.gov (United States)

    Scotti, Stephen J. (Inventor); Blosser, Max L. (Inventor); Camarda, Charles J. (Inventor)

    1993-01-01

    Various heat exchange apparatuses are described in which an oscillating flow of primary coolant is used to dissipate an incident heat flux. The oscillating flow may be imparted by a reciprocating piston, a double action twin reciprocating piston, fluidic oscillators or electromagnetic pumps. The oscillating fluid flows through at least one conduit in either an open loop or a closed loop. A secondary flow of coolant may be used to flow over the outer walls of at least one conduit to remove heat transferred from the primary coolant to the walls of the conduit.

  19. Modeling the spatial distribution of the parameters of the coolant in the reactor volume

    International Nuclear Information System (INIS)

    Nikonov, S.P.

    2011-01-01

    In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

  20. Main coolant pump testing at Ontario Hydro

    International Nuclear Information System (INIS)

    Hartlen, R.

    1991-01-01

    This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

  1. On-Line Coolant Chemistry Analysis

    International Nuclear Information System (INIS)

    LM Bachman

    2006-01-01

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level

  2. Leak detection device for reactor coolant

    International Nuclear Information System (INIS)

    Oshima, Koichiro.

    1990-01-01

    In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

  3. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  4. Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

    OpenAIRE

    Lefèvre, Grégory; Živković, Ljiljana S.; Jaubertie, Anne

    2012-01-01

    In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

  5. Cleaning of Sodium in the Cold Trap

    International Nuclear Information System (INIS)

    Kim, Byung Ho; Jeong, Ji Young; Kim, Jong Man; Choi, Byung Hae; Nam, Ho Yun

    2005-01-01

    The purpose of a cleaning process is to remove the residual sodium adhering to the component walls once it has been properly drained. It is necessary to clean and decontaminate a component, especially the large components of the primary coolant system; such as the intermediate heat exchangers and the primary pump. Improper and inadequate cleaning has in a number of cases resulted in problems in the storage, handling, and reuse of components. Several types of failures due to improper cleaning procedures have been defined in the past. Inadequate and incomplete removal of sodium results in residues which may contain metallic sodium and alkaline compounds such as sodium hydroxide, sodium oxide, sodium carbonate, and various types of alcoholates. Reinsertion of components containing these compounds into a high-temperature sodium system can result in either the intergranular penetration characteristic of a high- oxygen sodium or an accelerated corrosion due to oxygen. The methods used for cleaning sodium equipment depend on the condition and types of equipment to be cleaned and whether the equipment is to be reused. Cleaning methods are needed that will avoid a deleterious local overheating, material surface degradation or deposits, chemical, physical, or mechanical damage, and external effects. This paper discusses a steam-nitrogen gas cleaning method for the routine applications that permits the reuse of the cold trap in sodium

  6. Containment Sodium Chemistry Models in MELCOR.

    Energy Technology Data Exchange (ETDEWEB)

    Louie, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Humphries, Larry L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-04-01

    To meet regulatory needs for sodium fast reactors’ future development, including licensing requirements, Sandia National Laboratories is modernizing MELCOR, a severe accident analysis computer code developed for the U.S. Nuclear Regulatory Commission (NRC). Specifically, Sandia is modernizing MELCOR to include the capability to model sodium reactors. However, Sandia’s modernization effort primarily focuses on the containment response aspects of the sodium reactor accidents. Sandia began modernizing MELCOR in 2013 to allow a sodium coolant, rather than water, for conventional light water reactors. In the past three years, Sandia has been implementing the sodium chemistry containment models in CONTAIN-LMR, a legacy NRC code, into MELCOR. These chemistry models include spray fire, pool fire and atmosphere chemistry models. Only the first two chemistry models have been implemented though it is intended to implement all these models into MELCOR. A new package called “NAC” has been created to manage the sodium chemistry model more efficiently. In 2017 Sandia began validating the implemented models in MELCOR by simulating available experiments. The CONTAIN-LMR sodium models include sodium atmosphere chemistry and sodium-concrete interaction models. This paper presents sodium property models, the implemented models, implementation issues, and a path towards validation against existing experimental data.

  7. Liquid metal reactor development. Development of LMR coolant technology

    Energy Technology Data Exchange (ETDEWEB)

    Nam, H. Y.; Choi, S. K.; Hwang, J. s.; Lee, Y. B.; Choi, B. H.; Kim, J. M.; Kim, Y. G.; Kim, M. J.; Lee, S. D.; Kang, Y. H.; Maeng, Y. Y.; Kim, T. R.; Park, J. H.; Park, S. J.; Cha, J. H.; Kim, D. H.; Oh, S. K.; Park, C. G.; Hong, S. H.; Lee, K. H.; Chun, M. H.; Moon, H. T.; Chang, S. H.; Lee, D. N.

    1997-07-15

    Following studies have been performed during last three years as the 1.2 phase study of the mid and long term nuclear technology development plan. First, the small scale experiments using the sodium have been performed such as the basic turbulent mixing experiment which is related to the design of a compact reactor, the flow reversal characteristics experiment by natural circulation which is necessary for the analysis of local flow reversal when the electromagnetic pump is installed, the feasibility test of the decay heat removal by wall cooling and the operation of electromagnetic pump. Second, the technology of operation mechanism of sodium facility is developed and the technical analysis and fundamental experiments of sodium measuring technology has been performed such as differential pressure measuring experiment, local flow rate measuring experimenter, sodium void fraction measuring experiment, under sodium facility, the free surface movement experiment and the side orifice pressure drop experiment. A new bounded convection scheme was introduced to the ELBO3D thermo-hydraulic computer code designed for analysis of experimental result. A three dimensional computer code was developed for the analysis of free surface movement and the analysis model of transmission of sodium void fraction was developed. Fourth, the small scale key components are developed. The submersible-in-pool type electromagnetic pump which can be used as primary pump in the liquid metal reactor is developed. The SASS which uses the Curie-point electromagnet and the mock-up of Pantograph type IVTM were manufactured and their feasibility was evaluated. Fifth, the high temperature characteristics experiment of stainless steel which is used as a major material for liquid metal reactor and the material characteristics experiment of magnet coil were performed. (author). 126 refs., 98 tabs., 296 figs.

  8. A basic experimental study on combustion of suspended sodium droplet. 2

    International Nuclear Information System (INIS)

    Sato, Kenji

    1999-10-01

    For appropriate understanding and/or prediction of the combustion behavior of sodium, working as liquid coolant in fast breeder reactors, in case of leakage accident, phenomenological analyses of the behavior must be also important along with conventional engineering approach. Following our previous study in the last year, the major objective of this experimental research is to elucidate the effects of the initial temperature and diameter of droplet, and of the air flow velocity on ignition process of a sodium droplet, by exposing a suspended droplet to the air flow at room-temperature. In the experiments, a high-temperature droplet suspended from the end of a fine stainless steel nozzle of the liquid sodium supply system was exposed to an upward air flow, and the ignition and succeeding combustion phenomena were observed by using high-speed color video recording system. In the preliminary study, the effects of lighting and image data processing on obtaining pictures suitable to analyses were investigated with the apparatus used in the previous study. After the experimental apparatus was modified partially in order to expose the unreacted droplet to the air flow more quickly, main experiments were performed in synthetic dry air or oxygen-nitrogen mixture of 21% oxygen. Good quality pictures of the phenomena achieved under good conditions were recorded even for a few cases. The details of the ignition process of a sodium droplet, including the aspects of the surface and light emission, were examined, and the effects of the air flow velocity were discussed. Since number of performed experimental runs was small, the effects of the initial droplet temperature were not examined. (author)

  9. Hydrodynamics of heavy liquid metal coolant processes and filtering apparatus

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  10. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  11. Fission product release into the primary coolant

    International Nuclear Information System (INIS)

    Apperson, C.E.

    1977-01-01

    The analytic evaluation of steady state primary coolant activity is discussed. The reported calculations account for temperature dependent fuel failure in two particle types and arbitrary radioactive decay chains. A matrix operator technique implemented in the SUVIUS code is used to solve the simultaneous equations. Results are compared with General Atomic Company's published results

  12. 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

  13. 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

  14. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  15. Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

    International Nuclear Information System (INIS)

    DeMars, R.V.; Steinke, R.R.

    1975-01-01

    Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

  16. Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

    1978-01-01

    The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

  18. Simulation of IVR-ERVC and estimation method of coolant inflow to the cavity

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyunjin; Namgung, Ihn [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    In this study, the temperature distribution outside of RV wall and evaporation rate due to heat from core will be investigated. Using the universal analysis program ANSYS Fluent, the natural convection in the cavity for IVR-ERVC conditions were modelled and performed for heat transfer analysis. The aim of this study is to calculate the appropriate coolant flow so that coolant level in the cavity can be maintained at prescribed level and vessel wall temperature distribution, including RV outside wall temperature are also investigated. Reactor vessel and cavity in case of ex-vessel cooling for severe accident condition were modeled with and without insulators. The heat load into reactor vessel from corium inside of reactor lower head were obtained from MELCORE analysis and used as input B.C of CFD analysis. The Temperature gradient of reactor outer surface and evaporation rate of cooling eater was obtained from the analysis. These results can be used for further analysis of reactor vessel creep behavior and the estimate the coolant flow rate into the reactor cavity.. and The result can be used to verify the natural convection phenomena in the cavity and also to set the design parameters of cavity and coolant flow rate. The vessel outer surface temperature gradient can be also used to more accurate investigation of vessel creep behavior during severe accident condition, The result can also be used set up a strategy for severe accident managements.

  19. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  20. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants