WorldWideScience

Sample records for coolant liquid lead-lithium

  1. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

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

    Science.gov (United States)

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

    2017-10-01

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

  4. The effect of lead concentration on the corrosion susceptibility of 2 1/4Cr-1Mo steel in a lead-lithium liquid

    International Nuclear Information System (INIS)

    Wilkinson, B.D.; Edwards, G.R.; Hoffman, N.J.

    1982-01-01

    The intergranular penetration of 21/4Cr-1Mo steel by lead-lithium liquids containing 0, 17.6, and 53 w/o lead has been investigated at temperatures from 300 0 C to 600 0 C for times up to 1000 hours. Limited tests using a 99.3 w/o lead-lithium liquid were also conducted. Tempering was found to remove the susceptibility of as-quenched 21/4Cr-1Mo steel to penetration at 500 0 C by lead-lithium liquids containing up to 53 w/o lead. Penetration by the 99.3 w/o lead-lithium liquid in 1000 hours at 500 0 C was found to be negligible even when the steel was in the as-quenched condition. An Arrhenius analysis yielded the same low initial activation energy (approx. equal to25 kJ/mole) for liquids containing 0, 17.6, and 53 w/o lead. The initial penetration rate for lead-free lithium was significantly greater than that for the lead-bearing liquids, a factor thought to be related to the effect of lead on the wettability of the liquid. The same secondary activation energy (approx. equal to120 kJ/mole) was also found for the three liquids. Furthermore, the secondary penetration rate was found to be insensitive to lead content. Anomalous behavior at 500 0 C, observed in this study as well as in previous studies, is discussed, and a hypothetical explanation for the behavior is presented. (orig.)

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

    International Nuclear Information System (INIS)

    Malang, S.; Mattas, R.

    1994-06-01

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

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

    International Nuclear Information System (INIS)

    Simon, N.

    1992-06-01

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

  7. 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

  8. Insertion of lead lithium eutectic mixture in RELAP/SCDAPSIM Mod 4.0 for Fusion Reactor Systems

    International Nuclear Information System (INIS)

    Tiwari, Ashutosh; Allison, Brian; Hohorst, J.K.; Wagner, R.J.; Allison, Chris

    2012-01-01

    Highlights: ► Thermodynamic and transport properties of lead lithium eutectic mixture have been inserted in RELAP/SCDAPSIM MOD 4.0 code. ► Code results are verified for a simple pipe problem with lead lithium eutectic mixture flowing in it. ► Code is calculating the inserted properties of lead lithium eutectic mixture to a fairly good agreement. - Abstract: RELAP/SCDAPSIM Mod 4.0 code was developed by Innovative System Software (ISS) for the analysis of nuclear power plants (NPPs) cooled by light water and heavy water. Later on the code was expanded to analyze the NPPs cooled by liquid metal, in this sequence: lead bismuth eutectic mixture, liquid sodium and lead lithium eutectic mixture (LLE) are inserted in the code. This paper focuses on the insertion of liquid LLE as a coolant for NPPs in the RELAP/SCDAPSIM Mod 4.0 code. Evaluation of the code was made for a simple pipe problem connected with heat structures having liquid LLE as a coolant in it. The code is predicting well all the thermodynamic and transport properties of LLE.

  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. Method of charging instruments into liquid metal coolant

    International Nuclear Information System (INIS)

    Yamazaki, Hiroshi

    1980-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  12. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

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

  13. 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

  14. 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.

  15. 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

  16. Coolant Design System for Liquid Propellant Aerospike Engines

    Science.gov (United States)

    McConnell, Miranda; Branam, Richard

    2015-11-01

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

  17. 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

  18. 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

  19. 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.)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. 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

  2. Gas absorption and discharge behaviors of lead-lithium

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  3. 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.

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

    Science.gov (United States)

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

    2017-11-01

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

  5. Polonium release from an ATW burner system with liquid lead-bismuth coolant

    International Nuclear Information System (INIS)

    Li, N.; Yefimov, E.; Pankratov, D.

    1998-04-01

    The authors analyzed polonium release hazards in a conceptual pool-type ATW burner with liquid lead-bismuth eutectic (LBE) coolant. Simplified quantitative models are used based on experiments and real NPP experience. They found little Po contamination outside the burner under normal operating conditions with nominal leakage from the gas system. In sudden gas leak and/or coolant spill accidents, the P contamination level can reach above the regulation limit but short exposure would not lead to severe health consequences. They are evaluating and developing mitigation methods

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  7. 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

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

    Science.gov (United States)

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

    2009-01-01

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

  9. 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.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  11. Coolant circulation system for a liquid metal nuclear reactor

    International Nuclear Information System (INIS)

    DeLuca, R.A.; Garabedian, G.

    1988-01-01

    This patent describes a liquid metal circulation system comprising an electromagnetic pump comprised of: (a) an elongated cylindrical pump support housing; (b) a cylindrical pressure dome structure coaxially situated and supported within the pump support housing, having a closed, hemispherical upper end and an open, cylindrical lower end; (c) a cylindrical pump coaxially situated within the pressure dome structure including: (1) a central core body of laminated transformer steel having six peripherally equally spaced helical grooves on its outer surface extending the entire length of the central core body, (2) a multiplicity of square, ceramic insulated copper wires situated in the helical grooves, (3) electrical leads extending from the terminal ends of the square copper wires through the upper end of the pressure dome structure and to a three-phase low voltage/high amperage power source, (4) an austenitic stainless steel jacket covering the outer surface of the central core body and covering the helically coiled square copper wires, the outer stainless steel jacket and the inner surface of the pressure dome structure defining an annular flow passage

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

    Directory of Open Access Journals (Sweden)

    S. K. Pavlov

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    2003-01-01

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

  14. 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.

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

    International Nuclear Information System (INIS)

    Norajitra, P.

    1995-06-01

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

  16. 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

  17. 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

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

    International Nuclear Information System (INIS)

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

    1991-08-01

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

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

    Science.gov (United States)

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

    2015-03-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    2004-08-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Kiss, I.

    1967-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-01-01

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

  5. 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

  6. Loss of coolant accident mitigation for liquid metal cooled space reactors

    International Nuclear Information System (INIS)

    Georgevich, Vladimir; Best, Frederick; Erdman, Carl

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Casino, William A. Jr.

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  9. Feasibility of flooding the reactor cavity with liquid gallium coolant for IVR-ERVC strategy

    International Nuclear Information System (INIS)

    Park, Seong Dae; Bang, In Cheol

    2013-01-01

    Highlights: ► We investigate the feasibility of gallium liquid metal application for IVR-ERVC. ► We consider overall concerns to apply the liquid metal. ► Decay heat can be removed by flooding the reactor cavity with gallium liquid metal. -- Abstract: In this paper, a new approach replacing the ERVC coolant by a liquid metal instead of water is studied to avoid the heat removal limit of CHF during boiling of water. As the flooding material, gallium is used in terms of the melting and boiling points. Gallium has the enough low melting point of ∼29.7 °C to ensure to maintain liquid state within the containment building. A gallium storage tank for the new flooding system of the ERVC is located in higher position than one of the reactor cavity to make a passive system using the gravity for the event of a station blackout (SBO). While the decay heat from the reactor vessel is removed by gallium, the borated water which is coming out from the reactor system plays a role as the ultimate heat sink in this ERVC system. In the system, two configurations of gallium and borated water are devised depending on whether the direct contact between them occurs. In the first configuration, two fluids are separated by the block structure. The decay heat is transported from molten corium to gallium through the vessel wall. Then the heat is ultimately dissipated by boiling of water in the block structure surface facing the borated water. In the second configuration, the cavity is flooded with both borated water and gallium in the same reactor cavity space. As the result, two layers of the fluids are naturally formed by the density difference. Like the first configuration, finally the heat removal is achieved by boiling of water via gallium. The CFD analysis shows that the maximum temperature of gallium is much lower than its boiling point while the natural circulation is stably formed in two types of the configurations without any serious risk of thermal limit

  10. Liquid level measurement on coolant pipeline using Raman distributed temperature sensor

    International Nuclear Information System (INIS)

    Kasinathan, M.; Sosamma, S.; Babu Rao, C.; Murali, N.; Jayakumar, T.

    2011-01-01

    Optical fibre based Raman Distributed Temperature Sensor (RDTS) has been widely used for temperature monitoring in oil pipe line, power cable and environmental monitoring. Recently it has gained importance in nuclear reactor owing to its advantages like continuous, distributed temperature monitoring and immunity from electromagnetic interference. It is important to monitor temperature based level measurement in sodium capacities and in coolant pipelines for Fast Breeder Reactor (FBR). This particular application is used for filling and draining sodium in storage tank of sodium circuits of Fast breeder reactor. There are different conventional methods to find out the sodium level in the storage tank of sodium cooled reactors. They are continuous level measurement and discontinuous level measurement. For continuous level measurement, mutual inductance type level probes are used. The disadvantage of using this method is it needs a temperature compensation circuit. For discontinuous level measurement, resistance type discontinuous level probe and mutual inductance type discontinuous level probe are used. In resistance type discontinuous level probe, each level needs a separate probe. To overcome these disadvantages, RDTS is used for level measurement based distributed temperature from optical fibre as sensor. The feasibility of using RDTS for measurement of temperature based level measurement sensor is studied using a specially designed test set-up and using hot water, instead of sodium. The test set-up consist of vertically erected Stainless Steel (SS) pipe of length 2m and diameter 10cm, with provision for filling and draining out the liquid. Bare graded index multimode fibre is laid straight along the length of the of the SS pipe. The SS pipe is filled with hot water at various levels. The hot water in the SS pipe is maintained at constant temperature by insulating the SS pipe. The temperature profile of the hot water at various levels is measured using RDTS. The

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    International Nuclear Information System (INIS)

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-01

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

  14. Development of natural convection heat transfer correlation for liquid metal with overlying boiling coolant

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1999-01-01

    Experimental study was performed to investigate the natural convection heat transfer characteristics and the crust formation of the molten metal pool concurrent with forced convective boiling of the overlying coolant. Tests were performed under the condition of the bottom surface heating in the test section and the forced convection of the coolant being injected onto the molten metal pool. The constant temperature and constant heater input power conditions were adopted for the bottom heating. Test results showed that the temperature distribution and crust layer thickness in the metal layer are appreciably affected by the heated bottom surface temperature of the test section, but not much by the coolant injection rate. The relationship between the Nu number and Ra number in the molten metal pool region is determined and compared with the correlations in the literature, and the experiment without coolant boiling. A new correlation on the relationship between the Nu number and Ra number in the molten metal pool with crust formation is developed from the experimental data

  15. 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

  16. 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.)

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

    Science.gov (United States)

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

    1949-01-01

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

  18. Application of liquid chromatography techniques to the measurement of soluble transition metals in PWR primary coolant

    International Nuclear Information System (INIS)

    Amey, M.D.H.; Brown, G.R.

    1987-01-01

    Two chromatographic techniques have been developed, and evaluated for the on-line analysis of soluble transition metals, particularly cobalt, in PWR primary coolant. Automatic operation and control, together with data processing and storage has been achieved by interfacing a Dionex ion chromatograph to a microprocessor control system. An absolute detection limit of 0.1 ng cobalt has been obtained which, with on-line sample preconcentration (100 ml), has enabled measurements to be made down to part-per-trillion levels (0.001 ppb). Application of the techniques to PWR coolant analysis was demonstrated by a programme of work on the Half Megawatt Loop at Winfrith. During this work some aspects of the behaviour of soluble metal species have been studied in both de-oxygenated and hydrogenated conditions. The effects of changes in coolant chemistry, operating temperature, and sample line flowrates on circulating impurity levels are reported, together with the dramatic effects observed when part of the circuit pipework was replaced with new stainless steel tubing. (author)

  19. Steam generation device with heat exchange between a liquid metal coolant and the feedwater

    International Nuclear Information System (INIS)

    Malaval, C.

    1983-01-01

    The invention is particularly applicable to a liquid metal fast breeder reactor plant, the liquid metal being sodium. The steam generation device is described in detail, it allows to get an upper liquid metal level without turbulence and an easier passage for the shock wave towards the steam generator up to the liquid metal level without being laterally reflected back to the intermediate heat exchangers [fr

  20. Limets 2: a hot-cell test set-up for Liquid Metal Embrittlement (LME) studies in liquid lead alloys

    International Nuclear Information System (INIS)

    Van den Bosch, J.; Bosch, R.W.; Al Mazouzi, A.

    2008-01-01

    Full text of publication follows. In the nuclear energy sector one of the main candidate designs for the accelerator driven system (ADS) uses liquid lead or lead bismuth eutectic both as a coolant and as spallation target. In the fusion community liquid lead lithium eutectic is considered as a possible coolant for the blanket and as a tritium source. Therefore the candidate materials for such structural components should not only comply with the operating conditions but in addition need to guarantee chemical and physical integrity when coming into contact with the lead alloys. The latter phenomena can be manifested in terms of erosion/corrosion. and/or of the so called liquid metal embrittlement (LME). Thus the susceptibility to LME of the structural materials under consideration to be used in such applications should be investigated in contact with the various lead alloys. LME, if occurring in any solid metal/liquid meta] couple, is likely to increase with irradiation hardening as localised stresses and crack initiations can promote it. To investigate the mechanical response of irradiated materials in contact with a liquid metal under representative conditions, a dedicated testing facility has recently been developed and built at our centre. It consists of an instrumented hot cell. equipped with a testing machine that allows mechanical testing of active materials in contact with active liquid lead lithium and liquid lead bismuth under well controlled chemistry conditions. The specificity of the installation is to handle highly activated and contaminated samples. Also a dedicated dismantling set-up has been developed that allows to retrieve the samples from the irradiation rig without any supplementary damage. In this presentation we will focus on the technical design of this new installation, its special features that have been developed to allow testing in a hot environment and the modifications and actions that have been taken to allow testing in liquid lead-lithium

  1. 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)

  2. 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).

  3. 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)

  4. Self-cooled blanket concepts using Pb-17Li as liquid breeder and coolant

    International Nuclear Information System (INIS)

    Malang, S.; Deckers, H.; Fischer, U.; John, H.; Meyder, R.; Norajitra, P.; Reimann, J.; Reiser, H.; Rust, K.

    1991-01-01

    A blanket design concept using Pb-17Li eutectic alloy as both breeder material and coolant is described. Such a self-cooled blanket for the boundary conditions of a DEMO-reactor is under development at the Kernforschungszentrum Karlsruhe (KfK) in the frame of the European blanket development program. Results of investigations in the areas of design, neutronics, magneto-hydrodynamics, thermo-mechanics, ancillary loop systems, and safety are reported. Based on recent progress, it can be concluded that the boundary conditions of a DEMO-reactor can be met, tritium self-sufficiency can be obtained without using beryllium as an additional neutron multiplier, and tritium inventory and permeation are acceptably low. However, to complete judge the feasibility of the proposed concept, further studies are necessary to obtain a better understanding of the magneto-hydrodynamic phenomena and their effects on the thermal-hydraulic performance of a fusion reactor blanket. (orig.)

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

    Science.gov (United States)

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

    2011-01-01

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

  6. Numerical analysis of experiments with gas injection into liquid metal coolant

    International Nuclear Information System (INIS)

    Usov, E V; Lobanov, P D; Pribaturin, N A; Mosunova, N A; Chuhno, V I; Kutlimetov, A E

    2016-01-01

    Presented paper contains results of a numerical analysis of experiments with gas injection in water and liquid metal which have been performed at the Institute of Thermophysics Russian Academy of Science (IT RAS). Obtained experimental data are very important to predict processes that take place in the BREST-type reactor during the hypothetical accident with damage of the steam generator tubes, and may be used as a benchmark to validate thermo-hydraulic codes. Detailed description of models to simulate transport of gas phase in a vertical liquid column is presented in a current paper. Two-fluid model with closing relation for wall friction and interface friction coefficients was used to simulate processes which take place in a liquid during injection of gaseous phase. It has being shown that proposed models allow obtaining a good agreement between experimental data and calculation results. (paper)

  7. Development of a low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic

    International Nuclear Information System (INIS)

    Pawelko, R.; Shimada, M.; Katayama, K.; Fukada, S.; Terai, T.

    2014-01-01

    A new experimental system designed to investigate tritium mass transfer properties in materials important to fusion technology is operational at the Safety and Tritium Applied Research (STAR) facility located at the Idaho National Laboratory (INL). The tritium permeation measurement system was developed as part of the Japan/US TITAN collaboration to investigate tritium mass transfer properties in liquid lead lithium eutectic (LLE) alloy. The system is similar to a hydrogen/deuterium permeation measurement system developed at Kyushu University and also incorporates lessons learned from previous tritium permeation experiments conducted at the STAR facility. This paper describes the experimental system that is configured specifically to measure tritium mass transfer properties at low tritium partial pressures. We present preliminary tritium permeation results for α-Fe and α-Fe/LLE samples at 600degC and at tritium partial pressures between 1.0E-3 and 2.4 Pain helium. The preliminary results are compared with literature data. (author)

  8. Liquid metal MHD and heat transfer in a tokamak blanket slotted coolant channel

    International Nuclear Information System (INIS)

    Reed, C.B.; Hua, T.Q.; Black, D.B.; Kirillov, I.R.; Sidorenkov, S.I.; Shapiro, A.M.; Evtushenko, I.A.

    1993-01-01

    A liquid metal MHD (Magnetohydrodynamic)/heat transfer test was conducted at the ALEX (Argonne Liquid Metal Experiment) facility of ANL (Argonne National Laboratory), jointly between ANL and NIIEFA (Efremov Institute). The test section was a rectangular slotted channel geometry (meaning the channel has a high aspect ratio, in this case 10:1, and the long side is parallel to the applied magnetic field). Isothermal and heat transfer data were collected. A heat flux of ∼9 W/cm 2 was applied to the top horizontal surface (the long side) of the test section. Hartmann Numbers to 1050 (2 Tesla), interaction parameters to 9 x 10 3 , Peclet numbers of 10--200, based on the half-width of the small dimension (7mm), and velocities of 1--75 cm/sec. were achieved. The working fluid was NaK (sodium potassium eutectic). All four interior walls were bare, 300-series stainless steel, conducting walls

  9. Interaction of InGa liquid alloy coolant with gold coated optical materials

    International Nuclear Information System (INIS)

    Hulbert, S.L.

    1991-01-01

    A significant reaction is reported between gold surfaces and an indium-gallium eutectic liquid alloy often used to transfer heat away from optical elements in high power synchrotron radiation beamlines. This reaction proceeds at the slightly elevated temperatures (70 degrees C) typical of conservative bakeouts used to achieve ultrahigh vacuum in the chambers which house these optical elements (mirrors, gratings, and crystals) without damaging their highly precise figure and finish. The nature and extent of this reaction is discussed, based mostly on the experience gained recently during the vacuum commissioning of two mirror chambers for a VUV wiggler-based synchrotron radiation beamline. 2 refs., 5 figs

  10. Secondary coolant circuit for liquid-metal cooled reactor and steam generator for such a circuit

    International Nuclear Information System (INIS)

    Brachet, A.; Figuet, J.; Guidez, J.; Lions, N.; Traiteur, R.; Zuber, T.

    1984-01-01

    An upper buffer tank and downstream buffer tank are disposed inside the steam generators. The downstream briffer tank is annular and it surrounds and communicates with a zone of the steam generator through which the liquid metal flows towards the bottom between the exchange zone and the outlet nozzle. The pressure of the inert gas blanket in the downstream buffer volume is more important than this one in the upper buffer volume. The invention applies to fast neutron nuclear reactor cooled by sodium [fr

  11. Oxygen sensors for Heavy Liquid Metal coolants: Calibration and assessment of the minimum reading temperature

    Energy Technology Data Exchange (ETDEWEB)

    Bassini, S., E-mail: serena.bassini@enea.it; Antonelli, A.; Di Piazza, I.; Tarantino, M.

    2017-04-01

    Oxygen sensors for Heavy Liquid Metals (HLMs) such as lead and LBE (lead-bismuth eutectic) will be essential devices in future Lead Fast Reactor (LFR) and Accelerator Driven System (ADS). Potentiometric sensors based on solid electrolytes were developed in recent years to this purpose. Internal reference electrodes such as Pt-air and Bi/Bi{sub 2}O{sub 3} liquid metal/metal-oxide are among the most used but they both have a weak point: Pt-air sensor has a high minimum reading temperature around 400 °C whereas Bi/Bi{sub 2}O{sub 3} suffers from internal stresses induced by Bi volume variations with temperature, which may lead to the sensor failure in the long-term. The present work describes the performance of standard Pt-air and Bi/Bi{sub 2}O{sub 3} sensors and compares them with recent Cu/Cu{sub 2}O sensor. Sensors with Yttria Partially Stabilized Zirconia (YPSZ) electrolyte were calibrated in oxygen-saturated HLM between 160 and 550 °C and the electric potential compared to the theoretical one to define the accuracy and the minimum reading temperature. Standard Pt-air sensor were also tested using Yttria Totally Stabilized Zirconia (YTSZ) to assess the effect of a different electrolyte on the minimum reading temperature. The performance of Pt-air and Cu/Cu{sub 2}O sensors with YPSZ electrolyte were then tested together in low-oxygen HLM between 200 and 450 °C. The results showed that Pt-air, Bi/Bi{sub 2}O{sub 3} and Cu/Cu{sub 2}O sensors with YPSZ measured oxygen in HLMs down to 400 °C, 290 °C and 200 °C respectively. When the YTSZ electrolyte was used in place of the YPSZ, the Pt-air sensor measured correctly down to at least 350 °C thanks to the superior ionic conductivity of the YTSZ. When Cu/Cu{sub 2}O and Pt-air sensors were tested together in the same low-oxygen HLM between 200 and 450 °C, Cu/Cu{sub 2}O sensor worked predictably in the whole temperature range whereas Pt-air sensor exhibited a correct output only above 400 °C. - Highlights:

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  13. 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

  14. Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Heat transfer on the liquid-liquid interface between molten core pool and coolant. JAERI's nuclear research promotion program, H10-027-6. Contract research

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Saito, Yasushi

    2002-03-01

    Heat transfer experiments under steady and transient conditions were performed using molten Wood's metal and distilled water to study heat transfer on the liquid-liquid interface between molten fuel pool and coolant under severe accident conditions. In the steady state experiment, boiling curve was measured over the range from natural convection region to film boiling region. The boiling behavior was observed using a high-speed video camera. In the transient experiment, distilled water was poured onto the hot molten metal surface, and the boiling curve was obtained in the cooling process. Comparing the measured boiling curve with existing correlations and experimental data for solid-liquid and liquid-liquid systems, the following conclusions were drawn: (a) When the interface surge is negligible and oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface could be approximately reproduced by the heat transfer correlations for nucleate boiling and film boiling regions and the critical heat flux correlation for a liquid-solid system. (b) When no oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface moved towards higher wall superheat than that at the liquid-solid surface, as Novakovic et al. observed in their experiment using mercury. (c) Transient heat transfer coefficient for film boiling at the liquid-liquid surface was about 100% higher than that predicted by the heat transfer correlation for a solid-liquid system. (author)

  15. 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

  16. Numerical evaluation of various gas and coolant channel designs for high performance liquid-cooled proton exchange membrane fuel cell stacks

    International Nuclear Information System (INIS)

    Sasmito, Agus P.; Kurnia, Jundika C.; Mujumdar, Arun S.

    2012-01-01

    A careful design of gas and coolant channel is essential to ensure high performance and durability of proton exchange membrane (PEM) fuel cell stack. The channel design should allow for good thermal, water and gas management whilst keeping low pressure drop. This study evaluates numerically the performance of various gas and coolant channel designs simultaneously, e.g. parallel, serpentine, oblique-fins, coiled, parallel-serpentine and a novel hybrid parallel-serpentine-oblique-fins designs. The stack performance and local distributions of key parameters are investigated with regards to the thermal, water and gas management. The results indicate that the novel hybrid channel design yields the best performance as it constitutes to a lower pumping power and good thermal, water and gas management as compared to conventional channels. Advantages and limitation of the designs are discussed in the light of present numerical results. Finally, potential application and further improvement of the design are highlighted. -- Highlights: ► We evaluate various gas and coolant channel designs in liquid-cooled PEM fuel cell stack. ► The model considers coupled electrochemistry, channel design and cooling effect simultaneously. ► We propose a novel hybrid channel design. ► The novel hybrid channel design yields the best thermal, water and gas management which is beneficial for long term durability. ► The novel hybrid channel design exhibits the best performance.

  17. 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

  18. Use of microPCM fluids as enhanced liquid coolants in automotive EV and HEV vehicles. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mulligan, James C.; Gould, Richard D.

    2001-10-31

    Proof-of-concept experiments using a specific microPCM fluid that potentially can have an impact on the thermal management of automotive EV and HEV systems have been conducted. Samples of nominally 20-micron diameter microencapsulated octacosane and glycol/water coolant were prepared for testing. The melting/freezing characteristics of the fluid, as well as the viscosity, were determined. A bench scale pumped-loop thermal system was used to determine heat transfer coefficients and wall temperatures in the source heat exchanged. Comparisons were made which illustrate the enhancements of thermal performance, reductions of pumping power, and increases of heat transfer which occur with the microPCM fluid.

  19. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    Otte, J.N.A.; Liebmann, D.

    1989-01-01

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

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

    International Nuclear Information System (INIS)

    Buehler, L.; Horanyi, S.

    2009-01-01

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

  1. State-of-the-art for liquid-level measurements applied to in-vessel coolant level for nuclear reactors

    International Nuclear Information System (INIS)

    Anderson, R.L.

    1980-01-01

    The TMI-2 accident indicated that a direct indication of the liquid level in the reactor vessel would have told the operators that the core was being uncovered. This state-of-the-cost survey covered the following methods: heated thermocouple, differential pressure, ultrasonic, capacitance, microwave, time-domain reflectometry, and externally mounted radiation detectors

  2. Velocity profile measurement of lead-lithium flows by high-temperature ultrasonic doppler velocimetry

    International Nuclear Information System (INIS)

    Ueki, Y.; Kunugi, T.; Hirabayashi, Masaru; Nagai, Keiichi; Saito, Junichi; Ara, Kuniaki; Morley, N.B.

    2014-01-01

    This paper describes a high-temperature ultrasonic Doppler Velocimetry (HT-UDV) technique that has been successfully applied to measure velocity profiles of the lead-lithium eutectic alloy (PbLi) flows. The impact of tracer particles is investigated to determine requirements for HT-UDV measurement of PbLi flows. The HT-UDV system is tested on a PbLi flow driven by a rotating-disk in an inert atmosphere. We find that a sufficient amount of particles contained in the molten PbLi are required to successfully measure PbLi velocity profiles by HT-UDV. An X-ray diffraction analysis is performed to identify those particles in PbLi, and indicates that those particles were made of the lead mono-oxide (PbO). Since the specific densities of PbLi and PbO are close to each other, the PbO particles are expected to be well-dispersed in the bulk of molten PbLi. We conclude that the excellent dispersion of PbO particles enables in HT-UDV to obtain reliable velocity profiles for operation times of around 12 hours. (author)

  3. 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.)

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

    Science.gov (United States)

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

    2017-09-01

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

  5. 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.)

  6. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-05-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  7. Liquid argon as active shielding and coolant for bare germanium detectors. A novel background suppression method for the GERDA 0νββ experiment

    International Nuclear Information System (INIS)

    Peiffer, J.P.

    2007-01-01

    Two of the most important open questions in particle physics are whether neutrinos are their own anti-particles (Majorana particles) as required by most extensions of the StandardModel and the absolute values of the neutrino masses. The neutrinoless double beta (0νββ) decay, which can be investigated using 76 Ge (a double beta isotope), is the most sensitive probe for these properties. There is a claim for an evidence for the 0νββ decay in the Heidelberg-Moscow (HdM) 76 Ge experiment by a part of the HdM collaboration. The new 76 Ge experiment Gerda aims to check this claim within one year with 15 kg.y of statistics in Phase I at a background level of ≤10 -2 events/(kg.keV.y) and to go to higher sensitivity with 100 kg.y of statistics in Phase II at a background level of ≤10 -3 events/(kg.keV.y). In Gerda bare germanium semiconductor detectors (enriched in 76 Ge) will be operated in liquid argon (LAr). LAr serves as cryogenic coolant and as high purity shielding against external background. To reach the background level for Phase II, new methods are required to suppress the cosmogenic background of the diodes. The background from cosmogenically produced 60 Co is expected to be ∝2.5.10 -3 events/(kg.keV.y). LAr scintillates in UV (λ=128 nm) and a novel concept is to use this scintillation light as anti-coincidence signal for background suppression. In this work the efficiency of such a LAr scintillation veto was investigated for the first time. In a setup with 19 kg active LAr mass a suppression of a factor 3 has been achieved for 60 Co and a factor 17 for 232 Th around Q ββ = 2039 keV. This suppression will further increase for a one ton active volume (factor O(100) for 232 Th and 60 Co). LAr scintillation can also be used as a powerful tool for background diagnostics. For this purpose a new, very stable and robust wavelength shifter/reflector combination for the light detection has been developed, leading to a photo electron (pe) yield of as much as

  8. Liquid argon as active shielding and coolant for bare germanium detectors. A novel background suppression method for the GERDA 0{nu}{beta}{beta} experiment

    Energy Technology Data Exchange (ETDEWEB)

    Peiffer, J.P.

    2007-07-25

    Two of the most important open questions in particle physics are whether neutrinos are their own anti-particles (Majorana particles) as required by most extensions of the StandardModel and the absolute values of the neutrino masses. The neutrinoless double beta (0{nu}{beta}{beta}) decay, which can be investigated using {sup 76}Ge (a double beta isotope), is the most sensitive probe for these properties. There is a claim for an evidence for the 0{nu}{beta}{beta} decay in the Heidelberg-Moscow (HdM) {sup 76}Ge experiment by a part of the HdM collaboration. The new {sup 76}Ge experiment Gerda aims to check this claim within one year with 15 kg.y of statistics in Phase I at a background level of {<=}10{sup -2} events/(kg.keV.y) and to go to higher sensitivity with 100 kg.y of statistics in Phase II at a background level of {<=}10{sup -3} events/(kg.keV.y). In Gerda bare germanium semiconductor detectors (enriched in {sup 76}Ge) will be operated in liquid argon (LAr). LAr serves as cryogenic coolant and as high purity shielding against external background. To reach the background level for Phase II, new methods are required to suppress the cosmogenic background of the diodes. The background from cosmogenically produced {sup 60}Co is expected to be {proportional_to}2.5.10{sup -3} events/(kg.keV.y). LAr scintillates in UV ({lambda}=128 nm) and a novel concept is to use this scintillation light as anti-coincidence signal for background suppression. In this work the efficiency of such a LAr scintillation veto was investigated for the first time. In a setup with 19 kg active LAr mass a suppression of a factor 3 has been achieved for {sup 60}Co and a factor 17 for {sup 232}Th around Q{sub {beta}}{sub {beta}} = 2039 keV. This suppression will further increase for a one ton active volume (factor O(100) for {sup 232}Th and {sup 60}Co). LAr scintillation can also be used as a powerful tool for background diagnostics. For this purpose a new, very stable and robust wavelength

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

    International Nuclear Information System (INIS)

    2005-01-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  11. FY 2000 report on the results of the technology development of energy use reduction of machine tools, etc. Technology development of environmental load reduction related to water soluble lubricating oil, etc. (R and D of low energy coolant degradation prevention technology and waste liquid processing technology); 2000 nendo energy shiyo gorika kosaku kikai nado gijutsu kaihatsu seika hokokusho. Suiyosei junkatsuyu ni kakawaru kankyo fuka teigen nado gijutsu kanri (tei energy coolant fuhai boshi gijutsu oyobi haieki shori gijutsu no kenkyu kaihatsu)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    The R and D were carried out on a system by which the long life of coolant of machine tools can be achieved and a system to process waste efficiently, economically and with less environmental loads, and the FY 2000 results were summed up. In the R and D of a system to prevent degradation of low energy coolant, measurement was made of effects of the degradation prevention system at a laboratory level, and it was found out that propagation of bacteria causing the degradation can be prevented with pH kept high. Further, it was admitted that the alkali effect on metal formability was not very much. As to the coolant processing, in the present situation, most of the coolant is taken back by industrial waste processing dealers. So, the development of the low energy waste liquid processing system is earnestly desired. In the R and D of the low energy waste liquid processing system, test on characteristics evaluation was conducted about each method of systems. Subjects to be improved/solved were extracted such as the point that volatile organic matters are included in condensed water after evaporation of waste liquid and there seems to be a possibility of needing the secondary processing. (NEDO)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  13. 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

  14. 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)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  16. 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.)

  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. 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.)

  19. 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.)

  20. 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

  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. 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.)

  3. Development of lead-bismuth coolant technology for nuclear device

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  4. 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.)

  5. 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)

  6. 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)

  7. 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

  8. 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)

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

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

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

  10. Sloshing of coolant in a seismically isolated reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  11. Low-activation lead coolant for advanced small modular NPP

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. Some observations on simulated molten debris-coolant layer dynamics

    International Nuclear Information System (INIS)

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

    1983-04-01

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

  13. 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.)

  14. Interaction of alumina with liquid Pb{sub 83}Li{sub 17} alloy

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Uttam, E-mail: uttamj@barc.gov.in [Fusion Reactor Materials Section, Bhabha Atomic Research Centre, Mumbai 400085 (India); Mukherjee, Abhishek; Sonak, Sagar; Kumar, Sanjay [Fusion Reactor Materials Section, Bhabha Atomic Research Centre, Mumbai 400085 (India); Mishra, Ratikant [Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Krishnamurthy, Nagaiyar [Fusion Reactor Materials Section, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-11-15

    Highlights: • The role of oxygen in the interaction of alumina with Pb{sub 83}Li{sub 17} alloy was studied. • Li of Pb{sub 83}Li{sub 17} alloy undergoes oxidation even in flowing high pure argon atmosphere. • It was seen that alumina reacts with Pb{sub 83}Li{sub 17} alloy at 550 °C to form LiAlO{sub 2} compound. • The reaction is rapid in the presence of oxygen and happens more slowly in the presence of flowing argon. - Abstract: Eutectic lead lithium (Pb{sub 83}Li{sub 17}) alloy is being considered a coolant, neutron multiplier and tritium breeder for International Thermonuclear Experimental Reactor (ITER) and Fusion Power Reactors (FPR). In order to reduce the magneto-hydrodynamic drag (MHD) and to prevent corrosion of structural materials due to the flow of lead lithium (Pb{sub 83}Li{sub 17}) alloy, alumina (Al{sub 2}O{sub 3}) is proposed as a candidate ceramic coating material. Interaction of liquid Pb{sub 83}Li{sub 17} alloy with Al{sub 2}O{sub 3} at the operating temperature of these reactors is therefore an important issue. The present paper deals with the characterization of Pb{sub 83}Li{sub 17} alloy and its interaction with Al{sub 2}O{sub 3} at the reactor operating temperature. The interaction was studied using EPMA, XRD and thermal analysis technique. The result indicates that alumina can interact with Pb{sub 83}Li{sub 17} alloy at 550 °C even in high purity argon atmosphere. The role of oxygen in the interaction process has also been discussed.

  15. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Leigh, K.M.

    1980-01-01

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

  16. 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.)

  17. 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)

  18. 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

  19. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

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

  20. 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

  1. Design and qualification of an on-line permeator for the recovery of tritium from lead-lithium eutectic breeding alloy

    International Nuclear Information System (INIS)

    Veredas, G.; Fradera, J.; Fernandez, I.; Batet, L.; Penalva, I.; Mesquida, L.; Abella, J.; Sempere, J.; Martinez, I.; Herrazti, B.; Sedano, L.

    2011-01-01

    The fast and efficient recovery of bred tritium represents a major milestone of tritium breeding technologies R and D and is key for the demonstration of fusion reactor fuel self-sufficiency. For lead-lithium eutectic, diverse technologies are currently being investigated and qualified. Permeator Against Vacuum (PAV) solution represents a firm candidate because: (i) runs as a single-step process for tritium on-line recovery, (ii) works passively allowing to be thermally governed, (iii) can be easily in-pipe integrated in Pb15.7Li loop systems and (iv) can be conceived with high compactness. An optimal design of a PAV requires a detailed hydraulic design optimization for established operational ranges. An optimal PAV design is proposed and qualified by numerical simulation.

  2. 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

  3. 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.

  4. 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

  5. The 1994 loss of coolant incident at Pickering NGS

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  6. Method of eliminating cruds in the primary coolants of reactors

    International Nuclear Information System (INIS)

    Tamura, Takaaki.

    1984-01-01

    Purpose: To eliminate cruds in the primary coolants by using rind of onions or peanuts. Method: Since cruds contained in the reactor primary coolants increase the radioactive exposure to reactor operators, they have been intended to remove by ion exchange resins. In this invention, rind of onions or peanuts are crushed into an adequate particle size and packed into an absorption column instead of ion exchange resins into which primary coolants are circulated. The powderous onions or peanuts rind contain glucoside such as cosmosiin and has an effect of cationic exchanger, they satisfactorily catch heavy metals such as Fe and Cu. They have an excellent filtering effect even under a high pH condition and are excellent in economical point of view. They can be decrease the volume of the absorption column, reduce their devolume after use through corrosion and easily subjected to waste procession through oxidizing combustion in liquid. (Nakamoto, H.)

  7. EDF PWRs primary coolant purification strategies

    International Nuclear Information System (INIS)

    Gressier, Frederic; Mascarenhas, Darren; Taunier, Stephane; Le-Calvar, Marc; Bretelle, Jean-Luc; Ranchoux, Gilles

    2012-09-01

    In order to achieve a good physico-chemical quality of the primary coolant fluid, the primary water is continuously treated by the Chemical and Volume Control System (CVCS). This system is composed of a treatment chain containing filters and ion-exchange resins. In the EDF design, an upstream filter is placed before the resin so as to prevent it from being saturated with insoluble particles. Then, the fluid passes through several resin beds (up to 3 depending on the configuration) and again through a downstream filter that prevents resin fines dissemination into the reactor coolant. Much work has been conducted in the last 5 years on the homogenisation of products and usage on French EDF NPP primary coolant treatment, while taking into account the compromise between source term reduction, liquid and solid waste, and buying and disposal costs. Two national markets have been created, and two operational documents for chemists on site have been published: a filtration guideline and an ion-exchange resin guideline. Both documents give general information about the products used, how are they characterized and selected for national market (technical requirements, standards and tests), how they should be used and what are the change-out criteria. They are also periodically updated based on feedback from sites. The positive impact on resin and filter lifetime (extension of some, limitation of others), homogenisation of products and usage will be presented. Moreover, EDF is constantly in the process of improving the current purification methods, as well as researching the use of existing and novel technologies. In this field, recent experiments on short loading of resin during reactor shutdown has been tested on site with success. In addition, work is done on silica free filters, filter consumption and filter chemical release. An overview of these optimization methods will be given. (authors)

  8. 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

  9. 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.)

  10. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. A open-quotes zero wasteclose quotes coolant management strategy

    International Nuclear Information System (INIS)

    Kennicott, M.A.

    1994-01-01

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

  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. 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

  1. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

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

  2. 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

  3. 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)

  4. Susceptibility of 2 1/4 Cr-1Mo steel to liquid metal induced embrittlement by lithium-lead solutions

    International Nuclear Information System (INIS)

    Eberhard, B.A.; Edwards, G.R.

    1984-08-01

    An investigation has been conducted on the liquid metal induced embrittlement susceptibility of 2 1/4Cr-1Mo steel exposed to lithium and 1a/o lead-lithium at temperatures between 190 0 C and 525 0 C. This research was part of an ongoing effort to evaluate the compatibility of liquid lithium solutions with potential fusion reactor containment materials. Of particular interest was the microstructure present in a weld heat-affected zone, a microstructure known to be highly susceptible to corrosive attack by liquid lead-lithium solutions. Embrittlement susceptibility was determined by conducting tension tests on 2 1/4Cr-1Mo steel exposed to an inert environment as well as to a lead-lithium liquid and observing the change in tensile behavior. The 2 1/4Cr-1Mo steel was also given a base plate heat treatment to observe its embrittlement susceptibility to 1a/o lead-lithium. The base plate microstructure was severely embrittled at temperatures less than 500 0 C. Tempering the base plate was effective in restoring adequate ductility to the steel

  5. 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

  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. 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.

  8. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  9. 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

  10. 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.)

  11. 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

  12. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

    Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

  13. 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.)

  14. 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)

  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. 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.)

  17. 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)

  18. 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

  19. Calculational advance in the modeling of fuel-coolant interactions

    International Nuclear Information System (INIS)

    Bohl, W.R.

    1982-01-01

    A new technique is applied to numerically simulate a fuel-coolant interaction. The technique is based on the ability to calculate separate space- and time-dependent velocities for each of the participating components. In the limiting case of a vapor explosion, this framework allows calculation of the pre-mixing phase of film boiling and interpenetration of the working fluid by hot liquid, which is required for extrapolating from experiments to a reactor hypothetical accident. Qualitative results are compared favorably to published experimental data where an iron-alumina mixture was poured into water. Differing results are predicted with LMFBR materials

  20. 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.)

  1. 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

  2. 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.)

  3. Core design studies on various forms of coolants and fuel materials. 2. Studies on liquid heavy metal and gas cooled cores, small cores and evaluation of 4-type cores

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Sakashita, Yoshiyuki; Naganuma, Masayuki; Takaki, Naoyuki; Mizuno, Tomoyasu; Ikegami, Tetsuo

    2001-01-01

    Alternative concepts to sodium cooled fast reactors, such as heavy metal liquid cooled reactors and gas cooled fast reactors were studied in Phase-1 of the feasibility studies, aiming at simplification of the system, high thermal efficiency and enhancing safety. Fuel and core specifications and nuclear characteristics were surveyed to meet the targets for commercialization of fast reactor cycle. Nuclear characteristics of small fast reactor cores were also surveyed from the perspective of the possibility of multi-purpose use and dispersed power stations. The key points of the design study for each concept in Phase-2 were summarized from the aspect of the screening of the candidates for FR commercialization. (author)

  4. 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

  5. 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.

  6. 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

  7. 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)

  8. 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)

  9. 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

  10. 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

  11. Experimental simulation of low rate primary coolant leaks. For the case of vessel head penetrations affected by through wall cracking

    International Nuclear Information System (INIS)

    You, D.; Feron, D.; Turluer, G.

    2002-01-01

    An experimental simulation of primary coolant leaks was carried out to determine how the composition of the leaking liquid would change. The experiment used the EVA experimental setup, specially designed for quantitatively investigating concentration phenomena driven by evaporation. The test showed that the final composition, obtained from a solution representative of the primary coolant at the beginning of the cycle, is highly concentrated and slightly acid. The experimental results are compared with those obtained using the MULTEQ software. (authors)

  12. 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.)

  13. Membrane technology for treating of waste nanofluids coolant: A review

    Science.gov (United States)

    Mohruni, Amrifan Saladin; Yuliwati, Erna; Sharif, Safian; Ismail, Ahmad Fauzi

    2017-09-01

    The treatment of cutting fluids wastes concerns a big number of industries, especially from the machining operations to foster environmental sustainability. Discharging cutting fluids, waste through separation technique could protect the environment and also human health in general. Several methods for the separation emulsified oils or oily wastewater have been proposed as three common methods, namely chemical, physicochemical and mechanical and membrane technology application. Membranes are used into separate and concentrate the pollutants in oily wastewater through its perm-selectivity. Meanwhile, the desire to compensate for the shortcomings of the cutting fluid media in a metal cutting operation led to introduce the using of nanofluids (NFs) in the minimum quantity lubricant (MQL) technique. NFs are prepared based on nanofluids technology by dispersing nanoparticles (NPs) in liquids. These fluids have potentially played to enhance the performance of traditional heat transfer fluids. Few researchers have studied investigation of the physical-chemical, thermo-physical and heat transfer characteristics of NFs for heat transfer applications. The use of minimum quantity lubrication (MQL) technique by NFs application is developed in many metal cutting operations. MQL did not only serve as a better alternative to flood cooling during machining operation and also increases better-finished surface, reduces impact loads on the environment and fosters environmental sustainability. Waste coolant filtration from cutting tools using membrane was treated by the pretreated process, coagulation technique and membrane filtration. Nanomaterials are also applied to modify the membrane structure and morphology. Polyvinylidene fluoride (PVDF) is the better choice in coolant wastewater treatment due to its hydrophobicity. Using of polyamide nanofiltration membranes BM-20D and UF-PS-100-100, 000, it resulted in the increase of permeability of waste coolant filtration. Titanium dioxide

  14. Numerical simulations on a high-temperature particle moving in coolant

    International Nuclear Information System (INIS)

    Li Xiaoyan; Shang Zhi; Xu Jijun

    2006-01-01

    This study considers the coupling effect between film boiling heat transfer and evaporation drag around a hot-particle in cold liquid. Taking momentum and energy equations of the vapor film into account, a transient single particle model under FCI conditions has been established. The numerical simulations on a high-temperature particle moving in coolant have been performed using Gear algorithm. Adaptive dynamic boundary method is adopted during simulating to matching the dynamic boundary that is caused by vapor film changing. Based on the method presented above, the transient process of high-temperature particles moving in coolant can be simulated. The experimental results prove the validity of the HPMC model. (authors)

  15. 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.)

  16. 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.

  17. 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

  18. 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.)

  19. Dynamic response of INTOR/NET blankets after coolant tube rupture

    International Nuclear Information System (INIS)

    Klippel, H.T.

    1985-01-01

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

  20. Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead-lithium HCLL breeding blanket channels: Theory and code development

    Energy Technology Data Exchange (ETDEWEB)

    Batet, L., E-mail: lluis.batet@upc.edu [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Valls, E. Mas de les [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Heat Engines (DMMT), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, Fusion Technology Division, Av. Complutense 22, 28040 Madrid (Spain)

    2011-06-15

    Large helium (He) production rates in liquid metal breeding blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium-lead breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL breeding unit conditions.

  1. 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.

  2. 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.)

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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.)

  9. 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

  10. 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)

  11. 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.

  12. Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mirghaffari, Reza; Jahanfarnia, Gholamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-12-15

    Nanofluids have shown to be promising as an alternative for a PWR reactor coolant or as a safety system coolant to cover the core in the event of a loss of coolant accident. The nanoparticles distribution and neutronic parameters are intensively affected by the local boiling of nanofluid coolant. The main goal of this study was the physical-mathematical modeling of the nanoparticles distribution in the nucleate boiling of nanofluids within the viscous sublayer. Nanoparticles concentration, especially near the heat transfer surfaces, plays a significant role in the enhancement of thermal conductivity of nanofluids and prediction of CHF, Hide Out and Return phenomena. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux on the distribution of nanoparticles was studied. The steady state mass conservation equations for liquids, vapors and nanoparticles were written for the flow boiling within the viscous sublayer adjacent the fuel cladding surface. The derived differential equations were discretized by the finite difference method and were solved numerically. It was found out that by increasing the surface heat flux, the concentration of nanoparticles increased.

  13. 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.)

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. Reactor coolant pump seal response to loss of cooling

    International Nuclear Information System (INIS)

    Graham, T.; Metcalfe, R.; Burchett, P.

    2000-01-01

    This paper describes the results of a test done to determine the performance of a reactor coolant pump seal for a water cooled nuclear reactor under loss of all cooling conditions. Under these conditions, seal faces can lose their liquid lubricating film and elastomers can rapidly degrade. Temperatures in the seal-cartridge tester reached 230 o C in three hours, at which time the tester was stopped and the temperature increased to 265 o C for a further five hours before cooling was restored. Seal leakage was 'normal' throughout the test. Parts sustained minor damage with no effect on seal integrity. Plant operators were shown to have ample margin beyond their 15 minute allowable reaction time. (author)

  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. Progress on the development of H-concentration probes in eutectic lead-lithium: Synthesis and characterization of electrochemical sensor materials

    Energy Technology Data Exchange (ETDEWEB)

    Llivina, L.; Colominas, S. [Universitat Ramon Llull, ETS Institut Quimic de Sarria, Electrochemical Methods Laboratory - Analytical Chemistry Department Via Augusta, 390, 08017 Barcelona (Spain); Reyes, G. [Universitat Ramon Llull, ETS Institut Quimic de Sarria, Industrial Engineering Department, Via Augusta, 390, 08017 Barcelona (Spain); Abella, J., E-mail: jordi.abella@iqs.es [Universitat Ramon Llull, ETS Institut Quimic de Sarria, Electrochemical Methods Laboratory - Analytical Chemistry Department Via Augusta, 390, 08017 Barcelona (Spain)

    2012-08-15

    Dynamic tritium concentration measurement in lithium-lead eutectic (17% Li-83% Pb) is of major interest for a reliable tritium testing program in ITER TBM and for an experimental proof of tritium self-sufficiency in liquid metal breeding systems. Potentiometric hydrogen sensors for molten lithium-lead eutectic have been designed at the Electrochemical Methods Lab at Institut Quimic de Sarria (IQS) at Barcelona and are under development and qualification. The probes are based on the use of solid state electrolytes and works as Proton Exchange Membranes (PEM). In this work, the following compounds have been synthesized in order to be tested as PEM H-probes: BaCeO{sub 3}, BaCe{sub 0.9}Y{sub 0.1}O{sub 3-{delta}}, SrCe{sub 0.9}Y{sub 0.1}O{sub 3-{delta}} and Sr(Ce{sub 0.9}-Zr{sub 0.1}){sub 0.95}Yb{sub 0.05}O{sub 3-{delta}}. Potentiometric measurements of the synthesized ceramic elements have been performed at different hydrogen concentrations at 500 Degree-Sign C. In this campaign, a fixed and known hydrogen pressure has been used in the reference electrode. The sensors constructed using the proton conductor elements BaCeO{sub 3}, SrCe{sub 0.9}Y{sub 0.1}O{sub 3-{delta}} and Sr(Ce{sub 0.9}-Zr{sub 0.1}){sub 0.95}Yb{sub 0.05}O{sub 3-{delta}} exhibited quite stable output potential and its value was quite close to the theoretical value calculated with the Nernst equation (deviation less than 100 mV). Unstable measurement was obtained using BaCe{sub 0.9}Y{sub 0.1}O{sub 3-{delta}} as a solid state electrolyte in the sensor.

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

    International Nuclear Information System (INIS)

    Sharon, A.; Bankoff, S.G.

    1979-01-01

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

  2. Liquid metal level measurement

    International Nuclear Information System (INIS)

    Hale, J.C.; Leyland, K.S.

    1982-01-01

    A liquid metal level indicator is described which can be used to measure, in a stainless steel tank, the level of a nuclear reactor coolant such as sodium. The instrument, which is based on the eddy current induction effect, gives readings over substantially the full depth of the tank and indicates the sense of change of level. (U.K.)

  3. Nuclear data needs for subcritical reactors with heavy-metal coolant

    International Nuclear Information System (INIS)

    Ignatyuk, A.V.

    2001-01-01

    Requests on improvement of evaluated data files for minor actinides (MA) are briefly reviewed. New evaluations of neutron cross sections for Np-237, Am-241 and Am-243 after the corresponding tests and verifications should satisfy the required accuracies of data for developing MA-burners. More difficult problems arise for curium isotopes, evaluated data of which are strongly divergent. International expertise of available evaluations could be very desirable. Needs in data improvements for perspective heavy-metal liquid coolants are outlined. (author)

  4. Consequences in a long time of the forced loss of coolant in a pool type reactor

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1986-01-01

    The fuel and pool water temperatures are calculated as a function of time using unidimensional models of heat conduction and momentum conservation, to simulate the natural convection flow of the coolant. The reactor building pressure due to the pool water evaporation is calculated using a homogeneous model with thermal equilibrium. The heat loss from the three main components of the building volume (liquid water, air, and steam) to solid surfaces such as the building walls are taking into account. (Author) [pt

  5. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors

    International Nuclear Information System (INIS)

    Martin Munoz, M.; Alvarez Gonzalez, F.

    1969-01-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs

  6. Revised Mark 22 coolant temperature coefficients

    International Nuclear Information System (INIS)

    Graves, W.E.

    1987-01-01

    Coolant temperature coefficients for the Mark 22 charge published previously are non-conservative because of the neglect of a significant mechanism which has a positive contribution to reactivity. Even after correcting for this effect, dynamic tests made on a Mark VIB charge in the early 60's suggest the results are still non-conservative. This memorandum takes both of these sources of information into account in making a best estimate of the prompt (coolant plus metal) temperature coefficient. Although no safety issues arise from this work (the overall temperature coefficient still strongly contributes to reactor stability), it is obviously desirable to use best estimates for prompt coefficients in limits and other calculations

  7. Freeform Deposition Method for Coolant Channel Closeout

    Science.gov (United States)

    Gradl, Paul R. (Inventor); Reynolds, David Christopher (Inventor); Walker, Bryant H. (Inventor)

    2017-01-01

    A method is provided for fabricating a coolant channel closeout jacket on a structure having coolant channels formed in an outer surface thereof. A line of tangency relative to the outer surface is defined for each point on the outer surface. Linear rows of a metal feedstock are directed towards and deposited on the outer surface of the structure as a beam of weld energy is directed to the metal feedstock so-deposited. A first angle between the metal feedstock so-directed and the line of tangency is maintained in a range of 20-90.degree.. The beam is directed towards a portion of the linear rows such that less than 30% of the cross-sectional area of the beam impinges on a currently-deposited one of the linear rows. A second angle between the beam and the line of tangency is maintained in a range of 5-65 degrees.

  8. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  9. Recovery studies for plutonium machining oil coolant

    International Nuclear Information System (INIS)

    Navratil, J.D.; Baldwin, C.E.

    1977-01-01

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

  10. 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

  11. 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

  12. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kennicott, M.A.

    1994-04-01

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

  14. 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

  15. 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.

  16. 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

  17. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Harand, E.; Richter, G.; Tschoepel, G.

    1975-01-01

    A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

  18. Design of automotive engine coolant hoses

    Directory of Open Access Journals (Sweden)

    Hrishikesh D BACHCHHAV

    2018-03-01

    Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

  19. 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)

  20. Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Konovalenko, Alexander, E-mail: kono@kth.se; Karbojian, Aram, E-mail: karbojan@kth.se

    2017-04-01

    Highlights: • Steam explosion in stratified melt-coolant configuration is studied experimentally. • Different binary oxidic melt simulant materials were used. • Five spontaneous steam explosions were observed. • Instability of melt-coolant interface and formation of premixing layer was observed. • Explosion strength is influenced by melt superheat and water subcooling. - Abstract: Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.

  1. Safety and environmental impact of the dual coolant blanket concept. SEAL subtask 6.2, final report

    International Nuclear Information System (INIS)

    Kleefeldt, K.; Dammel, F.; Gabel, K.; Jordan, T.; Schmuck, I.

    1996-03-01

    The European Union has been engaged since 1989 in a programme to develop tritium breeding blankets for application in a fusion power reactor. There are four concepts under development, namely two of the solid breeder type and two of the liquid breeder type. At the Forschungszentrum Karlsruhe one blanket concept of each line has been pursued so far with the so-called dual coolant type representing the liquid breeder line. In the dual coolant concept the breeder material (Pb-17Li) is circulated to external heat exchangers to carry away the bulk of the generated heat and to extract the tritium. Additionally, the heavily loaded first wall is cooled by high pressure helium gas. The safety and environmental impact of the dual coolant blanket concept has been assessed as part of the blanket concept selection excercise, a European concerted action, aiming at selecting the two most promising concepts for futher development. The topics investigated are: (a) Blanket materials and toxic materials inventory, (b) energy sources for mobilisation, (c) fault tolerance, (d) tritium and activation products release, and (e) waste generation and management. No insurmountable safety problems have been identified for the dual coolant blanket. The results of the assessment are described in this report. The information collected is also intended to serve as input to the EU 'Safety and Environmental Assessment of Fusion longterm Programme' (SEAL). The unresolved issues pertaining to the dual coolant blanket which would need further investigations in future programmes are outlined herein. (orig.) [de

  2. The premixing and propagation phases of fuel-coolant interactions: a review of recent experimental studies and code developments

    International Nuclear Information System (INIS)

    Antariksawan, A.R.; Moriyama, Kiyofumi; Park, Hyun-sun; Maruyama, Yu; Yang, Yanhua; Sugimoto, Jun

    1998-09-01

    A vapor explosion (or an energetic fuel-coolant interactions, FCIs) is a process in which hot liquid (fuel) transfers its internal energy to colder, more volatile liquid (coolant); thus the coolant vaporizes at high pressure and expands and does works on its surroundings. Traditionally, the energetic fuel-coolant interactions could be distinguished in subsequent stages: premixing (or coarse mixing), triggering, propagation and expansion. Realizing that better and realistic prediction of fuel-coolant interaction consequences will be available understanding the phenomenology in the premixing and propagation stages, many experimental and analytical studies have been performed during more than two decades. A lot of important achievements are obtained during the time. However, some fundamental aspects are still not clear enough; thus the works are directed to that direction. In conjunction, the model/code development is pursuit. This is aimed to provide a scaling tool to bridge the experimental results to the real geometries, e.g. reactor pressure vessel, reactor containment. The present review intends to collect the available information on the recent works performed to study the premixing and propagation phases. (author). 97 refs

  3. The premixing and propagation phases of fuel-coolant interactions: a review of recent experimental studies and code developments

    Energy Technology Data Exchange (ETDEWEB)

    Antariksawan, A.R. [Reactor Safety Technology Research Center of BATAN (Indonesia); Moriyama, Kiyofumi; Park, Hyun-sun; Maruyama, Yu; Yang, Yanhua; Sugimoto, Jun

    1998-09-01

    A vapor explosion (or an energetic fuel-coolant interactions, FCIs) is a process in which hot liquid (fuel) transfers its internal energy to colder, more volatile liquid (coolant); thus the coolant vaporizes at high pressure and expands and does works on its surroundings. Traditionally, the energetic fuel-coolant interactions could be distinguished in subsequent stages: premixing (or coarse mixing), triggering, propagation and expansion. Realizing that better and realistic prediction of fuel-coolant interaction consequences will be available understanding the phenomenology in the premixing and propagation stages, many experimental and analytical studies have been performed during more than two decades. A lot of important achievements are obtained during the time. However, some fundamental aspects are still not clear enough; thus the works are directed to that direction. In conjunction, the model/code development is pursuit. This is aimed to provide a scaling tool to bridge the experimental results to the real geometries, e.g. reactor pressure vessel, reactor containment. The present review intends to collect the available information on the recent works performed to study the premixing and propagation phases. (author). 97 refs.

  4. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  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. Characterization of primary coolant purification system samples for assay of spent ion exchanger radionuclide inventor

    International Nuclear Information System (INIS)

    Sajin Prasad, S.; Pant, Amar; Sharma, Ranjit; Pal, Sanjit

    2018-01-01

    The primary coolant system water of a research reactor contains various fission and activation products and the water is circulated continuously through ion exchange resin cartridges, to reduce the radioactive ionic impurity present in it. The coolant purification system comprises of an ion exchange cooler, two micro filters, and a battery of six ion exchanger beds, associated valves, piping and instrumentation (Heavy water System Operating manual, 2014). The spent cartridge is finally disposed off as active solid waste which contains predominantly long lived fission and activation products. The heavy water coolant is also used to cool the structural assemblies after passing through primary heat exchanger and a metallic strainer, which accumulates the fission and activation products. When there is a reduction of coolant flow through these strainers, they are removed for cleaning and decontamination. This paper describes the characterization of ion exchange resin samples and liquid effluent generated during ultra sonic decontamination of strainer. The results obtained can be used as a methodology for the assay of the spent ion exchanger cartridges radionuclide inventory, during its disposal

  7. The effect of constraint on fuel-coolant interactions in a confined geometry

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.; Corradini, M.L. [Univ. of Wisconsin, Madison, WI (United States)

    1995-09-01

    A Fuel-Coolant Interaction (FCI or vapor explosion) is the phenomena in which a hot liquid rapidly transfers its internal energy into a surrounding colder and more volatile liquid. The energetics of such a complex multi-phase and multi-component phenomenon is partially determined by the surrounding boundary conditions. As one of the boundary conditions, we studied the effect of constraint on FCIs. The WFCI-D series of experiments were performed specifically to observe this effect. The results from these and our previous WFCI tests as well as those of other investigators are compared.

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

    International Nuclear Information System (INIS)

    Soli T. Khericha

    2006-01-01

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

  9. Interaction of hydrogen with Pb83Li17 eutectic alloy

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  10. Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

    International Nuclear Information System (INIS)

    Asher, R.C.; Davies, D.; Jones, P.G.

    1976-12-01

    A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

  11. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

  12. Microstructural characterization of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.

    1986-01-01

    Atom probe field-ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF 8 primary coolant pipe stainless steel after long term thermal aging. The cast duplex microstructure consisted of austenite with 15% delta-ferrite. Investigation of the aged material revealed that the ferrite spinodally decomposed into a fine scaled network of α and α'. A fine G-phase precipitate was also observed in the ferrite. The observed degradation in mechanical properties is probably a consequence of the spinodal decomposition in the ferrite

  13. 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

  14. Bulk coolant cavitation in LMFBR containment loading following a whole-core explosion

    International Nuclear Information System (INIS)

    Jones, A.V.

    1977-01-01

    An LMFBR core undergoing an explosion transmits energy to the containment in a series of pressure waves and the containment loading is determined by their cumulative effect. These pressure waves are modified by their interaction with the coolant through which they propagate. It is necessary to model both the induction of bulk cavitation by tension waves and the interaction of pressure waves with cavitated liquid in realistic containment loading calculations. This paper sets out the progress which has been achieved in such modelling and first indications for the effect of bulk coolant cavitation in LMFBR containment loading. Conclusions may be briefly summarised: 1) Bulk cavitation must be included in realistic containment loading calculations. 2) Phenomenological models of cavitated liquid without memory are inappropriate. The best approach is to model bubble dynamics directly, including at least momentum conservation and surface tension. 3) The containment loading resulting from a given explosion is sensitive to the state of preparation of the coolant. The number density of nucleation sites should therfore accompany the results of model tests. (Auth.)

  15. Design of the coolant system for the Large Coil Test Facility pulse coils

    International Nuclear Information System (INIS)

    Bridgman, C.; Ryan, T.L.

    1983-01-01

    The pulse coils will be a part of the Large Coil Test Facility in Oak Ridge, Tennessee, which is designed to test six large tokamak-type superconducting coils. The pulse coil set consists of two resistive coaxial solenoid coils, mounted so that their magnetic axis is perpendicular to the toroidal field lines of the test coil. The pulse coils provide transient vertical fields at test coil locations to simulate the pulsed vertical fields present in tokamak devices. The pulse coils are designed to be pulsed for 30 s every 150 s, which results in a Joule heating of 116 kW per coil. In order to provide this capability, the pulse coil coolant system is required to deliver 6.3 L/s (100 gpm) of subcooled liquid nitrogen at 10-atm absolute pressure. The coolant system can also cool down each pulse coil from room temperature to liquid nitrogen temperature. This paper provides details of the pumping and heat exchange equipment designed for the coolant system and of the associated instrumentation and controls

  16. 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)

  17. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

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

  18. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

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

  20. 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

  1. 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.

  2. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  3. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  4. Proposed model for fuel-coolant mixing during a core-melt accident

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1983-01-01

    If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  9. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  10. Filtering device for primary coolant circuits in BWR type reactors

    International Nuclear Information System (INIS)

    Tajima, Fumio; Yamamoto, Tetsuo.

    1985-01-01

    Purpose: To obtain a filtering device with a large filtering area and requiring less space. Constitution: A condensate inlet for introducing condensates to be filtered of primary coolant circuits, a filtrate exit, a backwash water exit and a bent tube are disposed to a container, and a plurality of hollow thread membrane modules are suspended in the container. The condensates are caused to flow through the condensate inlet, filtered through the hollow thread membrane and then discharged from the filtrate exit. When the filtering treatment is proceeded to some extent, since solid contents captured in the hollow thread membranes are accumulated, a differential pressure is produced between the condensate inlet and the filtrate exit. When the differential pressure reaches a predetermined value, the backwash is conducted to discharge the liquid cleaning wastes through the backwash exit. The bent tube disposed to the container body is used for water and air draining. The hollow thread membranes are formed with porous resin such as of polyethylene. (Kawakami, Y.)

  11. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  12. Slow coolant phaseout could worsen warming

    Science.gov (United States)

    Reese, April

    2018-03-01

    In the summer of 2016, temperatures in Phalodi, an old caravan town on a dry plain in northwestern India, reached a blistering 51°C—a record high during a heat wave that claimed more than 1600 lives across the country. Wider access to air conditioning (AC) could have prevented many deaths—but only 8% of India's 249 million households have AC. As the nation's economy booms, that figure could rise to 50% by 2050. And that presents a dilemma: As India expands access to a life-saving technology, it must comply with international mandates—the most recent imposed just last fall—to eliminate coolants that harm stratospheric ozone or warm the atmosphere.

  13. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-01-01

    An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

  14. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  15. 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)

  16. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  17. Coolant rate distribution in horizontal steam generator under natural circulation

    International Nuclear Information System (INIS)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A.

    1997-01-01

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered

  18. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  19. 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.

  20. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  1. Fuel coolant interaction experiment by direct electrical heating method

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Hirano, Kenmei

    1979-01-01

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

  2. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A; Leontieva, V; Mitrioukhin, A [St. Petersburg State Technical Univ. (Russian Federation)

    1998-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

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

    Science.gov (United States)

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

    2017-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  6. 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

  7. 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

  8. Device for preventing coolant in a reactor from being lost

    International Nuclear Information System (INIS)

    Maruyama, Hiromi; Matsumoto, Tomoyuki.

    1975-01-01

    Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

  9. 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

  10. Vessel coolant mass depletion during a 5% SBLOCA in the Semiscale Mod-2C facility

    International Nuclear Information System (INIS)

    Shaw, R.A.; Loomis, G.G.

    1985-01-01

    Experimental results are presented from two 5% small-break loss-of-coolant accident (SBLOCA) simulations in the Semiscale Mod-2C facility. In performing the simulated 5% SBLOCAs, boundary conditions scaled from a pressurized water reactor (PWR) were used. The experiment was run with initial conditions typical of a PWR (15.6 MPa pressure and 35 K core differential temperature). The Mod-2C facility represents the state-of-the-art in small facilities scaled from PWRs. Phenomena which occurred during the transient included: primary fluid saturation (change from subcooled to saturated blowdown), break uncovery (a centerline break was simulated), condensation-induced liquid hold-up in the steam generator primary tubes, pump suction liquid seal formation and core level depression with resulting core rod temperature excursion, pump suction liquid seal clearance, loop fluid mass redistribution, and gradual core rewet. The influence of core bypass flow is also discussed. 11 refs., 13 figs

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. 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.

  13. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Scott, D.

    1981-01-01

    An improved method of constructing the diagrid used to support fuel assemblies of liquid metal fast breeder reactors, is described. The functions of fuel assembly support and coolant plenum are performed by discrete components of the diagrid each of which can serve the function of the other in the event of failure of one of the components. (U.K.)

  14. 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

  15. MHD considerations for poloidal-toroidal coolant ducts of self-cooled blankets

    International Nuclear Information System (INIS)

    Hua, T.Q.; Walker, J.S.

    1990-01-01

    Magnetohydrodynamic flows of liquid metals through sharp elbow ducts with rectangular cross sections and with thin conducting walls in the presence of strong uniform magnetic fields are examined. The geometries simulate the poloidaltoroidal coolant channels in fusion tokamak blankets. Analysis for obtaining the three-dimensional numerical solutions are described. Results for pressure drop, velocity profiles and flow distribution are predicted for the upcoming joint ANL/KfK sharp elbow experiment. Results from a parametric study using fusion relevant parameters to investigate the three-dimensional pressure drop are presented for possible applications to blanket designs. 10 refs., 9 refs

  16. Analysis of an ultrasonic level device for in-core Pressurized Water Reactor coolant detection

    International Nuclear Information System (INIS)

    Johnson, K.R.

    1981-01-01

    A rigorous semi-empirical approach was undertaken to model the response of an ultrasonic level device (ULD) for application to in-core coolant detection in Pressurized Water Reactors (PWRs). An equation is derived for the torsional wave velocity v/sub t phi/ in the ULD. Existing data reduction techniques were analyzed and compared to results from use of the derived equation. Both methods yield liquid level measurements with errors of approx. 5%. A sensitivity study on probe performance at reactor conditions predicts reduced level responsivity from data at lower temperatures

  17. Breakup of jet and drops during premixing phase of fuel coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Haraldsson, Haraldur Oskar

    2000-05-01

    During the course of a hypothetical severe accident in a light water reactor, molten liquid may be introduced into a volatile coolant, which, under certain conditions, results in explosive interactions. Such fuel-coolant interactions (FCI) are characterised by an initial pre-mixing phase during which the molten liquid, metallic or oxidic in nature, undergoes a breakup (fragmentation) process which significantly increase the area available for melt-coolant contact, and thus energy transfer. Although substantial progress in the understanding of phenomenology of the FCI events has been achieved in recent years, there remain uncertainties in describing the primary and secondary breakup processes. The focus of this work is on the melt jet and drop breakup during the premixing phase of FCI. The objectives are to gain insight into the premixing phase of the FCI phenomena, to determine what fraction of the melt fragments and determine the size distribution. The approach is to perform experiments with various simulant materials, at different scales, different conditions and with variation of controlling parameters affecting jet and drop breakup processes. The analysis approach is to investigate processes at different level of detail and complexity to understand the physics, to rationalise experimental results and to develop and validate models. In the first chapter a brief introduction and review of the status of the FCI phenomena is performed. A review of previous and current experimental projects is performed. The status of the experimental projects and major findings are outlined. The first part of the second chapter deals with experimental investigation of jet breakup. Two series of experiments were performed with low and high temperature jets. The low temperature experiments employed cerrobend-70 as jet liquid. A systematic investigation of thermal hydraulic conditions and melt physical properties on the jet fragmentation and particle debris characteristics was

  18. Breakup of jet and drops during premixing phase of fuel coolant interactions

    International Nuclear Information System (INIS)

    Haraldsson, Haraldur Oskar

    2000-05-01

    During the course of a hypothetical severe accident in a light water reactor, molten liquid may be introduced into a volatile coolant, which, under certain conditions, results in explosive interactions. Such fuel-coolant interactions (FCI) are characterised by an initial pre-mixing phase during which the molten liquid, metallic or oxidic in nature, undergoes a breakup (fragmentation) process which significantly increase the area available for melt-coolant contact, and thus energy transfer. Although substantial progress in the understanding of phenomenology of the FCI events has been achieved in recent years, there remain uncertainties in describing the primary and secondary breakup processes. The focus of this work is on the melt jet and drop breakup during the premixing phase of FCI. The objectives are to gain insight into the premixing phase of the FCI phenomena, to determine what fraction of the melt fragments and determine the size distribution. The approach is to perform experiments with various simulant materials, at different scales, different conditions and with variation of controlling parameters affecting jet and drop breakup processes. The analysis approach is to investigate processes at different level of detail and complexity to understand the physics, to rationalise experimental results and to develop and validate models. In the first chapter a brief introduction and review of the status of the FCI phenomena is performed. A review of previous and current experimental projects is performed. The status of the experimental projects and major findings are outlined. The first part of the second chapter deals with experimental investigation of jet breakup. Two series of experiments were performed with low and high temperature jets. The low temperature experiments employed cerrobend-70 as jet liquid. A systematic investigation of thermal hydraulic conditions and melt physical properties on the jet fragmentation and particle debris characteristics was

  19. 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

  20. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  1. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  2. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  3. Characterization of reactor coolant by XRF

    Energy Technology Data Exchange (ETDEWEB)

    Legreid, G.; Beverskog, B. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  4. Characterization of reactor coolant by XRF

    International Nuclear Information System (INIS)

    Legreid, G.; Beverskog, B.

    2002-01-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  5. 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.)

  6. Reactor Coolant Temperature Measurement using Ultrasonic Technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, JaeCheon [KEPCO International Nuclear graduate School, Ulsan (Korea, Republic of); Seo, YongSun; Bechue, Nicholas [Krohne Messtechnik GmbH, Duisburg (Germany)

    2016-10-15

    In NPP, the primary piping temperature is detected by four redundant RTDs (Resistance Temperature Detectors) installed 90 degrees apart on the RCS (Reactor Coolant System) piping circumferentially. Such outputs however, if applied to I and C systems would not give balanced results. The discrepancy can be explained by either thermal stratification or improper arrangement of thermo-wells and RTDs. This phenomenon has become more pronounced in the hot-leg piping than in the cold-leg. Normally, the temperature difference among channels is in the range of 1°F in Korean nuclear power Plants. Consequently, a more accurate pipe average temperate measurement technique is required. Ultrasonic methods can be used to measure average temperatures with relatively higher accuracy than RTDs because the sound wave propagation in the RCS pipe is proportional to the average temperature around pipe area. The inaccuracy of RCS temperature measurement worsens the safety margin for both DNBR and LPD. The possibility of this discrepancy has been reported with thermal stratification effect. Proposed RCS temperature measurement system based on ultrasonic technology offers a countermeasure to cope with thermal stratification effect on hot-leg piping that has been an unresolved issue in NPPs. By introducing ultrasonic technology, the average internal piping temperature can be measured with high accuracy. The inaccuracy can be decreased less than ±1℉ by this method.

  7. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  8. 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.)

  9. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  10. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1984-11-01

    A review of French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all occurred leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by the compliance with the criteria defined in the operating technical specifications

  11. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1986-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

  12. Evaluation of filtration and distillation methods for recycling automotive coolant

    International Nuclear Information System (INIS)

    Randall, P.M.; Gavaskar, A.R.

    1992-01-01

    Government regulations and high waste disposal cost of spent automotive coolant have driven the vehicle maintenance industry to explore on-site recycling. The USEPA in cooperation with the New Jersey Department of Environmental Protection (NJDEP) and the New Jersey Department of Transportation (NJDOT) evaluated two commercially available technologies that have potential for reducing the volume of spent automotive coolant. The objective of this study was to evaluate the quality of the recycled coolant, the pollution prevention potential, and the economic feasibility of the technologies

  13. Liquid Wall Chambers

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W R

    2011-02-24

    The key feature of liquid wall chambers is the use of a renewable liquid layer to protect chamber structures from target emissions. Two primary options have been proposed and studied: wetted wall chambers and thick liquid wall (TLW) chambers. With wetted wall designs, a thin layer of liquid shields the structural first wall from short ranged target emissions (x-rays, ions and debris) but not neutrons. Various schemes have been proposed to establish and renew the liquid layer between shots including flow-guiding porous fabrics (e.g., Osiris, HIBALL), porous rigid structures (Prometheus) and thin film flows (KOYO). The thin liquid layer can be the tritium breeding material (e.g., flibe, PbLi, or Li) or another liquid metal such as Pb. TLWs use liquid jets injected by stationary or oscillating nozzles to form a neutronically thick layer (typically with an effective thickness of {approx}50 cm) of liquid between the target and first structural wall. In addition to absorbing short ranged emissions, the thick liquid layer degrades the neutron flux and energy reaching the first wall, typically by {approx}10 x x, so that steel walls can survive for the life of the plant ({approx}30-60 yrs). The thick liquid serves as the primary coolant and tritium breeding material (most recent designs use flibe, but the earliest concepts used Li). In essence, the TLW places the fusion blanket inside the first wall instead of behind the first wall.

  14. The second advanced lead lithium blanket concept using ODS steel as structural material and SiCf/SiC flow channel inserts as electrical and thermal insulators (Task PPA 2.5). Final report

    International Nuclear Information System (INIS)

    Norajitra, P.; Buehler, L.; Fischer, U.

    1999-12-01

    Preparatory work on the advanced dual coolant (A-DCL) blanket concept using SiC f /SiC flow channel inserts as electrical and thermal insulators has been carried out at the Forschungszentrum Karlsruhe in co-operation with CEA as a conceptual design proposal to the EU fusion power plant study planned to be launched in 2000 within the framework of the EU fusion programme with the main objective of specifying the characteristics of an attractive and viable commercial D-T fusion power plant. The basic principles and design characteristics of this A-DCL blanket concept are presented and its potential with regard to performance (neutron wall load, lifetime, availability) is discussed in this report. The results of this study show that the A-DCL blanket concept has a high potential for further development due to its high thermal efficiency and its simple concept solution. (orig.) [de

  15. Improving Coolant Effectiveness through Drill Design Optimization in Gundrilling

    Science.gov (United States)

    Woon, K. S.; Tnay, G. L.; Rahman, M.

    2018-05-01

    Effective coolant application is essential to prevent thermo-mechanical failures of gun drills. This paper presents a novel study that enhances coolant effectiveness in evacuating chips from the cutting zone using a computational fluid dynamic (CFD) method. Drag coefficients and transport behaviour over a wide range of Reynold numbers were first established through a series of vertical drop tests. With these, a CFD model was then developed and calibrated with a set of horizontal drilling tests. Using this CFD model, critical drill geometries that lead to poor chip evacuation including the nose grind contour, coolant hole configuration and shoulder dub-off angle in commercial gun drills are identified. From this study, a new design that consists a 20° inner edge, 15° outer edge, 0° shoulder dub-off and kidney-shaped coolant channel is proposed and experimentally proven to be more superior than all other commercial designs.

  16. 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

  17. 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.)

  18. Simulation of a loss of coolant accident

    International Nuclear Information System (INIS)

    1987-06-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

  19. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  20. Review on research of small break loss of coolant accident

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

    The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

  1. The purification of organic reactor coolants; La purification des refrigerants organiques

    Energy Technology Data Exchange (ETDEWEB)

    Hannaert, H; Lopes Cardozo, R [CCR EURATOM, Ispra, Varese (Italy)

    1967-01-01

    Among the main impurities present in irradiated and virgin coolants we have been particularly interested in chlorine, water, iron, oxygen and heavy elements. Our studies have been directed along two basic lines, namely: (1) the elimination of inorganic impurities, and (2) the elimination of organic impurities. The purpose of the studies on the elimination of inorganic impurities is to obtain a 'clean' coolant (virgin or used), this cleanness being marked by a low tendency to form deposits on the fuel element cladding (fouling corrosion). Careful attention has been paid to the problems of chlorine, water, iron and coated particles. Particular interest has been attached to research on deoxygenation of the organic liquid by catalytic hydrocracking, the oxygen (which comes from the dissociation of the water and the gases contained in the liquid) favouring polymerization and the formation of particles which are likely to be deposited on the hot walls. As regards the elimination of degradation products, many studies have been carried out with a view to permitting the maximum recycling of decomposed hydrocarbides and thus reducing the cost of make-up, the recycling rate being a function of the mean molecular weight and the viscosity of the recycled fractions. The technique based on solvent extraction has led to very satisfactory results. An extension of this technique, using counter-flow extraction, appears to be even more promising. (author)

  2. Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Callow, R.A.

    1989-01-01

    The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

  3. Simulation of steam explosion in stratified melt-coolant configuration

    International Nuclear Information System (INIS)

    Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

    2016-01-01

    Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

  4. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  5. Some experimental justifications of constructions of nuclear reactors with the use of solid coolant

    International Nuclear Information System (INIS)

    Deniskin, V.; Nalivaev, V.; Fedik, I.; Vishnevski, U.; Dmitriev, A.

    2003-01-01

    Full text: The work that has been conducted so far justifies a possibility of constructing a reactor with a non-traditional coolant to develop radically new reactors and their cycles with perfect architecture. A solid coolant, for example, the carbon-based one, allows to design the primary circuit of nuclear reactor without excess pressure. Such coolant withstands temperatures up to ∼4000 deg. K without a collapse. The analysis of theory and experiments produced requirements to be met by a solid coolant used in the primary circuit of nuclear reactor. One of the most important requirements is the arrangements for a continuous and homogeneous gravity flow of the coolant through all core sections taking into account the dust caused by wear and some amount of fractured particles. Therefore, the idea is that the mass of particles should resemble a liquid to a certain extend. The particles should be sphere like with average diameter from 0.5 to 2.0 mm and nonsphericity rate not more than 10%. 'Angle of repose' of particles to the horizon can be utilised as a validity criterion of particles which should not exceed 25 deg. The heat transfer coefficient should be increased up to the practical maximum value. In 1996 - 1997 the system of experimental facilities were built in the Scientific and Research Institute 'Luch' to prove the possibility to reliably cool a nuclear reactor with a flow of solid particles and to obtain a minimum set of data for the conceptual design of such reactor with solid coolant. The facility allows the research of the flow stability, heat mass transfer in the core, lifetime wearing of particles of the solid coolant. In 1994-1999 5 batches of particles of different size were fabricated in accordance to different technologies. Four batches were graphite-based and one was aluminium oxide-based (Al 2 O 3 ). The purpose was to verify how the heat transfer coefficient was changing as the particle size varied. The average diameter of graphite particles

  6. Corrosion particles in the primary coolant of VVER-440 reactors

    International Nuclear Information System (INIS)

    Vajda, N.; Molnar, Z.; Macsik, Z.; Szeles, E.; Hargittai, P.; Csordas, A.; Pinter, T.; Pinter, T.

    2010-01-01

    Corrosion and activity build-up processes are of major concern in ageing and life-extension of nuclear power reactors. Researches to study the migration of radioactive corrosion particles have been initiated at Paks Nuclear Power Plant (NPP), Hungary in order to better understand the corrosion of the primary circuit surfaces, the transport and activation of the particles of corrosion origin and their deposition on in-core and out-of-core surfaces. Radioactive corrosion particles were collected from the primary coolant and the steam generator surfaces of the 4 reactor units and subjected to detailed microanalytical and radioanalytical investigations. Scanning electron microscopy and energy dispersive X-ray microanalysis (SEM-EDX) were used to study the morphology and the composition of the matrix elements in the particles and the deposited corrosion layers. Particles identified by SEM-EDX were re-located under optical microscope by means of a coordinate transformation algorithm and were separated with a micromanipulator for further studies. Activities of γ emitting radionuclides were determined by high resolution γ spectrometry, and those of β decaying isotopes were measured by liquid scintillation (LS) spectrometry after radiochemical processing. High sensitivity of the nuclear measuring techniques allowed us to determine the low activity concentrations of the long-lived radionuclides, i.e. 60 Co, 54 Mn, 63 Ni, 55 Fe in the individual particles. Finally, high resolution sector-field inductively coupled plasma mass spectrometry (SF-ICP-MS) was applied to determine the ultralow concentrations of Co, Fe, Ni in the same particles. Specific activities of 60 Co/Co, 54 Mn/Fe, 55 Fe/Fe and 63 Ni/Ni were derived from the measured activity and concentration data. Specific activities of the radioactive corrosion products reveal the history of activity buildup processes in the particle. Typically, Fe-Cr-Ni oxide particles formed as a result of corrosion of the steel

  7. Liquid metal purification device

    International Nuclear Information System (INIS)

    Sakai, Takao; Shimoyashiki, Shigehiro.

    1992-01-01

    The device of the present invention concerns a liquid metal purification device for removing and purifying impuries in liquid metal sodium used as coolants of an FBR type reactor. A vessel having a group of pipes made of hydrogen permeable metal at the inside thereof is disposed to the inlet pipeline of a cold trap. The group of hydrogen permeable metal pipes is connected to an exhaust pipe and a vacuum pump, so that the inside of the pipes is exhausted. Liquid metal sodium branched from the main pipeline of a coolant system passes through the outer side of the group of the hydrogen permeable metal pipes. In this cae, hydrogen contained as impurities in the liquid metal sodium diffuses and permeates the hydrogen permeation metal pipes and enters into the pipe group and is discharged out of the system by the vacuum pump. This can mitigate the hydrogen removing burden of the cold trap, to extend the device life time. (I.N.)

  8. Decontamination of CANDU primary coolant system

    International Nuclear Information System (INIS)

    Pettit, P.J.

    1975-01-01

    Decontamination of radioactive systems is necessary to reduce personnel radiation exposures and also to reduce exposure during special work. Mechanical decontamination methods are sometimes useful, but most contaminated surfaces are inaccessible, so chemical decontamination often is preferred. The A-P Citrox method will remove most contaminants from CANDU systems, but is costly and long, damages components, and produces large quantities of radioactive liquid waste. The Redox cycling process is fast and inexpensive, produces only solid wastes, but removes small quantities of deposit from Monel only. The CAN-DECON process removes deposits from most materials including fuel cladding and has many other advantages. (author)

  9. 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

  10. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  11. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  12. Surface Treatment to Improve Corrosion Resistance in Lead-Alloy Coolants

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  13. Surface Treatment to Improve Corrosion Resistance in Lead-Alloy Coolants

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-08-29

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

  14. Actively controlling coolant-cooled cold plate configuration

    Science.gov (United States)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  15. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years

  16. Design and development of remotely operated coolant channel cutting machine

    International Nuclear Information System (INIS)

    Suthar, R.L.; Sinha, A.K.; Srikrishnamurty, G.

    1994-01-01

    One of the coolant tubes of Narora Atomic Power Station (NAPS) reactor needs to be removed. To remove a coolant tube, four cutting operations, (liner tube cutting, end-fitting cutting, machining of seal weld of bellow ring and finally coolant tube cutting) are required to be carried out. A remotely operated cutting machine to carry out all these operations has been designed and developed by Central Workshops. This machine is able to cut at the exact location because of numerically controlled axial and radial travel of tool. Only by changing the tool head and tool holder, same machine can be used for various types of cutting/machining operations. This report details the design, manufacture, assembly and testing work done on the machine. (author). 4 figs

  17. Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

    International Nuclear Information System (INIS)

    Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

    2016-01-01

    KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

  18. Physical model and calculation code for fuel coolant interactions

    International Nuclear Information System (INIS)

    Goldammer, H.; Kottowski, H.

    1976-01-01

    A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)

  19. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  20. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  1. 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

  2. Experimental interaction of magma and “dirty” coolants

    Science.gov (United States)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with ~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency and increase in

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  5. Verification of computer code for calculation of coolant radiolysis in the VVER reactor core with regard for boiling in its upper part

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, O.P.; Kabakchi, S.A. [OKB Gidropress, Podolsk, Moscow (Russian Federation)

    2010-07-01

    Code Bora for WWER coolant radiolysis calculation considering single jets boiling in the reactor core top part is developed on the basis of computer codes MOPABA-H2 (radiolysis of aqueous solutions) and SteamRad (radiolysis of vapor). Physico-chemical processes taking place in boiling core coolant are complex and diversified. Still, for the solution of certain problems their simulation can be simplified. The approach of reasonable simplification was used for development of code Bora: mathematical model assumed is purposed for simulation of phenomena only in the area of interest; the number of simulated chemical reactions and particles shall be reasonably minimum; complexity of interphase mass transfer calculation procedure shall be adequate to actually available accuracy of modeling. The analysis of new experimental initial yields of water radiolysis products data and kinetic parameters of elementary chemical reactions with their participation has been carried out. Some changes have been introduced in the mechanism of liquid water and aqueous solutions of ammonia radiolysis have been significantly revised on the basis of this analysis. Examples of the calculations provided for code Bora verification are presented. Despite of very simple simulation of interphase mass transfer, Bora allows to obtain average chemical composition of two-phase coolant at BWR core outlet with the accuracy sufficient for engineering calculations. The report also presents the results of two-phase coolant chemical composition test calculation for reactor core top part coolant boiling in pressurized water reactor. (author)

  6. User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

    1979-10-01

    The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

  7. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

  8. Reactor coolant pump shaft seal behavior during blackout conditions

    International Nuclear Information System (INIS)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

  9. Effects of different rod spacers (helical types) on coolant crossmixing

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sviridenko, E.Ya.; Matyukhin, N.M.; Rymkevich, K.S.; Ushakov, P.A.

    1981-11-01

    The results of investigations (electromagnetic measuring method) on coolant cross mixing in rod clusters with spiral wire spacers with different winding directions, with alternating unfinned and finned rods (case 'fin to rod'), as well as in rod clusters with much space between the rods, (case 'fin to fin') are reported. The local fluid dynamics parameters (distribution of the transversal and longitudinal velocity component) that define the physical processes of the coolant exchange in the rod clusters with helical spacers are explained. The investigation results for different helical spacer types are compared with each other. (orig.) [de

  10. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  11. Knock-limited performance of several internal coolants

    Science.gov (United States)

    Bellman, Donald R; Evvard, John C

    1945-01-01

    The effect of internal cooling on the knock-limited performance of an-f-28 fuel was investigated in a CFR engine, and the following internal coolants were used: (1) water, (2), methyl alcohol-water mixture, (3) ammonia-methyl alcohol-water mixture, (4) monomethylamine-water mixture, (5) dimethylamine-water mixture, and (6) trimethylamine-water mixture. Tests were run at inlet-air temperatures of 150 degrees and 250 degrees F. to indicate the temperature sensitivity of the internal-coolant solutions.

  12. Impedance calculations for power cables to primary coolant pump motors

    International Nuclear Information System (INIS)

    Hegerhorst, K.B.

    1977-01-01

    The LOFT primary system motor generator sets are located in Room B-239 and are connected to the primary coolant pumps by means of a power cable. The calculated average impedance of this cable is 0.005323 ohms per unit resistance and 0.006025 ohms per unit reactance based on 369.6 kVA and 480 volts. The report was written to show the development of power cable parameters that are to be used in the SICLOPS (Simulation of LOFT Reactor Coolant Loop Pumping System) digital computer program as written in LTR 1142-16 and also used in the pump coastdowns for the FSAR Analysis

  13. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  14. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

  15. Sensitivity calculation of the coolant temperature regarding the thermohydraulic parameters

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de; Silva, F.C. da; Thome Filho, Z.D.; Alvim, A.C.M.; Oliveira Barroso, A.C. de.

    1985-01-01

    It's studied the application of the Generalized Perturbation Theory (GPT) in the sensitivity calculation of thermalhydraulic problems, aiming at verifying the viability of the extension of the method. For this, the axial distribution, transient, of the coolant temperature in a PWR channel are considered. Perturbation expressions are developed using the GPT formalism, and a computer code (Tempera) is written, to calculate the channel temperature distribution and the associated importance function, as well as the effect of the thermalhydraulic parameters variations in the coolant temperature (sensitivity calculation). The results are compared with those from the direct calculation. (E.G.) [pt

  16. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  17. Prediction of loop seal formation and clearing during small break loss of coolant accident

    International Nuclear Information System (INIS)

    Lee, Suk Ho; Kim, Hho Jung

    1992-01-01

    Behavior of loop seal formation and clearing during small break loss of coolant accident is investigated using the RELAP5/MOD2 and /MOD3 codes with the test of SB-CL-18 of the LSTF(Large Scale Test Facility). The present study examines the thermal-hydraulic mechanisms responsible for early core uncovery includeing the manometric effect due to an asymmetric coolant holdup in the steam generator upflow and downflow side. The analysis with the RELAP5/ MOD2 demonstrates the main phenomena occuring in the depressurization transient including the loop seal formation and clearing with sufficient accuracy. Nevertheless, several differences regarding the evolution of phenomena and their timing have been pointed out in the base calculations. The RELAP5/MOD3 predicts overall phenomena, particularly the steam generator liquid holdup better than the RELAP5/MOD2. The nodalization study in the components of the steam generator U-tubes and the cross-over legs with the RELAP5/MOD3 results in good prediction of the loop seal clearing phenomena and their timing. (Author)

  18. Monitoring PWR reactor vessel liquid level with SPNDs during LOCAs

    International Nuclear Information System (INIS)

    Adams, J.P.

    1982-01-01

    Data from in-core self-powered neutron detectors taken during two nuclear loss-of-coolant accident simulations have been correlated with core moderator density changes. The detector current attenuation has been calculated during blowdown and reflood phases of the simulation. Based on these data, it is concluded that these detectors could be used to monitor reactor vessel liquid level during loss-of-coolant accidents in pressurized water reactors

  19. OPS liquid pathway generic study. Topical report No. 22A60

    International Nuclear Information System (INIS)

    1977-06-01

    An evaluation of the consequences of radioactivity released from an offshore nuclear plant to liquid pathways as a result of postulated accidents more severe than the design basis Loss-of-Coolant Accident is presented

  20. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    Improvements in the design of the thermally insulating material used to shield the concrete containment walls in liquid metal cooled nuclear reactors are described in detail. The insulating material is composed of two layers and is placed between the primary vessel (usually steel) and the steel lined concrete containment vault. The outer layer, which clads the inner wall surface of the vault, is generally impervious to liquid metal coolant whilst the inner layer is pervious to the coolant. In normal operation, both layers protect the concrete from heat radiated from the reactor. In the event of a breach of the containment vessel, the resulting leakage of liquid metal coolant permeates the inner layer of insulating material, provides a means of heat transfer by conduction and hence reduces the overall insulating properties of the two layers. The outer layer continues to protect the wall surface of the vault from substantial direct contact with the liquid metal. Thus the two apparently conflicting requirements of good thermal insulation during normal operation and of heat transfer during loss of coolant accidents are satisfied by this novel design. Suggestions are given for possible materials for use as the insulating layers. (U.K.)

  1. 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

  2. Enriched boric acid as an optimized neutron absorber in the EPR primary coolant

    International Nuclear Information System (INIS)

    Cosse, Christelle; Jolivel, Fabienne; Berger, Martial

    2012-09-01

    This paper focuses on one of the most important EPR PWR reactor design optimizations, through primary coolant conditioning by enriched boric acid (EBA). On PWRs throughout the world, boric acid has already been implemented in primary coolant and associated auxiliary systems for criticality control, due to its high Boron 10 neutron absorption cross section. Boric acid also allows primary coolant pH 300C control in combination with lithium hydroxide in many PWRs. The boric acid employed in the majority of existing PWRs is the 'natural' one, with a typical isotopic atomic abundance in Boron 10 about 19.8 at.%. However, EPR requirements for neutron management are more important, due to its fully optimized design compared to older PWRs. From the boron point of view, it means that criticality could be controlled either by increased 'natural' Boron concentrations or by using EBA. Comparatively to 'natural' boric acid, EBA allows for: - the use of smaller storage volumes for an identical total Boron concentration, or lower total Boron concentration if the tank volumes are kept identical. The latter also reduces the risks of boric acid crystallization, in spite of increased neutron-absorbing properties - the application of an evolutionary chemistry operating regime called Advanced pH Control, making it possible to maintain a constant pH 300C value at 7.2 in the primary coolant at nominal conditions throughout entire cycles. This optimized stability of pH 300C will contribute to reduce the consequences of contamination of the reactor coolant system by corrosion products, and consequently, all related issues - the reduction of borated liquid wastes, thanks to maximal recycling resulting from EPR design. The increased design costs associated with EBA are consequently compensated by a reduced total consumption of this chemical. Therefore, the basic design choice for the EPR is the use of EBA. For the Flamanville 3 EPR, according to the above

  3. 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)

  4. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  5. Nanofluid as coolant for grinding process: An overview

    Science.gov (United States)

    Kananathan, J.; Samykano, M.; Sudhakar, K.; Subramaniam, S. R.; Selavamani, S. K.; Manoj Kumar, Nallapaneni; Keng, Ngui Wai; Kadirgama, K.; Hamzah, W. A. W.; Harun, W. S. W.

    2018-04-01

    This paper reviews the recent progress and applications of nanoparticles in lubricants as a coolant (cutting fluid) for grinding process. The role of grinding machining in manufacturing and the importance of lubrication fluids during material removal are discussed. In grinding process, coolants are used to improve the surface finish, wheel wear, flush the chips and to reduce the work-piece thermal deformation. The conventional cooling technique, i.e., flood cooling delivers a large amount of fluid and mist which hazardous to the environment and humans. Industries are actively looking for possible ways to reduce the volume of coolants used in metal removing operations due to the economical and ecological impacts. Thus as an alternative, an advanced cooling technique known as Minimum Quantity Lubrication (MQL) has been introduced to the enhance the surface finish, minimize the cost, to reduce the environmental impacts and to reduce the metal cutting fluid consumptions. Nanofluid is a new-fangled class of fluids engineered by dispersing nanometre-size solid particles into base fluids such as water, lubrication oils to further improve the properties of the lubricant or coolant. In addition to advanced cooling technique review, this paper also reviews the application of various nanoparticles and their performance in grinding operations. The performance of nanoparticles related to the cutting forces, surface finish, tool wear, and temperature at the cutting zone are briefly reviewed. The study reveals that the excellent properties of the nanofluid can be beneficial in cooling and lubricating application in the manufacturing process.

  6. 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.

  7. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  8. A Design Study for Standard Nanofluid Coolants

    International Nuclear Information System (INIS)

    Bang, In Cheol; Heo, Gyun Young

    2007-01-01

    The experimental data for nanofluids in thermal-fluid systems have shown that the new fluids promise to become advanced heat transfer fluids in terms of thermal performance. While enhancing thermal characteristics, the solid-liquid mixtures present an unavoidable disadvantage in terms of pumping cost for economic operation of thermal-fluid systems. In addition, there is a lack of agreement between experimental data provided in the literature. We can find that this issue of nanofluids resembles that of designing new materials. Many nanofluids researchers tend to view the nanofluid field as a highly coupled 'tetrahedro' whose four vertices (performance, properties, structure, and processes) are interconnected to each other. The present design study has a big merit to systemize the nanofluid work and to reduce a lot of trial-error efforts. The present work found that there would be no comprehensible design strategy in developing nanofluids. In this work, the Axiomatic Design (AD) theory is applied to standardize the design of nanofluids in order to bring its practical use forward. According to the Independence Axiom of the AD theory, the excessive couplings between the functional requirements and the parameters of a nanofluid system prevent from meeting the functional goals of the entire system. At a parametric level, the design of a nanofluid system is inherently coupled due to the characteristics of thermal-fluid system; the design parameters physically affect each other sharing sub-level parameters for nanoparticles with making a feedback loop. Even though parts of the nanofluids are naturally coupled, it is possible to reduce and/or eliminate the degree of coupling by help of AD principles. From the perspective of AD, this implies that we are able to ascertain which nanofluid system is better one in the light of functional achievement

  9. Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins

    International Nuclear Information System (INIS)

    Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A.

    1979-01-01

    This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed (U,Pu)-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and (excess) liquid alkali metal (Na, K, Cs) were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate (equilibrium) phases has been analyzed. (orig./RW) [de

  10. Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

    International Nuclear Information System (INIS)

    Nguyen Van Thai; Kieu Ngoc Dung

    2015-01-01

    This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

  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. 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

  13. Responses to Small Break Loss of Coolant Accidents for SMART

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

    2004-01-01

    The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

  14. Fuel-Coolant Interactions: Visualization and Mixing Measurements

    International Nuclear Information System (INIS)

    Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

    2002-01-01

    Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

  15. 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

  16. Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

    International Nuclear Information System (INIS)

    Gaudard, G.; Gilles, B.; Mesnage, F.; Cattant, F.

    2002-01-01

    In the framework of a radiation exposure management program entitled >, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on 110m Ag, 122 Sb, 124 Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

  17. Modular Porous Plate Sublimator /MPPS/ requires only water supply for coolant

    Science.gov (United States)

    Rathbun, R. J.

    1966-01-01

    Modular porous plate sublimators, provided for each location where heat must be dissipated, conserve the battery power of a space vehicle by eliminating the coolant pump. The sublimator requires only a water supply for coolant.

  18. Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

    International Nuclear Information System (INIS)

    Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

    2010-01-01

    Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

  19. Loss-of-coolant accident analysis of the Savannah River new production reactor design

    International Nuclear Information System (INIS)

    Maloney, K.J.; Pryor, R.J.

    1990-11-01

    This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

  20. Development of liquid type TBM technology for ITER

    International Nuclear Information System (INIS)

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

    2012-03-01

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

  1. 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

  2. Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

    International Nuclear Information System (INIS)

    Aanstad, O.J.; Sklencar, A.M.

    1983-01-01

    A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

  3. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors; Determinacion espectrografica de impurezas metalicas en refrigerantes organicos para reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Martin Munoz, M; Alvarez Gonzalez, F

    1969-07-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs.

  4. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.; Allbeson, K.F.

    1984-01-01

    In a liquid metal cooled nuclear reactor with a nuclear fuel assembly in a coolant-containing primary vessel housed within a concrete containment vault, there is thermal insulation to protect the concrete, the insulation being disposed between vessel and concrete and being hung from metal structure secured to and projecting from the concrete, the insulation consisting of a plurality of adjoining units each unit incorporating a pack of thermal insulating material and defining a contained void co-extensive with said pack and situated between pack and concrete, the void of each unit being connected to the voids of adjoining units so as to form continuous ducting for a fluid coolant. (author)

  5. Liquids and liquid mixtures

    CERN Document Server

    Rowlinson, J S; Baldwin, J E; Buckingham, A D; Danishefsky, S

    2013-01-01

    Liquids and Liquid Mixtures, Third Edition explores the equilibrium properties of liquids and liquid mixtures and relates them to the properties of the constituent molecules using the methods of statistical thermodynamics. Topics covered include the critical state, fluid mixtures at high pressures, and the statistical thermodynamics of fluids and mixtures. This book consists of eight chapters and begins with an overview of the liquid state and the thermodynamic properties of liquids and liquid mixtures, including vapor pressure and heat capacities. The discussion then turns to the thermodynami

  6. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  7. Vision system for precision alignment of coolant channels

    International Nuclear Information System (INIS)

    Kar, S.; Rao, Y.V.; Valli Kumar; Joshi, D.G.; Chadda, V.K.; Nigam, R.K.; Kayal, J.N.; Panwar, S.; Sinha, R.K.

    1997-01-01

    This paper describes a vision system which has been developed for precision alignment of Coolant Channel Replacement Machine (CCRM) with respect to the front face of the coolant channel under repair/replacement. It has provisions for automatic as well as semi-automatic alignment. A special lighting scheme has been developed for providing illumination to the front face of the channel opening. This facilitates automatic segmentation of the digitized image. The segmented image is analysed to obtain the centre of the front face of the channel opening and thus the extent of misalignment i.e. offset of the camera with respect to the front face of the channel opening. The offset information is then communicated to the PLC to generate an output signal to drive the DC servo motors for precise positioning of the co-ordinate table. 2 refs., 5 figs

  8. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  9. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  10. Rapid thermal transient in a reactor coolant channel

    International Nuclear Information System (INIS)

    Cherubini, A.

    1986-01-01

    This report deals with the problem of one-dimensional thermo-fluid-dynamics in a reactor coolant channel, with the aim of determining the evolution in time of the coolant (H*L2O), in one-and/or two-phase regimes, subjected to a great and rapid increase in heat flux (accident conditions). To this aim, the following are set out: a) the physical model used; b) the equations inherent in the above model; c) the numerical methods employed to solve them by means of a computer programme called CABO (CAnale BOllente). Next a typical problem of rapid thermal transient resolved by CABO is reported. The results obtained, expressed in form of graphs, are fully discussed. Finally comments on possible developments of CABO follow

  11. 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)

  12. Qualification test of a main coolant pump for SMART pilot

    International Nuclear Information System (INIS)

    Park, Sang Jin; Yoon, Eui Soo; Oh, Hyong Woo

    2006-01-01

    SMART Pilot is a multipurpose small capacity integral type reactor. Main Coolant Pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of 310 .deg. C and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present work, a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and life-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP

  13. 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

  14. Device for preventing leakage of coolant in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kobayashi, Yukio; Sekiguchi, Mamoru; Yoshida, Hideo.

    1975-01-01

    Object: To prevent leakage of coolant from between lower tie plate and channel box without causing deformation of the channel box and also without the possibility of disturbing the installation and removal of the box by the provision of a thin plate provided with leakage holes for the lower tie plate. Structure: Static water pressure within the lower tie plate is adapted to act upon the bear side of a flat plate for leakage prevention through leakage holes formed in the tie plate, thus urging the flat plate against the channel box inner surface. Meanwhile, static water pressure having been led through the leakage holes in the flat plate is adapted to press the flat plate in the vertical direction, thus urging the flat plate against the channel box inner surface and thereby preventing leakage of the coolant through a gap between the channel box and lower tie plate. (Yoshino, Y.)

  15. Calculation of coolant temperature sensitivity related to thermohydraulic parameters

    International Nuclear Information System (INIS)

    Silva, F.C. da; Andrade Lima, F.R. de

    1985-01-01

    It is verified the viability to apply the generalized Perturbation Theory (GPT) in the calculation of sensitivity for thermal-hydraulic problems. It was developed the TEMPERA code in FORTRAN-IV to transient calculations in the axial temperature distribution in a channel of PWR reactor and the associated importance function, as well as effects of variations of thermalhydraulic parameters in the coolant temperature. The results are compared with one which were obtained by direct calculation. (M.C.K.) [pt

  16. Application of damage function analysis to reactor coolant circuits

    International Nuclear Information System (INIS)

    MacDonald, D.D.

    2002-01-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  17. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  18. Structural integrity analysis of reactor coolant pump flywheel(I)

    International Nuclear Information System (INIS)

    Kim, Young Jin

    1986-01-01

    A reactor coolant pump flywheel is an important machine element to provide the necessary rotational inertia in the event of loss of power to the pumps. This paper attempts to assess the influence of keyways on flywheel stresses and fracture behaviour in detail. The finite element method was used to determine stresses near keyways, including residual stresses, and to establish stress intensity factors for keyway cracks for use in fracture mechanics assessments. (Author)

  19. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

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

  20. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  1. Design of Reactor Coolant Pump Seal Online Monitoring System

    International Nuclear Information System (INIS)

    Ah, Sang Ha; Chang, Soon Heung; Lee, Song Kyu

    2008-01-01

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation

  2. Detection of coolant void in lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Wolniewicz, Peter; Håkansson, Ane; Jansson, Peter

    2015-01-01

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

  3. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  4. Assessment of Candidate Molten Salt Coolants for the Advanced High Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Williams, D.F.

    2006-03-24

    The Advanced High-Temperature Reactor (AHTR) is a novel reactor design that utilizes the graphite-matrix high-temperature fuel of helium-cooled reactors, but provides cooling with a high-temperature fluoride salt. For applications at temperatures greater than 900 C the AHTR is also referred to as a Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR). This report provides an assessment of candidate salts proposed as the primary coolant for the AHTR based upon a review of physical properties, nuclear properties, and chemical factors. The physical properties most relevant for coolant service were reviewed. Key chemical factors that influence material compatibility were also analyzed for the purpose of screening salt candidates. Some simple screening factors related to the nuclear properties of salts were also developed. The moderating ratio and neutron-absorption cross-section were compiled for each salt. The short-lived activation products, long-lived transmutation activity, and reactivity coefficients associated with various salt candidates were estimated using a computational model. Table A presents a summary of the properties of the candidate coolant salts. Certain factors in this table, such as melting point, vapor pressure, and nuclear properties, can be viewed as stand-alone parameters for screening candidates. Heat-transfer properties are considered as a group in Sect. 3 in order to evaluate the combined effects of various factors. In the course of this review, it became apparent that the state of the properties database was strong in some areas and weak in others. A qualitative map of the state of the database and predictive capabilities is given in Table B. It is apparent that the property of thermal conductivity has the greatest uncertainty and is the most difficult to measure. The database, with respect to heat capacity, can be improved with modern instruments and modest effort. In general, ''lighter'' (low-Z) salts tend to

  5. On-line real time gamma analysis of primary coolant

    International Nuclear Information System (INIS)

    Kalechstein, W.; Kupca, S.; Lipsett, J.J.

    1985-10-01

    The evolution of failed fuel monitoring at CANDU power stations is briefly summarized and the design of the latest system for failed fuel detection at a multi-unit power station is described. At each reactor, the system employs a germanium spectrometer combined with a novel spectrum analyzer that simultaneously accumulates the gamma-ray spectrum of the coolant and provides the control room with the concentration of radioisotope activity in the coolant for the gaseous fission products Xe-133, Xe-135, Kr-88 and I-131 in real time and with statistical precision independent of count rate. A gross gamma monitor is included to provide independent information on the level of radioactivity in the coolant and extend the measurement range at very high count rates. A central computer system archives spectra received from all four spectrum analyzers and provides both the activity concentrations and the release rates of specified isotopes. Compared with previous systems the current design offers improvements in that the activity concentrations are updated much more frequently, improved tools are provided for long term surveillance of the heat transport system and the monitor is more reliable and less costly

  6. 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

  7. Design criteria of primary coolant chemistry in SMART-P

    International Nuclear Information System (INIS)

    Choi, Byung Seon; Kim, Ah Young; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    SMART-P differs significantly from commercially designed PWRs. Materials inventories used in SMART-P differ from that at PWRs. All surfaces of the primary circuit with the primary coolant are either made from or plated with stainless steel. The material of steam generator (SG) is also different from that of the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. Also, SMART-P primary coolant technology differs from that in PWRs: ammonia is used as a pH raising agent and hydrogen formed due to radiolytic processes is kept in specific range by ammonia dosing. Nevertheless, main objectives of the SMART-P primary coolant are the same as at PWRs: to assure primary system pressure boundary integrity, fuel cladding integrity and to minimize out-of-core radiation buildup. The objective of this work is to introduce the design criteria for the primary water chemistry for SMART-P from the viewpoint of the system characteristics and the chemical design concept

  8. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  9. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  10. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

    There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

  11. 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%

  12. Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

    International Nuclear Information System (INIS)

    Almeida, G.L. de.

    1986-01-01

    The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

  13. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

    The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

  14. Reverse osmosis for the recovery of boric acid from the primary coolant at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Bártová, Šárka, E-mail: sarka.bartova@cvrez.cz [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); Kůs, Pavel [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); Skala, Martin [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic); University of Chemical Technology, Prague, Department of Chemical Engineering, Technická 5, Prague 166 28 (Czech Republic); Vonková, Kateřina [Research Centre Řež Ltd., Husinec-Řež 130, 250 68 Řež (Czech Republic)

    2016-04-15

    Highlights: • RO membranes tested for boric acid recovery from primary coolant of nuclear power plants. • Scanning electron microscopy was used for the characterization of the membranes. • Lab scale experiments performed under various operation conditions. • We proposed configuration of and operation conditions for RO unit in nuclear power plant. - Abstract: At nuclear power plants (NPP), evaporators are used for the treatment of primary coolant and other liquid radioactive waste containing H{sub 3}BO{sub 3}. Because the operation of evaporators is expensive, a number of more cost-effective alternatives has been considered, one of which is reverse osmosis. We tested reverse osmosis modules from several manufactures on a batch laboratory apparatus. SEM images of the tested membranes were taken to distinguish the differences between the membranes. Water permeability through membranes was evaluated from the experiments with pure water. The experiments were performed with feed solutions containing various concentrations of H{sub 3}BO{sub 3} in a range commonly occurring in radioactive waste. The pH of the feed solutions ranged from 5.2 to 11.2. Our results confirmed that the pH of the feed solution plays the most important role in membrane separation efficiency of H{sub 3}BO{sub 3}. Certain modifications to the pH of the feed solution were needed to enable the tested membranes to concentrate the H{sub 3}BO{sub 3} in the retentate stream, separate from the pure water in the permeate stream. On this basis, we propose the configuration of and operational conditions for a reverse osmosis unit at NPP.

  15. Development for LMR coolant technology - Development of a submersible-in-pool electromagnetic pump

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hee; Kim, Hee Reyoung; Lee, Sang Don; Seo, Chun Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)

    1995-08-01

    The conceptual and detailed designs of an annular linear induction electromagnetic pump of small scale submersible-in-pool type are performed for the purpose of domestic development of the pumps used for the high-temperature natrium coolant transportation in liquid metal reactors. The pump drawings for and input power of 1,100 VA, an input frequency of 17 Hz, a maximum flowrate of 60 l/min and a maximum operation temperature of 600 deg C are obtained from the optimum design analyses by solving MHD and equivalent circuit equations. The characteristics of pump materials in the high temperature and neutron irradiation environment are reflected in designing the pump, and theoretical analyses for improving the pump performance and efficiency are tried through calculations of magnetic flux and temperature distributions inside the pump. The present project contributes to the further design of engineering proto-type electromagnetic pump with higher capacity and the development of liquid metal reactor with innovative simplicity. 44 refs., 4 tabs., 33 figs. (author)

  16. 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)

  17. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  18. Numerical FEM Analyses of primary coolant system at NPP Temelin

    International Nuclear Information System (INIS)

    Junek, L.; Slovacek, M.; Ruzek, L.; Moulis, P.

    2003-01-01

    The main goal of this paper is to inform about the beginning and first steps of implementation of an aging management system at the Temelin NPP. The aging management system is important not only for achieving the current safety level but also for reaching operational reliability of a production unit equipment above the life time assumed by the original design, typically over 40 years. A method to locate the most prominent degradation regions is described. A global shell model of the primary coolant system including all loops and their components - reactor pressure vessel (RPV), steam generator (SG), main coolant pump (MCP), pressurizer, feed water and steam pipelines system is presented. The results of stress-strain analysis on the measured service parameters base are given. Validation of the results is very important and the method to compare the service measurement data with the numerical results is described. The global/local approach is mentioned and discussed. The effects of the complete global system on the individual components under monitoring are transformed into more accurate local spatial models. The local spatial models are used to analyze the gradual lifetime exhaustion of a facility during its service operation. Two spatial local models are presented, viz. feed water nozzle of SG and main coolant piping system T-brunch. The results of analysis of the local spatial models are processed by the neural network computing method, which is also described. The actual gradual damage of the material of the components under monitoring can be obtained based on the analyses performed and on the results from the neural network in combination with the knowledge of the real material characteristics. The procedures applied are included in the DIALIFE diagnostic system

  19. 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

  20. Modelling transient energy release from molten fuel coolant interaction debris

    International Nuclear Information System (INIS)

    Fletcher, D.F.

    1984-05-01

    A simple model of transient energy release in a Molten Fuel Coolant Interaction is presented. A distributed heat transfer model is used to examine the effect of heat transfer coefficient, time available for rapid energy heat transfer and particle size on transient energy release. The debris is assumed to have an Upper Limit Lognormal distribution. Model predictions are compared with results from the SUW series of experiments which used thermite-generated uranium dioxide molybdenum melts released below the surface of a pool of water. Uncertainties in the physical principles involved in the calculation of energy transfer rates are discussed. (author)

  1. In-operation diagnostic system for reactor coolant pump

    International Nuclear Information System (INIS)

    Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

    1996-01-01

    A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

  2. 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

  3. Graphite beds for coolant filtration at high temperature

    International Nuclear Information System (INIS)

    Heathcock, R.E.; Lacy, C.S.

    1978-01-01

    High temperature filtration will be provided for new Ontario Hydro CANDU heat transport systems. Filtration has been shown to effectively reduce the concentration of circulating corrosion products in our heat transport systems, hence, minimizing the processes of activity transport. This paper will present one option we have for this application; Deep Bed Granular Graphite Filters. The filter system is described by discussing pertinent aspects of its development programme. The compatibility of the filter and the heat transport coolant are demonstrated by results from loop tests, both out- and in-reactor, and by subsequent results from a large filter installation in the NPD NGS heat transport system. (author)

  4. Effects contributing to positive coolant void reactivity in CANDU

    International Nuclear Information System (INIS)

    Whitlock, J.J.; Garland, W.J.; Milgram, M.S.

    1995-01-01

    The lattice cell code WIMS-AECL (Ref. 3) is used to model a typical CANDU lattice cell, using nominal geometric bucklings, the PIJ option, and 69-group Winfrith library. The effect of cell voiding is modeled as a 100% instantaneous removal of coolant from the lattice. This is conservative because of the neglect of time dependence and partial core voiding, considered more plausible in CANDU. Results are grouped into three spectral groups: fast neutrons (0.821 to 10 MeV), epithermal neutrons (0.625 eV to 0.821 MeV), and thermal neutrons (<0.625 eV)

  5. COPDIRC - calculation of particle deposition in reactor coolants

    International Nuclear Information System (INIS)

    Reeks, M.W.

    1982-06-01

    A description is given of a computer code COPDIRC intended for the calculation of the deposition of particulate onto smooth perfectly sticky surfaces in a gas cooled reactor coolant. The deposition is assumed to be limited by transport in the boundary layer adjacent to the depositing surface. This implies that the deposition velocity normalised with respect to the local friction velocity, is an almost universal function of the normalised particle relaxation time. Deposition is assumed similar to deposition in an equivalent smooth perfectly absorbing pipe. The deposition is calculated using 2 models. (author)

  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. 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)

  8. Definition of loss-of-coolant accident radiation source

    International Nuclear Information System (INIS)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist

  9. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

    Kalafati, D.D.; Petrov, S.I.

    1978-01-01

    Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

  10. Burnup dependence of coolant void reactivity for ACR-1000 cell

    International Nuclear Information System (INIS)

    Le Tellier, R.; Marleau, G.; Hebert, A.; Roubstov, D.; Altiparmakov, D.; Irish, D.

    2008-01-01

    The Advanced Candu Reactor (ACR-1000) is light water cooled, fueled with enriched uranium and has a smaller lattice pitch than the Candu-6. As a result, the neutronic behavior of the ACR-1000 cell is expected to be somewhat different from that of the Candu-6 leading to a negative coolant void reactivity (CVR). Here we evaluate the CVR for the ACR-1000 cell using the lattice code DRAGON and compare our results with those obtained using the code WIMS-AECL. The differences observed between these two codes for the burnup dependence of the CVR is mainly explained in terms of the specific cell leakage model used by both codes. (authors)

  11. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  12. Void fraction calculation in a channel containing boiling coolant

    International Nuclear Information System (INIS)

    Norelli, F.

    1978-01-01

    The problem of void fraction calculation was studied for a channel containing boiling coolant, when a slip ratio correlation is used. Use of fitting (e.g. polinomial or rational algebraic) for slip ratio correlation and the characteristic method are proposed in this work. In this way we are reduced to some elementary quadrature problem. Another problem discussed in the present work concerns what we must consider as ''initial condition'' in any initial value problem, in order to take into account different error distributions in steady state and in successive time-dependent calculations

  13. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  14. Reactor coolant pump shaft seal stability during station blackout

    International Nuclear Information System (INIS)

    Rhodes, D.B.; Hill, R.C.; Wensel, R.G.

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries

  15. Reactor coolant pump shaft seal stability during station blackout

    Energy Technology Data Exchange (ETDEWEB)

    Rhodes, D B; Hill, R C; Wensel, R G

    1987-05-01

    Results are presented from an investigation into the behavior of Reactor Coolant Pump shaft seals during a potential station blackout (loss of all ac power) at a nuclear power plant. The investigation assumes loss of cooling to the seals and focuses on the effect of high temperature on polymer seals located in the shaft seal assemblies, and the identification of parameters having the most influence on overall hydraulic seal performance. Predicted seal failure thresholds are presented for a range of station blackout conditions and shaft seal geometries.

  16. 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

  17. Emergency cooling system for a liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Murata, Ryoichi; Fujiwara, Toshikatsu.

    1980-01-01

    Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

  18. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1976-06-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc

  19. Thermal management of Li-ion battery with liquid metal

    International Nuclear Information System (INIS)

    Yang, Xiao-Hu; Tan, Si-Cong; Liu, Jing

    2016-01-01

    Highlights: • Liquid metal is used for power battery pack thermal management. • Better cooling performance and more uniform module temperature is obtained. • Less power consumption is needed. • The proposed liquid metal cooling system is robust and can cope with stressful conditions. - Abstract: Thermal management especially cooling of electric vehicles (EVs) battery pack is of great significance for guaranteeing the performance of the cells as well as safety and high-efficiency working of the EVs. Liquid cooling is a powerful way to keep the battery temperature in a proper range. However, the efficiency of conventional liquid cooling is still limited due to the inherently low thermal conductivity of the coolant which is usually water or aqueous ethanol. In this paper, a new kind of coolant, liquid metal, is proposed to be used for the thermal management of the battery pack. Mathematical analysis and numerical simulations are conducted to evaluate the cooling capability, pump power consumption and module temperature uniformity of the liquid metal cooling system, in comparison with that of water cooling. The results show that under the same flow conditions, a lower and more uniform module temperature can be obtained and less pump power consumption are needed in the liquid metal cooling system. In addition, liquid metal has an excellent cooling capability coping with stressful conditions, such as high power draw, defects in cells, and high ambient temperature. This makes it a promising coolant for the thermal management of high driving force EVs and quick charge batteries.

  20. THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

    International Nuclear Information System (INIS)

    Muramatsu, Ken; Akimoto, Masayuki

    1982-08-01

    THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

  1. Design technology development of the main coolant pump for an integral reactor

    International Nuclear Information System (INIS)

    Park, J. S.; Lee, J. S.; Kim, M. H.; Kim, D. W.; Kim, J. I.

    2004-01-01

    All of the reactor coolant pump currently used in commercial nuclear power plant were imported from foreign country. Now, the developing program of design technology for the reactor coolant pump will be started in a few future by domestic researchers. At this stage, the design technology of the main coolant pump for an integral reactor is developed based on the regulation of domestic nuclear power plant facilities. The main coolant pump is a canned motor axial pump, which accommodates all constraints required from the integral reactor system. The main coolant pump does not have mechanical seal device because the rotor of motor and the shaft of impeller are the same one. There is no flywheel on the rotating shaft of main coolant pump so that the coastdown duration time is short when the electricity supply is cut off

  2. 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

  3. 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

  4. Preliminary design of reactor coolant pump canned motor for AC600

    International Nuclear Information System (INIS)

    Deng Shaowen

    1998-01-01

    The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

  5. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  6. 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)

  7. Multi-objective optimization of the reactor coolant system

    International Nuclear Information System (INIS)

    Chen Lei; Yan Changqi; Wang Jianjun

    2014-01-01

    Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

  8. 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

  9. Station blackout with reactor coolant pump seal leakage

    International Nuclear Information System (INIS)

    Evinay, A.

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

  10. Reactor coolant purification system circulation pumps (CUW pumps)

    International Nuclear Information System (INIS)

    Tsutsui, Toshiaki

    1979-01-01

    Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

  11. Coolant clean-up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  12. 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.)

  13. Reactor Coolant Pump Motor Maintenance Experience in Krsko NPP

    International Nuclear Information System (INIS)

    Vukovic, J.; Besirevic, A.; Boljat, Z.

    2016-01-01

    After thirty years of service as well as maintenance in Krsko NPP both original Reactor Coolant Pump (RCP) motors are remanufactured by original vendor Westinghouse and a new one was purchased. Design function of the RCP motor is to drive Reactor Coolant Pump and for coast-down feature during Design Basis Accident. This paper will give a view on maintenance issues of RCP motor during the thirty years of service and maintenance in Krsko NPP to be kept functionally operational. During the processes of remanufacturing inspection and disassembly it was made possible to get a deeper perspective in the motor condition and the wear or fatigue of the motor parts. Parameters like bearing & winding temperature, absolute and relative vibration greatly affect motor operation if not kept inside design margins. Rotational speed causes heat generation at the bearings which is then associated with oil temperatures and as a consequence bearing temperatures. That is why the most critical parts of the motor are the components of upper and lower bearing assembly. The condition of motor stator and rotor assembly technical characteristics shall be explained with respect to influence of demanding environmental conditions that the motor is exposed. Assessment shall be made how does the wear of critical RCP motor parts can influence reliable performance of the motor if not maintained in proper way. Information on upgrades that were done on RCP motor shall be shared: Oil Spillage Protection System (OSPS), Stator upgrades, Dynamic Port, etc. (author).

  14. Heat transfer properties of organic coolants containing high boiling residues

    International Nuclear Information System (INIS)

    Debbage, A.G.; Driver, M.; Waller, P.R.

    1964-01-01

    Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 10 4 to 10 5 . The data was correlated with the equation Nu = 0.015 Re b 0.85 Pr b 0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

  15. Utilization of DRUFAN 01/MOD 02 computer code for the depressurization phase analysis of a postulated loss of coolant accident in Angra 2/3 Nuclear Power Plants

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.; Figueiredo, P.J.M.

    1985-08-01

    The DRUFAN 01/Mod 2 developed by Gesellschaft fur Reaktorsicherheit (GRS) mbh to simulate thermohydraulic behavior of the primary circuit of PWR reactors, during the despressurization phase and initial refilling phase of loss of coolant accidents by great ruptures, is presented. The program simulates the system to be analysed by control volumes-concentrated parameters model - and it is based on numerical solution of conservation equations for mass of water, mass of vapor, quantities of motion and energy, and on the control volume homogeneity hypothesis. The possibilities of thermodynamic disequilibrium, determining mass transfer between liquid and vapor phases assuming that one saturated phase, are considered. The process of computer code implantation in the Honeywell Bull 64 DPS 7 system at CNEN, the modifications done into the program and the application to the despressurization phase analysis of a loss of coolant accident at Angra-2 and Angra-3 reactors are considered. (M.C.K.) [pt

  16. RELAP5 simulation of a large break Loss of Coolant Accident (LOCA) in the hot leg of the primary system in Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Andrade, Delvonei Alves de; Sabundjian, Gaiane

    2004-01-01

    The objective of this work is to present the simulation of a large break loss of coolant accident - LBLOCA in the hot leg of the primary loop in Angra 2, with RELAP5/MOD3.2.2g code. This accident is described in the Final Safety Report Analysis of Angra 2 - FSAR and consists basically of the hot leg total break, in loop 20 of the plant. The area considered for the rupture is 4480 cm 2 , which corresponds to 100% of the pipe flow area. Besides, this work also has the objective of verifying the efficiency of the emergency core coolant system - ECCS in case of accidents and transients. The thermal-hydraulic processes inherent to the accident phenomenology, such as hot leg vaporization and consequently core vaporization causing an inappropriate flow distribution in the reactor core, can lead to a reduction in the liquid level, until the ECCS is capable to reflood it

  17. Determination of blade-to-coolant heat-transfer coefficients on a forced-convection, water-cooled, single-stage turbine

    Science.gov (United States)

    Freche, John C; Schum, Eugene F

    1951-01-01

    Blade-to-coolant convective heat-transfer coefficients were obtained on a forced-convection water-cooled single-stage turbine over a large laminar flow range and over a portion of the transition range between laminar and turbulent flow. The convective coefficients were correlated by the general relation for forced-convection heat transfer with laminar flow. Natural-convection heat transfer was negligible for this turbine over the Grashof number range investigated. Comparison of turbine data with stationary tube data for the laminar flow of heated liquids showed good agreement. Calculated average midspan blade temperatures using theoretical gas-to-blade coefficients and blade-to-coolant coefficients from stationary-tube data resulted in close agreement with experimental data.

  18. Multi-state reliability for coolant pump based on dependent competitive failure model

    International Nuclear Information System (INIS)

    Shang Yanlong; Cai Qi; Zhao Xinwen; Chen Ling

    2013-01-01

    By taking into account the effect of degradation due to internal vibration and external shocks. and based on service environment and degradation mechanism of nuclear power plant coolant pump, a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation. Using this model, degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump. It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment. The results can provide a decision making basis for design changing and maintenance optimization. (authors)

  19. Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

    International Nuclear Information System (INIS)

    Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

    1976-01-01

    In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

  20. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1997-01-01

    This paper presents results of experimental studies on the heat transfer and solidifcation of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. As a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleight number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer

  1. Apparatus for the in situ inspection of tubes while submerged in a liquid

    International Nuclear Information System (INIS)

    Abell, G.E.; Plavsity, L.; Sattler, F.J.

    1979-01-01

    Apparatus is described for the in situ inspection of tubes which are submerged in a liquid such as the primary coolant of a nuclear reactor. A sensor is withdrawn from a tube by a cable. Means are provided for removing the liquid from and drying the cable. The liquid is returned to the tubes preventing the spread of deleterious liquids to otherwise benign environments and fouling of the drive mechanism used to control cable movements

  2. Influence of liquid holdup in steam generator U-tubes on small break LOCA severity

    International Nuclear Information System (INIS)

    Leonard, M.T.; Perryman, J.L.; Johnson, G.W.

    1983-01-01

    The severity of small cold leg break loss-of-coolant accidents has been shown to be influenced by liquid holdup in steam generator U-tubes during pump suction loop seal formation in two experiments performed in the Semiscale Mod-2A facility. The core coolant level can be depressed lower than previously thought possible due to a positive hydrostatic head across the steam generators caused by delayed drainage of liquid from the upflow side of the U-tubes. The significance of a lower core coolant level depression is the potential for a more severe temperature excursion occurring during the coolant boiloff phase subsequent to loop seal clearing and prior to accumulator injection. Presented in this paper are the experimental data analysis and supporting computer code calculations that led to these conclusions

  3. Concentration device for leak liquids

    International Nuclear Information System (INIS)

    Matsumoto, Kaname; Matsuda, Ken; Takabori, Ken-ichi.

    1987-01-01

    Purpose: To improve radioactivity recovery and volume-reducing rates, as well as enable safety and easy handling for leak liquids resulted from reptures in coolant circuits. Constitution: The device of the invention comprises an evaporation vessel filled with leak fluids to a predetermined level, an airtight vessel disposed in the evaporation vessel containing hydrophilic porous material partially immersed in the leak fluids and means for heating the hydrophilic material. In this device, leak liquids are absorbed in the hydrophilic porous material, a great amount of water is evaporated from the outer surface of the hydrophilic porous material exposed above the liquid surface, and salts and radioactive material are remained on the inside and the outer surface of the porous material. The evaporated water content is condensated and recovered in a cooler and the remaining salts, etc. are discarded together with the porous material. The volume-reducing property can be improved by constituting the porous material with burnable material. (Takahashi, M.)

  4. Loss of Coolant Accident Simulation for the Top-Slot break at Cold Leg Focusing on the Loop Seal Reformation under Long Term Cooling with the ATLAS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Rok; Park, Yu Sun; Bae, Byoung Uhn; Choi, Nam Hyun; Kang, Kyoung Ho; Choi, Ki Yong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In the present paper, loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS, which is a thermal-hydraulic integral effect test facility for evolutionary pressurized water reactors (PWRs) of an advanced power reactor of 1400 MWe (APR1400). The simulation was focused on the loop seal reformation under long term cooling condition. During a certain class of Loss of Coolant Accident (LOCA) in a PWR like an advanced power reactor of 1400 MWe (APR1400), the steam volume in the reactor vessel upper plenum and/or upper head may continue expanding until steam blows liquid out of the intermediate leg (U-shaped pump suction cold leg), called loop seal clearing (LSC), opening a path for the steam to be relieved from the break. Prediction of the LSC phenomena is difficult because they are varies for many parameters, which are break location, type, size, etc. This LSC is the major factor that affects the coolant inventory in the small break LOCA (SBLOCA) or intermediate break LOCA (IBLOCA). There is an issue about the loop seal reformation that liquid refills intermediate leg and blocks the steam path after LSC. During the SBLOCA or IBLOCA, the Emergency Core Cooling System (ECCS) is operated. For long term of the top slot small or intermediate break at cold leg, the primary steam condensation by SG heat transfer or SIP, SIT water flooding (reverse flow to loop seal) make loop seal reformation possibly. The primary pressure increase at the top core region due to the steam release blockage by loop seal reformation. And then core level decreases and partial core uncover may occur. The loss of coolant accident for the top-slot break at cold leg was simulated with the ATLAS. The loop seal clearing and loop seal reformation were occurred repeatedly.

  5. Corrosion of ferrous alloys in eutectic lead-lithium environments

    International Nuclear Information System (INIS)

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

    1983-09-01

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

  6. Bandwidth of reactor internals vibration resonance with coolant pressure oscillations

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    In a few decades a significant increase in a part of an electricity development on the NPP will require NPP to be operated in non full capacity modes and increase in operation time in transitive modes. Operating in such conditions as compared to the operation on a constant mode will lead to the increase in cyclic dynamical loading. In water cooled water moderated reactors these loading are realized as low-cyclic and high-cyclic loadings. High-cyclic loadings increases are caused by a raised vibration in non stationary modes of operation. It is known, that in some modes of a non full capacity reactor high-cyclic dynamic loadings can increase. It is obvious, that the development of management technologies is necessary for the life time management operation. In the context of this problem one of the main tasks are revealing and the prevention of the conditions of the occurrence of the operation leading to the resonant interaction of the coolant fluctuations and the equipment, reactor vessel (RV), fuel assemblies (FA) and reactor internals (RI) vibration. To prevent the appearance of the conditions for resonance interaction between the fluid flow and the equipments, it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of coolant outside of which there is no resonant interaction. The presented work is devoted to finding the solution of this problem. There are results of theoretical an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. The accordance of results had been calculated with had been measured are satisfied for practical purposes. These

  7. Zero waste machine coolant management strategy at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Carlson, B.; Algarra, F.; Wilburn, D.

    1998-01-01

    Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extended coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98

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

    Directory of Open Access Journals (Sweden)

    A. N. Shmelev

    2011-01-01

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

  9. Comparison of central axis and jet ring coolant supply for turbine disk cooling on a SSME-HPOTP model

    Science.gov (United States)

    Kim, Y. W.; Metzger, D. E.

    1992-01-01

    The test facility, test methods and results are presented for an experimental study modeling the cooling of turbine disks in the blade attachment regions with multiple impinging jets, in a configuration simulating the disk cooling method employed on the Space Shuttle Main Engine oxygen turbopump. The study's objective was to provide a comparison of detailed local convection heat transfer rates obtained for a single center-supply of disk coolant with those obtained with the present flight configuration where disk coolant is supplied through an array of 19 jets located near the disk outer radius. Specially constructed disk models were used in a program designed to evaluate possible benefits and identify any possible detrimental effects involved in employing an alternate disk cooling scheme. The study involved the design, construction and testing of two full scale rotating model disks, one plane and smooth for baseline testing and the second contoured to the present flight configuration, together with the corresponding plane and contoured stator disks. Local heat transfer rates are determined from the color display of encapsulated liquid crystals coated on the disk in conjunction with use of a computer vision system. The test program was composed of a wide variety of disk speeds, flowrates, and geometrical configurations, including testing for the effects of disk boltheads and gas ingestion from the gas path region radially outboard of the disk-cavity.

  10. Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    2002-01-01

    The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics

  11. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  12. Liquid metal cooled blanket concept for NET

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  13. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  14. An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

    International Nuclear Information System (INIS)

    Murphy, J.G.; Corradini, M.L.

    1997-01-01

    The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

  15. Reactor hydrodynamics during the reflood phase of a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gay, R.R.

    1977-01-01

    The thermohydraulics of a nuclear reactor during the reflood phase of a hypothetical loss-of-coolant accident can be represented by moving control volume methodology in which six control volumes are used to represent the downcomer, lower plenum, and reactor core. The one-dimensional, homogeneous, equilibrium constitutive equations for two-phase steam/water flow are solved in each control volume and connecting junctions. One of the three core control volumes represents the quench region; it changes size and position based on the axial location of the clad quench temperature and the condensed liquid level in the flow channel. The lengths of the remaining two core control volumes are determined by the position of the quench region. Simulation of actual reflood experiments demonstrates that the methodology predicts reflood-like flow oscillations and reproduces the correct trends in experimental data. The moving control volume methodology has proven itself as a valid concept for reflood hydrodynamics, but further development of the existing EFLOD code is required for simulation of actual reflood experiments

  16. Simulation of thermal phenomena expected in fuel coolant interactions in LMFBRs

    International Nuclear Information System (INIS)

    Yasin, J.

    1976-12-01

    High pressures and mechanical work may result when thermal energy is transferred from molten fuel to the coolant in a Liquid Metal Fast Breeder Reactor core meltdown accident. Two aspects of the interaction are examined in the thesis. First, the formation of high pressure pulses termed ''Vapor Explosions,'' and second, the distribution of the molten material into smaller particles, termed ''Fragmentation'', are studied. To understand the nature of the interaction simulant materials were used. Molten bismuth, molten tin and molten glass were dropped into water under various conditions. The interactions were recorded using multiflash and high speed photographing techniques. The pressure pulses were measured using transducers and the debris was examined by photographing them with an electron microscope. It was observed that vapor explosions have thresholds which depend on the material being dropped, its temperature and the bath conditions. The vapor explosions were enhanced by stratifying the bath. It was also noticed that the intensity of the vapor explosion depends on the way the molten drop fragmented in the initial stages of the interaction. The experiments with glass showed that the mode of fragmentation is important in determining when and if a vapor explosion is to be expected. The glass fragmented extensively but without any accompanying vapor explosion. The electron microscope photographs of the glass debris showed that thermal stress and surface tension phenomenon are apparently the cause of the fragmentation

  17. Development of LMR Coolant Technology - Development of a submersible-in-pool electromagnetic pump

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sang Hi; Kim, Hee Reyoung; Lee, Sang Don; Seo, Joon Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyoungki University, Suwon (Korea, Republic of)

    1997-07-15

    A submersible-in-pool type annular linear induction pumps of 60 l/min and 200 l/min, and 600 deg C has been designed with optimum geometrical and operating values found from MHD and circuit analyses reflecting the high-temperature characteristics of pump materials. Through the characteristics analyses inside the narrow flow channel of electromagnetic pump, the distribution of the time-varying flow field is calculated, and magnetic flux and force density are evaluated by end effects of linear induction electromagnetic pump and the instability analyses are carried out introducing one-dimensional linear perturbation. Testing the pump with the flow rate of 60 l/min in the suitably manufactured loop system shows a flow rate of 58 l/min at an input power of 1,377 VA with 60Hz. The design of a scaled-up pump is further taken into account LMR coolant system requiring increased capacity, and a basic analysis is carried out on the pump of 40,000 l/min for KALIMER. The present project contributes to the further design of engineering prototype electromagnetic pumps with higher capacity and to the development of liquid metal reactor with innovative simplicity. 89 refs., 8 tabs., 45 figs. (author)

  18. Boiling water reactor liquid radioactive waste processing system

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

  19. Database of thermophysical properties of liquid metal coolants for GEN-IV

    International Nuclear Information System (INIS)

    Sobolev, V.

    2011-01-01

    The report presents an Fe-Cr interatomic potential to model high-Cr ferritic alloys. The potential is fitted to thermodynamic and point-defect properties obtained from density functional theory (DFT) calculations and experiments. The developed potential is also benchmarked against other potentials available in literature. It shows particularly good agreement with the DFT obtained mixing enthalpy of the random alloy, the formation energy of intermetallics and experimental excess vibrational entropy and phase diagram. In addition, DFT calculated point-defect properties, both interstitial and substitutional, are well reproduced, as is the screw dislocation core structure. As a first validation of the potential, we study the precipitation hardening of Fe-Cr alloys via static simulations of the interaction between Cr precipitates and screw dislocations. It is concluded that the description of the dislocation core modification near a precipitate might have a significant influence on the interaction mechanisms observed in dynamic simulations.

  20. Experimental study of hydrodynamics of target system construction with liquid metallic coolant on water models

    International Nuclear Information System (INIS)

    Beznosov, A.; Davidov, D.; Khokhlov, D.

    2001-01-01

    This paper is about experimental study of hydrodynamics of target system construction with a spherical beam window (membrane). The visualization of current and current velocity fields in energy release area was considered. The findings are used to optimize the geometrical characteristics of target system. Also report contains the information about functioning accelerator-driven system, description it principle circuit and description of basic requirements what the target design should satisfy. (authors)

  1. An analysis of electron beam welds in a dual coolant liquid metal breeder blanket

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.; Kernforschungszentrum Karlsruhe GmbH

    1994-10-01

    Numerical simulation of electron beam welding of blanket segments was performed using non-linear finite element code ABAQUS. The thermal and stress fields were assumed uncoupled, while preserving the temperature dependency of all material parameters. The martensite-austenite and austenite-martensite transformations were taken into account through volume shrinking/expansion effects, which is consistent with available data. The distributions of post welding residual stress in a complex geometry of the first wall are obtained. Also, the effects of preheating and post-welding heat treatment were addressed. Time dependent temperature and stress-strain fields obtained provide good insight into the welding process. They may be used directly to support reliability and life-time studies of blanket structures. On the other hand, they provide useful hints about the feasibility of the geometrical configurations as proposed by different design concepts. (orig.) [de

  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. Corrosion of Structural Materials in Liquid Metals Used as Fast Reactor Coolants

    International Nuclear Information System (INIS)

    Balbaud-Célérier, F.; Courouau, J.L.; Martinelli, L.

    2013-01-01

    Conclusions: • Thermodynamic data give the stable state of the system, the compounds susceptible to form but no information on the kinetics of the process; • Need to perform corrosion tests in controlled conditions of temperature, chemistry, hydrodynamics; • Comparison of the materials behaviour: first selection of materials, optimisation of the composition; • Fundamental work on the understanding of the corrosion process to develop corrosion models and predictive laws to guarantee the long term behaviour

  4. 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

  5. Condensing heat transfer following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Krotiuk, W.J.; Rubin, M.B.

    1978-01-01

    A new method for calculating the steam mass condensation energy removal rates on cold surfaces in contact with an air-steam mixture has been developed. This method is based on the principles of mass diffusion of steam from an area of high concentration to the condensing surface, which is an area of low steam concentration. This new method of calculating mass condensation has been programmed into the CONTEMPT-LT Mod 26 computer code, which calculates the pressure and temperature transients inside a light water reactor containment following a loss-of-coolant accident. The condensing heat transfer coefficient predicted by the mass diffusion method is compared to existing semi-empirical correlations and to the experimental results of the Carolinas Virginia Tube Reactor Containment natural decay test. Closer agreement with test results is shown in the calculation of containment pressure, temperature, and heat sink surface temperature using the mass diffusion condensation method than when using any existing semi-empirical correlation

  6. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    International Nuclear Information System (INIS)

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs

  7. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  8. Reactor coolant pump shaft seal behavior during station blackout

    International Nuclear Information System (INIS)

    Kittmer, C.A.; Wensel, R.G.; Rhodes, D.B.; Metcalfe, R.; Cotnam, B.M.; Gentili, H.; Mings, W.J.

    1985-04-01

    A testing program designed to provide fundamental information pertaining to the behavior of reactor coolant pump (RCP) shaft seals during a postulated nuclear power plant station blackout has been completed. One seal assembly, utilizing both hydrodynamic and hydrostatic types of seals, was modeled and tested. Extrusion tests were conducted to determine if seal materials could withstand predicted temperatures and pressures. A taper-face seal model was tested for seal stability under conditions when leaking water flashes to steam across the seal face. Test information was then used as the basis for a station blackout analysis. Test results indicate a potential problem with an elastomer material used for O-rings by a pump vendor; that vendor is considering a change in material specification. Test results also indicate a need for further research on the generic issue of RCP seal integrity and its possible consideration for designation as an unresolved safety issue

  9. Fatigue check of nuclear safety class 1 reactor coolant pipe

    International Nuclear Information System (INIS)

    Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong

    2015-01-01

    Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)

  10. Fuel -coolant interactions in LWRs and LMFBRs: relationships and distinctions

    Energy Technology Data Exchange (ETDEWEB)

    Duffey, R B; Lellouche, G S [Nuclear Safety and Analysis Department, Electric Power Research Institute, Palo Alto, CA (United States)

    1979-10-15

    The question of fuel-coolant interaction and of potential vapor explosion is raised here. lt is the contention of the authors that there is in fact no need to study this question vis a vis Light Water Reactors (LWR) except from an academic point of view since it does not impact on safety considerations. As for LMFBRs, the design basis whole core accidents for LWRs are derived from the fundamental concern of maintaining core geometry to provide for convective cooling. However, the important distinction is that the core is in its most reactive configuration, and core and fuel rearrangement is therefore not of such concern. The author's thesis is that even if the probability of steam explosion following core melt were two orders of magnitude greater than currently assumed (10{sup -2}) the total LWR risk would increase only by a factor of 2-6 for BWRs and less a factor of 10 for PWRs

  11. Small break LOCA [loss of coolant accident] mitigation for Bellefonte

    International Nuclear Information System (INIS)

    Bayless, P.D.; Dobbe, C.A.

    1986-01-01

    Several 5-cm (2-in.) diameter cold leg break loss coolant accidents for the Bellefonte nuclear plant were analyzed as part of the Severe Accident Sequence Analysis Program. The transients assumed various system failures, and included the S 2 D sequence. Operator actions to mitigate the S 2 D transient were also investigated. The transients were analyzed until either core damage began or long-term decay heat removal was established. The S 2 D sequence was analyzed into the core damage phase of the transient. The analyses showed that the flow from one high pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were also able to prevent core damage for the S 2 D sequence

  12. Fusion Blanket Coolant Section Criteria, Methodology, and Results

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-02

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

  13. On-line monitoring of main coolant pump seals

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; Glass, S.W.; Sommerfield, G.A.; Harrison, D.

    1984-06-01

    The Babcock and Wilcox Company has developed and implemented a Reactor Coolant Pump Monitoring and Diagnostic System (RCPM and DS). The system has been installed at Toledo Edison Company's Davis-Besse Nuclear Power Station Unit 1. The RCPM and PS continuously monitors a number of indicators of pump performance and notifies the plant operator of out-of-tolerance conditions or pump performance trending toward out-of-tolerance conditions. Pump seal parameters being monitored include pump internal pressures, temperatures, and flow rates. Rotordynamic performanvce and plant operating conditions are also measured with a variety of dynamic sensors. This paper describes the implementation of the system and the results of on-line monitoring of four RC pumps

  14. Environmental radiological consequences of a loss of coolant accident

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1981-01-01

    The elaboration of a calculation model to determine safety areas, named Exclusion Zone and Low Population Zone for nuclear power plants, is dealt with. These areas are determined from a radioactive doses calculation for the population living around the NPP after occurence of a postulated ' Maximum Credible Accident' (MCA). The MCA is defined as an accident with complete loss of primary coolant and consequent fusion of a substantial portion of the reactor core. In the calculations carried out, data from NPP Angra I were used and the assumptions made were conservative, to be compatible with licensing requirements. Under the most pessimistic assumption (no filters) the values of 410m and 1000m were obtained for the Exclusion Zone and Low Population Zone radii, respectivily. (Author) [pt

  15. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

  16. Fatigue cycles evaluation of 500 MWe PHWR coolant channel sealdisc

    International Nuclear Information System (INIS)

    Chawla, D.S.; Vaze, K.K.; Kushwaha, H.S.; Gupta, K.S.; Bhambra, H.S.

    1998-07-01

    At each end of coolant channel there is one sealing plug assembly. The sealdisc is a part of sealing plug assembly. The sealdisc is used to avoid leakage of heavy water. The importance of sealdisc can be understood by the fact that there are 784 sealdiscs in one 500 MWe PHWR unit. During the life time of reactor the sealdisc will be subjected to cyclic loads due to reactor startup, shutdown, power setback and also due to refuelling operations. Excessive reversal of stresses may lead to fatigue failure. The sealdisc failure may cause loss of coolant accidents. Since sealdisc is safety class 1 component, it has to be qualified according to ASME Section III Division 1 NB. For cyclic loads, the fatigue analysis is essential to assess the allowable number of cycles and also to check the total usage factor due to different cyclic loads. To evaluate the allowable fatigue cycles, the analysis is carried out using finite element method. The present report deals with the fatigue cycles evaluation of 500 MWe PHWR sealdisc. The finite element model having eight noded axisymmetric elements is used for the analysis. The various loads considered in the analysis are mechanical loads arising due to refuelling operations and number of temperature-pressure transients. During refuelling, the sealdisc is removed and reinstalled back by use of fuelling machine ram which applies load at centre as well as at rocker point of sealdisc. The stress analysis is carried out for each stage of loading during refuelling and fatigue cycles are evaluated. For temperature transient, decoupled thermal analysis is carried out. At various instants of time, the stresses are computed using temperatures calculated in thermal analysis. The pressure variation is also considered along with temperature variation. The fatigue cycles are evaluated for each transient using maximum alternating stress intensities. The usage factors are calculated for various temperature/pressure transients and refuelling loads

  17. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D P; Lukas, M; Lynch, B K [Spectro Incorporated, Littleton, MA (United States)

    1998-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  18. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.P.; Lukas, M.; Lynch, B.K. [Spectro Incorporated, Littleton, MA (United States)

    1997-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  20. Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kabakchi, S.A.

    1995-01-01

    Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

  1. 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.

  2. 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.

  3. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  4. Operation diagnostics of the reactor coolant pumps in the Jaslovske Bohunice nuclear power plant, CSSR

    International Nuclear Information System (INIS)

    Bahna, J.; Jaros, I.; Oksa, G.

    1990-01-01

    The state of the art of the materials basis, the diagnostics methods used, organization of data collection and processing, and some results of routine and specific investigations concerned with diagnosis of the reactor coolant pump in the Jaslovske Bohunice NPP V-1 are presented. Some information is given about the reactor coolant pump monitor developed in the VUJE. (author)

  5. Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

    International Nuclear Information System (INIS)

    Hudcovsky, S.; Slanina, M.; Badiar, S.

    2001-01-01

    The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

  6. Liquid fuel concept benefits

    International Nuclear Information System (INIS)

    Hron, M.

    1996-01-01

    There are principle drawbacks of any kind of solid nuclear fuel listed and analyzed in the first part of the paper. One of the primary results of the analyses performed shows that the solid fuel concept, which was to certain degree advantageous in the first periods of a nuclear reactor development and operation, has guided this branch of a utilization of atomic nucleus energy to a death end. On the background of this, the liquid fuel concept and its benefits are introduced and briefly described in the first part of the paper, too. As one of the first realistic attempts to utilize the advantages of liquid fuels, the reactor/blanket system with molten fluoride salts in the role of fuel and coolant simultaneously, as incorporated in the accelerator-driven transmutation technology (ADTT) being proposed and currently having been under development in the Los Alamos National Laboratory, will be studied both theoretically and experimentally. There is a preliminary design concept of an experimental assembly LA-O briefly introduced in the paper which is under preparation in the Czech Republic for such a project. Finally, there will be another very promising concept of a small low power ADTT system introduced which is characterized by a high level of safety and economical efficiency. In the conclusion, the overall survey of principal benefits which may be expected by introducing liquid nuclear fuel in nuclear power and research reactor systems is given and critically analyzed. 7 refs, 4 figs

  7. 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)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  9. Radioactive corrosion products in circuit of fast reactor loop with dissociating coolant

    International Nuclear Information System (INIS)

    Dolgov, V.M.; Katanaev, A.O.

    1982-01-01

    The results of experimental investigation into depositions of radionuclides of corrosion origin on the surfaces of a reactor-in-pile loop facility with a dissociating coolant are presented. It is stated that the ratio of radionuclides in fixed depositions linearly decreases with decrease of the coolant temperature at the core-condenser section. The element composition of non-fixed compositions quantitatively and qualitatively differs from the composition of structural material, and it is more vividly displayed for the core-condenser section. The main mechanism of circuit contamination with radioactive corrosion products is substantiated: material corrosion in the zones of coolant phase transfer, their remove by the coolant in the core, deposition, activation and wash-out by the coolant from the core surfaces

  10. Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

    International Nuclear Information System (INIS)

    Ruger, C.J.; Luckas, W.J. Jr.

    1989-03-01

    Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

  11. 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

  12. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-02-01

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17LI is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Ph-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure.

  13. Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

    International Nuclear Information System (INIS)

    Hong, S. W.; Na, Y. S.; Hong, S. H.; Song, J. H.

    2014-01-01

    A molten material is injected into an interaction chamber by free gravitation fall. This type of fuel coolant interaction could happen to operating plants. However, the flooding of a reactor cavity is considered as SAM measures for new PWRs such as APR-1400 and AP1000 to assure the IVR of a core melt. In this case, a molten corium in a reactor is directly injected into water surrounding the reactor vessel without a free fall. KAERI has carried out fuel coolant interaction tests without a free fall using ZrO 2 and corium to simulate the reactor submerged conditions. There are four phases in a steam explosion. The first phase is a premixing phase. The premixing is described in the literature as follows: during penetration of melt into water, hydrodynamic instabilities, generated by the velocities and density differences as well as vapor production, induce fragmentation of the melt into particles; the particles fragment in turn into smaller particles until they reach a critical size such that the cohesive forces (surface tension) balance exactly the disruptive forces (inertial); and the molten core material temperature (>2500 K) is such that the mixing always occurs in the film boiling regime of the water: It is very important to qualify and quantify this phase because it gives the initial conditions for a steam explosion This paper mainly focuses on the observation of the premixing phase between a case with 1 m free fall and a case without a free fall to simulate submerged reactor condition. The premixing behavior between a 1m free fall case and reactor case submerged without a free fall is observed experimentally. The average velocity of the melt front passing through 1m water pool; - Case without a free fall: The average velocity of corium, 2.7m/s, is faster than ZrO 2 , 2.3m/s, in water. - Cases of with a 1 m free fall and without a free fall : The case without a free fall is about two times faster than a case with a 1 m free fall. Bubble characteristics; - Case

  14. Verification Test of Hydraulic Performance for Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Jun; Kim, Jae Shin; Ryu, In Wan; Ko, Bok Seong; Song, Keun Myung [Samjin Ind. Co., Seoul (Korea, Republic of)

    2010-01-15

    According to this project, basic design for prototype pump and model pump of reactor coolant pump and test facilities has been completed. Basic design for prototype pump to establish structure, dimension and hydraulic performance has been completed and through primary flow analysis by computational fluid dynamics(CFD), flow characteristics and hydraulic performance have been established. This pump was designed with mixed flow pump having the following design requirements; specific velocity(Ns); 1080.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 3115m{sup 3}/h, total head ; 26.3m, pump speed; 1710rpm, pump efficiency; 77.0%, Impeller out-diameter; 349mm, motor output; 360kw, design pressure; 17MPaG. The features of the pump are leakage free due to no mechanical seal on the pump shaft which insures reactor's safety and law noise level and low vibration due to no cooling fan on the motor which makes eco-friendly product. Model pump size was reduced to 44% of prototype pump for the verification test for hydraulic performance of reactor coolant pump and was designed with mixed flow pump and canned motor having the following design requirements; specific speed(NS); 1060.9(rpm{center_dot}m{sup 3}/m{center_dot}m), capacity; 539.4m{sup 3}/h, total head; 21.0m, pump speed; 3476rpm, pump efficiency; 72.9%, Impeller out-diameter; 154mm, motor output; 55kw, design pressure; 1.0MPaG. The test facilities were designed for verification test of hydraulic performance suitable for pump performance test, homologous test, NPSH test(cavitation), cost down test and pressure pulsation test of inlet and outlet ports. Test tank was designed with testing capacity enabling up to 2000m{sup 3}/h and design pressure 1.0MPaG. Auxiliary pump was designed with centrifugal pump having capacity; 1100m{sup 3}/h, total head; 42.0m, motor output; 190kw

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  16. Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

    2017-02-15

    This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

  17. 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

  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. 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. 

  20. Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

    International Nuclear Information System (INIS)

    Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

    1994-01-01

    Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab