WorldWideScience

Sample records for hot-spot-controlled critical heat

  1. Critical heat flux evaluation

    International Nuclear Information System (INIS)

    Banner, D.

    1995-01-01

    Critical heat flux (CHF) is of importance for nuclear safety and represents the major limiting factors for reactor cores. Critical heat flux is caused by a sharp reduction in the heat transfer coefficient located at the outer surface of fuel rods. Safety requires that this phenomenon also called the boiling crisis should be precluded under nominal or incidental conditions (Class I and II events). CHF evaluation in reactor cores is basically a two-step approach. Fuel assemblies are first tested in experimental loops in order to determine CHF limits under various flow conditions. Then, core thermal-hydraulic calculations are performed for safety evaluation. The paper will go into more details about the boiling crisis in order to pinpoint complexity and lack of fundamental understanding in many areas. Experimental test sections needed to collect data over wide thermal-hydraulic and geometric ranges are described CHF safety margin evaluation in reactors cores is discussed by presenting how uncertainties are mentioned. From basic considerations to current concerns, the following topics are discussed; knowledge of the boiling crisis, CHF predictors, and advances thermal-hydraulic codes. (authors). 15 refs., 4 figs

  2. Data bank of critical heat flux

    International Nuclear Information System (INIS)

    Balino, J.L.; Ruival, M.H.

    1985-01-01

    More than 13.000 measurements of critical heat flux are classified in a data bank. From each experiment the following information can be obtained: cooling medium (light water, freon 12 or freon 21), geometry of the test section and thermalhydraulic parameters. The data management is performed by a computer program called CHFTRAT. A brief study of the influence of different parameters in the critical heat flux is presented, as an example of how to use the program. (M.E.L.) [es

  3. Furan: A critical heat induced dietary contaminant

    DEFF Research Database (Denmark)

    Mariotti, María S.; Granby, Kit; Rozowski, Jaime

    2013-01-01

    The presence of furan in a broad range of heat processed foods (0-6000 μg kg-1) has received considerable attention due to the fact that this heat induced contaminant is considered as a "possible carcinogenic compound to humans". Since a genotoxic mode of action could be associated with furan...... of some critical factors such as heating conditions, pH and matrix microstructure are discussed in order to propose some potential methodologies for furan mitigation in a wide range of heated foods. © 2013 The Royal Society of Chemistry....

  4. Diameter effect on critical heat flux

    International Nuclear Information System (INIS)

    Tanase, A.; Cheng, S.C.; Groeneveld, D.C.; Shan, J.Q.

    2009-01-01

    The critical heat flux look-up table (CHF LUT) is widely used to predict CHF for various applications, including design and safety analysis of nuclear reactors. Using the CHF LUT for round tubes having inside diameters different from the reference 8 mm involves conversion of CHF to 8 mm. Different authors [Becker, K.M., 1965. An Analytical and Experimental Study of Burnout Conditions in Vertical Round Ducts, Aktiebolaget Atomenergie Report AE 177, Sweden; Boltenko, E.A., et al., 1989. Effect of tube diameter on CHF at various two phase flow regimes, Report IPE-1989; Biasi, L., Clerici, G.C., Garriba, S., Sala, R., Tozzi, A., 1967. Studies on Burnout, Part 3, Energia Nucleare, vol. 14, pp. 530-536; Groeneveld, D.C., Cheng, S.C., Doan, T., 1986. AECL-UO critical heat flux look-up table. Heat Transfer Eng., 7, 46-62; Groeneveld et al., 1996; Hall, D.D., Mudawar, I., 2000. Critical heat flux for water flow in tubes - II subcooled CHF correlations. Int. J. Heat Mass Transfer, 43, 2605-2640; Wong, W.C., 1996. Effect of tube diameter on critical heat flux, MaSC dissertation, Ottawa Carleton Institute for Mechanical and Aeronautical Engineering, University of Ottawa] have proposed several types of correlations or factors to describe the diameter effect on CHF. The present work describes the derivation of new diameter correction factor and compares it with several existing prediction methods

  5. Consideration of critical heat flux margin prediction by subcooled or low quality critical heat flux correlations

    International Nuclear Information System (INIS)

    Hejzlar, P.; Todreas, N.E.

    1996-01-01

    The accurate prediction of the critical heat flux (CHF) margin which is a key design parameter in a variety of cooling and heating systems is of high importance. These margins are, for the low quality region, typically expressed in terms of critical heat flux ratios using the direct substitution method. Using a simple example of a heated tube, it is shown that CHF correlations of a certain type often used to predict CHF margins, expressed in this manner, may yield different results, strongly dependent on the correlation in use. It is argued that the application of the heat balance method to such correlations, which leads to expressing the CHF margins in terms of the critical power ratio, may be more appropriate. (orig.)

  6. Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube

    International Nuclear Information System (INIS)

    Boscary, J.; Association Euratom-CEA, Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance

    1997-03-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author)

  7. Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow

    International Nuclear Information System (INIS)

    Boscary, J.

    1995-10-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs

  8. Critical heat flux, post dry-out and their augmentation

    International Nuclear Information System (INIS)

    Celata, G.P.; Mariani, A.

    1999-01-01

    The report shows the state of art review on the critical heat flux and the post-dryout heat transfer. The work, which is a merge of original researches carried out at the Institute of Thermal Fluid Dynamic of ENEA (National Agency for New Technology, Energy and the Environment) and a thorough review of the recent literature, is divided in four chapters: critical heat flux in subcooled flow boiling; critical heat flux in saturated flow boiling; post-dryout heat transfer; enhancement of critical heat flux and post-dryout heat transfer [it

  9. Prediction of critical heat flux using ANFIS

    Energy Technology Data Exchange (ETDEWEB)

    Zaferanlouei, Salman, E-mail: zaferanlouei@gmail.co [Nuclear Engineering and Physics Department, Faculty of Nuclear Engineering, Center of Excellence in Nuclear Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, Tehran (Iran, Islamic Republic of); Rostamifard, Dariush; Setayeshi, Saeed [Nuclear Engineering and Physics Department, Faculty of Nuclear Engineering, Center of Excellence in Nuclear Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, Tehran (Iran, Islamic Republic of)

    2010-06-15

    The prediction of Critical Heat Flux (CHF) is essential for water cooled nuclear reactors since it is an important parameter for the economic efficiency and safety of nuclear power plants. Therefore, in this study using Adaptive Neuro-Fuzzy Inference System (ANFIS), a new flexible tool is developed to predict CHF. The process of training and testing in this model is done by using a set of available published field data. The CHF values predicted by the ANFIS model are acceptable compared with the other prediction methods. We improve the ANN model that is proposed by to avoid overfitting. The obtained new ANN test errors are compared with ANFIS model test errors, subsequently. It is found that the ANFIS model with root mean square (RMS) test errors of 4.79%, 5.04% and 11.39%, in fixed inlet conditions and local conditions and fixed outlet conditions, respectively, has superior performance in predicting the CHF than the test error obtained from MLP Neural Network in fixed inlet and outlet conditions, however, ANFIS also has acceptable result to predict CHF in fixed local conditions.

  10. Prediction of critical heat flux using ANFIS

    International Nuclear Information System (INIS)

    Zaferanlouei, Salman; Rostamifard, Dariush; Setayeshi, Saeed

    2010-01-01

    The prediction of Critical Heat Flux (CHF) is essential for water cooled nuclear reactors since it is an important parameter for the economic efficiency and safety of nuclear power plants. Therefore, in this study using Adaptive Neuro-Fuzzy Inference System (ANFIS), a new flexible tool is developed to predict CHF. The process of training and testing in this model is done by using a set of available published field data. The CHF values predicted by the ANFIS model are acceptable compared with the other prediction methods. We improve the ANN model that is proposed by to avoid overfitting. The obtained new ANN test errors are compared with ANFIS model test errors, subsequently. It is found that the ANFIS model with root mean square (RMS) test errors of 4.79%, 5.04% and 11.39%, in fixed inlet conditions and local conditions and fixed outlet conditions, respectively, has superior performance in predicting the CHF than the test error obtained from MLP Neural Network in fixed inlet and outlet conditions, however, ANFIS also has acceptable result to predict CHF in fixed local conditions.

  11. Critical heat flux, post dry-out and their augmentation

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G.P.; Mariani, A. [ENEA, Centro Ricerche Casaccia, S. Maria di Galeria, RM (Italy). Dipt. Energia

    1999-07-01

    The report shows the state of art review on the critical heat flux and the post-dryout heat transfer. The work, which is a merge of original researches carried out at the Institute of Thermal Fluid Dynamic of ENEA (National Agency for New Technology, Energy and the Environment) and a thorough review of the recent literature, is divided in four chapters: critical heat flux in subcooled flow boiling; critical heat flux in saturated flow boiling; post-dryout heat transfer; enhancement of critical heat flux and post-dryout heat transfer. [Italian] Si passa in rassegna lo stato dell'arte sulla crisi termica e sullo scambio termico post-crisi, che compendia studi tradizionali condotti dall'ENEA. Il rapporto e' suddiviso in quattro parti: crisi termica in ebollizione sottoraffreddata; crisi termica in ebollizione satura; scambio termico dopo la crisi termica; incremento del flusso termico critico e dello scambio termico post-crisi.

  12. Critical heat flux, post dry-out and their augmentation

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G P; Mariani, A [ENEA, Centro Ricerche Casaccia, S. Maria di Galeria, RM (Italy). Dipt. Energia

    1999-07-01

    The report shows the state of art review on the critical heat flux and the post-dryout heat transfer. The work, which is a merge of original researches carried out at the Institute of Thermal Fluid Dynamic of ENEA (National Agency for New Technology, Energy and the Environment) and a thorough review of the recent literature, is divided in four chapters: critical heat flux in subcooled flow boiling; critical heat flux in saturated flow boiling; post-dryout heat transfer; enhancement of critical heat flux and post-dryout heat transfer. [Italian] Si passa in rassegna lo stato dell'arte sulla crisi termica e sullo scambio termico post-crisi, che compendia studi tradizionali condotti dall'ENEA. Il rapporto e' suddiviso in quattro parti: crisi termica in ebollizione sottoraffreddata; crisi termica in ebollizione satura; scambio termico dopo la crisi termica; incremento del flusso termico critico e dello scambio termico post-crisi.

  13. Heat transfer critical conditions in two-plase flow

    International Nuclear Information System (INIS)

    Assis, M.C.V. de.

    1980-02-01

    The critical heat flux for forced-convection flow of water inside an uniformly heated circular channel is analysed, taking into account several flow patterns usually met in this type of investigation. Comments about nomenclature, experimental methods and influence of operational parameters used in the description of this phenomenon are made. The experimental results from 187 tests of critical heat flux at low pressure are presented. One empirical correlation between the critical heat flux and the independent parameters, was developed. Some correlations developed in other laboratories in the same range of parameters are mentioned and compared with present one. (Author) [pt

  14. Measurement of Critical Heat Flux Using the Transient Inverse Heat Conduction Method in Spray cooling

    International Nuclear Information System (INIS)

    Kim, Yeung Chan

    2016-01-01

    A study on the measurement of critical heat flux using the transient inverse heat conduction method in spray cooling was performed. The inverse heat conduction method estimates the surface heat flux or temperature using a measured interior temperature history. The effects of the measuring time interval and location of temperature measurement on the measurement of critical heat flux were primarily investigated. The following results were obtained. The estimated critical heat flux decreased as the time interval of temperature measurement increased. Meanwhile, the effect of measurement location on critical heat flux was not significant. It was also found, from the experimental results, that the critical superheat increased as the measurement location of thermocouple neared the heat transfer surface.

  15. Measurement of Critical Heat Flux Using the Transient Inverse Heat Conduction Method in Spray cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeung Chan [Andong Nat’l Univ., Andong (Korea, Republic of)

    2016-10-15

    A study on the measurement of critical heat flux using the transient inverse heat conduction method in spray cooling was performed. The inverse heat conduction method estimates the surface heat flux or temperature using a measured interior temperature history. The effects of the measuring time interval and location of temperature measurement on the measurement of critical heat flux were primarily investigated. The following results were obtained. The estimated critical heat flux decreased as the time interval of temperature measurement increased. Meanwhile, the effect of measurement location on critical heat flux was not significant. It was also found, from the experimental results, that the critical superheat increased as the measurement location of thermocouple neared the heat transfer surface.

  16. Steady state and transient critical heat flux examinations

    International Nuclear Information System (INIS)

    Szabados, L.

    1978-02-01

    In steady state conditions within the P.W.R. parameter range the critical heat flux correlations based on local parameters reproduce the experimental data with less deviations than those based on system parameters. The transient experiments were restricted for the case of power transients. A data processing method for critical heat flux measurements has been developed and the applicability of quasi steady state calculation has been verified. (D.P.)

  17. Critical heat flux in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2015-01-01

    This Brief concerns the important problem of critical heat flux in flow boiling in microchannels. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to “Heat Transfer and Pressure Drop in Flow Boiling in Microchannels,” by the same author team, this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.

  18. Predicting critical heat flux in slug flow regime of uniformly heated ...

    African Journals Online (AJOL)

    Numerical computation code (PWR-DNBP) has been developed to predict Critical Heat Flux (CHF) of forced convective flow of water in a vertical heated channel. The code was based on the liquid sub-layer model, with the assumption that CHF occurred when the liquid film thickness between the heated surface and vapour ...

  19. Work and power fluctuations in a critical heat engine

    Science.gov (United States)

    Holubec, Viktor; Ryabov, Artem

    2017-09-01

    We investigate fluctuations of output work for a class of Stirling heat engines with working fluid composed of interacting units and compare these fluctuations to an average work output. In particular, we focus on engine performance close to a critical point where Carnot's efficiency may be attained at a finite power as reported by M. Campisi and R. Fazio [Nat. Commun. 7, 11895 (2016), 10.1038/ncomms11895]. We show that the variance of work output per cycle scales with the same critical exponent as the heat capacity of the working fluid. As a consequence, the relative work fluctuation diverges unless the output work obeys a rather strict scaling condition, which would be very hard to fulfill in practice. Even under this condition, the fluctuations of work and power do not vanish in the infinite system size limit. Large fluctuations of output work thus constitute inseparable and dominant element in performance of the macroscopic heat engines close to a critical point.

  20. Work and power fluctuations in a critical heat engine.

    Science.gov (United States)

    Holubec, Viktor; Ryabov, Artem

    2017-09-01

    We investigate fluctuations of output work for a class of Stirling heat engines with working fluid composed of interacting units and compare these fluctuations to an average work output. In particular, we focus on engine performance close to a critical point where Carnot's efficiency may be attained at a finite power as reported by M. Campisi and R. Fazio [Nat. Commun. 7, 11895 (2016)2041-172310.1038/ncomms11895]. We show that the variance of work output per cycle scales with the same critical exponent as the heat capacity of the working fluid. As a consequence, the relative work fluctuation diverges unless the output work obeys a rather strict scaling condition, which would be very hard to fulfill in practice. Even under this condition, the fluctuations of work and power do not vanish in the infinite system size limit. Large fluctuations of output work thus constitute inseparable and dominant element in performance of the macroscopic heat engines close to a critical point.

  1. Economic analysis of electric heating based on critical electricity price

    Science.gov (United States)

    Xie, Feng; Sun, Zhijie; Zhou, Xinnan; Fu, Chengran; Yang, Jie

    2018-06-01

    The State Grid Corporation of China proposes an alternative energy strategy, which will make electric heating an important task in the field of residential electricity consumption. This article takes this as the background, has made the detailed introduction to the inhabitant electric heating technology, and take the Zhangjiakou electric panels heating technology as an example, from the expense angle, has carried on the analysis to the electric panels heating economy. In the field of residential heating, electric panels operating costs less than gas boilers. After customers implying energy-saving behavior, electric panels operating cost is even lower than coal-fired boilers. The critical price is higher than the execution price, which indicates that the economic performance of the electric panels is significantly higher than that of the coal boiler.

  2. Thermalhydraulic behavior of electrically heated rods during critical heat flux transients

    International Nuclear Information System (INIS)

    Lima, Rita de Cassia Fernandes de

    1997-01-01

    In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed utilizing heated rods to simulate operational and accidental conditions of nuclear fuel rods, with special attention to the phenomenon of boiling crisis. The use of mechanisms which detect the abrupt temperature rise allows the electric power switch off. These facts prevent the test section from damage. During the critical heat flux phenomenon the axial heat conduction becomes very important. The study of the dryout and rewetting fronts yields the analysis, planning and following of critical heat flux experiments. These facts are important during the reflooding of nuclear cores at severe accidents. In the present work it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step or power slope from steady state condition. After the occurrence of critical heat flux, it is predicted the drying front propagation. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. In all these transients the coolant pressure is 13,5 MPa. For one of them, comparisons are done with a pressure of 8,00 MPa. Mass flow and enthalpy influences on the fronts velocities are also analysed. These results show that mass flow has more importance on the drying front velocities whereas the pressure alters strongly the rewetting ones. (author)

  3. Critical heat flux determination in an annulus section

    International Nuclear Information System (INIS)

    Reyes C, C.A.

    1997-01-01

    The present report explains the phenomenon of Critical heat flux. The study of this physical phenomenon is carried out during the boiling of a liquid and is of supreme importance for the calculation and operation of a nuclear reactor even in the moderns generators of steam (thermoelectric and nucleoelectrics), industrial cooling and in all those industrial process that use a liquid subject to sources of heating and to conditions of work excessively high (temperatures and pressures) so that stay in operation in an appropriate manner and sure. Once well-known this value, the equipment used in these process works with a maximum heat that is smaller than the Critical Heat Flux. The study of the Critical Heat Flux has achieved important advances in the last years, mainly for the enormous obligation that in this moment involved the safety to world level, this has forced to researchers and designers of this type of equipment to center their attention in the obtaining of a correlation which of general way explains it. In this reports two correlations will be compared that they contribute to the evaluation of the Critical Heat Flux in annulus and that they try to be generals in this type of geometry, the Shah correlation's and the Katto correlation's. The same as most of the correlations, these have been calculated so that the fluid of work is water, although they have also been proven with others fluids. The results obtained in this report only will show the degree of advance which the investigation of this phenomenon has achieved in annulus and to low amounts of flow of liquid, like which they are in the Experimental Heat Transfer Circuit located in the Department of Physics of the National Institute of Nuclear Research. (Author)

  4. Evaluation of Criticality of Self-Heating of Polymer Composites by Estimating the Heat Dissipation Rate

    Science.gov (United States)

    Katunin, A.

    2018-03-01

    The critical self-heating temperature at which the structural degradation of polymer composites under cyclic loading begins is evaluated by analyzing the heat dissipation rate. The method proposed is an effective tool for evaluating the degradation degree of such structures.

  5. Thermalhydraulic behavior of electrically heated rod during a critical heat flux transient

    International Nuclear Information System (INIS)

    Lima, Rita de Cassia Fernandes de; Carajilescov, Pedro

    1997-01-01

    In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, a theoretical analysis of the drying and rewetting front propagation is performed during a critical heat flux experiment, starting with the application of slope of electrical power from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies done with several values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one. (author)

  6. Critical heat flux experiments in tight lattice core

    Energy Technology Data Exchange (ETDEWEB)

    Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  7. Critical heat flux experiments in tight lattice core

    International Nuclear Information System (INIS)

    Kureta, Masatoshi

    2002-01-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  8. Critical heat flux in subcooled and low quality boiling

    International Nuclear Information System (INIS)

    Maroti, L.

    1976-06-01

    A semi-empirical relationship for critical heat flux prediction in a light water cooled power reactor core is developed. The method of developing this relationship is the extension of the analysis of pool boiling crisis for forced convective boiling. In the calculations the energy conservation equation is used together with additional condition for the crisis. Assuming that in the vicinity of the crisis the heat is transported only by the latent heat of the vapour this condition for the crisis can be characterized by the maximum departure velocity of the vapour. Because only flow boiling crisis associating with bubbling at the heating surface is considered the model could be applied only for low quality boiling crisis. The calculated results are compared to the available experimental ones. (Sz.N.Z.)

  9. A comparison of critical heat flux in tubes and bilaterally heated annuli

    Energy Technology Data Exchange (ETDEWEB)

    Doerffer, S.; Groeneveld, D.C.; Cheng, S.C. [Univ. of Ottawa (Canada)

    1995-09-01

    This paper examines the critical heat flux (CHF) behaviour for annular flow in bilaterally heated annuli and compares it to that in tubes and unilaterally heated annuli. It was found that the differences in CHF between bilaterally and unilaterally heated annuli or tubes strongly depend on pressure and quality. the CHF in bilaterally heated annuli can be predicted by tube CHF prediction methods for the simultaneous CHF occurrence at both surfaces, and the following flow conditions: pressure 7-10 MPa, mass flux 0.5-4.0 Mg/m{sup 2}s and critical quality 0.23-0.9. The effect on CHF of the outer-to-inner surface heat flux ratio, was also examined. The prediction of CHF for bilaterally heated annuli was based on the droplet-diffusion model proposed by Kirillov and Smogalev. While their model refers only to CHF occurrence at the inner surface, we extended it to cases where CHF occurs at the outer surface, and simultaneously at both surfaces, thus covering all cases of CHF occurrence in bilaterally heated annuli. From the annuli CHF data of Becker and Letzter, we derived empirical functions required by the model. the proposed equations provide good accuracy for the CHF data used in this study. Moreover, the equations can predict conditions at which CHF occurs simultaneously at both surfaces. Also, this method can be used for cases with only one heated surface.

  10. POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

    Directory of Open Access Journals (Sweden)

    Vojtěch Caha

    2016-12-01

    Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

  11. Further comparisons of critical heat flux correlations for vertical tubes

    International Nuclear Information System (INIS)

    Govan, A.H.

    1986-11-01

    An earlier report by Govan (1984, AERE-R11298), described a data-bank of critical heat flux measurements in vertical upflow in tubes, and compared the predictions of the Harwell Annular Flow Model with two previously reported correlations. In this report two further correlations, those of Biasi [1967, Studies on burnout, Part 3] and Zuber [1961, Int. Devel. Heat Transfer, Part 2, PB230-236]/ Griffith,[1977, Nucl. Safety vol 18, no3] have been tested. These two correlations are used extensively in reactor design. Overall comparisons are given between all the correlations tested so far. (author)

  12. Effect of heated length on the Critical Heat Flux of subcooled flow boiling. 2. Effective heated length under axially nonuniform heating condition

    International Nuclear Information System (INIS)

    Kinoshita, Hidetaka; Yoshida, Takuya; Nariai, Hideki; Inasaka, Fujio

    1998-01-01

    Effect of heated length on the Critical Heat Flux (CHF) of subcooled flow boiling with water was experimentally investigated by using direct current heated tube made of stainless steel a part of whose wall thickness was axially cut for realizing nonuniform heat flux condition. The higher enhancement of the CHF was derived for shorter tube length. The effective heated length was determined for the tube under axially nonuniform heat flux condition. When the lower heat flux part below the Net Vapor Generation (NVG) heat flux exists at the middle of tube length, then the effective heated length becomes the tube length downstream the lower heat flux parts. However, when the lower heat flux part is above the NVG, then the effective heated length is full tube length. (author)

  13. Critical heat flux correlation for thin rectangular channels

    International Nuclear Information System (INIS)

    Tanaka, Futoshi; Mishima, Kaichiro; Hibiki, Takashi

    2007-01-01

    The effect of heated length on Critical heat flux (CHF) in thin rectangular channels was studied based on CHF data obtained under atmospheric pressure. CHF in small channels has been widely studied in the past decades but most of the studies are related to CHF in round tubes. Although basic mechanisms of burnout in thin rectangular channels are similar to tubes, applicability of CHF correlations for tubes to rectangular channels are questionable since CHF in rectangular channels are affected by the existence of non-heated walls and the non-circular geometry of channel circumference. Several studies of CHF in thin rectangular channels have been reported in relation to thermal hydraulic design of research reactors and neutron source targets and CHF correlations have been proposed, but the studies mostly focus on CHFs under geometrical conditions of the application of interest. In his study, existing CHF data obtained in thin rectangular channels were collected and the effect of heated length on CHF was examined. Existing CHF correlations were verified with positive quality flow CHF data but none of the correlations successfully reproduced the CHF for a wide range of heated length. A new CHF correlation for qualify region applicable to a wide range of heated length was developed based on the collected data. (author)

  14. Inlet effect induced ''upstream'' critical heat flux in smooth tubes

    International Nuclear Information System (INIS)

    Kitto, J.B. Jr.

    1986-01-01

    An unusual form of ''upstream'' critical heat flux (CHF) has been observed and directly linked to the inlet flow pattern during an experimental study of high pressure (17 - 20 MPa) water flowing through a vertical 38.1 mm ID smooth bore tube with uniform axial and nonuniform circumferential heating. These upstream CHF data were characterized by temperature excursions which initially occurred at a relatively fixed axial location in the middle of the test section while the outlet and inlet heated lengths experienced no change. A rifled tube inlet flow conditioner could be substituted for a smooth tube section to generate the desired swirling inlet flow pattern. The upstream CHF data were found to match data from a uniformly heated smooth bore tube when the comparison was made using the peak local heat flux. The mechanism proposed to account for the upstream CHF observations involves the destructive interference between the decaying swirl flow and the secondary circumferential liquid flow field resulting from the one-sided heating

  15. A study on critical heat flux in gap

    International Nuclear Information System (INIS)

    Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs

  16. A study on critical heat flux in gap

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs.

  17. Prediction of critical heat flux in vertical pipe flow

    International Nuclear Information System (INIS)

    Levy, S.; Healzer, J.M.; Abdollahian, D.

    1981-01-01

    A previously developed semi-empirical model for adiabatic two-phase annular flow ix extended to predict the critical heat flux (CHF) in a vertical pipe. The model exhibits a sharply declining curve of CHF versus steam quality (X) at low X, and is relatively independent of the heat flux distribution. In this region, vaporization of the liquid film controls. At high X, net deposition upon the liquid film becomes important and CHF versus X flattens considerably. In this zone, CHF is dependent upon the heat flux distribution. Model predictions are compared to test data and an empirical correlation. The agreement is generally good if one employs previously reported mass transfer coefficients. (orig.)

  18. Occurrence of critical heat flux during blowdown with flow reversal

    International Nuclear Information System (INIS)

    Leung, J.C.M.

    1976-04-01

    A small-scale experiment using Freon-11 at 130 0 F and 65 psia in a well-instrumented transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 2 ft length. Inlet and exit void fractions were measured by a capacitance technique. Flow regime transition was observed with high speed photography. A 1-hr contact time between Freon-11 and nitrogen at 130 0 F and 60 psig was found to greatly affect the steady-state subcooled boiling initial conditions. Delay in bubble growth was observed in adiabatic blowdown runs. This was caused by the thermodynamic nonequilibrium conditions required for the unstable bubble growth. For the diabatic runs, equilibrium was more closely approached in the test section during the early phase of blowdown. Critical heat flux did not occur immediately during the flow decay in an approximately 60 msec reversal period. The first or early CHF which occurred at about 400 msec was independent of the blowdown volume and did not propagate upward. An annular flow pattern appeared at the onset of this CHF which occurred only at the lower 8 in. of the heated zone

  19. Surface wettability effects on critical heat flux of boiling heat transfer using nanoparticle coatings

    KAUST Repository

    Hsu, Chin-Chi

    2012-06-01

    This study investigates the effects of surface wettability on pool boiling heat transfer. Nano-silica particle coatings were used to vary the wettability of the copper surface from superhydrophilic to superhydrophobic by modifying surface topography and chemistry. Experimental results show that critical heat flux (CHF) values are higher in the hydrophilic region. Conversely, CHF values are lower in the hydrophobic region. The experimental CHF data of the modified surface do not fit the classical models. Therefore, this study proposes a simple model to build the nexus between the surface wettability and the growth of bubbles on the heating surface. © 2012 Elsevier Ltd. All rights reserved.

  20. Experimental investigation of pool boiling heat transfer and critical heat flux on a downward facing surface

    International Nuclear Information System (INIS)

    Gocmanac, M.; Luxat, J.C.

    2012-01-01

    A separate effects experimental study of heat transfer and Critical Heat Flux (CHF) on a downward facing plate in subcooled water pool boiling is described. Two geometries of downwards facing surfaces are studied. The first is termed the 'confined' study in which bubble motion is restricted to the heated surface. The second is termed the 'unconfined' study where individual bubbles are free to move along the heated surface and vent in any direction. The method used in the confined study is novel and involves the placement of a lip surrounding the heated surface. The CHF as a function of angle of inclination of the surface is presented and is in good agreement with other experimental data from somewhat different test geometries. (author)

  1. Experimental study of critical heat flux in inclined rectangular gap

    International Nuclear Information System (INIS)

    Kim, S.J.; Kim, Y.H.; Noh, S.W.; Suh, K.Y.; Rempe, J.L.; Cheung, F.B.; Kim, S.B.

    2003-01-01

    In the TMI-2 accident, the lower part of the reactor pressure vessel was overheated and then rather rapidly cooled down, as was later found out in a vessel investigation project. This accounted for the possibility of gap cooling feasibility. For this reason, a great deal of investigations was performed to determine the critical heat flux (CHF) from the standpoint of in-vessel retention (IVR). As part of a joint Korean-U.S. International Nuclear Energy Research Initiative (INERI) project, Tests were conducted to examine the critical heat flux (CHF) on the one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under the atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10 mm, and the surface orientation angles from the downward-facing position (180deg) to the vertical position (90deg), respectively. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. However, in downward-facing position (180deg), somewhat differing results were detected relative to previous reports. For a certain gap size having a similar dimension with vapor layer thickness, more efficient heat transfer was detected and this may be interpreted by characteristic property such as the vapor layer thickness of water. In consistency with several studies reported in the literature, it was found that there exists a transition angle above that the CHF changes with a rapid slope. (author)

  2. Anomalies and other concerns related to the critical heat flux

    International Nuclear Information System (INIS)

    Groeneveld, D.C.

    2009-01-01

    This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes. They include the following: Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated tubes. This phenomenon has been observed at high flows in both water and Freon. Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30 - 90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment-controlled and the deposition-controlled region and causes problems for prediction methods of the form CHF=f(X). Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, thus diminishing the effects of flow regime/heat transfer regime transitions responsible for the above unusual CHF trends. This will lead to a more gradually decreasing CHF vs. X curve. Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a dryout temperature excursion) to a positive slope. Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the shape of the axial temperature distribution of a heated tube experiencing CHF or heat transfer deterioration, when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical. CHF fluid-to-fluid modeling: differences in CHF behavior at certain conditions between refrigerants and water at equivalent conditions. The mechanisms responsible for these trends and the implications for predicting CHF for bundle geometries

  3. Using a thermalhydraulics system code to estimate heat transfer coefficients for a critical heat flux experiment

    International Nuclear Information System (INIS)

    Statham, B.A.

    2009-01-01

    RELAP5/SCDAPSIM MOD 3.4 is used to predict wall temperature before and after critical heat flux (CHF) is reached in a vertical, uniformly heated tube using light water as the working fluid. The heated test section is modeled as a 1 m long Inconel 600 tube having an OD of 6.35 mm and ID of 4.57 mm with a 0.5 m long unheated development length at the inlet. Simulations are performed at pressures of 0.5 to 2.0 MPa with mass fluxes from 500 to 2000 kg m -2 s -1 and inlet qualities ranging from -0.2 to 0. Loss of flow simulations are performed with flow reduction rates of 10, 20, 50, and 100 kg m -2 s -2 . Inlet mass flux at CHF was nominally independent of rate in the model; this may or may not be realistic. (author)

  4. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-01-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

  5. Anomalies and other concerns related to the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: thermal@magma.ca [Researcher Emeritus, Chalk River Laboratories, Atomic Energy of Canada Ltd, Chalk River (Canada) and University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada)

    2011-11-15

    This paper summarizes various unusual trends in the critical heat flux (CHF) that have been observed experimentally in tubes or bundle subassemblies. They include the following: Bullet Occurrence of a minimum in the CHF vs. quality (X) curve at high flows - leading to an initial upstream CHF occurrence in uniformly heated channels. This phenomenon has been observed at high flows in both water and Freon. Bullet Occurrence of a limiting quality region on the CHF vs. X curve where the CHF drops by 30-90% for a nearly constant quality. This is thought to correspond to the boundary between the entrainment controlled and the deposition controlled region and causes problems for prediction methods of the form CHF = f(X). Bullet Impact of flow obstructions on the occurrence of upstream CHF and the limiting quality region. The additional mixing by grid spacers or bundle appendages results in a more homogeneous phase distribution, and diminishes the effects of flow regime/heat transfer regime transitions responsible for some of the unusual CHF trends, and results in a more gradually decreasing CHF vs. X curve. Bullet Absence of a CHF temperature excursion at high flows and high qualities - this is found to be caused by a change in slope of the transition boiling part of the boiling curve from a negative value (usual trend that results in a temperature excursion) to a positive slope. Bullet Gradual disappearance of the sharp temperature excursion at CHF when increasing the pressure towards and beyond the critical pressure - no drastic change is observed in the axial temperature distribution of a heated tube experiencing CHF when, for constant mass flux and inlet temperature, the pressure is gradually increased from subcritical to supercritical. Bullet CHF fluid-to-fluid modelling: differences in CHF trends at certain conditions between refrigerants and water at equivalent conditions. The mechanisms responsible for these trends and the implications for bundle geometries are

  6. Critical heat flux in tubes and tight hexagonal rod lattices

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Cheng Xu; Zeggel, W.

    1994-01-01

    The critical heat flux (CHF) in small-diameter tubes and in tight hexagonal 7-rod and 37-rod bundles was investigated in the KRISTA test facility, using Freon 12 as the working fluid. The measurements in tubes showed that the influence of the tube diameter on CHF cannot be described as suggested by earlier publications with sufficient accuracy. CHF in bundles is lower than in tubes under comparable conditions. The influence of spacers (grid spacers, wire wraps) on CHF was found to be governed by local steam qualities. A comparison of the test results with some CHF prediction methods showed that the look-up table method reproduces the test results in circular tubes most accurately. Combined with CHF look-up tables, subchannel analysis and Ahmad's fluid-to-fluid scaling law, Freon experiments have proven to be a suitable tool for CHF prediction in water-cooled rod bundles. (orig.) [de

  7. Experiments on Critical Heat Flux for CAREM -25 Reactor

    International Nuclear Information System (INIS)

    Mazufri, C.M

    2000-01-01

    The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data.Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions.Correlations found in the open literature are not sufficiently verified for the thermal hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities.To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions was carried out.The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation.A short description of facilities, details of the experimental program and some preliminary results obtained are presented in this work

  8. Development of low flow critical heat flux correlation for HANARO

    International Nuclear Information System (INIS)

    Park, Cheol; Chae, Hee Taek; Hang, Gee Yang.

    1997-07-01

    A low flow CHF correlation was developed for the safe operation of HANARO during the natural circulation cooling and the assessment of safety during the low flow condition of accident. The analytical model was applied to estimate the heat flux and the temperature distributions along the periphery of the fin at CHF conditions, and the predicted wall temperature at the sheath between the fins by the model agreed well with the measured one. The parametric trends of the CHF data for the finned geometry agreed with the general understanding from the previous studies for the unfinned annulus or tube geometries. It is revealed that the fin does not affect the CHF for low flow condition, although it increase the critical power due to larger heat transfer area. As the existing CHF correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental CHF data with RMS errors of 13.7 %. (author). 19 refs., 3 tabs., 23 figs

  9. Development of low flow critical heat flux correlation for HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Chae, Hee Taek; Hang, Gee Yang

    1997-07-01

    A low flow CHF correlation was developed for the safe operation of HANARO during the natural circulation cooling and the assessment of safety during the low flow condition of accident. The analytical model was applied to estimate the heat flux and the temperature distributions along the periphery of the fin at CHF conditions, and the predicted wall temperature at the sheath between the fins by the model agreed well with the measured one. The parametric trends of the CHF data for the finned geometry agreed with the general understanding from the previous studies for the unfinned annulus or tube geometries. It is revealed that the fin does not affect the CHF for low flow condition, although it increase the critical power due to larger heat transfer area. As the existing CHF correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental CHF data with RMS errors of 13.7 %. (author). 19 refs., 3 tabs., 23 figs.

  10. Review of the critical heat flux correlations for liquid metals

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Han, H. D.; Chang, W. P.; Kwon, Y. M.

    1999-09-01

    The CHF phenomenon in the two-phase convective flows has been an important issue in the fields of design and safety analysis of light water reactor (LWR) as well as sodium cooled liquid metal reactor (LMR). Especially in the LWR application, many physical aspects of the CHF phenomenon are understood and reliable correlations and mechanistic models to predict the CHF condition have been proposed over the past three decades. Most of the existing CHF correlations have been developed for light water reactor core applications. Compared with water, liquid metals show a divergent picture of boiling pattern. This can be attributed to the consequence that special CHF conditions obtained from investigations with water cannot be applied to liquid metals. Numerous liquid metal boiling heat transfer and two-phase flow studies have put emphasis on development of models and understanding of the mechanism for improving the CHF predictions. Thus far, no overall analytical solution method has been obtained and the reliable prediction method has remained empirical. The principal objectives of the present report are to review the state of the art in connection with liquid metal critical heat flux under low pressure and low flow conditions and to discuss the basic mechanisms. (author)

  11. Azimuthal critical heat flux in narrow rectangular channels

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Hoon; Noh, Sang Woo; Kim, Sung Joong; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-07-01

    Tests were conducted to examine the critical heat flux (CHF) on the one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under the atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10mm, and the surface orientation angles from the downward-facing position (180{sup o}) to the vertical position (90{sup o}), respectively. Also, the CHF experiments were performed for pool boiling with varying heater surface orientations in the unconfined space at the atmospheric pressure using the rectangular test section. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. In consistency with several studies reported in the literature, it was found that there exists a transition angle above which the CHF changes with a rapid slope. An engineering correlation is developed for the CHF during natural convective boiling in the inclined, confined rectangular channels with the aid of dimensional analysis.

  12. Occurrence of critical heat flux during blowdown with flow reversal

    International Nuclear Information System (INIS)

    Leung, J.C.M.

    1977-01-01

    A small-scale experiment using Freon-11 at 130 0 F (54.4 0 C) and 65 psia (0.45 MPa) in a well-instrumented, transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blowdown with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 61-cm length. Inlet and exit void fractions were measured by a capacitance technique. Flow-regime transition was observed with high-speed photography. A 1-hr contact time between Freon-11 and nitrogen at 130 0 F (54.4 0 C) and 60 psig (0.517 MPa) was found to greatly affect the steady-state subcooled-boiling initial conditions. Delay in bubble growth was observed in adiabatic blowdown runs. This was caused by the conditions of thermodynamic nonequilibrium required for the unstable bubble growth. For the diabatic runs, equilibrium was more closely approached in the test section during the early phase of blowdown

  13. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

    1997-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  14. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G R; Bullock, D E; Inch, W [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  15. Critical heat flux and exit film flow rate in a flow boiling system

    International Nuclear Information System (INIS)

    Ueda, Tatsuhiro; Isayama, Yasushi

    1981-01-01

    The critical heat flux in a flowing boiling system is an important problem in the evaporating tubes with high thermal load such as nuclear reactors and boilers, and gives the practical design limit. When the heat flux in uniformly heated evaporating tubes is gradually raised, the tube exit quality increases, and soon, the critical heat flux condition arises, and the wall temperature near tube exit rises rapidly. In the region of low exit quality, the critical heat flux condition is caused by the transition from nucleating boiling, and in the region of high exit quality, it is caused by dry-out. But the demarcation of both regions is not clear. In this study, for the purpose of obtaining the knowledge concerning the critical heat flux condition in a flowing boiling system, the relation between the critical heat flux and exit liquid film flow rate was examined. For the experiment, a uniformly heated vertical tube supplying R 113 liquid was used, and the measurement in the range of higher heating flux and mass velocity than the experiment by Ueda and Kin was carried out. The experimental setup and experimental method, the critical heat flux and exit quality, the liquid film flow rate at heating zone exit, and the relation between the critical heat flux and the liquid film flow rate at exit are described. (Kako, I.)

  16. Boiling Heat Transfer Coefficients of Nanofluids Containing Carbon Nanotubes up to Critical Heat Fluxes

    International Nuclear Information System (INIS)

    Park, Ki Jung; Lee, Yohan; Jung, Dong Soo; Shim, Sang Eun

    2011-01-01

    In this study, the nucleate pool boiling heat transfer coefficients (HTCs) and critical heat flux (CHF) for a smooth and square flat heater in a pool of pure water with and without carbon nanotubes (CNTs) dispersed at 60 .deg. C were measured. Tested aqueous nanofluids were prepared using CNTs with volume concentrations of 0.0001%, 0.001%, and 0.01%. The CNTs were dispersed by chemically treating them with an acid in the absence of any polymers. The results showed that the pool boiling HTCs of the nanofluids are higher than those of pure water in the entire nucleate boiling regime. The acid-treated CNTs led to the deposition of a small amount of CNTs on the surface, and the CNTs themselves acted as heat-transfer-enhancing particles, owing to their very high thermal conductivity. There was a significant increase in the CHF- up to 150%-when compared to that of pure water containing CNTs with a volume concentration of 0.001%. This is attributed to the change in surface characteristics due to the deposition of a very thin layer of CNTs on the surface. This layer delays nucleate boiling and causes a reduction in the size of the large vapor canopy around the CHF. This results in a significant increase in the CHF

  17. Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube; Transfert thermique et flux critique dans un ecoulement helicoidal en tube chauffe asymetriquement

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Sciences de la Matiere; [Association Euratom-CEA, Centre d` Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee

    1997-03-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author) 197 refs.

  18. Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow; Transfert thermique et flux critique dans un ecoulement helicoidal en tube chauffe asymetriquement

    Energy Technology Data Exchange (ETDEWEB)

    Boscary, J

    1995-10-01

    The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs.

  19. A Critical Review of OSHA Heat Enforcement Cases: Lessons Learned.

    Science.gov (United States)

    Arbury, Sheila; Lindsley, Matthew; Hodgson, Michael

    2016-04-01

    The aim of the study was to review the Occupational Safety and Health Administration's (OSHA) 2012 to 2013 heat enforcement cases, using identified essential elements of heat illness prevention to evaluate employers' programs and make recommendations to better protect workers from heat illness. (1) Identify essential elements of heat illness prevention; (2) develop data collection tool; and (3) analyze OSHA 2012 to 2013 heat enforcement cases. OSHA's database contains 84 heat enforcement cases in 2012 to 2013. Employer heat illness prevention programs were lacking in essential elements such as providing water and shade; adjusting the work/rest proportion to allow for workload and effective temperature; and acclimatizing and training workers. In this set of investigations, most employers failed to implement common elements of illness prevention programs. Over 80% clearly did not rely on national standard approaches to heat illness prevention.

  20. Application of the Critical Heat Flux Look-Up Table to Large Diameter Tubes

    Directory of Open Access Journals (Sweden)

    M. El Nakla

    2013-01-01

    Full Text Available The critical heat flux look-up table was applied to a large diameter tube, namely 67 mm inside diameter tube, to predict the occurrence of the phenomenon for both vertical and horizontal uniformly heated tubes. Water was considered as coolant. For the vertical tube, a diameter correction factor was directly applied to the 1995 critical heat flux look-up table. To predict the occurrence of critical heat flux in horizontal tube, an extra correction factor to account for flow stratification was applied. Both derived tables were used to predict the effect of high heat flux and tube blockage on critical heat flux occurrence in boiler tubes. Moreover, the horizontal tube look-up table was used to predict the safety limits of the operation of boiler for 50% allowable heat flux.

  1. Correlation of critical heat flux data for uniform tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jafri, T.; Dougherty, T.J.; Yang, B.W. [Columbia Univ., New York, NY (United States)

    1995-09-01

    A data base of more than 10,000 critical heat flux (CHF) data points has been compiled and analyzed. Two regimes of CHF are observed which will be referred to as the high CHF regime and the low CHF regime. In the high CHF regime, for pressures less than 110 bar, CHF (q{sub c}) is a determined by local conditions and is adequately represented by q{sub c} = (1.2/D{sup 1/2}) exp[-{gamma}(GX{sub t}){sup 1/2}] where the parameter {gamma} is an increasing function of pressure only, X{sub t} the true mass fraction of steam, and all units are metric but the heat flux is in MWm{sup -2}. A simple kinetic model has been developed to estimate X{sub t} as a function of G, X, X{sub i}, and X{sub O}, where X{sub i} is the inlet quality and X{sub O} represents the quality at the Onset of Significant Vaporization (OSV) which is estimated from the Saha-Zuber (S-Z) correlation. The model is based on a rate equation for vaporization suggested by, and consistent with, the S-Z correlation and contains no adjustable parameters. When X{sub i}X{sub O}, X{sub t} depends on X{sub i}, a nonlocal variable, and, in this case, CHF, although determined by local conditions, obeys a nonlocal correlation. This model appears to be satisfactory for pressures less than 110 bar, where the S-Z correlation is known to be reliable. Above 110 bar the method of calculating X{sub O}, and consequently X{sub t}, appears to fail, so this approach can not be applied to high pressure CHF data. Above 35 bar, the bulk of the available data lies in the high CHF regime while, at pressures less than 35 bar, almost all of the available data lie in the low CHF regime and appear to be nonlocal.

  2. Heat stress and public health: a critical review.

    Science.gov (United States)

    Kovats, R Sari; Hajat, Shakoor

    2008-01-01

    Heat is an environmental and occupational hazard. The prevention of deaths in the community caused by extreme high temperatures (heat waves) is now an issue of public health concern. The risk of heat-related mortality increases with natural aging, but persons with particular social and/or physical vulnerability are also at risk. Important differences in vulnerability exist between populations, depending on climate, culture, infrastructure (housing), and other factors. Public health measures include health promotion and heat wave warning systems, but the effectiveness of acute measures in response to heat waves has not yet been formally evaluated. Climate change will increase the frequency and the intensity of heat waves, and a range of measures, including improvements to housing, management of chronic diseases, and institutional care of the elderly and the vulnerable, will need to be developed to reduce health impacts.

  3. Measurements of Critical Heat Flux using Mass Transfer System

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seung Hyun; Chung Bum Jin [Kyunghee University, Yongin (Korea, Republic of)

    2016-05-15

    In a severe accident, the reactor vessel is heated by the decay heat from core melts and the outer surface of reactor vessel is cooled by the natural convection of water pool. When the heat flux increases, boiling will start. Further increase of the heat flux may result in the CHF, which is generated by the bubble combinations. The CHF means that the reactor vessel was separated with coolant and wall temperature is raised rapidly. It may damage the reactor vessel. Also the CHF indicates the maximum cooling capability of the system. Therefore, the CHF has been used as a criterion for the regulatory and licensing. Mechanism of hydrogen vapor bubbles generated and combined can be simulated water bubbles mechanism. And also the both heat and mass transfer mechanism of CHF can be identified in the same methods. Therefore, the CHF phenomena can be simulated enough by mass transfer.

  4. Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, S.; Holmes, D.H.

    1977-04-01

    An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

  5. Critical heat flux detection in rods simulating fuel elements by using dilation method

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1993-01-01

    In out-reactor heat transfer experiments, fuel elements are often simulated by electrically heated rods. In order to prevent the heating rod from being damaged by burnout, when the critical heat flux occurs a safety system is provided which checks the axial thermal expansion of the rod. In case of sudden temperature increase, the corresponding elongation causes a fast interruption of the electrical power supply. The experiments presented here show that this method is more effective than one that uses thermocouples. (author)

  6. Experimental study and technique for calculation of critical heat fluxes in helium boiling in tubes

    International Nuclear Information System (INIS)

    Arkhipov, V.V.; Kvasnyuk, S.V.; Deev, V.I.; Andreev, V.K.

    1979-01-01

    Studied is the effect of regime parameters on critical heat loads in helium boiling in a vertical tube in the range of mass rates of 80 2 xc) and pressures of 100<=p<=200 kPa for the vapor content range corresponding to the heat exchange crisis of the first kind. The method for calculating critical heat fluxes describing experimental data with the error less than +-15% is proposed. The critical heat loads in helium boiling in tubes reduce with the growth of pressure and vapor content in the regime parameter ranges under investigation. Both positive and negative effects of the mass rate on the critical heat flux are observed. The calculation method proposed satisfactorily describes the experimental data

  7. Tabular method of critical heat flux description in square packing rod bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.; Smogalev, I.P.

    2003-01-01

    Elaborations of harnessing tabular method for the description and calculation of critical heat fluxes in square packing rod bundles are presented. The tabular method for fuel rod triangular assemblies derived from using basic table for critical heat fluxes in triangular fuel assemblies demonstrates good results. For the harnessing tabular method in square packing rod bundles correction functions reflecting specific geometry were found. Comparative evaluations of calculated values for the critical heat fluxes with experimental ones are presented. Good agreement of calculations with experiments is noted in all range of parameters [ru

  8. Experimental result of BWR post-CHF tests. Critical heat flux and post-CHF heat transfer coefficient. Contract research

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Anoda, Yoshinari

    2002-02-01

    Authors performed post-CHF experiments under wider pressure ranges of 2 MPa - 18 MPa, wider mass flux ranges of 33 kg/m 2 s - 1651 kg/m 2 s and wider superheat of heaters up to 500 K in comparison to experimental ranges at previous post-CHF experiments. Data on boiling transition, critical heat flux and post-CHF heat transfer coefficient were obtained. Used test section was 4x4-rod bundle with heaters, which diameter and length were the same as those of BWR nuclear fuels. As the result of the experiments, it was found that the boiling transition occurred just below several grid spacers, and that the fronts of the boiling transition region proceeded lower with increase of heated power. Heat transfer was due to nucleate boiling above grid spacers, while it was due to film boiling below grid spacers. Consequently, critical heat flux is affected on the distance from the grid spacers. Critical heat flux above the grid spacers was about 15% higher than that below the grid spacers, by comparing them under the same local condition. Heat transfer by steam turbulent flow was dominant to post-CHF heat transfer, when superheat of heaters was sufficiently high. Then, post-CHF heat transfer coefficient was predicted with heat transfer correlations for single-phase flow. On the other hand, when superhead of heaters was not sufficiently high, post-CHF heat transfer coefficient was higher than the prediction with heat transfer correlations for single-phase flow. Mass flux effect on post-CHF heat transfer coefficient was described by standardization of post-CHF heat transfer coefficient with the prediction for single-phase flow. However, pressure effect, superheat effect and effect of position were not described. Authors clarified that those effects could be described with functions of heater temperature and position. Post-CHF heat transfer coefficient was lowest just blow the grid spacers, and it increased with the lower positions. It increased by about 30% in one span of the grid

  9. Experimental result of BWR post-CHF tests. Critical heat flux and post-CHF heat transfer coefficient. Contract research

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi; Anoda, Yoshinari [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwaki, Chikako [Toshiba Corp., Tokyo (Japan)

    2002-02-01

    Authors performed post-CHF experiments under wider pressure ranges of 2 MPa - 18 MPa, wider mass flux ranges of 33 kg/m{sup 2}s - 1651 kg/m{sup 2}s and wider superheat of heaters up to 500 K in comparison to experimental ranges at previous post-CHF experiments. Data on boiling transition, critical heat flux and post-CHF heat transfer coefficient were obtained. Used test section was 4x4-rod bundle with heaters, which diameter and length were the same as those of BWR nuclear fuels. As the result of the experiments, it was found that the boiling transition occurred just below several grid spacers, and that the fronts of the boiling transition region proceeded lower with increase of heated power. Heat transfer was due to nucleate boiling above grid spacers, while it was due to film boiling below grid spacers. Consequently, critical heat flux is affected on the distance from the grid spacers. Critical heat flux above the grid spacers was about 15% higher than that below the grid spacers, by comparing them under the same local condition. Heat transfer by steam turbulent flow was dominant to post-CHF heat transfer, when superheat of heaters was sufficiently high. Then, post-CHF heat transfer coefficient was predicted with heat transfer correlations for single-phase flow. On the other hand, when superhead of heaters was not sufficiently high, post-CHF heat transfer coefficient was higher than the prediction with heat transfer correlations for single-phase flow. Mass flux effect on post-CHF heat transfer coefficient was described by standardization of post-CHF heat transfer coefficient with the prediction for single-phase flow. However, pressure effect, superheat effect and effect of position were not described. Authors clarified that those effects could be described with functions of heater temperature and position. Post-CHF heat transfer coefficient was lowest just blow the grid spacers, and it increased with the lower positions. It increased by about 30% in one span of

  10. Prediction of critical heat flux for water in uniformly heated vertical ...

    African Journals Online (AJOL)

    Keywords: CHF - Heat transfer - Water vapor - Porous coated tubes. Auteur correspondant ... electrical and mechanical characteristics were well validated. Figure. 1 shows ... resistance to vapor filtration from the heating wall to the liquid bulk.

  11. A new facility for the determination of critical heat flux in nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Fortman, R A; Hadaller, G I; Hamilton, R C; Hayes, R C; Shin, K S; Stern, F [Stern Laboratories Inc., Hamilton, ON (Canada)

    1993-11-01

    A facility for the determination of critical heat flux in simulated reactor fuel assemblies has been constructed at Stern Laboratories for CANDU Owners` Group. This paper describes the facility and method of testing. 9 figs.

  12. Hydrodynamics of double phase under high pressure: evolutions of flow configurations until critical heating

    International Nuclear Information System (INIS)

    Raisson, Claude

    1968-01-01

    This research thesis reports the experimental study of flows and of their evolution until critical heating by using appropriate measurement instruments. The objective is to understand how flow evolution may condition critical heating. After a recall of some notions and values related to the study of two-phase flows, and an overview of published works on flow configurations and on critical heating, the author describes test installation and measurement devices, presents the typical test process, reports instrument calibration, and flow configuration tests with water-air flow under low pressure. Results are reported. The author proposes explanations regarding observed phenomena, and a possible scheme to explain the flow evolution until critical heating [fr

  13. Flow instability and critical heat flux in a ribbed annulus

    International Nuclear Information System (INIS)

    Yang, B.W.; Dougherty, T.; Fighetti, C.; Kokolis, S.; Reddy, G.D.; McAssey, E.V. Jr.; Coutts, A.

    1993-01-01

    An experimental program has been conducted to determine the onset of flow instability point in a heated annulus which is divided into four sub channels by non-conducting ribs. The onset of flow instability is identified by the minimum point in the pressure drop-velocity curve. Comparison with a ribless annulus show that the presence of ribs increases the minimum point velocity. In addition, data are presented which show that under certain conditions premature CHF can be induced by the ribs

  14. Critical review of hydraulic modeling on atmospheric heat dissipation

    International Nuclear Information System (INIS)

    Onishi, Y.; Brown, S.M.

    1977-01-01

    Objectives of this study were: to define the useful roles of hydraulic modeling in understanding the predicting atmospheric effects of heat dissipation systems; to assess the state-of-the-art of hydraulic modeling of atmospheric phenomena; to inventory potentially useful existing hydraulic modeling facilities both in the United States and abroad; and to scope hydraulic model studies to assist the assessment of atmospheric effects of nuclear energy centers

  15. Laminar forced convective heat transfer to near-critical water in a tube

    International Nuclear Information System (INIS)

    Lee, Sang Ho

    2003-01-01

    Numerical modeling is carried out to investigate forced convective heat transfer to near-critical water in developing laminar flow through a circular tube. Due to large variations of thermo-physical properties such as density, specific heat, viscosity, and thermal conductivity near thermodynamic critical point, heat transfer characteristics show quite different behavior compared with pure forced convection. With flow acceleration along the tube unusual behavior of heat transfer coefficient and friction factor occurs when the fluid enthalpy passes through pseudocritical point of pressure in the tube. There is also a transition behavior from liquid-like phase to gas-like phase in the developing region. Numerical results with constant heat flux boundary conditions are obtained for reduced pressures from 1.09 to 1.99. Graphical results for velocity, temperature, and heat transfer coefficient with Stanton number are presented and analyzed

  16. Heat Capacity of Room-Temperature Ionic Liquids: A Critical Review

    Science.gov (United States)

    Paulechka, Yauheni U.

    2010-09-01

    Experimental data on heat capacity of room-temperature ionic liquids in the liquid state were compiled and critically evaluated. The compilation contains data for 102 aprotic ionic liquids from 63 literature references and covers the period of time from 1998 through the end of February 2010. Parameters of correlating equations for temperature dependence of the heat capacities were developed.

  17. Prediction of Critical Heat Flux under Rolling Motion

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Jinseok; Lee, Yeongun; Park, Gooncherl [Seoul National Univ., Seoul (Korea, Republic of)

    2013-05-15

    The aim to this paper may be summarized as follows: identify the flow regime compare with existing void-quality relationship and void fraction at OAF derived from the vapor superficial velocity obtained by the churn-to annular flow criterion, develop and evaluate the correlation for accurate prediction of CHF ratio under rolling motion. Experimentally measured CHF results from the previous study were not well-predicted by existing CHF correlations developed for wide range of pressure under rolling motion in vertical tube. Specifically, existing correlations do not account for the dynamic motion parameter, such as tangential and centrifugal force. This study reviewed some existing correlation and experimental studies related to reduction and enhancement of CHF and heat transfer and flow behavior under heaving and rolling motion, and developed a CHF ratio correlation for upward flow vertical tube under rolling motion. Based upon dimensionless groups, equations and interpolation factor, an empirical CHF correlation has been developed which is consistent with experimental data for uniformly heated tubes internally cooled by R-134 under rolling motion. Flow regime was determined through the prediction method for annular flow. Non-dimensional number and function were decided by CHF mechanism of each region. Interaction of LFD and DNB regions is taken into account by means of power interpolation which is reflected void fraction at OAF. The suggested correlation predicted the CHF Ratio with reasonable accuracy, showing an average error of -0.59 and 2.51% for RMS. Rolling motion can affect bubble motion and liquid film behavior complexly by combination of tangential and centrifugal forces and mass flow than heaving motion. Through a search of literature and a comparison of previous CHF ratio results, this work can contribute to the study of boiling heat transfer and CHF for the purpose of enhancement or reduction the CHF of dynamic motion system, such as marine reactor.

  18. Heat transfer in pool boiling liquid neon, deuterium and hydrogen, and critical heat flux in forced convection of liquid neon

    International Nuclear Information System (INIS)

    Astruc, J.M.

    1967-12-01

    In the first part, free-convection and nucleate pool boiling heat transfer (up to burn-out heat flux) between a platinum wire of 0.15 mm in diameter in neon, deuterium and hydrogen has been studied at atmospheric pressure. These measurements were continued in liquid neon up to 23 bars (Pc ≅ 26.8 b). Film boiling heat transfer coefficients have been measured in pool boiling liquid neon at atmospheric pressure with three heating wires (diameters 0.2, 0.5, 2 mm). All the results have been compared with existing correlations. The second part is devoted to measurements of the critical heat flux limiting heat transfer with small temperature differences between the wall and the liquid neon flowing inside a tube (diameters 3 x 3.5 mm) heated by joule effect on 30 cm of length. Influences of flow stability, nature of electrical current, pressure, mass flow rate and subcooling are shown. In conclusion, the similarity of the heat transfer characteristics in pool boiling as well as in forced convection of liquid neon and hydrogen is emphasized. (author) [fr

  19. A review on critical heat flux in horizontal tubes

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Gaikwad, Avinash; Prabhu, S.V.

    2015-01-01

    Coolant channels of PHWR during accident similar to loss of coolant accident (LOCA) may experience different flow transients with low pressure and low flow conditions. In the advanced PHWRs it is desired to have small amount of positive quality at the exit of the coolant channel to increase the thermal efficiency. Investigation on pressure drop and heat transfer coefficient under subcooled boiling condition is important in the design and operation of the PHWRs. Understanding of thermal hydraulic phenomena associated with horizontal flow is also important in the safety and accident management in these reactors. A detailed experimental investigation on the important thermal hydraulic phenomena of horizontal tubes under low pressure and low flow conditions is carried out. The phenomena covered in this work are measurement of diabatic single phase and subcooled boiling pressure drop and local heat transfer coefficients, steady state CHF, effect of upstream flow restrictions on flow transients and CHF, CHF under oscillatory flow and flow decreasing transients. A detailed literature review is carried out on CHF in horizontal channels to take stock of the works being carried out along with current state of the art and to justify the motivation for the experimental study. This paper presents the review of available literature on horizontal CHF with the results of the experimental work. (author)

  20. Critical heat fluxes and liquid distribution in annular channels in the dispersion-annular flow

    International Nuclear Information System (INIS)

    Boltenko, Eh.A.; Pomet'ko, R.S.

    1984-01-01

    On the basis of using the dependence of intensity of total mass transfer between the flux nucleus and wall film obtained for tubes with uniform heat release and taking into account the peculiarities of mass transfer between the flux nucleus and wall film in annular channels the technique for calculating the liquid distribution and critical capacity of annular channels with internal, external and bilateral heating at uniform and non-uniform heat release over the length is proposed. The calculation of annular channels critical capacity according to the suggested technique is performed. A satisfactory agreement of calculation results with the experimental data is attained

  1. Critical heat flux with subcooled boiling of water at low pressure

    International Nuclear Information System (INIS)

    Chen Yuzhou; Zhou Runbin; Hao Laomi; Chen Haiyan

    1997-01-01

    The critical heat flux experiment has been performed in round tubes of 10 and 16 mm in diameter with different heating length, covering the range of pressure 1.5-16.7 bar, velocity 1.4-15.4 m/s and exit subcooling 30-136 K. The experimental data and empirical correlations are presented. Based on the results an evaluation of some correlations and 1995 CHF look-up table is made. For the conditions tested the effect of diameter on the critical heat flux is found to be related to the liquid velocity. (author)

  2. Critical condition for current-driven instability excited in turbulent heating of TRIAM-1 tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Y; Watanabe, T; Nagao, A; Nakamura, K; Kikuchi, M; Aoki, T; Hiraki, N; Itoh, S [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics; Mitarai, O

    1982-02-01

    Critical condition for current-driven instability excited in turbulently heated TRIAM-1 tokamak plasma is investigated experimentally. Resistive hump in loop voltage, plasma density fluctuation and rapid increase of electron temperature in a skin layer are simultaneously observed at the time when the electron drift velocity amounts to the critical drift velocity for low-frequency ion acoustic instability.

  3. Experiments of Pool Boiling Performance (Boiling Heat Transfer and Critical Heat Flux) on Designed Micro-Structures

    International Nuclear Information System (INIS)

    Kim, Seol Ha; Kang, Jun Young; Lee, Gi Chol; Kiyofumia, Moriyama; Kim, Moo Hwan; Park, Hyun Sun

    2015-01-01

    In general, the evaluation of the boiling performance mainly focuses on two physical parameters: boiling heat transfer (BHT) and critical heat flux (CHF). In the nuclear power plants, both BHT and CHF contribute the nuclear system efficiency and safety, respectively. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on Pin-fin effect analysis. In terms of CHF, critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on the roughness ratio. The extended heat transfer area contributes the boiling heat transfer increase on the structured surface, and its quantitative analysis has been performed. In terms of CHF, the critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. We suggested a capillary limit to CHF delay for modeling capillary induced liquid inflow through microstructured surfaces. The critical size of the capillary limit on the prepared structured surface, determined by a model, could be reasonable explanation points for the experimental results (optimal size for CHF delay). The present experimental results also showed clearly the critical size (10 - 20 μm) for CHF delay, predicted by capillary limit analysis. This study provides fundamental insight into BHT and CHF enhancement of structured surfaces, and an optimal design guide for the required CHF and boiling heat-transfer performance. Finally, this study can contribute the basic understanding of the boiling on designed microstructure surface, and it also suggest the optimal micro scaled structured surface of boiling

  4. Surface wettability effects on critical heat flux of boiling heat transfer using nanoparticle coatings

    KAUST Repository

    Hsu, Chin-Chi; Chen, Ping-Hei

    2012-01-01

    This study investigates the effects of surface wettability on pool boiling heat transfer. Nano-silica particle coatings were used to vary the wettability of the copper surface from superhydrophilic to superhydrophobic by modifying surface topography

  5. Transient critical heat flux under flow coast-down in vertical annulus with non-uniform heat flux distribution

    International Nuclear Information System (INIS)

    Moon, S.K.; Chun, S.Y.; Choi, K.Y.; Yang, S.K.

    2001-01-01

    An experimental study on transient critical heat flux (CHF) under flow coast-down has been performed for water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady state CHF. The transient CHF experiments have been performed for three kinds of flow transient modes based on the coast-down data of the Kori 3/4 nuclear power plant reactor coolant pump. Most of the CHFs occurred in the annular-mist flow regime. Thus, it means that the possible CHF mechanism might be the liquid film dryout in the annular-mist flow regime. For flow transient mode with the smallest flow reduction rate, the time-to-CHF is the largest. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to-CHF becomes large as the heat flux decreases. Usually, the critical mass flux is large for slow flow reduction. There is a pressure effect on the ratio of the transient CHF data to steady state CHF data. Some conventional correlations show relatively better CHF prediction results for high system pressure, high quality and slow transient modes than for low system pressure, low quality and fast transient modes. (author)

  6. Nuclear boiling heat transfer and critical heat flux in titanium dioxide-water nanofluids

    International Nuclear Information System (INIS)

    Okawa, Tomio; Takamura, Masahiro; Kamiya, Takahito

    2011-01-01

    Nucleate boiling heat transfer was experimentally studied for saturated pool boiling of water-based nanofluids. Since significant nanoparticle deposition on the heated surface was observed after the nucleate boiling in nanofluids, measurement of CHF was also carried out using the nanoparticle deposited heated surface; pure water was used in the CHF measurement. In the present work, the heated surface was a 20 mm diameter cupper surface, and titanium-dioxide was selected as the material of nanoparticles. Experiments were performed for upward- and downward-facing surfaces. Although the CHFs for the downward-facing surface were generally lower than those for the upward-facing surface, the CHFs for the nanoparticle deposited surface were about 1.9 times greater than those for the bare surface in both the configurations. The CHF improvement corresponded well to the reduction of the surface contact angle. During the nucleate boiling in nanofluids, the boiling heat transfer showed peculiar behavior; it was first deteriorated, then improved, and finally approached to an equilibrium state. This observation indicated that the present nanofluid had competing effects to deteriorate and improve the nucleate boiling heat transfer. It was assumed that the wettability and the roughness of the heated surface were influenced by the deposited nanoparticles to cause complex variation of the number of active nucleation sites. During the nucleate boiling of pure water using the downward-facing surface, a sudden increase in the wall temperature was observed stochastically probably due to the accumulation of bubbles beneath the heated surface. Such behavior was not observed when the pure water was replaced by the nanofluid. (author)

  7. Critical heat flux (CHF) phenomenon on a downward facing curved surface

    Energy Technology Data Exchange (ETDEWEB)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical Engineering

    1997-06-01

    This report describes a theoretical and experimental study of the boundary layer boiling and critical heat flux phenomena on a downward facing curved heating surface, including both hemispherical and toroidal surfaces. A subscale boundary layer boiling (SBLB) test facility was developed to measure the spatial variation of the critical heat flux and observe the underlying mechanisms. Transient quenching and steady-state boiling experiments were performed in the SBLB facility under both saturated and subcooled conditions to obtain a complete database on the critical heat flux. To complement the experimental effort, an advanced hydrodynamic CHF model was developed from the conservation laws along with sound physical arguments. The model provides a clear physical explanation for the spatial variation of the CHF observed in the SBLB experiments and for the weak dependence of the CHF data on the physical size of the vessel. Based upon the CHF model, a scaling law was established for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water. The scaling law, which compares favorably with all the available local CHF data obtained for various vessel sizes, can be used to predict the local CHF limits on large commercial-size vessels. This technical information represents one of the essential elements that is needed in assessing the efficacy of external cooling of core melt by cavity flooding as a severe accident management strategy. 83 figs., 3 tabs.

  8. Critical heat flux (CHF) phenomenon on a downward facing curved surface

    International Nuclear Information System (INIS)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C.

    1997-06-01

    This report describes a theoretical and experimental study of the boundary layer boiling and critical heat flux phenomena on a downward facing curved heating surface, including both hemispherical and toroidal surfaces. A subscale boundary layer boiling (SBLB) test facility was developed to measure the spatial variation of the critical heat flux and observe the underlying mechanisms. Transient quenching and steady-state boiling experiments were performed in the SBLB facility under both saturated and subcooled conditions to obtain a complete database on the critical heat flux. To complement the experimental effort, an advanced hydrodynamic CHF model was developed from the conservation laws along with sound physical arguments. The model provides a clear physical explanation for the spatial variation of the CHF observed in the SBLB experiments and for the weak dependence of the CHF data on the physical size of the vessel. Based upon the CHF model, a scaling law was established for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water. The scaling law, which compares favorably with all the available local CHF data obtained for various vessel sizes, can be used to predict the local CHF limits on large commercial-size vessels. This technical information represents one of the essential elements that is needed in assessing the efficacy of external cooling of core melt by cavity flooding as a severe accident management strategy. 83 figs., 3 tabs

  9. Study of critical dependence of stable phases in Nitinol on heat treatment using electrical resistivity probe

    International Nuclear Information System (INIS)

    Uchil, J.; Mohanchandra, K.P.; Kumara, K.G.; Mahesh, K.K.

    1998-01-01

    Phase transformations in 40% cold-worked Nitinol as a function of heat treatment have been studied using electrical resistivity variation with temperature. The stabilisation of austenitic, rhombohedral and martensitic phases is shown to critically depend on the temperatures of heat treatment by the analysis of temperature dependence of electrical resistivity in heating and cooling parts of the cycle. Characteristic values of electrical resistivity of the stable phases are determined. The R-phase has been found to form continuously with increasing heat-treatment temperature starting from room temperature and to suddenly disappear beyond heat-treatment at 683 K. The observed presence or absence of R-phase is confirmed by heat capacity measurements as a function of temperature. (orig.)

  10. Radiant heat increases piglets’ use of the heated creep area on the critical days after birth

    DEFF Research Database (Denmark)

    Larsen, Mona Lilian Vestbjerg; Thodberg, Karen; Pedersen, Lene Juul

    2017-01-01

    The aim of the present study was to investigate how piglets’ use of a creep area is affected by using radiant heat compared to an incandescent light bulb. It was hypothesised that radiant heat would increase the use of the creep area. Twenty litters were randomly assigned to one of two heat sources...... in the creep area: (1) an incandescent light bulb (STANDARD, n=10) or (2) a radiant heat source (RADIANT, n=10) with five of each type of heat source in each of two batches. Observations on piglets’ position in the pen were made by scan sampling every ten minutes in a 4-hour period from 1100 to 1500 h on day 1......–7, 14 and 21 post partum. A higher percentage of piglets in the creep area was seen for RADIANT litters compared to STANDARD litters on day 2 (P=0.002) and day 3 (P=0.005), and percentage of piglets in the creep area increased for RADIANT litters from day 1 to 2 (P

  11. Critical heat-flux experiments under low-flow conditions in a vertical annulus

    International Nuclear Information System (INIS)

    Mishima, K.; Ishii, M.

    1982-03-01

    An experimental study was performed on critical heat flux (CHF) at low flow conditions for low pressure steam-water upward flow in an annulus. The test section was transparent, therefore, visual observations of dryout as well as various instrumentations were made. The data indicated that a premature CHF occurred due to flow regime transition from churn-turbulent to annular flow. It is shown that the critical heat flux observed in the experiment is essentially similar to a flooding-limited burnout and the critical heat flux can be well reproduced by a nondimensional correlation derived from the previously obtained criterion for flow regime transition. The observed CHF values are much smaller than the standard high quality CHF criteria at low flow, corresponding to the annular flow film dryout. This result is very significant, because the coolability of a heater surface at low flow rates can be drastically reduced by the occurrence of this mode of CHF

  12. Improvement of the skeleton tables for calculation of the critical heat load

    International Nuclear Information System (INIS)

    Gotovskij, M.A.; Kvetnyj, M.A.

    2002-01-01

    Paper presents analysis of drawbacks of the skeleton tables of the critical heat flows applied in calculated heat and hydraulic codes. Paper demonstrates the necessity to take account of specific nature of mechanisms of dryout crisis, of boiling crisis at slow mass rates and the range of small underheatings up to temperature of saturation. Attention is drawn to necessity of detailed account of the natural limitations of the application field of the skeleton tables [ru

  13. An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements

    International Nuclear Information System (INIS)

    Bowditch, F.H.; Mogford, D.J.

    1987-02-01

    This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m 2 s, and coolant inlet temperature between 150 and 320 0 C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)

  14. Thermohydraulics in rod bundles and critical heat flux in transient conditions in a tube

    International Nuclear Information System (INIS)

    Courtaud, M.; Roumy, R.

    1975-01-01

    After the determination of the scaling factor of Stevens's similitude for the pressure range of pressurized water vectors by comparison of critical heat flux data in from and in water, some examples of studies performed with freon are shown. The efficiency of the mixing vanes of spacer grids has been determined on the mixing phenomenon in single phase on critical heat flux. A calculation performed with the code FLICA using subchannel analysis on freon data transposed in water is in good agreement with the experiment. The influence of the number of spacer grids has been also shown. Critical heat fluxes have been determined in water at 140 bar in steady state and transient conditions on two tubular test sections. During the transient tests the flow rate was reduced by half in 0.5 seconds and the reincreased heat flux and inlet temperature remaining constant. These tests have shown the validity of the method which consists in using a critical heat flux correlation determined in steady state conditions applied with local transient conditions of enthalpy and mass velocity computed with the FLICA code [fr

  15. General correlation for prediction of critical heat flux ratio in water cooled channels

    Energy Technology Data Exchange (ETDEWEB)

    Pernica, R.; Cizek, J.

    1995-09-01

    The paper present the general empirical Critical Heat Flux Ration (CHFR) correlation which is valid for vertical water upflow through tubes, internally heated concentric annuli and rod bundles geometries with both wide and very tight square and triangular rods lattices. The proposed general PG correlation directly predicts the CHFR, it comprises axial and radial non-uniform heating, and is valid in a wider range of thermal hydraulic conditions than previously published critical heat flux correlations. The PG correlation has been developed using the critical heat flux Czech data bank which includes more than 9500 experimental data on tubes, 7600 data on rod bundles and 713 data on internally heated concentric annuli. Accuracy of the CHFR prediction, statistically assessed by the constant dryout conditions approach, is characterized by the mean value nearing 1.00 and the standard deviation less than 0.06. Moverover, a subchannel form of the PG correlations is statistically verified on Westinghouse and Combustion Engineering rod bundle data bases, i.e. more than 7000 experimental CHF points of Columbia University data bank were used.

  16. Application of the Bowring correlation for calculating the critical heat flux

    International Nuclear Information System (INIS)

    Borges, R.C.; Freitas, R.L.

    1986-01-01

    The evaluation of the critical heat flux is of great importance for the nuclear reactor project, because it permits the verification of the safety margin with respect to fuel rod damage. This work presents a comparison of the original critical heat flux correlation proposed by Bowring with an alternative form derived from it presented in several papers. Very different results have been encountered from the application of the two correlation forms. Therefore, a criterious choice of the correlation form must be done avoid the violation of the project's safety margin. (Author) [pt

  17. Evaluation of subcooled critical heat flux correlations for tubes with and without internal twisted tapes

    International Nuclear Information System (INIS)

    Inasaka, F.; Nariai, H.

    1996-01-01

    Eleven correlations and models for critical heat flux (CHF) of subcooled flow boiling in water were evaluated. Both a direct substitution method (DSM) and a heat balance condition method (HBM) were compared in the evaluations. The HBM was recommended as a better prediction method in the present study. For straight tubes under uniform heating conditions, the correlations of the Gunther, Knoebel, modified Tong, W-2, and Tong-75, and also the Celata and Weisman-Pei models were confirmed to give reasonably good predictions. Among them, the Celata model was the best with respect to accuracy. For swirl flow under uniform heating conditions, Tong-75-I (involving modification of the water velocity parameter) and Nariai-Inasaka correlations were confirmed to give reasonably good predictions, even though their predictions were too low for the CHF under non-uniform heating conditions. (orig.)

  18. Critical heat flux and post-critical heat flux performance of a 6-m, 37-element fully segmented bundle cooled by Freon-12

    International Nuclear Information System (INIS)

    Nickerson, J.R.

    1982-05-01

    A 6-m, 37-element, electrically heated bundle with full end plate simulation, cooled by Freon-12, has been tested for CHF (critical heat flux) and post-CHF conditions in the MR-3 Freon loop. The bundle was tested in a horizontal attitude and had a uniform axial heat flux distribution and radial heat flux depression. A total of 110 CHF points have been collected over the following range of water equivalent conditions: exit pressure 8.27 - 11.03 MPa, mass flux 1.38 - 8.14 Mg.m -2 .s -1 , inlet subcooling 0 - 500 kJ.kg -1 , outlet quality 10% - 37%. The data have been correlated on both a systems and local conditions basis over a limited mass flux range to within 2.8% rms. Significant CHF increases over smooth bundle results have been observed along with significant CHF improvement over a two end plate bundle simulation in the lower mass flux ranges. A satisfactory axial drypatch spreading correlation has been determined and extensive drypatch wall superheat mapping has been performed

  19. Evaluation of critical temperatures for heat damage in northern highbush blueberry

    Science.gov (United States)

    Overhead sprinklers are often used to cool blueberry fields in the Pacific Northwest, but more information is needed to determine exactly when cooling is needed. The objective of this study was to identify the critical temperatures for heat damage in northern highbush blueberry (Vaccinium corymbosum...

  20. Fine structure in the inter-critical heat-affected zone of HQ130 super ...

    Indian Academy of Sciences (India)

    Unknown

    †Key Laboratory of Liquid Structure and Heredity of Materials, Ministry of Education, ... The microstructure in the inter-critical heat-affected zone (ICHAZ) of HQ130 steel, has been .... Ac3. The microhardness was measured by using the.

  1. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    International Nuclear Information System (INIS)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D.; Driemeyer, D.E.; Kubik, D.L.; Slattery, K.T.; Hellwig, T.H.

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles

  2. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    Energy Technology Data Exchange (ETDEWEB)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D. [Sandia National Labs., Albuquerque, NM (United States); Driemeyer, D.E. Kubik, D.L.; Slattery, K.T.; Hellwig, T.H. [McDonnell Douglas Aerospace, St. Louis, MO (United States)

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles.

  3. Thermal and mechanical behavior of APWR-claddings under critical heat flux conditions

    International Nuclear Information System (INIS)

    Diegele, E.; Rust, K.

    1986-10-01

    Helical grid spacers, such as three or six helical fins as integral part of the claddings, are regarded as a more convenient design for the very tight lattice of an advanced pressurized water reactor (APWR) than grid spacers usually used. Furthermore, it is expected that this spacer design allows an increased safety margin against the critical heat flux (CHF), the knowledge of which is important for design, licensing, and operation of water cooled reactors. To address the distribution of the heat flux density at the outer circumference of the cladding geometry under investigation, the temperature fields in claddings without as well with fins were calculated taking into consideration nuclear and electrically heated rods. Besides the thermal behavior of the claddings, the magnitude and distribution of thermal stresses were determined additionally. A locally increased surface heat flux up to about 40 percent was calculated for the fin bases of nuclear as well as indirect electrically heated claddings with six such helical fins. For all investigated cases, the VON MISES stresses are clearly lower than 200 MPa, implying that no plastic deformations are to be expected. The aim of this theoretical analysis is to allow a qualitative assessment of the finned tube conception and to support experimental investigations concerning the critical heat flux. (orig.) [de

  4. Critical heat fluxes in tubular fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    Subbotin, V.I.; Alekseev, G.V.; Peskov, O.L.

    1974-01-01

    The results of the experiments carried out show that with appropriate choice of tube, type and dimensions of intensifier the attainment of critical conditions at certain parameters is not accompanied by sharp or considerable increases in temperature of the heat removing surface. Increase in power to above critical under these conditions does not lead to considerable variation in temperature either. Thus, it appears possible to change from heat removal by steam-water mixture to convective heat removal by wet steam without manifestation of intolerable temperature conditions of the heating surface (Fig. 6). A change to convective heat removal by wet steam is possible at different levels of heat fluxes which depend during constant conditions at the inlet on tube length and the degree of the disturbing influence on the flow. This is especially important since in principle the possibility arises for developing a power reactor with tubular fuel elements, in which a once-through cycle with steam superheat involving no intermediate separation can be realised

  5. Critical heat flux for downward-facing pool boiling on CANDU calandria tube surface

    Energy Technology Data Exchange (ETDEWEB)

    Behdadi, Azin, E-mail: behdada@mcmaster.ca; Talebi, Farshad; Luxat, John

    2017-04-15

    Highlights: • Pressure tube-calandria tube contact may challenge fuel channel integrity in CANDU. • Critical heat flux variation is predicted on the outer surface of CANDU calandria tube. • A two-phase boundary layer flow driven by buoyancy is modeled on the surface. • Different slip ratios and flow regimes are considered inside the boundary layer. • Subcooling effects are added to the model using wall heat flux partitioning. - Abstract: One accident scenario in CANDU reactors that can challenge the integrity of the primary pressure boundary is a loss of coolant accident, referred to as critical break LOCA, in which the pressure tube (PT) can undergo thermal creep strain deformation and contact its calandria tube (CT). In such case, rapid redistribution of stored heat from PT to CT, leads to a large spike in heat flux to the moderator which can cause bubble accumulation and dryout on the CT surface. A challenge to fuel channel integrity is posed if critical heat flux occurs on the surface of the CT and results in sustained film boiling. If the post-dryout temperature becomes sufficiently high then continued creep strain of the PT and CT may lead to fuel channel failure. In this study, a mechanistic model is developed to predict the critical heat flux variations along the downward facing outer surface of CT. The hydrodynamic model considers a liquid macrolayer beneath an elongated vapor slug on the surface. Local dryout is postulated to occur whenever the fresh liquid supply to the macrolayer is not sufficient to compensate for the liquid depletion. A boundary layer analysis is performed, treating the two phase motion as an external buoyancy driven flow. The model shows good agreement with the available experimental data and has been modified to take into account the effect of subcooling.

  6. An experimental study on critical heat flux in a hemispherical narrow gap

    International Nuclear Information System (INIS)

    Park, R.J.; Lee, S.J.; Kang, K.H.; Kim, J.H.; Kim, S.B.; Kim, H.D.; Jeong, J.H.

    2000-01-01

    An experimental study of CHFG (Critical Heat Flux in Gap) has been performed to investigate the inherent cooling mechanism using distilled water and Freon R-113 in hemispherical narrow gaps. As a separate effect test of the CHFG test, a CCFL (Counter Current Flow Limit) test has been also performed to confirm the mechanism of the CHF in narrow annular gaps with large diameter. The CHFG test results have shown that an increase in the gap thickness leads to an increase in critical power. The pressure effect on the critical power was found to be much milder than predictions by CHF correlations of other studies. In the CCFL experiment, the occurrence of CCFL was correlated with the Wallis parameter, which was assumed to correspond to the critical power in the CHFG experiment. The measured values of critical power in the CHFG tests are much lower than CCFL experimental data and the predictions made by empirical CHF correlations. (author)

  7. Development of the heated length to diameter correction factor on critical heat flux using the artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Ho; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    With using artificial neural networks (ANNs), an analytical study related to the heated length effect on critical heat flux (CHF) has been carried out to make an improvement of the CHF prediction accuracy based on local condition correlations or table. It has been carried out to suggest a feasible criterion of the threshold length-to-diameter (L/D) value in which heated length could affect CHF. And within the criterion, a L/D correction factor has been developed through conventional regression. In order to validate the developed L/D correction factor, CHF experiments for various heated lengths have been carried out under low and intermediate pressure conditions. The developed threshold L/D correlation provides a new feasible criterion of L/D threshold value. The developed correction factor gives a reasonable accuracy for the original database, showing the error of -2.18% for average and 27.75% for RMS, and promising results for new experimental data. 7 refs., 12 figs., 1 tab. (Author)

  8. In-pile critical heat flux and post-dryout heat transfer measurements – A historical perspective

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com

    2017-06-15

    In the 1960s’ and 1970s’ Canada was a world leader in performing in-reactor heat transfer experiments on fuel bundles instrumented with miniature sheath thermocouples. Several Critical Heat Flux (CHF) and Post-CHF experiments were performed in Chalk River’s NRU and NRX reactors on water-cooled 3-, 18-, 19-, 21-, and 36-element fuel bundles. Most experiments were obtained at steady-state conditions, where the power was raised gradually from single-phase conditions up to the CHF and beyond. Occasionally, post-dryout temperatures up to 600 °C were maintained for several hours. In some tests, the fuel behaviour during loss-of-flow and blowdown transients was investigated – during these transients sheath temperatures could exceed 2000 °C. Because of the increasingly more stringent licensing requirements for in-pile heat transfer tests on instrumented fuel bundles, no in-pile CHF and post-dryout tests on fuel bundles have been performed anywhere in the world for the past 40 years. This paper provides details of these unique in-pile experiments and describes some of their heat transfer results.

  9. Development of the heated length to diameter correction factor on critical heat flux using the artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Ho; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    With using artificial neural networks (ANNs), an analytical study related to the heated length effect on critical heat flux (CHF) has been carried out to make an improvement of the CHF prediction accuracy based on local condition correlations or table. It has been carried out to suggest a feasible criterion of the threshold length-to-diameter (L/D) value in which heated length could affect CHF. And within the criterion, a L/D correction factor has been developed through conventional regression. In order to validate the developed L/D correction factor, CHF experiments for various heated lengths have been carried out under low and intermediate pressure conditions. The developed threshold L/D correlation provides a new feasible criterion of L/D threshold value. The developed correction factor gives a reasonable accuracy for the original database, showing the error of -2.18% for average and 27.75% for RMS, and promising results for new experimental data. 7 refs., 12 figs., 1 tab. (Author)

  10. Flow-Boiling Critical Heat Flux Experiments Performed in Reduced Gravity

    Science.gov (United States)

    Hasan, Mohammad M.; Mudawar, Issam

    2005-01-01

    Poor understanding of flow boiling in microgravity has recently emerged as a key obstacle to the development of many types of power generation and advanced life support systems intended for space exploration. The critical heat flux (CHF) is perhaps the most important thermal design parameter for boiling systems involving both heatflux-controlled devices and intense heat removal. Exceeding the CHF limit can lead to permanent damage, including physical burnout of the heat-dissipating device. The importance of the CHF limit creates an urgent need to develop predictive design tools to ensure both the safe and reliable operation of a two-phase thermal management system under the reduced-gravity (like that on the Moon and Mars) and microgravity environments of space. At present, very limited information is available on flow-boiling heat transfer and the CHF under these conditions.

  11. Exploratory heat transfer studies on critical elements of a proposed 6 GeV synchrotron

    International Nuclear Information System (INIS)

    Kuzay, T.M.; Knapp, G.S.

    1985-11-01

    Certain types of insertion devices for angiography, can produce extraordinarily large heat fluxes on critical components of a synchrotron beam line and its optics. The shutters, beam splitters, filters, and the first-stage monochromators all are subjected to large fluxes of radiation. The cooling requirements of such beam line components are approached in a comprehensive manner to identify the governing parameters from first principles. Analytical techniques have been used to study various methods of handling the heat loads using both liquid metal and water coolants for various potential heated geometries. It is found that when properly designed, liquid metal cooling can be much more efficient. In addition, composites and low Z surfaces have been considered. Also investigated are the heat transfer problems of the optical stages and rotating monochromators

  12. Exploratory heat transfer studies on critical elements of a proposed 6 GeV synchrotron

    International Nuclear Information System (INIS)

    Kuzay, T.M.; Knapp, G.S.

    1986-01-01

    Certain types of insertion devices for angiography can produce extraordinarily large heat fluxes on critical components of a synchrotron beam line and its optics. The shutters, beam splitters, filters, and the first-stage monochromators all are subjected to large fluxes of radiation. The cooling requirements of such beam line components are approached in a comprehensive manner to identify the governing parameters from first principles. Analytical techniques have been used to study various methods of handling the heat loads using both liquid metal and water coolants for various potential heated geometries. It is found that when properly designed, liquid metal cooling can be much more efficient. In addition, composites and low Z surfaces have been considered. Also investigated are the heat transfer problems of the optical stages and rotating monochromators

  13. Critical heat flux of subcooled flow boiling in narrow rectangular channels

    International Nuclear Information System (INIS)

    Kureta, Masatoshi; Akimoto, Hajime

    1999-01-01

    In relation to the high-heat-load devices such as a solid-target cooling channel of a high-intensity neutron source, burnout experiments were performed to obtain critical heat flux (CHF) data systematically for vertical upward flow in one-side heated rectangular channels. One of the objectives of this study was to study an extensibility of existing CHF correlations and models, which were proposed for a round tube, to rectangular channels for design calculation. Existing correlations and models were reviewed and compared with obtained data. Sudo's thin liquid layer dryout model, Griffel correlation and Bernath correlation were in good agreement with the experimental data for short-heated-length and low inlet water temperature conditions. (author)

  14. Does attenuated skin blood flow lower sweat rate and the critical environmental limit for heat balance during severe heat exposure?

    Science.gov (United States)

    Cramer, Matthew N; Gagnon, Daniel; Crandall, Craig G; Jay, Ollie

    2017-02-01

    What is the central question of this study? Does attenuated skin blood flow diminish sweating and reduce the critical environmental limit for heat balance, which indicates maximal heat loss potential, during severe heat stress? What is the main finding and its importance? Isosmotic hypovolaemia attenuated skin blood flow by ∼20% but did not result in different sweating rates, mean skin temperatures or critical environmental limits for heat balance compared with control and volume-infusion treatments, suggesting that the lower levels of skin blood flow commonly observed in aged and diseased populations may not diminish maximal whole-body heat dissipation. Attenuated skin blood flow (SkBF) is often assumed to impair core temperature (T c ) regulation. Profound pharmacologically induced reductions in SkBF (∼85%) lead to impaired sweating, but whether the smaller attenuations in SkBF (∼20%) more often associated with ageing and certain diseases lead to decrements in sweating and maximal heat loss potential is unknown. Seven healthy men (28 ± 4 years old) completed a 30 min equilibration period at 41°C and a vapour pressure (P a ) of 2.57 kPa followed by incremental steps in P a of 0.17 kPa every 6 min to 5.95 kPa. Differences in heat loss potential were assessed by identifying the critical vapour pressure (P crit ) at which an upward inflection in T c occurred. The following three separate treatments elicited changes in plasma volume to achieve three distinct levels of SkBF: control (CON); diuretic-induced isosmotic dehydration to lower SkBF (DEH); and continuous saline infusion to maintain SkBF (SAL). The T c , mean skin temperature (T sk ), heart rate, mean laser-Doppler flux (forearm and thigh; LDF mean ), mean local sweat rate (forearm and thigh; LSR mean ) and metabolic rate were measured. In DEH, a 14.2 ± 5.7% lower plasma volume resulted in a ∼20% lower LDF mean in perfusion units (PU) (DEH, 139 ± 23 PU; CON, 176 ± 22 PU; and SAL

  15. Effect of pressure on critical heat flux for water in an internally heated annulus

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Hibiki, Takashi; Nishihara, Hideaki

    2004-01-01

    It was pointed out earlier that existing CHF correlations based upon data for annuli at high pressures did not reproduce CHF very well at the atmospheric pressure. It appears to be necessary to investigate CHF at intermediate pressures to interpret the apparent discrepancy between CHFs at high and low pressures. In view of this an experiment was performed to obtain more information on CHF at intermediate pressures and the effect of pressure was discussed in the present study. It was revealed from this study that the effect of pressure on the CHF in the range from 0.1 to 1 MPa could be explained by the annular flow boundary and the critical quality. (author)

  16. Critical density and disruptions in α-heated thermonuclear Tokamak discharges

    International Nuclear Information System (INIS)

    Cotsaftis, M.; Firestone, M.; Wang, P.K.C.

    1985-02-01

    The study of existence of a critical density limit has been extended to the case of thermonuclear α-particle heated regime. To proceed, a 0-D model including sources and sinks affecting the evolution of ion and electron temperatures and of electron and α-particle densities with auxiliary neutral injected power has been developed. It is mainly shown when considering a Tokamak machine adapted for thermonuclear performances that, like in previous case, there is a critical density above which no other equilibrium point than 0 does exist. Temperatures then drop down the 0 past this critical value, leading to disruption. Analytic expression for critical density is given in terme of auxiliary projected power Psup(N). For Psup(N)=0, critical density value is low, but it increases fast enough for small Psup(N) to give a large safety margin once Psup(N) is moderate, much below the power required for reaching thermonuclear regime. So it is only at shutdown power periods that critical density can be crossed. But in this case, the heat content of particles in the discharge can significantly contribute to smooth out the temperature drop off. This typically operates up to the point where, due to change in magnetic islands configuration resulting from profile modification due to energy release at critical density crossing, heat transport doubles. Then on a fast thermal diffusion time scale, temperature drops now to a new equilibrium value, which can be made above the limiting value for which position control system of the plasma cannot forbid the plasma current to drop off itself, which is the important phenomenon of disruption. So on top of controls previously discussed, it is possible to use the α-particles themselves as a new preventive control against disruptions, making this phenomenon less dangerous for thermonuclear regime operation

  17. Critical heat flux predictions for the Sandia Annular Core Research Reactor

    International Nuclear Information System (INIS)

    Rao, D.V.; El-Genk, M.S.

    1994-08-01

    This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value of 2 MWt to 4 MWt. These predictions are based on the University of New Mexico (UNM) - CHF correlation, originally developed for uniformly heated vertical annuli. The UNM-CHF correlation is applicable to low-flow and low-pressure conditions, which are typical of those in the ACRR. The three hypotheses that examined the effect of the nonuniform axial heat flux distribution in the ACRR core are (1) the local conditions hypotheses, (2) the total power hypothesis, and (3) the global conditions hypothesis. These hypotheses, in conjunction with the UNM-CHF correlation, are used to estimate the CHF and CHFR in the ACRR. Because the total power hypothesis predictions of power per rod at CHF are approximately 15%-20% lower than those corresponding to saturation exit conditions, it can be concluded that the total power hypothesis considerably underestimates the CHF for nonuniformly heated geometries. This conclusion is in agreement with previous experimental results. The global conditions hypothesis, which is more conservative and more accurate of the other two, provides the most reliable predictions of CHF/CHFR for the ACRR. The global conditions hypothesis predictions of CHFR varied between 2.1 and 3.9, with the higher value corresponding to the lower water inlet temperature of 20 degrees C

  18. Experimental study of the critical density of heat flux in open channels cooled with helium - II

    International Nuclear Information System (INIS)

    Pron'ko, V.G.; Gorokhov, V.V.; Saverin, V.N.

    1981-01-01

    Experimental values of the critical density of a heat flux qsub(cr) in uniformly heated open channels cooled with helium-2 are reported for the first time. The experimental test bench and experimental element are described. Experimental data are obtained in cylindrical channels of 12Kh18N1OT steel with inner diameter d=0.8, 1.8; 2.8 mm and ratio l/d=20.8, 44, 85. The channel orientation has varied from vertical to horizontal position, the immersion depth - from 100, to 600 mm. It has been found that the heat transfer crisis propagation over the whole length of the channel with He-2 occurs practically instantaneously. The qsub(cr) value depends essentially on the bath liquid temperature, angle of inclivnation and relative length (l/d) of the channel with qsub(cr) approximately (l/d)sup(-1.5) being independent of the depth of channel immersion. The obtained values of critical density of a heat flux in channels are papproximately by an order less than those found for a great bulk of He-2. The results presented may be used for designing various types of devices cooled with He-2 and development of heat exchange theory in it [ru

  19. The Effect of Inclination Angle on Critical Heat Flux in a Locally Heated Liquid Film Moving Under the Action of Gas Flow in a Mini-Channel

    Directory of Open Access Journals (Sweden)

    Tkachenko Egor M.

    2016-01-01

    Full Text Available Intensively evaporating liquid films moving under the action of the cocurrent gas flow in a microchannel are promising for the use in modern cooling systems of semiconductor devices with high local heat release. This work has studied the dependence of the critical heat flux on the inclination angle of the channel. It has been found that the inclination angle in the plane parallel to the flow has no significant effect on the critical heat flux. Whereas the inclination angle in the plane perpendicular to the flow, on the contrary, significantly changes the value of the critical heat flux. However, for a given flow rate of fluid there is a threshold gas velocity at which the critical heat flux does not differ from the case of zero inclination of the channel. Thus, it can be concluded that the cooling system based on shear-driven liquid films can be potentially used when direction of the gravity changes.

  20. 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

  1. Study on tube critical heat flux data treatment with artificial neural networks

    International Nuclear Information System (INIS)

    Han Lang; Shan Jianqiang

    2005-01-01

    Prediction of the Critical Heat Flux (CHF) are analyzed by Artificial Neural Networks (ANN) to a CHF database for upward flow of water in uniformly heated vertical round tubes. The analysis is performed with three viewpoints hypothesis, i.e. for fixed inlet condition, fixed exit condition and local condition. Half of 6941 from CHF database data is trained through ANN, the trained ANN predicts the total CHF data better than any other conventional correlations, showing RMS error of 6.6%, 10.39% and 21.39%, respectively. (author)

  2. Identification of critical equipment and determination of operational limits in helium refrigerators under pulsed heat load

    Science.gov (United States)

    Dutta, Rohan; Ghosh, Parthasarathi; Chowdhury, Kanchan

    2014-01-01

    Large-scale helium refrigerators are subjected to pulsed heat load from tokamaks. As these plants are designed for constant heat loads, operation under such varying load may lead to instability in plants thereby tripping the operation of different equipment. To understand the behavior of the plant subjected to pulsed heat load, an existing plant of 120 W at 4.2 K and another large-scale plant of 18 kW at 4.2 K have been analyzed using a commercial process simulator Aspen Hysys®. A similar heat load characteristic has been applied in both quasi steady state and dynamic analysis to determine critical stages and equipment of these plants from operational point of view. It has been found that the coldest part of both the cycles consisting JT-stage and its preceding reverse Brayton stage are the most affected stages of the cycles. Further analysis of the above stages and constituting equipment revealed limits of operation with respect to variation of return stream flow rate resulted from such heat load variations. The observations on the outcome of the analysis can be used for devising techniques for steady operation of the plants subjected to pulsed heat load.

  3. Effect of orientation on critical heat flux in a 3-rod bundle cooled by Freon-12

    International Nuclear Information System (INIS)

    Dimmick, G.R.

    1979-06-01

    Critical heat flux measurements have been made in a segmented 3-rod test section cooled by Freon-12. Three test section orientations were used: vertical, inclined at 11 deg to the vertical, and horizontal. It was found that at flows of less than 2.5 Mg.m -2 .s -1 the transverse gravity force on the inclined and horizontal orientations reduced the magnitude of the critical heat flux and also changed the location of initial dryout when compared to the vertical data. To account for the effect of orientation during correlation of the data, the Reynolds number was modified to include a transverse gravity term. The minimum standard deviation for the data from the three orientations combined was 3.4 percent and less than 3.7 percent for the three orientations separately. (author)

  4. Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry

    International Nuclear Information System (INIS)

    Staron, E.

    1996-01-01

    Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs

  5. Effect of flow obstacles with various leading and trailing edges on critical heat flux

    International Nuclear Information System (INIS)

    Pioro, I.L.; Groeneveld, D.C.; Groeneveld, D.C.; Cheng, S.C.; Antoshko, Y.V.

    2001-01-01

    A joint investigation has been performed by the University of Ottawa and Chalk River Laboratories that examined the effect of the shape of the leading and trailing edges of the turbulence enhancing devices ('flow obstacles') on critical heat flux (CHF). The objective of this study was to gain a better overall understanding of the limit of CHF improvement for various obstacle designs and the impact of flow conditions on the improvements. (author)

  6. Critical heat flux and flow pattern for water flow in annular geometry

    International Nuclear Information System (INIS)

    Park, Jae Wook; Baek, Won Pil; Chang, Soon Heung

    1996-01-01

    An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced-circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m, inner diameter = 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for upward flow

  7. An improved mechanistic critical heat flux model for subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer`s equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error. 9 refs., 5 figs. (Author)

  8. An improved mechanistic critical heat flux model for subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer`s equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error. 9 refs., 5 figs. (Author)

  9. Electron heating caused by parametrically driven turbulence near the critical density

    International Nuclear Information System (INIS)

    Mizuno, K.; DeGroot, J.S.; Estabrook, K.G.

    1986-01-01

    Microwave-driven experiments and particle simulation calculations are presented that model s-polarized laser light incident on a pellet. In the microwave experiments, the incident microwaves are observed to decay into ion and electron waves near the critical density if the microwave power is above a well-defined threshold. Significant absorption, thermal electron heating, and hot electron generation are observed for microwave powers above a few times threshold. Strong absorption, strong profile modification, strongly heated hot electrons with a Maxwellian distribution, a hot-electron temperature that increases slowly with power, and a hot-electron density that is almost constant, are all observed in both the microwave experiments and simulation calculations for high powers. In addition, the thermal electrons are strongly heated for high powers in the microwave experiments

  10. Prediction of critical heat flux by a new local condition hypothesis

    International Nuclear Information System (INIS)

    Im, J. H.; Jun, K. D.; Sim, J. W.; Deng, Zhijian

    1998-01-01

    Critical Heat Flux(CHF) was predicted for uniformly heated vertical round tube by a new local condition hypothesis which incorporates a local true steam quality. This model successfully overcame the difficulties in predicted the subcooled and quality CHF by the thermodynamic equilibrium quality. The local true steam quality is a dependent variable of the thermodynamic equilibrium quality at the exit and the quality at the Onset of Significant Vaporization(OSV). The exit thermodynamic equilibrium quality was obtained from the heat balance, and the quality at OSV was obtained from the Saha-Zuber correlation. In the past CHF has been predicted by the experimental correlation based on local or non-local condition hypothesis. This preliminary study showed that all the available world data on uniform CHF could be predicted by the model based on the local condition hypothesis

  11. Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

    International Nuclear Information System (INIS)

    Hwang, J. S.; Suh, K. Y.

    2009-01-01

    The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

  12. Measurement of critical heat flux in narrow gap with two-dimensional slices

    International Nuclear Information System (INIS)

    Kim, Yong Hoon; Kim, Sung Joong; Noh, Sang Woo; Suh, Kune Y.

    2002-01-01

    A cooling mechanism due to boiling in a gap between the debris crust and the reactor pressure vessel (RPV) wall was proposed for the TMI-2 reactor accident analysis. If there is enough heat transfer through the gap to cool the outer surface of the debris and the inner surface of the wall, the RPV wall may preserve its integrity during a severe core melt accident. If the heat removal through gap cooling relative to the counter-current flow limitation (CCFL) is pronounced, the safety margin of the reactor can be far greater than what had been previously known in the severe accident management arena. Should a severe accident take place, the RPV integrity will be maintained because of the inherent nature of degraded core coolability inside the lower head due to boiling in a narrow gap between the debris crust and the RPV wall. As a defense-in-depth measure, the heat removal capability by gap cooling coupled with external cooling can be examined for the Korean Standard Nuclear Power Plant (KSNPP) and the Advanced Power Reactor 1400MWe (APR1400) in light of the TMI-2 vessel survival. A number of studies were carried out to investigate the complex heat transfer mechanisms for the debris cooling in the lower plenum. However, these heat transfer mechanisms have not been clearly understood yet. The CHFG (Critical Heat Flux in Gap) experiments at KAERI were carried out to develop the critical heat flux (CHF) correlation in a hemispherical gap, which is the upper limit of the heat transfer. According to the CHFG experiments performed with a pool boiling condition, the CHF in a parallel gap was reduced by 1/30 compared with the value measured in the open pool boiling condition. The correlation developed from the CHFG experiment is based on the fact that the CHF in a hemispherical gap is governed by the CCFL and a Kutateladze type CCFL parameter correlates CCFL data well in hemispherical gap geometry. However, the results of the CHFG experiments appear to be limited in their

  13. Influence of variable heat transfer coefficient of fireworks and crackers on thermal explosion critical ambient temperature and time to ignition

    Directory of Open Access Journals (Sweden)

    Guo Zerong

    2016-01-01

    Full Text Available To study the effect of variable heat transfer coefficient of fireworks and crackers on thermal explosion critical ambient temperature and time to ignition, considering the heat transfer coefficient as the power function of temperature, mathematical thermal explosion steady state and unsteady-state model of finite cylindrical fireworks and crackers with complex shell structures are established based on two-dimensional steady state thermal explosion theory. The influence of variable heat transfer coefficient on thermal explosion critical ambient temperature and time to ignition are analyzed. When heat transfer coefficient is changing with temperature and in the condition of natural convection heat transfer, critical ambient temperature lessen, thermal explosion time to ignition shorten. If ambient temperature is close to critical ambient temperature, the influence of variable heat transfer coefficient on time to ignition become large. For firework with inner barrel in example analysis, the critical ambient temperature of propellant is 463.88 K and the time to ignition is 4054.9s at 466 K, 0.26 K and 450.8s less than without considering the change of heat transfer coefficient respectively. The calculation results show that the influence of variable heat transfer coefficient on thermal explosion time to ignition is greater in this example. Therefore, the effect of variable heat transfer coefficient should be considered into thermal safety evaluation of fireworks to reduce potential safety hazard.

  14. An investigation of critical heat fluxes in vertical tubes internally cooled by Freon-12. Part I - Critical heat flux experiments with axially uniform and non-uniform heating and comparisons of data with selected correlations

    International Nuclear Information System (INIS)

    Green, W.J.; Stevens, J.R.

    1981-08-01

    Experiments have been performed using vertical heated tubes, cooled internally by Freon-12, to determine critical heat fluxes (CHFs) for both a uniformly heated section and an exit region with a separately controlled power supply. Heated lengths of the main separately were 2870 mm (8.48 and 16.76 mm tube bores) and 3700 mm (for 21.34 mm tube bore); heated length of the exit section was 230 mm. Coolant pressures, exit qualities and mass fluxes were in the range 0.9 to 1.3 MPa, 0.19 to 0.86, and 380 to 2800 kg m -2 s -1 , respectively. The data have been compared with published empirical correlations specifically formulated to predict CHFs in Freon-cooled, vertical tubes; relevant published CHF data have also been compared with these correlations. These comparisons show that, even over the ranges of conditions for which the correlations were developed, predicted values are only accurate to within +-20 per cent. Moreover, as mass fluxes increase above 3500 kg m -2 s -1 , the modified Groeneveld correlation becomes increasingly inadequate, and the Bertoletti and modified Bertoletti correlations under-predict CHF values by increasing amounts. At mass fluxes below 750 kg m -2 s -1 the Bertoletti correlations exhibit increasing inaccuracy with a decrease in mass flux. For non-uniform heating, the correlations are at variance with the experimental data

  15. Critical heat flux analysis on change of plate temperature and cooling water flow rate for rectangular narrow gap with bilateral-heated cases

    International Nuclear Information System (INIS)

    M Hadi Kusuma; Mulya Juarsa; Anhar Riza Antariksawan

    2013-01-01

    Boiling heat transfer phenomena on rectangular narrow gap was related to the safety of nuclear reactors. Research done in order to study the safety of nuclear reactors in particular relating to boiling heat transfer and useful on the improvement of next-generation reactor designs. The research focused on calculation of the heat flux during the cooling process in rectangular narrow gap size 1.0 mm. with initial temperatures 200°C. 400°C, and 600°C, also the flow rates of cooling water 0,1 liters/second. 0,2 liters/second. and 0,3 liters/second. Experiments carried out by injecting water at a certain flow rate with the water temperature 85°C. Transient temperature measurement data recorded by the data acquisition system. Transient temperature measurement data is used to calculate the flux of heat gain is then used to obtain the heat transfer coefficient. This research aimed to obtain the correlation between critical heat flux and heat transfer coefficient to changes in temperatures and water flow rates for bilaterally-heated cases on rectangular narrow gap. The results obtained for a constant cooling water flow rate, critical heat flux will increase when hot plate temperature also increased. While on a constant hot plate temperature, coefficient heat transfer will increase when cooling water flow rate also increased. Thus it can be said that the cooling water flow rate and temperature of the hot plate has a significant effect on the critical heat flux and heat transfer coefficient resulted in quenching process of vertical rectangular narrow gap with double-heated cases. (author)

  16. Adaptation of a Freon-12 critical heat flux correlation to correlate water data from uniformly heated vertical tubes. Part I: Based on critical heat flux data for water at pressures of 3 to 14 MPa

    International Nuclear Information System (INIS)

    Green, W.J.

    1981-12-01

    Comparisons have been made between experimental critical heat flux (CHF) data for upflow of water in uniformly heated vertical tubes and values calculated from an empirical CHF correlation developed from Freon-12 data. When this correlation is re-evaluated to account for vapour Prandtl number effects, very good agreement is obtained between experimental data and calculated values over a wide range of coolant conditions. Comparison of values calculated from the revised correlation with 2063 sets of CHF data obtained from experiments with water in vertical, uniformly heated tubes shows a mean ratio of the calculated to experimental CHF of 0.82 and an r.m.s. error of 5.8 per cent for the following coolant conditions: (1) local pressure of 3.4 to 12 MPa; (2) mass flux greater than approx. 300 kg s -1 m -2 , and (3) thermal equilibrium value of exit quality greater than 0.1

  17. Computational multi-fluid dynamics predictions of critical heat flux in boiling flow

    International Nuclear Information System (INIS)

    Mimouni, S.; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.

    2016-01-01

    Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune_CFD code. • The model has been validated against 150 tests. • Neptune_CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.

  18. Computational multi-fluid dynamics predictions of critical heat flux in boiling flow

    Energy Technology Data Exchange (ETDEWEB)

    Mimouni, S., E-mail: stephane.mimouni@edf.fr; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.

    2016-04-01

    Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune-CFD code. • The model has been validated against 150 tests. • Neptune-CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.

  19. A new mechanistic model of critical heat flux in forced-convection subcooled boiling

    International Nuclear Information System (INIS)

    Alajbegovic, A.; Kurul, N.; Podowski, M.Z.; Drew, D.A.; Lahey, R.T. Jr.

    1997-10-01

    Because of its practical importance and various industrial applications, the process of subcooled flow boiling has attracted a lot of attention in the research community in the past. However, the existing models are primarily phenomenological and are based on correlating experimental data rather than on a first-principle analysis of the governing physical phenomena. Even though the mechanisms leading to critical heat flux (CHF) are very complex, the recent progress in the understanding of local phenomena of multiphase flow and heat transfer, combined with the development of mathematical models and advanced Computational Fluid Dynamics (CFD) methods, makes analytical predictions of CHF quite feasible. Various mechanisms leading to CHF in subcooled boiling have been investigated. A new model for the predictions of the onset of CHF has been developed. This new model has been coupled with the overall boiling channel model, numerically implemented in the CFX 4 computer code, tested and validated against the experimental data of Hino and Ueda. The predicted critical heat flux for various channel operating conditions shows good agreement with the measurements using the aforementioned closure laws for the various local phenomena governing nucleation and bubble departure from the wall. The observed differences are consistent with typical uncertainties associated with CHF data

  20. Critical heat flux for flow boiling of water in mini-channels

    International Nuclear Information System (INIS)

    Zhang, Weizhong; Mishima, Kaichiro; Hibiki, Takashi

    2007-01-01

    Critical heat flux (CHF) is a limiting factor when flow boiling is applied to dissipate high heat flux in mini-channels. In view of practical importance of critical heat flux correlations in engineering design and prediction, this study presents an evaluation of existing CHF correlations for flow boiling of water with available databases taken from small-diameter tubes, and then develops a new, simple CHF correlation for flow boiling in mini-channel. Three correlations by Bowring, Katto and Shah are evaluated with available CHF data in the literature for saturated flow boiling, and three correlations by Inasaka-Nariai, Celata et al. and Hall-Mudawar evaluated with the CHF data for subcooled flow boiling. The Hall-Mudawar correlation and the Shah correlation appear to be the most reliable tools for CHF prediction in the subcooled and saturated flow boiling regions, respectively. In order to avoid the defect of predictive discontinuities often encountered when applying previous correlations, a simple, nondimensional, inlet conditions dependent CHF correlation for saturated flow boiling has been formulated. Its functional form is determined by application of the artificial neural network and parametric trend analyses to the collected database. Superiority of this new correlation has been verified by the collected database. It has a mean deviation of 16.8% for this collected databank, smallest among all tested correlations. Compared to many inordinately complex correlations, this new correlation consists only of one single equation. (author)

  1. Flow visualization and critical heat flux measurement of a boundary layer pool boiling process

    International Nuclear Information System (INIS)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C.; Shiah, S.W.

    1998-01-01

    As part of the effort to evaluate the concept of external passive cooling of core melt by cavity flooding under severe accident conditions, a subscale boundary layer boiling (SBLB) facility, consisting of a pressurized water tank with a condenser unit, a heated hemispherical test vessel, and a data acquisition/photographic system, was developed to simulate the boiling process on the external bottom surface of a fully submerged reactor vessel. Transient quenching and steady-state boiling experiments were conducted in the facility to measure the local critical heat flux (CHF) and observe the underlying mechanisms under well controlled saturated and subcooled conditions. Large elongated vapor slugs were observed in the bottom region of the vessel which gave rise to strong upstream influences in the resulting two-phase liquid-vapor boundary layer flow along the vessel outer surface. The local CHF values deduced from the transient quenching data appeared to be very close to those obtained in the steady-state boiling experiments. Comparison of the SBLB data was made with available 2-D full-scale data and the differences were found to be rather small except in a region near the bottom center of the vessel. The angular position of the vessel outer surface and the degree of subcooling of water had dominant effects on the local critical heat flux. They totally dwarfed the effect of the physical dimensions of the test vessels. (author)

  2. Critical heat flux tests for a 12 finned-element assembly

    Energy Technology Data Exchange (ETDEWEB)

    Yang, J., E-mail: Jun.Yang@cnl.ca; Groeneveld, D.C.; Yuan, L.Q.

    2017-03-15

    Highlights: • CHF tests for a 12 finned-fuel-element assembly at highly subcooled conditions. • Test approach to maximize experimental information and minimize heater failures. • Three series of tests were completed in vertical upward light water flow. • Bundle simulators of two axial power profiles and three heated lengths were tested. • Results confirm that the prediction method predicts lower CHF values than measured. - Abstract: An experimental study was undertaken to provide relevant data to validate the current critical heat flux (CHF) prediction method of the NRU driver fuel for safety analysis, i.e., to confirm no CHF occurrence below the predicted values. The NRU driver fuel assembly consists of twelve finned fuel elements arranged in two rings – three in the inner ring and nine in the outer ring. To satisfy the experimental objective tests at very high heat fluxes, very high mass velocities, and high subcoolings were conducted where the CHF mechanism is the departure from nucleate boiling (DNB). Such a CHF experiment can be very difficult, costly and time consuming since failure of the heating surface due to rupture or melting (physical burnout) is expected when the DNB type of CHF is reached. A novel experimental approach has been developed to maximize the amount of relevant experimental information on safe operating conditions in the tests, and to minimize any possible heater failures that inherently accompany the CHF occurrence at these conditions. Three series of tests using electrically heated NRU driver fuel simulators with three heated lengths and two axial power profiles (or axial heat flux distribution (AFD)) were completed in vertical upward light water flow. Each series of tests covered two mass flow rates, several heat flux levels, and local subcoolings that bound the ranges of interest for the analysis of postulated slow loss-of-regulation accident (LORA) and loss-of-flow accident (LOFA) scenarios. Tests for each mass flow rate of

  3. Critical heat flux for APR1400 lower head vessel during a severe accident

    International Nuclear Information System (INIS)

    Noh, Sang W.; Suh, Kune Y.

    2013-01-01

    Highlights: ► Studied boiling on downward-facing hemispherical vessel with asymmetric thermal insulator. ► Scaled the APR1400 lower head linearly down by 1/10 including ICI tubes and shear keys. ► Performed thermal analysis using ANSYS V11.0 to determine the internal temperature and heat flux. ► Performed tests to obtain the CHF with saturated demineralized water at atmospheric pressure. ► Measured CHF accounting for 3D random flow effect expected in the APR1400 application. -- Abstract: Corium Ablation Stopper Apparatus (CASA) has a downward-facing hemispherical vessel and geometrically asymmetric thermal insulator of the Advanced Power Reactor 1400 MWe (APR1400) scaled linearly down by 1/10, as well as sixty-one in-core instrumentation (ICI) tubes and four shear keys. The heated vessel plays a pivotal role in CASA depending on the configuration of the oxide pool and metal layer to bring about the focusing effect expected of a molten pool in the lower head during a severe accident. The heated vessel was designed through a trial-and-error method and thermal analysis. Thermal analysis was performed using ANSYS V11.0 to investigate the effect of the internal temperature and heat flux on the integral hemispherical copper vessel. The CASA tests were carried out to obtain the critical heat flux (CHF) with saturated and demineralized water at the atmospheric pressure (0.1 MPa). The CHF in the metal layer through the hemispherical channel was found to be lower than that in the ULPU-2400 configuration V data through the streamlined thermal insulator. The experimental CHF was measured and obtained through the CASA hemispherical heated surface accounting for the three-dimensional random flow effect expected in the APR1400 application

  4. An experimental study on the flow instabilities and critical heat flux under natural circulation

    International Nuclear Information System (INIS)

    Kim, Yun Il

    1993-02-01

    This study has been carried out to investigate the hydrodynamic stabilities of natural circulation and to analyze Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations and the average mass flux have been observed. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. The results show that the flow instability has strongly dependent on geometric conditions and operating parameters, the inlet temperature and the heat flux of test section. It was found that unstable region for the heat flux and the inlet temperature exists between the single-phase stable region of low heat and low inlet temperature and the two-phase stable region of very high heat flux and high inlet temperature. The CHF data from the natural and forced circulation experiments have been compared each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions were low flow less than 70 kg/m 2 s of water in vertical round tube with diameter of 0.008m at near atmospheric pressure. In this study, no difference in CHF values is observed between natural and fored circulation. Since low flow usually has the oscillation characteristic of relatively low amplitude and high frequency, the effect of the flow instabilities on the CHF seems to be negligible

  5. Effects of Microencapsulated Phase Change Material (MPCM) on Critical Heat Flux in Pool Boiling

    International Nuclear Information System (INIS)

    Park, Sung Dae; Kim, Seong Man; Kang, Sarah; Lee, Seung Won; Seo, Han; Bang, In Cheol

    2011-01-01

    Thermal power is limited by critical heat flux (CHF) in the nuclear power plant. And the in-vessel retention by external reactor vessel cooling (IVR-ERVC) is applied in some nuclear power plants; AP600, AP1000, Loviisa and APR1400. The heat removal capacity of IVR-ERVC is also restricted by CHF. So, it is essential to get CHF margin to improve an economics and a safety of the plant. There are some typical approaches to enhance CHF: vibrating the heater or fluid, coating with porous media on the heater surface, applying an electric field. The recent study related to the CHF is focus on using the nanofluid. In this paper, the new approach was investigated by using the microencapsulated phase change material (MPCM). MPCM is the particles whose diameter is from 0.1μm to 1000μm. The MPCM consists of the core material and the shell material. The core material can be solid, liquid, gas or even the mixture. The solid paraffin is the best candidate as the core material due to its stable chemical and thermal properties. And the shell material is generally synthesized polymer of about several micrometers in thickness. The most interesting feature of the MPCM is that the latent heat associated with the solid-liquid phase change is related to the heat transfer. When the MPCM is dispersed into the carrier fluid, a kind of suspension named as microencapsulated phase change slurry (MPCS) is formed. The study on the MPCS was conducted in field of both the heat transfer fluids and energy storage media. It is inspired by the fact that the latent heat can serve distribution to the additional CHF margin. The purpose of this work is to confirm whether or not the CHF is enhanced

  6. Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

    2017-05-15

    Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

  7. Study on subcooled-forced flow boiling heat transfer and critical heat flux of solid particle-water two-phase mixture

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Mochizuki, Manabu; Ohtake, Hiroyasu

    1999-01-01

    The effect of solid particle introduction on forced flow boiling and the critical heat flux was examined for the mixture of subcooled-water and 0.6 mm glass beads. When the particles were introduced, the growth on of a superheated layer near a wall seemed to be suppressed and the onset of nucleate boiling was delayed. The particles tempted for bubbles to condense at nucleation sites, and then the initiation of net vapor generation was also delayed and sifted to a high wall-superheat region. The nucleate boiling heat transfer was augmented by the particles, which considered to be caused by the combination of the suppression of the superheated layer growth and the promotion of the condensation and dissipation of the bubbles. The wall superheat at the critical heat flux condition was sifted to a high wall superheat region and the critical heat flux itself was also elevated a little. (author)

  8. Heat adaptation of bioabsorbable craniofacial plates: a critical review of science and technology.

    Science.gov (United States)

    Pietrzak, William S

    2009-11-01

    Bioabsorbable fixation plates often require adaptation to the bone. This is typically accomplished by heating the plates to above the glass transition temperature and placing the softened plates against the bone or a prebent template until cool. Upon cooling, the plates regain stiffness and can be attached to bone to obtain anatomic fixation. This procedure is both efficient and effective and has been used throughout the craniofacial skeleton. There are many types of equipment available to heat the plates, each with advantages and disadvantages. Although a conceptually simple process, there are several nuances that have been reported in the literature, including transient effects on plate mechanical properties, memory effects, differences between wet and dry heating, and others. Upon the backdrop of the overwhelming clinical success of heat adaptation, this review critically evaluates the method and provides a comprehensive examination and explanation of the basic science and technology involved. This should help give surgeons a better understanding of the process that can help improve their use and further advance the technology.

  9. Critical heat flux concerns during the flow instability phase of a DEGB LOCA

    International Nuclear Information System (INIS)

    Shadday, M.A. Jr.

    1990-08-01

    Arguments are presented that support the proposal that a separate burnout risk analysis, for the Flow Instability (FI) phase of a LOCA, not be required for reactor restart. With expected reactor power limits, flow instability will occur before critical heat flux (CHF). Since FI power limits preclude the occurrence of flow instability in a bounding accident, a DEGB LOCA, the risk of CHF and attendant burnout is negligible. A review of RDAP data revealed that in the past reactor assemblies operated at flow and power conditions similar to those expected in a LOCA without burnout occurring. This is strong bounding empirical evidence, without the scaling concerns of laboratory experiments. A bounding analysis of the influences of assembly non-idealities on CHF, power tilts, and channel eccentricity, is included. The margin between operating heat fluxes, during the postulated LOCA, and CHF was quantified by scoping calculations. Based on measured azimuthal power variations, the local heat flux would have to be more than 20 standard deviations above the calculated mean heat flux for CHF to occur

  10. Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; McRae, G.A.; Coleman, C.E.; Nitheanandan, T.; Sanderson, D.B.

    1999-10-01

    In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

  11. Doubling of critical heat flux using a grapheme oxide nanofluid and its repeatabiltiy

    International Nuclear Information System (INIS)

    Moon, Sung Bo; Bang, In Cheol

    2013-01-01

    CHF(Critical Heat Flux : heat flux which makes dramatic increase of temperature on heater surface) is one of the most important phenomena in the thermal hydraulic system. High CHF makes more thermal margin of heat transfer. This makes high efficiency and safety of power plant especially in nuclear power plant. Much smaller danger can be concerned to public society like radioactive material leakage in the accidents. Graphene Oxide which can be deposited on the heater surface makes nano-scale structures with enhancing thermal limit of heater. Three major models of enhancing limit of heater have been concerned in many heat transfer studies. In this study, wettability that is about ability to wet on surface and thermal activity which is about thermal property of coated layer are concerned to analyze the mechanism of CHF enhancing. Also, chemical reduction of Graphene Oxide(GO) to Reduced Graphene Oxide(RGO) on the surface will be concerned with one reason of changing wettability of nano-scale structure on the heater surface. We used GO nanofluid 0.001 volume percent. Two models are compared to explain how CHF is enhanced. Results show wettability increased with slightly reduced GO and structure. And in thermal activity model, the most powerful term, thickness of layer, is too small to affect thermal activity. It has low ability to explain how GO nanofluid can enhance CHF

  12. Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, X., E-mail: xavier.courtois@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, F. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint-Paul-Lez-Durance (France); Richou, M.; Cantone, V. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Constans, S. [AREVA-NP, Le Creusot (France)

    2013-10-15

    Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear. Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.

  13. Critical heat flux analysis and R and D for the design of the ITER divertor

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Merola, M.; Tivey, R.; Vieider, G.; Schlosser, J.; Driemeyer, D.; Escourbiac, F.; Grigoriev, S.; Youchison, D.

    1999-01-01

    The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW/m 2 over ∼10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R and D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R and D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of ∼1.4 or more for all divertor components. (orig.)

  14. Flow induced vibration characteristics in 2X3 bundle critical heat flux experiment

    International Nuclear Information System (INIS)

    Kim, Dae Hun; Chang, Soon Heung

    2005-01-01

    Above a certain heat flux, the liquid can no longer permanently wet the heater surface. This situation leads to an inordinate decrease in the surface heat transfer. This heat flux is commonly referred to as the critical heat flux (CHF). The CHF in nuclear reactors is one of the important thermal hydraulic parameters limiting the available power. Flow induced vibration (FIV) is the vibration caused by a fluid flowing around a body. In the fluid flowing system, FIV occurred by structures and flow condition. Many structures in nuclear power plant system are designed to prevent from structure failure due to FIV. Recently, Hibiki and Ishii (1998) carried out an experimental investigation on the effect of flow-induced vibration (FIV) on two-phase flow structure in vertical tube and reported that the FIV drastically changed the void fraction profiles. The void fraction profiles is one of the important parameter for determining CHF. Therefore, the investigation on the effect of FIV on CHF are needed. The research on FIV characteristics detection during CHF experiment in 2X3 bundle using R-134a has been carried out in KAIST. Using the results new FIV correlation in 2-pahse turbulent flow are suggested after finding out relation between CHF and dynamic pressure fluctuation value

  15. A theoretical prediction of critical heat flux in saturated pool boiling during power transients

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Nelson, R.A.; Gunnerson, F.S.

    1987-01-01

    Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions is of fundamental interest in the design, operation, and safety of boiling and two-phase flow devices. Presented within this paper are the results of a comprehensive theoretical study specifically conducted to model transient CHF behavior in saturated pool boiling. Thermal energy conduction within a heating element and its influence on the CHF are also discussed. The resultant theory provides new insight into the basic physics of the CHF phenomenon and indicates favorable agreement with the experimental data from cylindrical heaters with small radii. However, the flat-ribbon heater data compared poorly with the present theory, although the general trend was predicted. Finally, various factors that affect the discrepency between the data and the theory are listed

  16. A formal approach for the prediction of the critical heat flux in subcooled water

    Energy Technology Data Exchange (ETDEWEB)

    Lombardi, C. [Polytechnic of Milan (Italy)

    1995-09-01

    The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon.

  17. Study on critical heat flux based on wavelet transform in rectangular narrow channels

    International Nuclear Information System (INIS)

    Zhou Tao; Ju Zhongyun; Zhang Lei; Li Jingjing; Sheng Cheng; Xiao Zejun

    2014-01-01

    Critical heat flux is very important for the safety of nuclear reactor, and observing temperature rise rate is a feasible method. The wavelet transform is used to analyze the CHF temperature rise curves in rectangular narrow channels, which can remove relative weaker interference and effectively judge CHF. Rectangular narrow channel can strengthen heat transfer and reduce CHF, whose characteristics are proved by temperature rise curves analyzed by wavelet transform. Respectively applying Daubechies function and Haar function is to guarantee the accuracy of the wavelet analysis, and Daubechies function is more accurate than Haar function in the detail signal processing from results. While the wavelet analysis and experimental results are compared and found in good agreement with the experimental results. (authors)

  18. Study on critical heat flux based on wavelet transform in rectangular narrow channels

    International Nuclear Information System (INIS)

    Zhou Tao; Ju Zhongyun; Zhang Lei; Li Jingjing; Sheng Cheng; Xiao Zejun

    2014-01-01

    Critical heat flux is very important for nuclear reactor safety, and observing temperature rise rate is a feasible method. Through using the wavelet transform to analyze the CHF temperature rise curves in rectangular narrow channels, it can remove relative weaker interference and effectively judge CHF. Rectangular narrow channel can strengthen heat transfer and reduce CHF, whose characteristics are proved by, temperature rise curves analyzed by wavelet transform. Respectively applying Daubechies function and Haar function is for guarantee the accuracy of the wavelet analysis, and Daubechies function is more accurate than Haar function in the detail signal processing from results. While the wavelet analysis and experimental results are compared and found in good agreement with the experimental results. (authors)

  19. Critical heat flux of forced convection boiling in an oscilating acceleration field. Pt. 1

    International Nuclear Information System (INIS)

    Otsuji, T.; Kurosawa, A.

    1982-01-01

    The influence of periodically varying acceleration on critical heat flux (CHF) of Freon-113 flowing upward in a uniformly heated vertical annular channel has been studied experimentally. The freon loop was oscillated vertically to determine the ratio of CHF in the oscillating acceleration field to the corresponding stationary value. The amplitude of inlet flow oscillation induced by variation of acceleration, which causes early CHF, is proportional to the acceleration amplitude. The dependence of inlet flow rate on the oscillating acceleration decreases with increasing inlet subcooling, and no oscillation of inlet flow is observed in the case of negative exit quality (subcooled boiling). Nevertheless the degradation of CHF is more remarkable in the low quality region. This result suggests the necessity to introduce an other mechanism of early CHF than flow oscillation. (orig.)

  20. A continuum self organized critically model of turbulent heat transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Tangri, V; Das, A; Kaw, P; Singh, R [Institute for Plasma Research, Gandhinagar (India)

    2003-09-01

    Based on the now well known and experimentally observed critical gradient length (R/L{sub Te} = RT/{nabla}T) in tokamaks, we present a continuum one dimensional model for explaining self organized heat transport in tokamaks. Key parameters of this model include a novel hysteresis parameter which ensures that the switch of heat transport coefficient {chi} upwards and downwards takes place at two different values of R/L{sub Te}. Extensive numerical simulations of this model reproduce many features of present day tokamaks such as submarginal temperature profiles, intermittent transport events, 1/f scaling of the frequency spectra, propagating fronts, etc. This model utilises a minimal set of phenomenological parameters, which may be determined from experiments and/or simulations. Analytical and physical understanding of the observed features has also been attempted. (author)

  1. A formal approach for the prediction of the critical heat flux in subcooled water

    International Nuclear Information System (INIS)

    Lombardi, C.

    1995-01-01

    The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon

  2. Review of issues relating to enhancement of critical heat flux in nuclear reactors - 15220

    International Nuclear Information System (INIS)

    Hasan, S.M.K.; Chowdhury, M.A.Z.; Sarkar, M.A.R.

    2015-01-01

    Critical heat flux (CHF) is manifested by sharp reduction in heat transfer coefficient and a drastic rise in fuel rod surface temperature, which may lead to failure of the fuel cladding and is of immense importance in reactor design and operation. In this paper, the existing literature on CHF is studied and analyzed. Different modeling techniques such as geometry modeling and fluid modeling are highlighted. CHF enhancement techniques such as swirl flow, porous coating, nano-fluids or mechanical vibrations are analyzed and their suitability for different applications is discussed. The advances in prediction methods for tubes and rod bundles are also presented. CHF correlations and their applications have been compiled to estimate the CHF for fluids other than water or other geometries. Finally, future research needs are identified and presented. (authors)

  3. Influence of test tube material on subcooled flow boiling critical heat flux in short vertical tube

    International Nuclear Information System (INIS)

    Hata, Koichi; Shiotsu, Masahiro; Noda, Nobuaki

    2007-01-01

    The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u=4.0 to 13.3 m/s), the inlet subcoolings (ΔT sub,in =48.6 to 154.7 K), the inlet pressure (P in =735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tube of inner diameter (d=6 mm), heated length (L=66 mm) and L/d=11 with the inner surface of rough finished (Surface roughness, Ra=3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tube of d=6 mm, L=60 mm and L/d=10 with Ra=0.18 μm and the Platinum (Pt) test tubes of d=3 and 6 mm, L=66.5 and 69.6 mm, and L/d=22.2 and 11.6 respectively with Ra=0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcoolings. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (author)

  4. Influence of Test Tube Material on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube

    International Nuclear Information System (INIS)

    Koichi Hata; Masahiro Shiotsu; Nobuaki Noda

    2006-01-01

    The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcooling (ΔT sub,in = 48.6 to 154.7 K), the inlet pressure (P in = 735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/t), t = 10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tubes of inner diameters (d = 6 mm), heated lengths (L = 66 mm) and L/d = 11 with the inner surface of rough finished (Surface roughness, R a = 3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tubes of d = 6 mm, L = 60 mm and L/d = 10 with R a = 0.18 μm and the Platinum (Pt) test tubes of d = 3 and 6 mm, L = 66.5 and 69.6 mm, and L/d 22.2 and 11.6 respectively with R a = 0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcooling. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (authors)

  5. Critical heat flux for free convection boiling in thin rectangular channels

    International Nuclear Information System (INIS)

    Cheng, Lap Y.; Tichler, P.R.

    1991-01-01

    A review of the experimental data on free convection boiling critical heat flux (CHF) in vertical rectangular channels reveals three mechanisms of burnout. They are the pool boiling limit, the circulation limit, and the flooding limit associated with a transition in flow regime from churn to annular flow. The dominance of a particular mechanism depends on the dimensions of the channel. Analytical models were developed for each free convection boiling limit. Limited agreement with data is observed. A CHF correlation, which is valid for a wide range of gap sizes, was constructed from the CHFs calculated according to the three mechanisms of burnout. 17 refs., 7 figs

  6. Experimental study on the critical heat flux in a varying acceleration field, (1)

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Otsuji, Tomoo; Ikawa, Masahiro; Kurosawa, Akira.

    1988-12-01

    It is very important for the thermohydraulic design and for the safety assesement of marine reactors, to understand the effect of varying acceleration induced by ship motion on critical heart flux. The purpose of this joint study is to examine quantitatively the influence of varying acceleration on the behavior of bubbles. In the experiment, FREON-113 was used as working fluid. This report describes some experimental results; measurements of void fraction and bubble velocity near the heat transfer surface, measurement of bubble size under stationary acceleration field and observation of bubble behavior under varying acceleration field. (author)

  7. Critical heat flux correlation analysis for PWR reactors with low mass flow

    International Nuclear Information System (INIS)

    Carajilescov, Pedro

    1996-01-01

    The major limit in the thermalhydraulic design of water cooled reactors consists in the occurrence of critical heat flux, which is verified by correlation of large range of validity. In the present work, the major design correlations were analyzed, through comparisons with experimental data, for utilization in PWR with low mass flux in the core. The results show that the EPRI correlation, with modifications, gives conservative results, from the safety point of view, with lower data spreading, being the most indicated for the reactor thermal design. (author)

  8. Critical heat flux measurements in small-diameter tubes using R12 as model fluid

    International Nuclear Information System (INIS)

    Mueller-Menzel, T.

    1987-01-01

    Results of critical heat flux measurements are reported for vertical upflow of Refrigerant 12 at high mass fluxes and high pressures in small diameter tubes. The data are transformed into water data using a scaling law, which is verified by means of a new analysis. An error estimation includes the error of the scaling law. Special phenomena ('limiting quality', 'upstream boiling crisis') are explained by theoretical models. The applicability of existing correlations is checked and a new CHF-table for small diameter tubes is presented. With 41 figs., 12 tabs [de

  9. Rethinking Heat Injury in the SOF Multipurpose Canine: A Critical Review.

    Science.gov (United States)

    Baker, Janice L; Hollier, Paul J; Miller, Laura; Lacy, Ward A

    2012-01-01

    Heat injury is a significant concern of the Special Operations Forces Multipurpose Canine (SOF MPC). The unique athletic abilities and working environment of the SOF MPC differ from that of companion dogs or even conventional military working dogs. This should be considered in the prevention, diagnosis, and treatment of heat injury of the SOF MPC. A critical review of the literature on canine heat injury as it pertains to working dogs demonstrates limited scientific evidence on best practices for immediate clinical management of heat injury in SOF MPCs. A majority of management guidelines for heat injury in veterinary reference books and journals are based on review articles or professional opinion of the author vs. evidence from original research. In addition, guidelines are written primarily for companion animal populations vs. SOF MPCs and focus on measures to be undertaken in a clinical setting vs. point of injury. The phenomenon of ?circular referencing? is also prevalent in the heat injury literature. Current guidelines supported by review articles and textbooks often provide no citation or cite other review articles for clinical standards such as normal temperature ranges, treatment methods, and recurrence of heat injury. This ?circular referencing? phenomenon misrepresents anecdotal evidence and professional opinion as scientifically validated, reinforcing concepts and recommendations that are not truly supported by the evidence. Further study is needed to fully understand heat injury in SOF MPCs and how this applies to prevention, diagnosis and treatment guidelines. In order to provide SOF canine programs with best clinical advice and care, SOF Veterinarians must make clinical judgments based on evaluation of the most accurate and valid information possible. Clinical guidelines are fluid and should be reviewed regularly for relevance to the defined population in question. Clinical Guidelines should also be utilized as guiding principles in conjunction with

  10. A prediction method of the effect of radial heat flux distribution on critical heat flux in CANDU fuel bundles

    International Nuclear Information System (INIS)

    Yuan, Lan Qin; Yang, Jun; Harrison, Noel

    2014-01-01

    Fuel irradiation experiments to study fuel behaviors have been performed in the experimental loops of the National Research Universal (NRU) Reactor at Atomic Energy of Canada Limited (AECL) Chalk River Laboratories (CRL) in support of the development of new fuel technologies. Before initiating a fuel irradiation experiment, the experimental proposal must be approved to ensure that the test fuel strings put into the NRU loops meet safety margin requirements in critical heat flux (CHF). The fuel strings in irradiation experiments can have varying degrees of fuel enrichment and burnup, resulting in large variations in radial heat flux distribution (RFD). CHF experiments performed in Freon flow at CRL for full-scale bundle strings with a number of RFDs showed a strong effect of RFD on CHF. A prediction method was derived based on experimental CHF data to account for the RFD effect on CHF. It provides good CHF predictions for various RFDs as compared to the data. However, the range of the tested RFDs in the CHF experiments is not as wide as that required in the fuel irradiation experiments. The applicability of the prediction method needs to be examined for the RFDs beyond the range tested by the CHF experiments. The Canadian subchannel code ASSERT-PV was employed to simulate the CHF behavior for RFDs that would be encountered in fuel irradiation experiments. The CHF predictions using the derived method were compared with the ASSERT simulations. It was observed that the CHF predictions agree well with the ASSERT simulations in terms of CHF, confirming the applicability of the prediction method in fuel irradiation experiments. (author)

  11. Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel

    Energy Technology Data Exchange (ETDEWEB)

    Limin, Zheng [Shanghai Nuclear Engineering Research and Design Inst., SH (China); Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime

    1997-08-01

    The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms{sup -1} and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D{sub hy}=0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm{sup -2}s{sup -1} (water flow velocity, 5-15 ms{sup -1}); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within {+-}18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within {+-}16%. (author)

  12. Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel

    International Nuclear Information System (INIS)

    Zheng Limin; Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime.

    1997-08-01

    The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms -1 and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D hy =0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm -2 s -1 (water flow velocity, 5-15 ms -1 ); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within ±18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within ±16%. (author)

  13. Regions of existence of two forms of the critical void fraction dependence on heat flux density at burnout

    International Nuclear Information System (INIS)

    Smolin, V.N.

    1981-01-01

    On the basis of the available experimental data considered is the burnout during the movement of steam-water flow in vertical heated tubes with internal diameter from 8 to 40 mm. Critical steam content Xsub(cr) dependences on the critical heat flux qsub(cr) in different tubes and under different pressure are analyzed. Two main regions of the weak and strong dependences Xsub(cr)=f(qsub(cr)) at burnout are found out [ru

  14. Critical heat flux data in a vertical tube at low and medium pressures

    Energy Technology Data Exchange (ETDEWEB)

    Teyssedou, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Olekhnowitch, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Tapucu, A [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Champagne, P [Institut de Genie Nucleaire, Ecole Polytechnique, C.P. 6079, succ. Centre-ville, Montreal, Quebec H3C 3A7 (Canada); Groeneveld, D [Chalk River Laboratories, AECL Research, Chalk River (Canada)

    1994-09-01

    AECL Research and Ecole Polytechnique have been cooperating on the validation of the critical heat flux (CHF) look-up table (D.C. Groeneveld et al., Heat Transfer Eng. 7(1-2) (1986) 46-62). For low and medium pressures the values in the table have been obtained by extrapolation and curve fitting; therefore, errors could be expected. To reduce these possible extrapolation errors, CHF experiments are being carried out in water cooled 8mm internal diameter (ID) tubes, at conditions where the data are scarce. This paper presents some of the experimental CHF data obtained for vertical up flow in an 8mm ID test section, for a wide range of exit qualities (5-70%) and the exit pressure ranging from 5 to 30bar. The experiments were carried out for heated lengths of 0.75, 1, 1.4 and 1.8m. In general, the collected data show parametric trends similar to those described in the open literature. However, it was observed that for low pressure conditions CHF depends on the heated length; this dependence begins to disappear for exit pressure of about 30bar. The CHF data have also been compared with predictions of well-known correlations (L. Biasi et al., Energia Nucl. 14(9) (1967) 530-536; R. Bowring, Br. Report AEEW-R789, Winfrith, UK, 1972; Y. Khatto and H. Ohno, Int. J. Heat Mass Transfer 27 (1984) 1641-1648) and those of the look-up table given by Groeneveld et al. For low pressures and low mass fluxes the look-up table seems to yield better predictions of the CHF than the correlations. However, for medium pressures and mass fluxes the correlations perform better than the look-up table; among those tested, Katto and Ohno's correlation gives the best results. ((orig.))

  15. Critical heat flux on micro-structured zircaloy surfaces for flow boiling of water at low pressures

    International Nuclear Information System (INIS)

    Haas, C.; Miassoedov, A.; Schulenberg, T.; Wetzel, T.

    2012-01-01

    The influence of surface structure on critical heat flux for flow boiling of water was investigated for Zircaloy tubes in a vertical annular test section. The objectives were to find suitable surface modification processes for Zircaloy tubes and to test their critical heat flux performance in comparison to the smooth tube. Surface structures with micro-channels, porous layer, oxidized layer, and elevations in micro- and nano-scale were produced on a section of a Zircaloy cladding tube. These modified tubes were tested in an internally heated vertical annulus with a heated length of 326 mm and an inner and outer diameter of 9.5 and 18 mm. The experiments were performed with mass fluxes of 250 and 400 kg/(m 2 s), outlet pressures between 120 and 300 kPa, and constant inlet subcooling enthalpy of 167 kJ/kg. Only a small influence of modified surface structures on critical heat flux was observed for the pressure of 120 kPa in the present test section geometry. However, with increasing pressure the critical heat flux could increase up to 29% using the surface structured tubes with micro-channels, porous and oxidized layers. Capillary effects and increased nucleation site density are assumed to improve the critical heat flux performance. (authors)

  16. Generated forces and heat during the critical stages of friction stir welding and processing

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, Sadiq Aziz; Tahir, Abd Salam Md; Izamshah, R. [University Teknikal Malaysia Melaka, Malacca (Malaysia)

    2015-10-15

    The solid-state behavior of friction stir welding process results in violent mechanical forces that should be mitigated, if not eliminated. Plunging and dwell time are the two critical stages of this welding process in terms of the generated forces and the related heat. In this study, several combinations of pre-decided penetration speeds, rotational speeds, tool designs, and dwell time periods were used to investigate these two critical stages. Moreover, a coupled-field thermal-structural finite element model was developed to validate the experimental results and the induced stresses. The experimental results revealed the relatively large changes in force and temperature during the first two stages compared with those during the translational tool movement stage. An important procedure to mitigate the undesired forces was then suggested. The model prediction of temperature values and their distribution were in good agreement with the experimental prediction. Therefore, the thermal history of this non-uniform heat distribution was used to estimate the induced thermal stresses. Despite the 37% increase in these stresses when 40 s dwell time was used instead of 5 s, these stresses showed no effect on the axial force values because of the soft material incidence and stir effects.

  17. Critical heat flux of subcooled flow boiling in a narrow tube

    International Nuclear Information System (INIS)

    Inasaka, Fujio; Nariai, Hideki; Shimura, Toshiya.

    1986-01-01

    The critical heat flux (CHF) of subcooled flow boiling in a narrow tube was investigated experimentally using water as a coolant. Experiments were conducted at nearly ambient pressure under the following conditions: tube inside diameter: 1 ∼ 3 mm, tube length: 10 ∼ 100 mm, and water mass velocity: 7000 - 20000 kg/(m 2 · s). The critical heat flux increases the shorter the tube length and the smaller the tube inside diameter, at the same water mass velocity and exit quality. Experimental data were compared with empirical correlations, such as the Griffel and Knoebel correlations for subcooled boiling at low pressure, the Tong correlation for subcooled boiling at high pressure, and the Katto correlation. The existence of two parameter regions was confirmed. The first is the low CHF region in which experimental data can be predicted well by Griffel and Knoebel correlations, and the second is the high CHF region in which experimental data are higher than the predictions by the above two correlations. (author)

  18. Evaluated experimental database on critical heat flux in WWER FA models

    International Nuclear Information System (INIS)

    Artamonov, S.; Sergeev, V.; Volkov, S.

    2015-01-01

    The paper presents the description of the evaluated experimental database on critical heat flux in WWER FA models of new designs. This database was developed on the basis of the experimental data obtained in the years of 2009-2012. In the course of its development, the database was reviewed in terms of completeness of the information about the experiments and its compliance with the requirements of Rostekhnadzor regulatory documents. The description of the experimental FA model characteristics and experimental conditions was specified. Besides, the experimental data were statistically processed with the aim to reject incorrect ones and the sets of experimental data on critical heat fluxes (CHF) were compared for different FA models. As a result, for the fi rst time, the evaluated database on CHF in FA models of new designs was developed, that was complemented with analysis functions, and its main purpose is to be used in the process of development, verification and upgrading of calculation techniques. The developed database incorporates the data of 4183 experimental conditions obtained in 53 WWER FA models of various designs. Keywords: WWER reactor, fuel assembly, CHF, evaluated experimental data, database, statistical analysis. (author)

  19. A Critical Heat Generation for Safe Nuclear Fuels after a LOCA

    Directory of Open Access Journals (Sweden)

    Jae-Yong Kim

    2014-01-01

    Full Text Available This study applies a thermo-elasto-plastic-creep finite element procedure to the analysis of an accidental behavior of nuclear fuel as well as normal behavior. The result will be used as basic data for the robust design of nuclear power plant and fuels. We extended the range of mechanical strain from small or medium to large adopting the Hencky logarithmic strain measure in addition to the Green-Lagrange strain and Almansi strain measures, for the possible large strain situation in accidental environments. We found that there is a critical heat generation after LOCA without ECCS (event category 5, under which the cladding of fuel sustains the internal pressure and temperature for the time being for the rescue of the power plant. With the heat generation above the critical value caused by malfunctioning of the control rods, the stiffness of cladding becomes zero due to the softening by high temperature. The weak position of cladding along the length continuously bulges radially to burst and to discharge radioactive substances. This kind of cases should be avoid by any means.

  20. Generated forces and heat during the critical stages of friction stir welding and processing

    International Nuclear Information System (INIS)

    Hussein, Sadiq Aziz; Tahir, Abd Salam Md; Izamshah, R.

    2015-01-01

    The solid-state behavior of friction stir welding process results in violent mechanical forces that should be mitigated, if not eliminated. Plunging and dwell time are the two critical stages of this welding process in terms of the generated forces and the related heat. In this study, several combinations of pre-decided penetration speeds, rotational speeds, tool designs, and dwell time periods were used to investigate these two critical stages. Moreover, a coupled-field thermal-structural finite element model was developed to validate the experimental results and the induced stresses. The experimental results revealed the relatively large changes in force and temperature during the first two stages compared with those during the translational tool movement stage. An important procedure to mitigate the undesired forces was then suggested. The model prediction of temperature values and their distribution were in good agreement with the experimental prediction. Therefore, the thermal history of this non-uniform heat distribution was used to estimate the induced thermal stresses. Despite the 37% increase in these stresses when 40 s dwell time was used instead of 5 s, these stresses showed no effect on the axial force values because of the soft material incidence and stir effects

  1. Critical heat flux and flow pattern for water flow in annular geometry

    International Nuclear Information System (INIS)

    Park, J.-W.; Baek, W.-P.; Chang, S.H.

    1997-01-01

    An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced-circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m, inner diameter 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, churn-to-annular flow transition and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for the upward flow. In addition to the experiment, selected CHF correlations for annuli are assessed based on 1156 experimental data from various sources. The Doerffer et al. (1994); Barnett (1966); Jannsen and Kervinen (1963); Levitan and Lantsman (1977) correlations show reasonable predictions for wide parameter ranges, among which the Doerffer et al. (1994) correlation shows the widest parameter ranges and a possibility of further improvement. However, there is no correlation predicting the low-pressure, low-flow CHF satisfactorily. (orig.)

  2. Critical heat flux phenomena in flow boiling during step wise and ramp wise power transients

    International Nuclear Information System (INIS)

    Celata, G.P.; Cumo, M.; D'Annibale, F.; Farello, G.E.; Abou Said, S.

    1987-01-01

    The present paper deals with the results of an experimental investigation of the forced flow critical heat flux during power transients in a vertically heated channel. Experiments were carried out with a Refrigerant-12 1oop employing a circular test section which was electrically and uniformly heated. The power transients were performed with the step-wise and ramp-wise increase of the power to the test section. The test parameters included several values of the initial power (before the transient) and the final power (at the end of the transient) in the case of step-wise transients and the slope of the ramp in the case of ramp-wise transients. The pressure and specific mass flow rate, which were kept constant during the power transient,were varied from 1.2 to 2.7 MPa and 850 to 1500 Kg/sm 2 , respectively. Correlations of the experimental data for the time-to-crisis in terms of the independent parameters of the system are also proposed and verified for different values of pressure,mass flow rate, and inlet subcooling

  3. Parametric trends analysis of the critical heat flux based on artificial neural networks

    International Nuclear Information System (INIS)

    Moon, S.K.; Baek, W.P.; Chang, S.H.

    1996-01-01

    Parametric trends of the critical heat flux (CHF) are analyzed by applying artificial neural networks (ANNs) to a CHF data base for upward flow of water in uniformly heated vertical round tubes. The analyses are performed from three viewpoints, i.e., for fixed inlet conditions, for fixed exit conditions, and based on local conditions hypothesis. Katto's and Groeneveld et al. dimensionless parameters are used to train the ANNs with the experimental CHF data. The trained ANNs predict the CHF better than any other conventional correlations, showing RMS errors of 8.9%, 13.1% and 19.3% for fixed inlet conditions, for fixed exit conditions, and for local conditions hypothesis, respectively. The parametric trends of the CHF obtained from those trained ANNs show a general agreement with previous understanding. In addition, this study provides more comprehensive information and indicates interesting points for the effects of the tube diameter, the heated length, and the mass flux. It is expected that better understanding of the parametric trends is feasible with an extended data base. (orig.)

  4. RELAP5/MOD2 benchmarking study: Critical heat flux under low-flow conditions

    International Nuclear Information System (INIS)

    Ruggles, E.; Williams, P.T.

    1990-01-01

    Experimental studies by Mishima and Ishii performed at Argonne National Laboratory and subsequent experimental studies performed by Mishima and Nishihara have investigated the critical heat flux (CHF) for low-pressure low-mass flux situations where low-quality burnout may occur. These flow situations are relevant to long-term decay heat removal after a loss of forced flow. The transition from burnout at high quality to burnout at low quality causes very low burnout heat flux values. Mishima and Ishii postulated a model for the low-quality burnout based on flow regime transition from churn turbulent to annular flow. This model was validated by both flow visualization and burnout measurements. Griffith et al. also studied CHF in low mass flux, low-pressure situations and correlated data for upflows, counter-current flows, and downflows with the local fluid conditions. A RELAP5/MOD2 CHF benchmarking study was carried out investigating the performance of the code for low-flow conditions. Data from the experimental study by Mishima and Ishii were the basis for the benchmark comparisons

  5. Neutralized wettability effect of superhydrophilic Cr-layered surface on pool boiling critical heat flux

    International Nuclear Information System (INIS)

    Son, Hong Hyun; Jeong, Ui Ju; Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong

    2016-01-01

    The former method is deemed challenging due to longer development period and license issue. In this regard, FeCrAl, Cr, and SiC have been received positive attention as ATF coating materials because they are highly resistant to high temperature steam reaction causing massive hydrogen generation. In this study, Cr was selected as a target deposition material on the metal substrate because we found that Cr-layered surface becomes superhydrophilic, favorable to delaying the triggering of the critical heat flux (CHF). Thus in order to investigate the effect of Cr-layered superhydrophilic surfaces (under explored coating conditions) on pool boiling heat transfer, pool boiling experiment was conducted in the saturated deionized water under atmospheric pressure. As a physical vapor deposition (PVD) method, the DC magnetron sputtering technique was introduced to develop Cr-layered nanostructure. As a control variable of DC sputtering, substrate temperature was selected. Surface wettability and nanostructure were analyzed as major surface parameters on the CHF. We believe that highly dense micro/nano structure without nucleation cavities and inner pores neutralized the wettability effect on the CHF. Moreover, superhydrophilic surface with deficient cavity density rather hinders active nucleation. This emphasizes the importance of micro/nano structure surface for enhanced boiling heat transfer.

  6. Influence of the heater material on the critical heat load at boiling of liquids on surfaces with different sizes

    Science.gov (United States)

    Anokhina, E. V.

    2010-05-01

    Data on critical heat loads q cr for the saturated and unsaturated pool boiling of water and ethanol under atmospheric pressure are reported. It is found experimentally that the critical heat load does not necessarily coincide with the heat load causing burnout of the heater, which should be taken into account. The absolute values of q cr for the boiling of water and ethanol on copper surfaces 65, 80, 100, 120, and 200 μm in diameter; tungsten surface 100 μm in diameter; and nichrome surface 100 μm in diameter are obtained experimentally.

  7. Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

    International Nuclear Information System (INIS)

    Williams, C.L.; Beus, S.G.

    1980-05-01

    Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values

  8. Pool boiling characteristics and critical heat flux mechanisms of microporous surfaces and enhancement through structural modification

    Science.gov (United States)

    Ha, Minseok; Graham, Samuel

    2017-08-01

    Experimental studies have shown that microporous surfaces induce one of the highest enhancements in critical heat flux (CHF) during pool boiling. However, microporous surfaces may also induce a very large surface superheat (>100 °C) which is not desirable for applications such as microelectronics cooling. While the understanding of the CHF mechanism is the key to enhancing boiling heat transfer, a comprehensive understanding is not yet available. So far, three different theories for the CHF of microporous surfaces have been suggested: viscous-capillary model, hydrodynamic instability model, and dryout of the porous coatings. In general, all three theories account for some aspects of boiling phenomena. In this study, the theories are examined through their correlations with experimental data on microporous surfaces during pool boiling using deionized (DI) water. It was found that the modulation of the vapor-jet through the pore network enables a higher CHF than that of a flat surface based on the hydrodynamic instability theory. In addition, it was found that as the heat flux increases, a vapor layer grows in the porous coatings described by a simple thermal resistance model which is responsible for the large surface superheat. Once the vapor layer grows to fill the microporous structure, transition to film boiling occurs and CHF is reached. By disrupting the formation of this vapor layer through the fabrication of channels to allow vapor escape, an enhancement in the CHF and heat transfer coefficient was observed, allowing CHF greater than 3500 kW/m2 at a superheat less than 50 °C.

  9. Critical heat flux experiments for high conversion light water reactor, (3)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

    1990-03-01

    As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

  10. Experimental study of critical heat flux enhancement with hypervapotron structure under natural circulation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Chang, Huajian [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Zhao, Yufeng, E-mail: zhaoyufeng@snptc.com.cn [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Zhang, Ming; Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei [State Power Investment Corporation, Beijing (China)

    2017-05-15

    Highlights: • Natural circulation tests are performed to study the effect of hypervapotron on CHF. • Hypervapotron structure improves CHF under natural circulation conditions. • Visualization data illustrate vapor blanket behavior under subcooled flow conditions. - Abstract: The enhancement of critical heat flux with a hypervapotron structure under natural circulation conditions is investigated in this study. Subcooled flow boiling CHF experiments are performed using smooth and hypervapotron surfaces at different inclination angles under natural circulation conditions. The experimental facility, TESEC (Test of External Vessel Surface with Enhanced Cooling), is designed to conduct CHF experiments in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. The two-phase flow of subcooled flow boiling on both smooth and hypervapotron heating plates is observed and analyzed by the high-speed visualization technology. The results show that both smooth surface and hypervapotron surface CHF data exhibit a similar trend against inclination angles compared with the CHF results under forced flow condition on the same facility in earlier studies. However, the CHF enhancement of the hypervapotron structure is evidently more significant than the one under forced flow conditions. The experiments also indicate that the natural flow rates are higher with hypervapotron structure. The initiation of CHF is analyzed under transient subcooling and flow rate conditions for both smooth and hypervapotron heating surfaces. An explanation is given for the significant enhancement effect caused by the hypervapotron surface under natural circulation conditions. The visualization data are exhibited to demonstrate the behavior of the vapor blanket at various inclination angles and on different surfaces. The geometric data of the vapor blanket are quantified by an image post-processing method. It is found that the thickness of the vapor blanket

  11. Heat treatment control of Bi-2212 coils: I. Unravelling the complex dependence of the critical current density of Bi-2212 wires on heat treatment

    Science.gov (United States)

    Shen, Tengming; Li, Pei; Ye, Liyang

    2018-01-01

    A robust and reliable heat treatment is crucial for developing superconducting magnets from several superconductors especially Bi-2212. An improper heat treatment may significantly reduce the critical current density Jc of a Bi-2212 superconducting coil, even to zero, since the Jc of Bi-2212 wires is sensitive to parameters of its heat treatment (partial melt processing). To provide an essential database for heat treating Bi-2212 coils, the dependence of Jc on heat treatment is studied systematically in 11 industrial Bi-2212 wires, revealing several common traits shared between these wires and outlier behaviors. The dependence of the Jc of Bi-2212 on heat treatment is rather complex, with many processing parameters affecting Jc, including the peak processing temperature Tp, the time at the peak temperature tp, the time in the melt tmelt, the rate at which Bi-2212 melt is initially cooled CR1, the rate at which the solidification of Bi-2212 melt occurs CR2, and the temperature Tq at which the cooling rate switches from CR1 to CR2. The role of these parameters is analyzed and clarified, in the perspective of heat treating a coil. Practical advices on heat treatment design are given. The ability of a Bi-2212 coil to follow the prescribed recipe decreases with increasing coil sizes. The size of a coil that can be properly heat treated is determined.

  12. Critical heat flux near the critical pressure in heater rod bundle cooled by R-134A fluid: Effects of unheated rods and spacer grid

    International Nuclear Information System (INIS)

    Chun, Se-Y.; Shin, C.W.; Hong, S. D.; Moon, S. K.

    2007-01-01

    A supercritical-pressure light water reactor (SCWR) is currently investigated as the next generation nuclear reactors. The SCWR, which is operated above the thermodynamic critical point of water (647 K, 22.1 MPa), have advantages over conventional light water reactors in terms of thermal efficiency as well as in compactness and simplicity. Many experimental studies have been performed on heat transfer in the boiler tubes of supercritical fossil fire power plants (FPPs). However, the thermal-hydraulic conditions of the SCWR core are different from those of the FPP boiler. In the SCWR core, the heat transfer to the cooling water occurs on the outside surface of fuel rods in rod bundle with spacers. In addition, the experimental studies in which the critical heat flux (CHF) has been carefully measured near the critical pressure have never yet been carried out, as far as we know. Therefore, we have recently conducted the CHF experiments with a vertical 5x5 heater rod bundle cooled by R- 134a fluid. The purpose of this work is to find out some novel knowledge for the CHF near the critical pressure, based on more careful experiments. The outer diameter, heated length and rod pitch of the heater rods are 9.5, 2000 and 12.85 mm, respectively. The critical power has been measured in a range of the pressure of 2.474.03 MPa (the critical pressure of R-134a is 4.059 MPa), the mass flux 502000 kg/m 2 s, and the inlet subcooling 4084 kJ/kg. For the mass fluxes of not less than 550 kg/m 2 s, the critical power decreases monotonously up to the pressure of about 3.63.8 MPa with increasing pressure, and then fall sharply at about 3.83.9 MPa as if the values of the critical power converge on zero at the critical pressure. For the low mass fluxes of 50 to 250 kg/m 2 , the sharp decreasing trend of the critical power near the critical pressure is not observed. The CHF phenomenon near the critical pressure no longer leads to an inordinate increase in the heated wall temperature such as

  13. Investigation of Body Force Effects on Flow Boiling Critical Heat Flux

    Science.gov (United States)

    Zhang, Hui; Mudawar, Issam; Hasan, Mohammad M.

    2002-01-01

    The bubble coalescence and interfacial instabilities that are important to modeling critical heat flux (CHF) in reduced-gravity systems can be sensitive to even minute body forces. Understanding these complex phenomena is vital to the design and safe implementation of two-phase thermal management loops proposed for space and planetary-based thermal systems. While reduced gravity conditions cannot be accurately simulated in 1g ground-based experiments, such experiments can help isolate the effects of the various forces (body force, surface tension force and inertia) which influence flow boiling CHF. In this project, the effects of the component of body force perpendicular to a heated wall were examined by conducting 1g flow boiling experiments at different orientations. FC-72 liquid was boiled along one wall of a transparent rectangular flow channel that permitted photographic study of the vapor-liquid interface at conditions approaching CHF. High-speed video imaging was employed to capture dominant CHF mechanisms. Six different CHF regimes were identified: Wavy Vapor Layer, Pool Boiling, Stratification, Vapor Counterflow, Vapor Stagnation, and Separated Concurrent Vapor Flow. CHF showed great sensitivity to orientation for flow velocities below 0.2 m/s, where very small CHF values where measured, especially with downflow and downward-facing heated wall orientations. High flow velocities dampened the effects of orientation considerably. Figure I shows representative images for the different CHF regimes. The Wavy Vapor Layer regime was dominant for all high velocities and most orientations, while all other regimes were encountered at low velocities, in the downflow and/or downward-facing heated wall orientations. The Interfacial Lift-off model was modified to predict the effects of orientation on CHF for the dominant Wavy Vapor Layer regime. The photographic study captured a fairly continuous wavy vapor layer travelling along the heated wall while permitting liquid

  14. Testing plan for critical heat flux measurement during in-vessel retention

    International Nuclear Information System (INIS)

    Aoki, Kazuyoshi; Iwaki, Chikako; Sato, Hisaki; Mimura, Satoshi; Kanamori, Daisuke

    2015-01-01

    In-Vessel Retention (IVR) is a method to maintain molten debris in a reactor vessel (RV) by RV outer surface cooling. Structural integrity of RV and cooling capacity on RV outer surface are important to verify IVR strategy. Critical Heat Flux (CHF) data is necessary to estimate cooling capacity on the RV outer surface. And there are some CHF data to estimate cooling capacity on the RV outer surface. However, these data were obtained for specific plants. Thus, the objective of this study is developing a CHF correlation for various PWR plants. The objectives of this paper are developing test equipment and testing plan for the CHF correlation. Firstly, plant conditions during severe accidents were organized. Then, ranges of testing parameters were estimated with the plant conditions. And specifications of the test equipment were set to cover the range of parameters. Secondly, testing cases were set based on design of experiments. The test cases are suitable to develop experimental correlations. (author)

  15. A theoretical prediction of critical heat flux in subcooled pool boiling during power transients

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Nelson, R.A.; Gunnerson, F.S.

    1988-01-01

    Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions are of fundamenatal interest in the design, operation, safety of boiling and two-phase flow devices. This paper discusses the results of a comprehensive theoretical study made specifically to model transient CHF behavior in subcooled pool boiling. This study is based upon a simplified steady-state CHF model in terms of the vapor mass growth period. The results obtained from this theory indicate favorable agreement with the experimental data from cylindrical heaters with small radii. The statistical nature of the vapor mass behavior in transient boiling also is considered and upper and lower limits for the current theory are established. Various factors that affect the discrepancy between the data and the theory are discussed

  16. Optimizing critical heat flux enhancement through nano-particle-based surface modifications

    International Nuclear Information System (INIS)

    Truong, B.; Hu, L. W.; Buongiorno, J.

    2008-01-01

    Colloidal dispersions of nano-particles, also known as nano-fluids, have shown to yield significant Critical Heat Flux (CHF) enhancement. The CHF enhancement mechanism in nano-fluids is due to the buildup of a porous layer of nano-particles upon boiling. Unlike microporous coatings that had been studied extensively, nano-particles have the advantages of forming a thin layer on the substrate with surface roughness ranges from the sub-micron to several microns. By tuning the chemical properties it is possible to coat the nano-particles in colloidal dispersions onto the desired surface, as has been demonstrated in engineering thin film industry. Building on recent work conducted at MIT, this paper illustrates the maximum CHF enhancement that can be achieved based on existing correlations. Optimization of the CHF enhancement by incorporation of key factors, such as the surface wettability and roughness, will also be discussed. (authors)

  17. Classification and prediction of the critical heat flux using fuzzy theory and artificial neural networks

    International Nuclear Information System (INIS)

    Moon, Sang Ki; Chang, Soon Heung

    1994-01-01

    A new method to predict the critical heat flux (CHF) is proposed, based on the fuzzy clustering and artificial neural network. The fuzzy clustering classifies the experimental CHF data into a few data clusters (data groups) according to the data characteristics. After classification of the experimental data, the characteristics of the resulting clusters are discussed with emphasis on the distribution of the experimental conditions and physical mechanism. The CHF data in each group are trained in an artificial neural network to predict the CHF. The artificial neural network adjusts the weight so as to minimize the prediction error within the corresponding cluster. Application of the proposed method to the KAIST CHF data bank shows good prediction capability of the CHF, better than other existing methods. ((orig.))

  18. Global, decaying solutions of a focusing energy-critical heat equation in R4

    Science.gov (United States)

    Gustafson, Stephen; Roxanas, Dimitrios

    2018-05-01

    We study solutions of the focusing energy-critical nonlinear heat equation ut = Δu - | u|2 u in R4. We show that solutions emanating from initial data with energy and H˙1-norm below those of the stationary solution W are global and decay to zero, via the "concentration-compactness plus rigidity" strategy of Kenig-Merle [33,34]. First, global such solutions are shown to dissipate to zero, using a refinement of the small data theory and the L2-dissipation relation. Finite-time blow-up is then ruled out using the backwards-uniqueness of Escauriaza-Seregin-Sverak [17,18] in an argument similar to that of Kenig-Koch [32] for the Navier-Stokes equations.

  19. Application of tube critical heat flux tables to annuli and rod bundles

    International Nuclear Information System (INIS)

    Ulrych, G.

    1985-01-01

    The purpose of this paper is to show that tables for the critical heat flux (CHF) in tubes have a much wider range of applicability than only to tubes. With the proper choice of a characteristic length replacing the tube diameter as a parameter the validity of the tables can be expanded to more complex geometries. The paper describes how the tables must be applied to annuli or rod bundles. The data base for comparisons is mainly taken from the open literature. For rod bundles the proposed methodology was checked for very different geometries including rod bundles from very tight hexagonal to extremely open square bundle arrays. It is concluded that the tables give reasonable results for a wide range of hydraulic diameters

  20. Increase in VVER type reactor critical heat fluxes due to placing the mixing grids

    International Nuclear Information System (INIS)

    Bezrukov, Y.; Lisenkov, E.; Vasilchenko, I.

    2011-01-01

    The report deals with the results of studies of critical heat fluxes (CHF) on the models of VVER type reactor fuel assembly models equipped with the 'Vihr' intensifiers-grids. The models are the seven-rod bundles with the uniform and non-uniform axial power that correspond to two periods of FA operation i.e. beginning of cycle and end of cycle. The experiments performed showed that the mixing grids of this type are capable of increasing the FA burnout power. The power ascension rate depends on both coolant pressure and steam quality value in the CHF point. Placing the mixing grids in the bundle upper spans results in shifting the point of DNB occurrence downward along the FA height. The experimental data obtained will be used to develop the correlations for determining the CHF in the FA equipped with the mixing grids. (authors)

  1. Critical heat flux in bottom heated two-phase thermosyphon. Improvement in critical heat flux due to concentric tube; Katan shuchu kanetsugata niso netsu syphon no genkai netsu ryusoku. Nijukan ni yoru genkai netsu ryusoku no kaizen

    Energy Technology Data Exchange (ETDEWEB)

    Monde, M.; Mitsutake, Y. [Saga University, Saga (Japan). Faculty of Science and Engineering

    2000-02-25

    An experiment has been carried out to elucidate the critical heat flux (CHF) of an open two-phase thermosyphon with a bottom heated chamber in which heat is absorbed by evaporation of liquid. Another objective is to enhance the CHF using a concentric-tube by which counter-current flow of vapor and liquid in the throat of the chamber can be controlled well. The CHF data are measured for the saturated liquid of R 113 at a different pressure and different configuration of concentric tubes. The CHF data without the inner tube are in good agreement with the existing correlation and analytical result. The CHF increases by as much as several times of the CHF without the inner tube with an increase in the inner tube diameter up to a certain diameter of the inner tube and then decreases continuously as the inner tube diameter approaches the outer tube diameter. The optimum diameter of inner tube exists at which the CHF is maximum. (author)

  2. Two-phase flow regimes and mechanisms of critical heat flux under subcooled flow boiling conditions

    International Nuclear Information System (INIS)

    Le Corre, Jean-Marie; Yao, Shi-Chune; Amon, Cristina H.

    2010-01-01

    A literature review of critical heat flux (CHF) experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available experimental information. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. Even though the selected concept has not received much attention (in term or theoretical developments and applications) as compared to other more popular DNB models, its basis have often been cited by experimental investigators and is considered by the authors as the 'most-likely' mechanism based on the literature review and analysis performed in this work. The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow and has been numerically implemented and validated in bubbly flow and coupled with one- and three-dimensional (CFD) two-phase flow codes, in a companion paper. [Le Corre, J.M., Yao, S.C., Amon, C.H., in this issue. A mechanistic model of critical heat flux under subcooled flow boiling conditions for application to one and three-dimensional computer codes. Nucl. Eng. Des.].

  3. An experimental study on the flow instabilities and critical heat flux under natural circulation

    International Nuclear Information System (INIS)

    Kim, Yun II; Chang, Soon Heung

    2004-01-01

    This study has been carried out to investigate the hydrodynamic stabilities and Critical Heat Flux (CHF) characteristics for the natural and forced circulation. A low pressure experimental loop was constructed, and experiments under various conditions have been performed. In the experiments of the natural circulation, flow oscillations has been observed and the average mass flux under flow oscillation have been measured. Several parameters such as heat flux, the inlet temperature of test section, friction valve opening and riser length have been varied in order to investigate their effects on the flow stability of the natural circulation system. And the CHF data from low flow experiments, namely the natural and forced circulation, have been compared with each other to identify the effects of the flow instabilities on the CHF for the natural circulation mode. The test conditions for the CHF experiments were a low flow of less than 70 kg/m 2 s of water in a vertical round tube with diameter of 0.008 m at near atmospheric pressure. (author)

  4. Critical energy in the cyclotron heating of ions in a mirror machine

    International Nuclear Information System (INIS)

    Gutierrez T, C.; Hernandez A, O.

    2002-01-01

    The problem of heating in the plasma sources where the geometry of the magnetic field forms a magnetic mirror as it is the case of the Ecr sources type, for maintaining the reload, it continues being an actual important problem. There are two methods for the analysis of this problem. The first of these methods is the stochastic mechanism of a particle where it is considered the existence of three characteristic frequencies as the cyclotron frequency, the electromagnetic field frequency and the transit frequency. The second method is that related with the non linear interaction of waves where the collective effects of the particles are the most important. In this work, in the Hamiltonian formalism, the stochastic mechanism in the cyclotron heating is analysed. It is considered the particular case of a plasma source with an external magnetic field, type mirror where a TE 11 electromagnetic wave is injected. The critical energy in the resonance mixing is calculated by the Poincare mapping method. The heterogeneity of the magnetic field is analysed. (Author)

  5. One-dimensional critical heat flux concerning surface orientation and gap size effects

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Hoon; Suh, Kune Y. E-mail: kysuh@snu.ac.kr

    2003-12-01

    Tests were conducted to examine the critical heat flux (CHF) on a one-dimensional downward heating rectangular channel having a narrow gap by changing the orientation of the copper test heater assembly in a pool of saturated water under atmospheric pressure. The test parameters include both the gap sizes of 1, 2, 5 and 10 mm, and the surface orientation angles from the downward-facing position (180 deg.) to the vertical position (90 deg.), respectively. Also, the CHF experiments were performed for pool boiling with varying heater surface orientations in the unconfined space at atmospheric pressure using the rectangular test section. It was observed that the CHF generally decreases as the surface inclination angle increases and as the gap size decreases. In consistency with several studies reported in the literature, it was found that there exists a transition angle at which the CHF changes with a rapid slope. An engineering correlation is developed for the CHF during natural convective boiling in the inclined, confined rectangular channels with the aid of dimensional analysis. This correlation agrees with the experimental data of this study within {+-}20%.

  6. An investigation into the nature and signifiance of a new critical heat flux correlation

    International Nuclear Information System (INIS)

    Green, W.J.; Beattie, D.R.H.

    1983-01-01

    An empirical critical heat flux correlation, which is based upon dimensionsless groups, and which was developed from a wide range of experimental data for vertical upflow in uniformly heated tubes, has been further examined to determine if, because of its accuracy and generality, it might provide an insight into the mechanisms of boiling crisis. A parameter survey using the correlation showed that it was well able to predict the so-called 'crisis of the second kind' without needing to define any separate or distinct flow regimes. Comparison of the empirical correlation with a general form of theoretical correlation, developed from a combination of several simple physical models which occur during the crisis phenomenon, shows a strong similarity in the form of the dimensionless groups. It also indicates that a further dimensionless group may need to be incorporated in the empirical correlation to achieve complete generality. Rearrangements of the dimensionless groups and the form of the empirical correlation, together with some minor approximations, indicate that boiling crisis is influenced by local hydrodynamic and thermal phenomena and can be related to pre- and post-crisis coolant conditions. (orig.)

  7. Phenomenological modeling of critical heat flux: The GRAMP code and its validation

    International Nuclear Information System (INIS)

    Ahmad, M.; Chandraker, D.K.; Hewitt, G.F.; Vijayan, P.K.; Walker, S.P.

    2013-01-01

    Highlights: ► Assessment of CHF limits is vital for LWR optimization and safety analysis. ► Phenomenological modeling is a valuable adjunct to pure empiricism. ► It is based on empirical representations of the (several, competing) phenomena. ► Phenomenological modeling codes making ‘aggregate’ predictions need careful assessment against experiments. ► The physical and mathematical basis of a phenomenological modeling code GRAMP is presented. ► The GRAMP code is assessed against measurements from BARC (India) and Harwell (UK), and the Look Up Tables. - Abstract: Reliable knowledge of the critical heat flux is vital for the design of light water reactors, for both safety and optimization. The use of wholly empirical correlations, or equivalently “Look Up Tables”, can be very effective, but is generally less so in more complex cases, and in particular cases where the heat flux is axially non-uniform. Phenomenological models are in principle more able to take into account of a wider range of conditions, with a less comprehensive coverage of experimental measurements. These models themselves are in part based upon empirical correlations, albeit of the more fundamental individual phenomena occurring, rather than the aggregate behaviour, and as such they too require experimental validation. In this paper we present the basis of a general-purpose phenomenological code, GRAMP, and then use two independent ‘direct’ sets of measurement, from BARC in India and from Harwell in the United Kingdom, and the large dataset embodied in the Look Up Tables, to perform a validation exercise on it. Very good agreement between predictions and experimental measurements is observed, adding to the confidence with which the phenomenological model can be used. Remaining important uncertainties in the phenomenological modeling of CHF, namely the importance of the initial entrained fraction on entry to annular flow, and the influence of the heat flux on entrainment rate

  8. Improvement of critical heat flux correlation for research reactors using plate-type fuel

    International Nuclear Information System (INIS)

    Kaminaga, Masanori; Yamamoto, Kazuyoshi; Sudo, Yukio

    1998-01-01

    In research reactors, plate-type fuel elements are generally adopted so as to produce high power densities and are cooled by a downward flow. A core flow reversal from a steady-state forced downward flow to an upward flow due to natural convection should occur during operational transients such as Loss of the primary coolant flow'. Therefore, in the thermal hydraulic design of research reactors, critical heat flux (CHF) under a counter-current flow limitation (CCFL) or a flooding condition are important to determine safety margins of fuel against CHF during a core flow reversal. The authors have proposed a CHF correlation scheme for the thermal hydraulic design of research reactors, based on CHF experiments for both upward and downward flows including CCFL condition. When the CHF correlation scheme was proposed, a subcooling effect for CHF correlation under CCFL condition had not been considered because of a conservative evaluation and a lack of enough CHF data to determine the subcooling effect on CHF. A too conservative evaluation is not appropriate for the design of research reactors because of construction costs etc. Also, conservativeness of the design must be determined precisely. In this study, therefore, the subcooling effect on CHF under the CCFL conditions in vertical rectangular channels heated from both sides were investigated quantitatively based on CHF experimental results obtained under uniform and non-uniform heat flux conditions. As a result, it was made clear that CHF in this region increase linearly with an increase of the channel inlet subcooling and a new CHF correlation including the effect of channel inlet subcooling was proposed. The new correlation could be adopted under the conditions of the atmospheric pressure, the inlet subcooling less than 78K, the channel gap size between 2.25 to 5.0mm, the axial peaking factor between 1.0 to 1.6 and L/De between 71 to 174 which were the ranges investigated in this study. (author)

  9. Experimental determination of heat transfer critical conditions in water forced convection at low pressure in a circular channel

    International Nuclear Information System (INIS)

    Fernandes, M.P.

    1973-02-01

    An experimental determination was made of heat transfer critical conditions in a circular channel, uniformly heated, and internally cooled by water in ascending forced convection, under a pressure slightly above atmospheric pressure. Measurements were made of water flow, pressure, electric power temperature and heating, and a systematic analysis was made of the system's parameters. The values obtained for the heat critical flux are circa 50% lower than those predicted by Becker and Biasi and this is accounted to flowing instabilities of thermo-hydrodynamic nature. It is suggested that the flowing channels of circuits aiming at the study of the boiling crisis phenomenon be expanded in its upper extremity, and that the coolant circulation be kept through a pump with a pressure X flow characteristic as vertical as possible

  10. Experimental data and calculation studies of critical heat fluxes at local disturbances of geometry of WWER fuel assemblies

    International Nuclear Information System (INIS)

    Kobzar, L.L.; Oleksyuk, D.A.

    2001-01-01

    The results of experiments executed in RRC 'Kurchatov Institute on the thermal-physical critical facility SVD are presented herein. The experiments modeled the drawing of two fuel rods to each other till touching WWER-1000 reactor in FA. The experimental model is a 7-rod bundle with the heated length of 1 m. The primary goal of experiments was to acquire the quantitative factors of the reduction in the critical heat fluxes as contrasted to the basic model (without disturbances of FA geometry) at the expense of local disturbance of a rod bundle geometry. As it follows from the experiment, the effect of decrease of the critical heat rate depends on combination of regime parameters and it makes 15% in the most unfavorable case (Authors)

  11. Method for calculating the critical heat flux in mixed rod assemblies based on the tables of crisis in bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.

    2000-01-01

    The method for calculating the critical heat flux in the mixed rod assemblies, for example RBMK, containing three-four angle and peripheral macrocells, is presented. The method is based on generalization of experimental data in form of tables for the rods beams. It is recommended for the areas of parameters both provided for by experimental data and for others, where the data are absent. The advantages of the table method as follows: it is acceptable within a wide range of parameters and provides for smooth description of dependence of critical heat fluxes on these parameters; it is characterized by clearness, high reliability and accuracy and is easy in application [ru

  12. Critical heat flux of R134A and R245FA in a 2.2 mm circular tube

    Energy Technology Data Exchange (ETDEWEB)

    Tibirica, Cristiano Bigonha; Ribatski, Gherhardt [Universidade de Sao Paulo (EESC/USP), Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia Mecanica], E-mails: bigonha@sc.usp.br, ribatski@sc.usp.br; Szczukiewicz, Sylwia; Thome, John Richard [Ecole Polytechnique Federale de Lausanne (LTCM/EPFL) (Switzerland). Lab. of Heat and Mass Transfer], Emails: sylwia.szczukiewicz@epfl.ch, john.thome@epfl.ch

    2010-07-01

    Critical heat flux (CHF) during flow boiling is generally related to a drastic decrease in the heat transfer coefficient and it is the maximum operational heat flux that can be achieved under safe operation. Due to such a fact, this topic has attracted great attention of the academic society dealing with boiling heat transfer and also in the industrial sector involved with the dissipation of high heat flux densities. In the specific case of high heat flux densities, micro-channel flow boiling is a promising technique for pursuing this objective. The boundary where microscale effects start in flow boiling is still an open issue in the literature and a 3 mm internal diameter (ID) threshold value, as suggested by Kandlikar and Grande (2003) is frequently adopted to characterize this point. Considering the needs for a better understanding of the micro/macro transition, this paper presents new experimental critical heat flux results in saturated flow boiling conditions for a macro/micro-scale tube. The data were obtained in a horizontal 2.20 mm ID stainless steel tube with heating lengths of 361 and 154 mm, R134a and R245fa as working fluids, mass velocities ranging from 100 to 1500 kg/m{sup 2s}, critical heat fluxes from 25 to 300 kW/m2, exit saturation temperatures of 25, 31 and 35 degree C, and critical vapor qualities ranging from 0.55 to 1. The experimental results show that critical heat flux increases with increasing mass velocity and inlet subcooling but decreases with increasing saturation temperature and heated length. The data also indicated a higher CHF for R245fa when compared with R134a at similar conditions. The experimental data were compared against the following CHF predictive methods: Katto and Ohno (1984), Shah (1987), Zhang et al. (2006) and Ong and Thome(2010). Katto and Ohno (1984) and Ong and Thome (2010) best predicted the database with a mean average error smaller than 15%. Both correlations include low and high pressure fluids in their

  13. A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

    International Nuclear Information System (INIS)

    Ha, Sang Jun; No, Hee Cheon

    1997-01-01

    This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variation in pressure, tube diameter and length, mass flux and inlet subcooling

  14. Experimental study of heat exchange coefficients, critical heat flux and charge losses, using water-steam mixtures in turbulent flow in a vertical tube

    International Nuclear Information System (INIS)

    Perroud, P.; De La Harpe, A.; Rebiere, J.

    1960-12-01

    Two stainless steel tubes were used (with diameters of 5 and 10 mm, lengths 400 and 600 mm respectively), heated electrically (50 Hz). The mixture flows from top to bottom. The work was carried out mainly on mixtures of high concentration (x > 0.1), at pressures between 50 and 60 kg/cm 2 , flowing as a liquid film on the walls of the tube with droplets suspended in the central current of steam. By analysis of the heat transfer laws the exchange mechanisms were established, and the conditions under which the critical heat flux may be exceeded without danger of actual burnout were determined. In this way high output concentrations (x s > 0.9) may be obtained. An attempt has been made to find out to what extent existing correlation formulae can be used to account for the phenomena observed. It is shown that those dealing with exchange coefficients can only be applied in a first approximation in cases where exchange by convection is preponderant, and only below the critical flux. The formulae proposed by WAPD and CISE do not give a satisfactory estimation of the critical heat flux, and the essential reasons for this inadequacy are explained. Lastly, the Martinelli and Nelson method may be used to an approximation of 30 per cent for the calculation of charge losses. (author) [fr

  15. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    International Nuclear Information System (INIS)

    Rao, Y.F.; Cheng, Z.; Waddington, G.M.

    2014-01-01

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

  16. Critical heat flux in a vertical annulus under low upward flow and near atmospheric pressure

    International Nuclear Information System (INIS)

    Schoesse, T.; Aritomi, Masanori; Lee, Sang-Ryoul; Kataoka, Yoshiaki; Yoshioka, Yuzuru; Chung, Moon-Ki.

    1997-01-01

    As future boiling water reactors (BWR), concepts of evolutional ABWR (ABWR-IER) and natural circulation BWR (JSBWR) have been investigated in order to reduce their construction cost and simplify their maintenance and inspection procedures. One of the promised features of the design of the evolutional ABWR is to reduce the number of internal pumps and to remove the Motor Generation (MG) sets. These design changes may induce boiling transition in the fuel rods of reactor core during a pump trip transient due to the more rapid flow coastdown characteristics than these of the present design. In addition, the understanding of critical heat flux (CHF) is one important subject to grasp safety margin during the start-up for the natural circulation BWR and to establish the rational start-up procedure in which thermo-hydraulic instabilities can be suppressed. The present study is to clarify CHF characteristics under low velocity conditions. CHF measurements were conducted in a vertical upward annulus channel composed of an inner heated rod and an outer tube made of glass. CHF data were obtained repeatedly under the condition of stable inlet flow to examine statistically their reproducibility. The flow regime was investigated from flow observation and measurement of differential pressure fluctuation. The CHF data are correlated with the flow regime transition. It was clear from the obtained flow pattern and the CHF data that the CHF behavior could be classified into specified regions by the mass flux and inlet subcooling conditions. A CHF correlation was developed and agreed with other researchers' data within acceptable error. (author)

  17. Critical heat flux determination in an annulus section; Determinacion del flujo critico de calor en una seccion anular

    Energy Technology Data Exchange (ETDEWEB)

    Reyes C, C A

    1997-04-01

    The present report explains the phenomenon of Critical heat flux. The study of this physical phenomenon is carried out during the boiling of a liquid and is of supreme importance for the calculation and operation of a nuclear reactor even in the moderns generators of steam (thermoelectric and nucleoelectrics), industrial cooling and in all those industrial process that use a liquid subject to sources of heating and to conditions of work excessively high (temperatures and pressures) so that stay in operation in an appropriate manner and sure. Once well-known this value, the equipment used in these process works with a maximum heat that is smaller than the Critical Heat Flux. The study of the Critical Heat Flux has achieved important advances in the last years, mainly for the enormous obligation that in this moment involved the safety to world level, this has forced to researchers and designers of this type of equipment to center their attention in the obtaining of a correlation which of general way explains it. In this reports two correlations will be compared that they contribute to the evaluation of the Critical Heat Flux in annulus and that they try to be generals in this type of geometry, the Shah correlation`s and the Katto correlation`s. The same as most of the correlations, these have been calculated so that the fluid of work is water, although they have also been proven with others fluids. The results obtained in this report only will show the degree of advance which the investigation of this phenomenon has achieved in annulus and to low amounts of flow of liquid, like which they are in the Experimental Heat Transfer Circuit located in the Department of Physics of the National Institute of Nuclear Research. (Author).

  18. External glass peening of zircaloy calandria tubes to increase the critical heat flux

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Coleman, C.E.; Nitheanandan, T.; Kroeger, V.D.; Moyer, R.G.; Sanderson, D.B.; Root, J.H.; Rogge, R.B.

    1997-12-01

    Glass-peening the outside surfaces of Zircaloy calandria tubes increases the nucleation sites available for boiling heat transfer and has been demonstrated to enhance the critical heat flux (CHF) in pool-boiling experiments. The objective of this study is to optimise the heat-transfer enhancement by glass peening while ensuring that the microstructure of the peened tube is acceptable for reactor use. Pool-boiling tests were done using small Zircaloy tubes with as-received ('smooth') surfaces and variously peened surfaces, to evaluate two peening parameters, glass-bead size and the coverage of peened surface. Our results showed that the maximum enhancement of CHF (by 60% compared with as-received tubes) was obtained using a glass-bead size of 90-125 μm with a coverage of 100%. The CHF enhancement was found to be insensitive to glass-bead size over a wide range (from 60-90 μm to 125-180 μm). Using a fixed glass-bead size of 125-180 μm to evaluate the influence of peening coverage, the maximum effect on the CHF response was obtained with a coverage of 1 00%. The microstructures of the peened tubes were evaluated using light microscopy, X-ray and neutron diffraction, and mechanical tests. After peening, the microstructure in the subsurface layer (-30 μm) consisted of deformed α-Zr grains, and the crystallographic texture of the grains changed slightly. After stress-relieving at 500 degrees C for 1 h, some recrystallisation had occurred and the residual strains remaining in the tube were low. The tensile and burst properties of glass-peened and stress-relieved tubes were similar to those of as-received tubes. The microstructures introduced by peening and stress relieving were judged to have little effect on creep and growth behaviour. Since there are no deleterious consequences of the glass-peening treatment, the peened and stress-relieved tubes are found to be acceptable for reactor use. (author)

  19. Investigation of Critical Heat Flux in Reduced Gravity Using Photomicrographic Techniques

    Science.gov (United States)

    Mudawar, Issam; Zhang, Hui

    2003-01-01

    Experiments were performed to examine the effects of body force on flow boiling critical heat flux (CHF). FC-72 was boiled along one wall of a transparent rectangular flow channel that permitted photographic study of the vapor-liquid interface just prior to CHF. High-speed video imaging techniques were used to identify dominant CHF mechanisms corresponding to different flow orientations and liquid velocities. Six different CHF regimes were identified: Wavy Vapor Layer, Pool Boiling, Stratification, Vapor Counterflow, Vapor Stagnation, and Separated Concurrent Vapor Flow. CHF showed significant sensitivity to orientation for flow velocities below 0.2 m/s, where extremely low CHF values where measured, especially with downward-facing heated wall and downflow orientations. High flow velocities dampened the effects of orientation considerably. The CHF data were used to assess the suitability of previous CHF models and correlations. It is shown the Interfacial Lift-off Model is very effective at predicting CHF for high velocities at all orientations. The flooding limit, on the other hand, is useful at estimating CHF at low velocities and for downflow orientations. A new method consisting of three dimensionless criteria is developed for determining the minimum flow velocity required to overcome body force effects on near-saturated flow boiling CHF. Vertical upflow boiling experiments were performed in pursuit of identifying the trigger mechanism for subcooled flow boiling CHF. While virtually all prior studies on flow boiling CHF concern the prediction or measurement of conditions that lead to CHF, this study was focused on events that take place during the CHF transient. High-speed video imaging and photomicrographic techniques were used to record the transient behavior of interfacial features from the last steady-state power level before CHF until the moment of power cut-off following CHF. The video records show the development of a wavy vapor layer which propagates

  20. Flow regimes and mechanistic modeling of critical heat flux under subcooled flow boiling conditions

    Science.gov (United States)

    Le Corre, Jean-Marie

    Thermal performance of heat flux controlled boiling heat exchangers are usually limited by the Critical Heat Flux (CHF) above which the heat transfer degrades quickly, possibly leading to heater overheating and destruction. In an effort to better understand the phenomena, a literature review of CHF experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available data. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. It is postulated that a high local wall superheat occurs locally in a dry area of the heated wall, due to a cyclical event inherent to the considered CHF two-phase flow regime, preventing rewetting (Leidenfrost effect). The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow. A numerical model using a two-dimensional transient thermal analysis of the heater undergoing nucleation was developed to mechanistically predict CHF in the case of a bubbly flow regime. In this type of CHF two-phase flow regime, the high local wall superheat occurs underneath a nucleating bubble at the time of bubble departure. The model simulates the spatial and temporal heater temperature variations during nucleation at the wall, accounting for the stochastic nature of the boiling phenomena. The model has also the potential to evaluate

  1. A study on the critical heat flux for annuli and round tubes under low pressure conditions

    International Nuclear Information System (INIS)

    Park, Jae Wook

    1997-02-01

    This study aims to reveal the characteristics of the critical heat flux (CHF) of internally heated concentric annuli and vertical round tubes in low-pressure and low-flow (LPLF) conditions. Although many efforts have been devote to the subject of the CHF during the last forty years, the information on the CHF phenomenon for LPLF conditions is still very limited. The applicable ranges of the CHF correlations for annuli and round tubes are concentrate on the operating conditions of nuclear power plant (NPP), namely high-pressure and high-flow (HPHF) conditions. these facts promoted to collect the reliable CHF data for LPLF conditions for both annuli and round tubes. The critical heat flux data for vertical flow boiling of water in annuli and round tubes at low pressures and low mass fluxes show the following trends: The observed CHF mechanism for annuli was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. The observed parametric trends for annuli are consistent with the previous understanding except that the CHF for downward flow is considerably lower (up to 40%) than that for upward flow. The critical quality is much lower than that for round tubes at the same inlet conditions. The observed parametric trends for round tubes are generally consistent with the previous understanding except for system pressure an tube diameter effect. For the system pressure effect, it is observed that the pressure effect is complicated but not so large, whereas the existing CHF correlations do not present the parametric trend exactly. For tube diameter effect, the decreasing trends of CHF with respect to tube diameter was the general understanding so far, but in this region the CHF show a increasing trend of tube diameter. The prediction and the parametric trend analyses are performed by two view points, I.e., for fixed inlet conditions and for local

  2. Improving the understanding of thermal-hydraulics and heat transfer for super critical water cooled reactors

    International Nuclear Information System (INIS)

    Bilbao y Leon, S.; Aksan, N.

    2010-01-01

    Ensuring the exchange of information and fostering the collaboration among Member States on the development of technology advances for future nuclear power plants are among the key roles of the IAEA. There is high interest internationally in both developing and industrialized countries in the design of innovative super-critical water-cooled reactors (SCWRs). This interest arises from the high thermal efficiencies (44-45%) and improved economic competitiveness promised by for this concept, utilizing and building on the recent developments of highly efficient fossil power plants. The SCWR is one of the six concepts included in the Generation-IV International Forum (GIF). Following the advice of the IAEA Nuclear Energy Dept.'s Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA is working on a Coordinated Research Project (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The second Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna (Austria)) in August 2009. This paper summarizes the current status of the CRP, as well as the major achievements to date. (authors)

  3. On the look-up tables for the critical heat flux in tubes (history and problems)

    International Nuclear Information System (INIS)

    Kirillov, P.L.; Smogalev, I.P.

    1995-01-01

    The complication of critical heat flux (CHF) problem for boiling in channels is caused by the large number of variable factors and the variety of two-phase flows. The existence of several hundreds of correlations for the prediction of CHF demonstrates the unsatisfactory state of this problem. The phenomenological CHF models can provide only the qualitative predictions of CHF primarily in annular-dispersed flow. The CHF look-up tables covered the results of numerous experiments received more recognition in the last 15 years. These tables are based on the statistical averaging of CHF values for each range of pressure, mass flux and quality. The CHF values for regions, where no experimental data is available, are obtained by extrapolation. The correction of these tables to account for the diameter effect is a complicated problem. There are ranges of conditions where the simple correlations cannot produce the reliable results. Therefore, diameter effect on CHF needs additional study. The modification of look-up table data for CHF in tubes to predict CHF in rod bundles must include a method which to take into account the nonuniformity of quality in a rod bundle cross section

  4. Prediction of critical heat flux in fuel assemblies using a CHF table method

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; Hwang, Dae Hyun; Bang, Je Geon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Baek, Won Pil; Chang, Soon Heung [Korea Advance Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor. 8 refs., 3 figs., 3 tabs. (Author)

  5. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

    International Nuclear Information System (INIS)

    Jeong, Yong Hoon; Chang, Soon Heung; Baek, Won-Pil

    2005-01-01

    The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m 2 .s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m 2 .s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux ( 2 .s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water

  6. A critical heat flux correlation for advanced pressurized light water reactor application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Hame, W.

    1982-05-01

    Many CHF-correlations have been developed for water cooled rod clusters representing typical PWR or BWR fuel element geometries with relative wide rod lattices. However the fuel elements of an Advanced Pressurized Water Reactor (APWR) have a tight fuel rod lattice, in view of increasing the fuel utilization. It was therefore decided to produce a new CHF-correlation valid for rod bundles with tight lattices. The already available WSC-2 correlation was chosen as a basis. The geometry dependent parameters of this correlation were determined again with the method of the root mean square fitting from the experimental data of the CHF-tests performed in the frame of the Light Water Breeder Reactor programme at the Bettis Laboratory. These tests include triangular array rod bundles with very tight lattices. Furthermore the effect of spiral spacer ribs was investigated on the basis of experimental data from the Columbia University. Application of the new CHF-correlation to conditions typical for an APWR shows that the predicted critical heat fluxes are much smaller than those calculated with the usual PWR-CHF-correlations, but they are higher than those predicted by the B+W-VPI+SU correlation. (orig.) [de

  7. Prediction of critical heat flux in fuel assemblies using a CHF table method

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Tae Hyun; Hwang, Dae Hyun; Bang, Je Geon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Baek, Won Pil; Chang, Soon Heung [Korea Advance Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor. 8 refs., 3 figs., 3 tabs. (Author)

  8. On the look-up tables for the critical heat flux in tubes (history and problems)

    Energy Technology Data Exchange (ETDEWEB)

    Kirillov, P.L.; Smogalev, I.P. [Institute of Physics and Power Engineering, Kaluga (Russian Federation)

    1995-09-01

    The complication of critical heat flux (CHF) problem for boiling in channels is caused by the large number of variable factors and the variety of two-phase flows. The existence of several hundreds of correlations for the prediction of CHF demonstrates the unsatisfactory state of this problem. The phenomenological CHF models can provide only the qualitative predictions of CHF primarily in annular-dispersed flow. The CHF look-up tables covered the results of numerous experiments received more recognition in the last 15 years. These tables are based on the statistical averaging of CHF values for each range of pressure, mass flux and quality. The CHF values for regions, where no experimental data is available, are obtained by extrapolation. The correction of these tables to account for the diameter effect is a complicated problem. There are ranges of conditions where the simple correlations cannot produce the reliable results. Therefore, diameter effect on CHF needs additional study. The modification of look-up table data for CHF in tubes to predict CHF in rod bundles must include a method which to take into account the nonuniformity of quality in a rod bundle cross section.

  9. Investigation on premature occurrence of critical heat flux under oscillatory flow and power conditions

    International Nuclear Information System (INIS)

    Vishnoi, A.K.; Dasgupta, A.; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.

    2015-01-01

    Two-phase natural circulation loops have extensive applications in nuclear and process industries. One of the major concerns with natural circulation is the occurrence of the various types of flow instabilities, which can cause premature boiling crisis due to flow and power oscillations. In this work a transient computer code COPCOS (Code for Prediction of CHF under Oscillating flow and power condition) has been developed to predict the premature occurrence of CHF (critical heat flux) under oscillating flow and power. The code incorporates conduction equation of the fuel and coolant energy equation. For CHF prediction, CHF look-up table developed by Groeneveld is used. A facility named CHF and Instability Loop (CHIL) has been set up to study the effect of oscillatory flow on CHF. CHF and Instability Loop (CHIL) is a simple rectangular loop having a 10.5 mm ID and 1.2 m long test section. The flow through the test section is controlled by a canned motor pump using a Variable Frequency Drive (VFD). This leads to the ability of having a very precise control over flow oscillations which can be induced in the test section. The effect of frequency and amplitude of flow oscillation on occurrence of premature CHF has been investigated in this facility using COPCOS. Full paper covers details of COPCOS code, description of the facility and effect of frequency and the effect of oscillatory flow on CHF in the facility. (author)

  10. The critical role of extreme heat for maize production in the United States

    Science.gov (United States)

    Lobell, David B.; Hammer, Graeme L.; McLean, Greg; Messina, Carlos; Roberts, Michael J.; Schlenker, Wolfram

    2013-05-01

    Statistical studies of rainfed maize yields in the United States and elsewhere have indicated two clear features: a strong negative yield response to accumulation of temperatures above 30°C (or extreme degree days (EDD)), and a relatively weak response to seasonal rainfall. Here we show that the process-based Agricultural Production Systems Simulator (APSIM) is able to reproduce both of these relationships in the Midwestern United States and provide insight into underlying mechanisms. The predominant effects of EDD in APSIM are associated with increased vapour pressure deficit, which contributes to water stress in two ways: by increasing demand for soil water to sustain a given rate of carbon assimilation, and by reducing future supply of soil water by raising transpiration rates. APSIM computes daily water stress as the ratio of water supply to demand, and during the critical month of July this ratio is three times more responsive to 2°C warming than to a 20% precipitation reduction. The results suggest a relatively minor role for direct heat stress on reproductive organs at present temperatures in this region. Effects of elevated CO2 on transpiration efficiency should reduce yield sensitivity to EDD in the coming decades, but at most by 25%.

  11. An analysis of critical heat flux on the external surface of the reactor vessel lower head

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Baek, Won Pil; Chang, Soon Heung

    1999-01-01

    CHF (Critical heat flux) on the external surface of the reactor vessel lower head is major key in the evaluation on the feasibility of IVR-EVC (In-Vessel Retention through External Vessel Cooling) concept. To identify the CHF on the external surface, considerable works have been performed. Through the review on the previous works related to the CHF on the external surface, liquid subcooling, induced flow along the external surface, ICI (In-Core Instrument) nozzle and minimum gap are identified as major parameters. According to the present analysis, the effects of the ICI nozzle and minimum gap on CHF are pronounced at the upstream of test vessel: on the other hand, the induced flow considerably affects the CHF at downstream of test vessel. In addition, the subcooling effect is shown at all of test vessel, and decreases with the increase in the elevation of test vessel. In the real application of the IVR-EVC concept, vertical position is known as a limiting position, at which thermal margin is the minimum. So, it is very important to precisely predict the CHF at vertical position in a viewpoint of gaining more thermal margins. However, the effects of the liquid subcooling and induced flow do not seem to be adequately included in the CHF correlations suggested by previous works, especially at the downstream positions

  12. Particle swarm optimization-based least squares support vector regression for critical heat flux prediction

    International Nuclear Information System (INIS)

    Jiang, B.T.; Zhao, F.Y.

    2013-01-01

    Highlights: ► CHF data are collected from the published literature. ► Less training data are used to train the LSSVR model. ► PSO is adopted to optimize the key parameters to improve the model precision. ► The reliability of LSSVR is proved through parametric trends analysis. - Abstract: In view of practical importance of critical heat flux (CHF) for design and safety of nuclear reactors, accurate prediction of CHF is of utmost significance. This paper presents a novel approach using least squares support vector regression (LSSVR) and particle swarm optimization (PSO) to predict CHF. Two available published datasets are used to train and test the proposed algorithm, in which PSO is employed to search for the best parameters involved in LSSVR model. The CHF values obtained by the LSSVR model are compared with the corresponding experimental values and those of a previous method, adaptive neuro fuzzy inference system (ANFIS). This comparison is also carried out in the investigation of parametric trends of CHF. It is found that the proposed method can achieve the desired performance and yields a more satisfactory fit with experimental results than ANFIS. Therefore, LSSVR method is likely to be suitable for other parameters processing such as CHF

  13. A study of critical heat flux in the fuel assembly dummies with various types of mixing grids

    International Nuclear Information System (INIS)

    Bezrukov, Yu. A.; Lisenkov, E. A.; Astakhov, V. I.; Vasilchenko, I. N.

    2013-01-01

    The report deals with the results of a study The report deals with the results of a study of critical heat fluxes (CHF) on the rod bundles equipped with mixing grids – intensifier of heat exchange. The study was carried out on FA dummies equipped with various kinds of mixing grids: “Vikhr” and “Sector screw die”. The tests were carried out at the following parameters: pressure 14 and 16 MPa; mass velocity 2000, 3000 and 4000 kg/(m2⋅s); relative enthalpy from minus 0,07 up to 0,18. Bundles with uniform and non-uniform axial heat generation were tested. The results of the experiments are intended for the elaboration of a correlation to be used in design development.of critical heat fluxes (CHF) on the rod bundles equipped with mixing grids – intensifier of heat exchange. The study was carried out on FA dummies equipped with various kinds of mixing grids: “Vikhr” and “Sector screw die”. The tests were carried out at the following parameters: pressure 14 and 16 MPa; mass velocity 2000, 3000 and 4000 kg/(m 2 ⋅s); relative enthalpy from minus 0,07 up to 0,18. Bundles with uniform and non-uniform axial heat generation were tested. The results of the experiments are intended for the elaboration of a correlation to be used in design development. (authors)

  14. New Westinghouse correlation WRB-1 for predicting critical heat flux in rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Motley, F.E.; Hill, K.W.; Cadek, F.F.; Shefcheck, J.

    1976-07-01

    A new critical heat flux (CHF) correlation, based on local fluid conditions, has been developed from Westinghouse rod bundle data. This correlation applies to both 0.422 inch and 0.374 inch rod O.D. geometries. It accounts for typical cell and thimble cell effects, uniform and non-uniform heat flux profiles, variations in rod heated length and in grid spacing. The correlation predicts CHF for 1147 data points with a sample mean and standard deviation of measured-to-predicted heat flux ratio of 1.0043 and 0.0873, respectively. It was concluded that to meet the reactor design criterion the minimum DNBR should be 1.17

  15. Critical heat flux and flow instability in an advanced light water reactor

    International Nuclear Information System (INIS)

    Dae-Hyun Hwang; Kyong-Won Seo; Chung-Chan Lee; Sung-Kyun Zee

    2005-01-01

    Full text of publication follows: An advanced light water reactor concept has been continuously studied in KAERI with an output in the range of about 60 to 300 MW th . The reactor is purposed to be utilized as an energy source for seawater desalination as well as small scale power generation. In order to achieve the intrinsic safety and enhanced operational flexibility, some specific design considerations such as low power density and soluble boron free operation have been incorporated in the multiple-parallel-channel type reactor core. The low power density can be achieved by adopting fuel assemblies with tightly spaced non-square lattice rod array. The allowable core operating region should be primarily limited by the two design parameters; the critical heat flux(CHF) and the flow instabilities in the multiple parallel fuel assembly channels. The characteristics of CHF and flow instability have been investigated through experimental and analytical works. The CHF prediction model was established on the basis of experimental data obtained from 19-rod test bundles. The CHF experiments have been conducted for various test bundles with different heated lengths, uniform and non-uniform radial and axial power distributions, water and Freon as the working fluids, and different number of unheated rods. The parametric ranges of CHF experiments covers the pressure from 6 to 18 MPa, the mass flux from 150 to 2000 kg/m 2 /s, and the inlet subcooling from 10 to 120 deg. C. The flow instabilities due to density wave oscillations were investigated by conducting experiments with two parallel channels under the pressure ranges from 6 to 16 MPa. The parametric behavior of flow instability was examined for the test sections with different lengths of adiabatic risers, different axial power shapes, different inlet restrictions, and different channel cross sections. The stability boundary was experimentally determined by increasing channel inlet temperature or reducing the flow rate

  16. An analytic model of pool boiling critical heat flux on an immerged downward facing curved surface

    International Nuclear Information System (INIS)

    He, Hui; Pan, Liang-ming; Wu, Yao; Chen, De-qi

    2015-01-01

    Highlights: • Thin liquid film and supplement of liquid contribute to the CHF. • CHF increases from the bottom to the upper of the lowerhead. • Evaporation of thin liquid film is dominant nearby bottom region. • The subcooling has significant effects on the CHF. - Abstract: In this paper, an analytical model of the critical heat flux (CHF) on the downward facing curved surface for pool boiling has been proposed, which hypothesizes that the CHF on the downward facing curved is composed of two parts, i.e. the evaporation of the thin liquid film underneath the elongated bubble adhering to the lower head outer surface and the depletion of supplement of liquid due to the relative motion of vapor bubbles along with the downward facing curved. The former adopts the Kelvin–Helmholtz instability analysis of vapor–liquid interface of the vapor jets which penetrating in the thin liquid film. When the heat flux closing to the CHF point, the vapor–liquid interface becomes highly distorted, which block liquid to feed the thin liquid film and the thin liquid film will dry out gradually. While the latter considers that the vapor bubbles move along with the downward facing curved surface, and the liquid in two-phase boundary layer enter the liquid film that will be exhausted when the CHF occurs. Based on the aforementioned mechanism and the energy balance between the thin liquid film evaporation and water feeding, and taking the subcooling of the bulk water into account, the mathematic model about the downward facing curved surface CHF has been proposed. The CHF of the downward facing curved surface for pool boiling increases along with the downward facing orientation except in the vicinity of bottom center region, because in this region the vapor bubble almost stagnates and the evaporation of the thin liquid film is dominant. In addition, the subcooling has significant effect on the CHF. Comparing the result of this model with the published experimental results show

  17. Prediction of critical heat flux in narrow rectangular channels using an artificial neural network

    International Nuclear Information System (INIS)

    Zhou Lei; Yan Xiao; Huang Yanping; Xiao Zejun; Yu Jiyang

    2011-01-01

    The concept of Critical heat flux (CHF) and its importance are introduced and the meaning to research CHF in narrow rectangular channels independently is emphasized. This paper is the first effort to predict CHF in NRCs using aritificial neural network. The mathematical structure of the artificial neural network and the error back-propagation algorithm are introduced. To predict CHF, the four dimensionless groups are inputted to the neural network and the output is the dimensionless CHF. As the hidden nodes increased, the training error decreases while the testing error decreases firstly and then transition occurs. Based on this, the hidden nodes are set as 5 and the trained network predicts all of the training and testing data points with RMS=0.0016 and μ=1.0003, which is better than several well-known existing correlations. Based on the trained network, the effect of several parameters on CHF are simulated and discussed. CHF increases almost linearly as the inlet subcooling increases. And larger mass flux enhances the effect of the inlet subcooling. CHF increases with the mass flux increasing. And the effect seems to be a little stronger for relatively low system pressure. CHF decreases almost linearly as the system pressure increases for the fixed inlet condition. The slope of the curve also increases with higher mass flux. This observation is limited to the ranges of the experimental database. CHF decreases as the heated length is increased and the gradients of the curves become very sharp for relatively short channel. CHF increases slightly with the diameter increasing with the variance of the gap limited within 1 to 3 mm. For relatively low mass flux, the effect of the equivalent diameter on CHF is insignificant. As the width of the channel is large enough, the effect of the gap is quite the same as that of the equivalent diameter. A BPNN is successfully trained based on near 500 CHF data points in NRCs, which has much better performances than the

  18. Electron critical gradient scale length measurements of ICRF heated L-mode plasmas at Alcator C-Mod tokamak

    Science.gov (United States)

    Houshmandyar, S.; Hatch, D. R.; Horton, C. W.; Liao, K. T.; Phillips, P. E.; Rowan, W. L.; Zhao, B.; Cao, N. M.; Ernst, D. R.; Greenwald, M.; Howard, N. T.; Hubbard, A. E.; Hughes, J. W.; Rice, J. E.

    2018-04-01

    A profile for the critical gradient scale length (Lc) has been measured in L-mode discharges at the Alcator C-Mod tokamak, where electrons were heated by an ion cyclotron range of frequency through minority heating with the intention of simultaneously varying the heat flux and changing the local gradient. The electron temperature gradient scale length (LTe-1 = |∇Te|/Te) profile was measured via the BT-jog technique [Houshmandyar et al., Rev. Sci. Instrum. 87, 11E101 (2016)] and it was compared with electron heat flux from power balance (TRANSP) analysis. The Te profiles were found to be very stiff and already above the critical values, however, the stiffness was found to be reduced near the q = 3/2 surface. The measured Lc profile is in agreement with electron temperature gradient (ETG) models which predict the dependence of Lc-1 on local Zeff, Te/Ti, and the ratio of the magnetic shear to the safety factor. The results from linear Gene gyrokinetic simulations suggest ETG to be the dominant mode of turbulence in the electron scale (k⊥ρs > 1), and ion temperature gradient/trapped electron mode modes in the ion scale (k⊥ρs < 1). The measured Lc profile is in agreement with the profile of ETG critical gradients deduced from Gene simulations.

  19. Critical heat flux under zero flow conditions in a vertical 3 X 3 rod bundle with a non-uniform axial heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seok; Chun, Se Young; Moon, Sang Ki; Baek, Won Pil

    2003-11-01

    KAERI has performed an experimental study of water Critical Heat Flux (CHF) under zero flow conditions with a non-uniformly heated 3 by 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.5 to 15.0 MPa and inlet water subcooling enthalpies from 67.5 to 351.5 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 by 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a consistent parametric trend. The CHFs occur in the upper region of the heated section, but the vertical distances of the detected CHFs from the bottom of the heated section are reduced as the system pressure increases. Even though the effects of the inlet water subcooling enthalpies and system pressure in the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations.

  20. Study on Enhancement of Sub-Cooled Flow Boiling Heat Transfer and Critical Heat Flux of Solid-Water Two-Phase Mixture

    International Nuclear Information System (INIS)

    Yasuo Koizumi; Hiroyasu Ohtake; Tomoyuki Suzuki

    2002-01-01

    The influence of particle introduction into a subcooled water flow on boiling heat transfer and critical heat flux (CHF) was examined. When the water velocity was low, the particles crowded on the bottom wall of the flow channel and flowed just like sliding on the wall. When the water velocity was high, the particles were well dispersed in the water flow. In the non-boiling region, the heat transfer was augmented by the introduction of the particles into the water flow. As the introduction of the particles were increased, the augmentation was also increased in the high water flow rate region. However, it was independent upon the particle introduction rate in the low water flow rate region. The onset of boiling was delayed by the particle inclusion. The boiling heat transfer was enhanced by the particles. However, it was rather decreased in the high heat flux fully-developed-boiling region. The CHF was decreased by the particle inclusion in the low water flow region and was not affected in the high water flow region. (authors)

  1. An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the Heat Transfer Research Facility

    International Nuclear Information System (INIS)

    Min Lee

    1994-01-01

    Critical heat flux (CHF) bundle data from the Heat Transfer Research Facility of Columbia University are used to check the validity of the CHF approaches used in thermal-hydraulic system analysis codes for light water reactors. The CHF approaches assessed include the Biasi et al. correlation of TRAC, the Groeneveld et al. CHF table lookup approach of RELAP5/MOD3, the CHF table lookup approach of CATHARE, and the CHF approach of RETRAN. Depending on system pressure, RETRAN uses the B and W2, Barnett, and modified Barnett correlations and a linear interpolation scheme to predict CHF. Results show that among these CHF approaches, the Groeneveld et al. approach has the best prediction accuracy and the smallest uncertainty in the estimation of the HTRF bundle data. On the average, the Groeneveld et al. approach overpredicts the uniform axial heat flux distribution by 3.6% and the nonuniform axial heat flux distribution by 0.9%. The performance of the RETRAN approach is comparable with that of the Groenevel et al. Approach for uniform axial heat flux. In general, the accuracy and the uncertainty of all the approaches, except that of CATHARE, are worse under a nonuniform axial heat distribution than under a uniform axial heat distribution. All the CHF approaches assessed have a tendency to overpredict the HTRF bundle data at low pressure, low measured CHF, and high CHF quality. The performance of the Groenevel et al. approach is improved through a CHF table update and modification of the bundle correction factor using the HTRF bundle data

  2. Measurements of convective heat transfer to vertical upward flows of CO{sub 2} in circular tubes at near-critical and supercritical pressures

    Energy Technology Data Exchange (ETDEWEB)

    Zahlan, H., E-mail: hussamzahlan@gmail.com [Canadian Nuclear Laboratories, Chalk River, K0J 1J0 (Canada); Department of Mechanical Engineering, University of Ottawa, Ottawa, ON K1N 6N5 (Canada); Groeneveld, D. [Canadian Nuclear Laboratories, Chalk River, K0J 1J0 (Canada); Department of Mechanical Engineering, University of Ottawa, Ottawa, ON K1N 6N5 (Canada); Tavoularis, S. [Department of Mechanical Engineering, University of Ottawa, Ottawa, ON K1N 6N5 (Canada)

    2015-08-15

    Highlights: • We present and discuss results of thermal–hydraulic measurements in CO{sub 2} for the near critical and supercritical pressure region. • We report the full heat transfer and pressure drop database. - Abstract: An extensive experimental program of heat transfer measurements has been completed recently at the University of Ottawa's supercritical pressure test facility (SCUOL). Thermal–hydraulics tests were performed for vertical upflow of carbon dioxide in directly heated tubes with inner diameters of 8 and 22 mm, at high subcritical, near-critical and supercritical pressures. The test conditions, when converted to water-equivalent values, correspond to conditions of interest to current Super-Critical Water-Cooled Reactor designs, and include many measurements under conditions for which few data are available in the literature. These data significantly complement the existing experimental database and are being used for the derivation and validation of a new heat transfer prediction method in progress at the University of Ottawa. The same data are also suitable for the assessment of the accuracy of other heat transfer prediction methods and fluid-to-fluid scaling laws for near-critical and supercritical pressures. In addition, they permit further examination of previously suggested relationships describing the critical heat flux and post-dryout heat transfer coefficient at high subcritical pressures and the boundaries of the deteriorated/enhanced heat transfer regions for near-critical and supercritical pressures. The measurements reported in this paper cover several subcritical heat transfer modes, including single phase liquid heat transfer, nucleate boiling, critical heat flux, post-dryout heat transfer and superheated vapor heat transfer; they also cover several supercritical heat transfer modes, including heat transfer to liquid-like supercritical fluid and heat transfer to vapor-like supercritical fluid, which occurred in the

  3. Heat capacities and asymmetric criticality of the (liquid + liquid) coexistence curves for {dimethyl carbonate + n-undecane, or n-tridecane}

    International Nuclear Information System (INIS)

    Chen, Zhiyun; Shi, Aiqin; Liu, Shixia; Yin, Tianxiang; Shen, Weiguo

    2014-01-01

    Highlights: • Coexistence curves of dimethyl carbonate + n-undecane (or + n-tridecane) were measured. • Isobaric heat capacity per unit volume of critical binary solutions dimethyl carbonate + n-undecane (or + n-tridecane) were determined. • The critical exponent β are consistent with the 3D-Ising value. • The asymmetry of the coexistence curves were discussed by the complete scaling theory. - Abstract: The (liquid + liquid) coexistences and the critical behavior of isobaric heat capacity per unit volume for critical binary solutions {dimethyl carbonate + n-undecane, or n-tridecane} have been studied. The critical exponents β and α were deduced and found to be consistent with the 3D-Ising values. The critical amplitudes were determined and used to test the asymmetric criticality of coexistence curves. It was found that the heat capacity does play an important role in describing the asymmetric criticality of the coexistence curves

  4. Study on the effect of the CANFLEX-NU fuel element bowing on the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Cho, Moon Sung; Jeon, Ji Su

    2001-01-01

    The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance.

  5. Generalized correlation of latent heats of vaporization of coal liquid model compounds between their freezing points and critical points

    Energy Technology Data Exchange (ETDEWEB)

    Sivaraman, A.; Kobuyashi, R.; Mayee, J.W.

    1984-02-01

    Based on Pitzer's three-parameter corresponding states principle, the authors have developed a correlation of the latent heat of vaporization of aromatic coal liquid model compounds for a temperature range from the freezing point to the critical point. An expansion of the form L = L/sub 0/ + ..omega..L /sub 1/ is used for the dimensionless latent heat of vaporization. This model utilizes a nonanalytic functional form based on results derived from renormalization group theory of fluids in the vicinity of the critical point. A simple expression for the latent heat of vaporization L = D/sub 1/epsilon /SUP 0.3333/ + D/sub 2/epsilon /SUP 0.8333/ + D/sub 4/epsilon /SUP 1.2083/ + E/sub 1/epsilon + E/sub 2/epsilon/sup 2/ + E/sub 3/epsilon/sup 3/ is cast in a corresponding states principle correlation for coal liquid compounds. Benzene, the basic constituent of the functional groups of the multi-ring coal liquid compounds, is used as the reference compound in the present correlation. This model works very well at both low and high reduced temperatures approaching the critical point (0.02 < epsilon = (T /SUB c/ - T)/(T /SUB c/- 0.69)). About 16 compounds, including single, two, and three-ring compounds, have been tested and the percent root-mean-square deviations in latent heat of vaporization reported and estimated through the model are 0.42 to 5.27%. Tables of the coefficients of L/sub 0/ and L/sub 1/ are presented. The contributing terms of the latent heat of vaporization function are also presented in a table for small increments of epsilon.

  6. Preliminary review of critical shutdown heat removal items for common cause failure susceptibility on LMFBR's. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Allard, L.T.; Elerath, J.G.

    1976-02-01

    This document presents a common cause failure analysis for Critical LMFBR Shutdown Heat Removal Systems. The report is intended to outline a systematic approach to defining areas with significant potential for common causes of failure, and ultimately provide inputs to the reliability prediction model. A preliminary evaluation of postulatd single initiating causes resulting in multiple failures of LMFBR-SHRS items is presented in Appendix C. This document will be periodically updated to reflect new information and activity.

  7. The current status of theoretically based approaches to the prediction of the critical heat flux in flow boiling

    International Nuclear Information System (INIS)

    Weisman, J.

    1991-01-01

    This paper reports on the phenomena governing the critical heat flux in flow boiling. Inducts which vary with the flow pattern. Separate models are needed for dryout in annular flow, wall overheating in plug or slug flow and formation of a vapor blanket in dispersed flow. The major theories and their current status are described for the annular and dispersed regions. The need for development of the theoretical approach in the plug and slug flow region is indicated

  8. Critical heat flux in vertical flows at low pressures; Flux de chaleur critique en ecoulements verticaux aux pressions faibles

    Energy Technology Data Exchange (ETDEWEB)

    Olekhnowitch, A [Ecole Polytechnique, Montreal, PQ (Canada)

    1994-12-31

    This paper presents some critical heat flux (CHF) data obtained for vertical upflow of water in an 8 mm test section, for exit pressures ranging from 5 to 30 bar. The experiments were carried out for heated lengths of 0.75, 1, 1.4 and 1.8 m. In general, the collected data show trends similar to those described in the open literature. However, it was observed that for low pressures CHF depends on the heated length; this dependence begins to disappear for exit pressure of about 30 bar. The data have been compared with a look-up table and predictions of well known correlations. For low pressures and low mass fluxes, the look-up table seems to give better predictions, but for medium pressures and mass fluxes, the correlations perform better. 19 refs., 5 figs.

  9. A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Sang Jun; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling. 16 refs., 6 figs., 1 tab. (Author)

  10. A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Sang Jun; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling. 16 refs., 6 figs., 1 tab. (Author)

  11. Critical temperature gradient length signatures in heat wave propagation across internal transport barriers in the Joint European Torus

    International Nuclear Information System (INIS)

    Casati, Alessandro; Mantica, P.; Eester, D. van; Hawkes, N.; De Vries, P.; Imbeaux, F.; Joffrin, E.; Marinoni, A.; Ryter, F.; Salmi, A.; Tala, T.

    2007-01-01

    New results on electron heat wave propagation using ion cyclotron resonance heating power modulation in the Joint European Torus (JET) [P. H. Rebut et al., Nucl. Fusion 25, 1011 (1985)] plasmas characterized by internal transport barriers (ITBs) are presented. The heat wave generated outside the ITB, and traveling across it, always experiences a strong damping in the ITB layer, demonstrating a low level of transport and loss of stiffness. In some cases, however, the heat wave is strongly inflated in the region just outside the ITB, showing features of convective-like behavior. In other cases, a second maximum in the perturbation amplitude is generated close to the ITB foot. Such peculiar types of behavior can be explained on the basis of the existence of a critical temperature gradient length for the onset of turbulent transport. Convective-like features appear close to the threshold (i.e., just outside the ITB foot) when the value of the threshold is sufficiently high, with a good match with the theoretical predictions for the trapped electron mode threshold. The appearance of a second maximum is due to the oscillation of the temperature profile across the threshold in the case of a weak ITB. Simulations with an empirical critical gradient length model and with the theory based GLF23 [R. E. Waltz et al., Phys. Plasmas, 4, 2482 (1997)] model are presented. The difference with respect to previous results of cold pulse propagation across JET ITBs is also discussed

  12. A technical basis for the flux corrected local conditions critical heat flux correlation

    International Nuclear Information System (INIS)

    Luxat, J.C.

    2008-01-01

    The so-called 'flux-corrected' local conditions CHF correlation was developed at Ontario Hydro in the 1980's and was demonstrated to successfully correlate the Onset of Intermittent Dryout (OID) CHF data for 37-element fuel with a downstream-skewed axial heat flux distribution. However, because the heat flux correction factor appeared to be an ad-hoc, albeit a successful modifying factor in the correlation, there was reluctance to accept the correlation more generally. This paper presents a thermalhydraulic basis, derived from two-phase flow considerations, that supports the appropriateness of the heat flux correction as a local effects modifying factor. (author)

  13. Critical evaluation of molybdenum and its alloys for use in space reactor core heat pipes

    International Nuclear Information System (INIS)

    Lundberg, L.B.

    1981-01-01

    The choice of pure molybdenum as the prime candidate material for space reactor core heat pipes is examined, and the advantages and disadvantages of this material are brought into focus. Even though pure molybdenum heat pipes have been built and tested, this metal's high ductile-brittle transition temperature and modest creep strength place significant design restrictions on a core heat pipe made from it. Molybdenum alloys are examined with regard to their promise as potential replacements for pure molybdenum. The properties of TZM and molybdenum-rhenium alloys are examined, and it appears that Mo-Re alloys with 10 to 15 wt % rhenium offer the most advantage as an alternative to pure molybdenum in space reactor core heat pipes

  14. Numerical simulation and experimental validation of internal heat exchanger influence on CO{sub 2} trans-critical cycle performance

    Energy Technology Data Exchange (ETDEWEB)

    Rigola, Joaquim; Ablanque, Nicolas; Perez-Segarra, Carlos D.; Oliva, Assensi [Centre Tecnologic de Transferencia de Calor (CTTC), Universitat Politecnica de Catalunya (UPC), ETSEIAT, C. Colom 11, 08222 Terrassa (Barcelona) (Spain)

    2010-06-15

    The present paper is a numerical and experimental comparative study of the whole vapour compression refrigerating cycle in general, and reciprocating compressors in particular, with the aim of showing the possibilities that CO{sub 2} offers for commercial refrigeration, considering a single-stage trans-critical cycle using semi-hermetic reciprocating compressors under small cooling capacity systems. The present work is focussed on the influence of using an internal heat exchanger (IHX) in order to improve the cycle performance under real working conditions. In order to validate the numerical results, an experimental unit specially designed and built to analyze trans-critical refrigerating equipments considering IHX has been built. Both numerical results and experimental data show reasonable good agreement, while the comparative global values conclude the improvement of cooling capacity and COP when IHX is considered in the CO{sub 2} trans-critical cycle. (author)

  15. Influence of the Particle Length of Carbon Nanotube for Pool Boiling Critical Heat Flux Enhancement of Nanofluids

    International Nuclear Information System (INIS)

    Park, Sung Seek; Kim, Yong Hwan; Kim, Nam Jin

    2013-01-01

    The results of this experiment were that the CHF of the two nanofluids increased along with the volumetric fraction until 0.001 vol%, and the two types of nanofluids are the highest CHF at 0.001 vol%. Also, the results show clearly that the rate of CHF increase of the CM-100 MWCNT nanofluid with longer-length nanoparticles is higher than that of the CM-95 MWNCT nanofluid. These results indicate that the length of carbon nanotube influences the pool boiling CHF of carbon nanotube nanofluid and that long-length MWCNT, as above-noted, offers a superior effect in this regard. Boiling heat transfer is used in a variety of industrial processes and applications, such as refrigeration, power generation, heat exchangers, cooling of high-power electronics components and cooling of nuclear reactors. The critical heat flux (CHF) phenomenon is the thermal limit during a boiling heat transfer phase change; at the CHF point the heat transfer is maximised, followed by a drastic degradation after the CHF point. The consequence is a substantial increase in wall temperature which may result in physical failure phenomenon of heat transfer systems. Therefore, the CHF is important being considered in the cooling device design, such as nuclear reactor and nuclear fuels, steam generators, high-density electronic component, etc. And, CHF enhancement is essential for safety of heat transfer system. Recently, CHF reported increased when applied to the nanofluids, with its high (higher-than-base-fluid) thermal characteristic in the nuclear power plant system. Therefore, in this study, carried out the pool boiling CHF experiments by the particle length using carbon nanotube nanofluids, and the results are compared and analyzed for the CHF enhancement. The pool boiling CHF of experiments of carbon nanotube nanofluids carried out by the length of particles and the various concentrations

  16. Mechanism of subcooled water flow boiling critical heat flux in a circular tube at high liquid Reynolds number

    International Nuclear Information System (INIS)

    Hata, K.; Fukuda, K.; Masuzaki, S.

    2014-01-01

    The subcooled boiling heat transfer and the steady state critical heat flux (CHF) in a vertical circular tube for the flow velocities (u=3.95 to 30.80 m/s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter (d=6 mm) and heated length (L=59.5 mm) is used in this work. The outer surface temperatures of the SUS304 test tube with heating are observed by an infrared thermal imaging camera and a video camera. The subcooled boiling heat transfers for SUS304 test tube are compared with the values calculated by other workers' correlations for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated into details based on the experimental data. Nucleate boiling surface superheats at the CHF are close to the lower limit of the heterogeneous spontaneous nucleation temperature and the homogeneous spontaneous nucleation temperature. The dominant mechanism of the subcooled flow boiling CHF on the SUS304 circular tube is discussed at high liquid Reynolds number. On the other hand, theoretical equations for k-ε turbulence model in a circular tube of a 3 mm in diameter and a 526 mm long are numerically solved for heating of water on heated section of a 3 mm in diameter and a 67 mm long with various thicknesses of conductive sub-layer by using PHOENICS code under the same conditions as the experimental ones previously obtained considering the temperature dependence of thermo-physical properties concerned. The Platinum (Pt) test tube of inner diameter (d=3 mm) and heated length (L=66.5 mm) was used in this experiment. The thicknesses of conductive sub-layer from non-boiling regime to CHF are clarified. The thicknesses of conductive sub-layer at the CHF point are evaluated for various flow velocities. The experimental values of the CHF are also compared with the corresponding

  17. Effect of inter-critically reheating temperature on microstructure and properties of simulated inter-critically reheated coarse grained heat affected zone in X70 steel

    International Nuclear Information System (INIS)

    Zhu, Zhixiong; Kuzmikova, Lenka; Li, Huijun; Barbaro, Frank

    2014-01-01

    This study investigated the influence of the inter-critical reheating temperature on the microstructure and mechanical properties of a coarse grained heat affected zone (CGHAZ) in an API 5L grade X70 pipeline steel seam weld. A Gleeble 3500 thermo-mechanical simulator was employed to duplicate particular weld thermal cycles in order to accurately assess specific regions of the weld HAZ. Detailed microstructural analysis, including investigation of the martensite–austenite (M–A) constituent, was performed using optical microscope (OM), scanning electron microscope (SEM) and selective etching techniques. It is shown that the fracture toughness of the CGHAZ is significantly reduced following exposure to a subsequent inter-critical thermal cycle. Fracture toughness gradually improves as the inter-critical temperature is increased, but does not return to the value of the original CGHAZ due to the presence of isolated large M–A particles and coarse microstructure. Significance of M–A particles to the HAZ fracture toughness is first related to the location of particles along prior austenite grain boundaries, followed by the size of individual M–A particles

  18. Evaluation of subcooled critical heat flux correlations using the PU-BTPFL CHF database for vertical upflow of water in a uniformly heated round tube

    International Nuclear Information System (INIS)

    Hall, D.D.; Mudawar, I.

    1997-01-01

    A simple methodology for assessing the predictive ability of critical heat flux (CHF) correlations applicable to subcooled flow boiling in a uniformly heated vertical tube is developed. Popular correlations published in handbooks and review articles as well as the most recent correlations are analyzed with the PU-BTPFL CHF database, which contains 29,718 CHF data points. This database is the largest collection of CHF data (vertical upflow of water in a uniformly heated round tube) ever cited in the world literature. The parametric ranges of the CHF database are diameters from 0.3 to 45 mm, length-to-diameter ratios from 2 to 2484, mass velocities from 0.01 x 10 3 to 138 x 10 3 kg/m 2 ·s, pressures from 1 to 223 bars, inlet subcoolings from 0 to 347 C, inlet qualities from -2.63 to 0.00, outlet subcoolings from 0 to 305 C, outlet qualities from -2.13 to 1.00, and CHFs from 0.05 x 10 6 to 276 x 10 6 W/m 2 . The database contains 4,357 data points having a subcooled outlet condition at CHF. A correlation published elsewhere is the most accurate in both low- and high-mass velocity regions, having been developed with a larger database than most correlations. In general, CHF correlations developed from data covering a limited range of flow conditions cannot be extended to other flow conditions without much uncertainty

  19. A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility

    International Nuclear Information System (INIS)

    Lee, M.

    2000-01-01

    The critical heat flux (CHF) approach using CHF look-up tables has become a widely accepted CHF prediction technique. In these approaches, the CHF tables are developed based mostly on the data bank for flow in circular tubes. A set of correction factors was proposed by Groeneveld et al. [Groeneveld, D.C., Cheng, S.C., Doan, T. (1986)] to extend the application of the CHF table to other flow situations including flow in rod bundles. The proposed correction factors are based on a limited amount of data not specified in the original paper. The CHF approach of Groeneveld and co-workers is extensively used in the thermal hydraulic analysis of nuclear reactors. In 1996, Groeneveld et al. proposed a new CHF table to predict CHF in circular tubes [Groeneveld, D.C., et al., 1996. The 1995 look-up table for Critical Heat Flux. Nucl. Eng. Des. 163(1), 23]. In the present study, a set of correction factors is developed to extend the applicability of the new CHF table to flow in rod bundles of square array. The correction factors are developed by minimizing the statistical parameters of the ratio of the measured and predicted bundle CHF data from the Heat Transfer Research Facility. The proposed correction factors include: the hydraulic diameter factor (K hy ), the bundle factor (K bf ), the heated length factor (K hl ), the grid spacer factor (K sp ), the axial flux distribution factors (K nu ), the cold wall factor (K cw ) and the radial power distribution factor (K rp ). The value of constants in these correction factors is different when the heat balance method (HBM) and direct substitution method (DSM) are adopted to predict the experimental results of HTRF. With the 1995 Groeneveld CHF Table and the proposed correction factors, the average relative error is 0.1 and 0.0% for HBM and DSM, respectively, and the root mean square (RMS) error is 31.7% in DSM and 17.7% in HBM for 9852 square array data points of HTRF. (orig.)

  20. A dry-spot model of critical heat flux and transition boiling in pool and subcooled forced convection boiling

    International Nuclear Information System (INIS)

    Ha, Sang Jun

    1998-02-01

    A new dry-spot model for critical heat flux (CHF) is proposed. The new concept for dry area formation based on Poisson distribution of active nucleation sites and the critical active site number is introduced. The model is based on the boiling phenomena observed in nucleate boiling such as Poisson distribution of active nucleation sites and formation of dry spots on the heating surface. It is hypothesized that when the number of bubbles surrounding one bubble exceeds a critical number, the surrounding bubbles restrict the feed of liquid to the microlayer under the bubble. Then a dry spot of vapor will form on the heated surface. As the surface temperature is raised, more and more bubbles will have a population of surrounding active sites over the critical number. Consequently, the number of the spots will increase and the size of dry areas will increase due to merger of several dry spots. If this trend continues, the number of effective sites for heat transport through the wall will diminish, and CHF and transition boiling occur. The model is applicable to pool and subcooled forced convection boiling conditions, based on the common mechanism that CHF and transition boiling are caused by the accumulation and coalescences of dry spots. It is shown that CHF and heat flux in transition boiling can be determined without any empirical parameter based on information on the boiling parameters such as active site density and bubble diameter, etc., in nucleate boiling. It is also shown that the present model well represents actual phenomena on CHF and transition boiling and explains the mechanism on how parameters such as flow modes (pool or flow) and surface wettability influence CHF and transition boiling. Validation of the present model for CHF and transition boiling is achieved without any tuning parameter always present in earlier models. It is achieved by comparing the predictions of CHF and heat flux in transition boiling using measured boiling parameters in nucleate

  1. Critical Heat Flux Phenomena at HighPressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States); Wu, Qiao [Oregon State Univ., Corvallis, OR (United States)

    2015-04-30

    This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m2s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

  2. Integrating artificial neural networks and empirical correlations for the prediction of water-subcooled critical heat flux

    International Nuclear Information System (INIS)

    Mazzola, A.

    1997-01-01

    The critical heat flux (CHF) is an important parameter for the design of nuclear reactors, heat exchangers and other boiling heat transfer units. Recently, the CHF in water-subcooled flow boiling at high mass flux and subcooling has been thoroughly studied in relation to the cooling of high-heat-flux components in thermonuclear fusion reactors. Due to the specific thermal-hydraulic situation, very few of the existing correlations, originally developed for operating conditions typical of pressurized water reactors, are able to provide consistent predictions of water-subcooled-flow-boiling CHF at high heat fluxes. Therefore, alternative predicting techniques are being investigated. Among these, artificial neural networks (ANN) have the advantage of not requiring a formal model structure to fit the experimental data; however, their main drawbacks are the loss of model transparency ('black-box' character) and the lack of any indicator for evaluating accuracy and reliability of the ANN answer when 'never-seen' patterns are presented. In the present work, the prediction of CHF is approached by a hybrid system which couples a heuristic correlation with a neural network. The ANN role is to predict a datum-dependent parameter required by the analytical correlation; ; this parameter was instead set to a constant value obtained by usual best-fitting techniques when a pure analytical approach was adopted. Upper and lower boundaries can be possibly assigned to the parameter value, thus avoiding the case of unexpected and unpredictable answer failure. The present approach maintains the advantage of the analytical model analysis, and it partially overcomes the 'black-box' character typical of the straight application of ANNs because the neural network role is limited to the correlation tuning. The proposed methodology allows us to achieve accurate results and it is likely to be suitable for thermal-hydraulic and heat transfer data processing. (author)

  3. Transfer laws between water and freon 113 for average volumetric steam quality, pressure drop, and critical heat flux

    International Nuclear Information System (INIS)

    Nabizadeh, H.

    1977-01-01

    Simulation of the thermohydraulic processes of the steady-state reactor operation with boiling water and typical fuel element geometries leads to considerable increase of the heat rates to be tranferred and thus to an increase of the experimental cost which can hardly be justified. By proper choice of a model fluid with low heat of evaporation the system parameters like pressure, temperature, and heat rate, while retaining the original geometry, may be reduced to a fraction of those of the original fluid water. This permits not only a decrease in experimental cost but also a modification of the existing calculation data under more favorable experimental conditions. Starting from these considerations the cooling medium R113 was used as model fluid in carrying out the experiments. The necessary knowledge of the thermodynamical laws of simularity, however, have to be determined first of all in simple geometries and the scaling factors are then derived from them. In this connection the following experimental studies have been carried out with R113: a) average volumetric steam quality; b) two-phase pressure drop; c) critical heat flux. (orig.) [de

  4. A theoretical critical heat flux model for low-pressure, low-mass-flux, and low-steam quality conditions

    International Nuclear Information System (INIS)

    Weihsiao Ho; Kuanchywan Tu; Baushei Pei; Chinjang Chang

    1993-01-01

    The critical heat flux (CHF) is the maximum heat flux just before a boiling crisis; its importance as a measurement of nuclear reactor power capability design as well as in the safety of reactors has been recognized. With emphasis on CHF behavior under subcooled and low-quality (i.e., 2 ·s), an improved model that uses the sublayer dry out theory has been developed. Based on experimental observations of CHF, the model assumes that CHF under such conditions is of the departure from nucleate boiling type. Based on the postulation that CHF is triggered by Helmholtz instability in the sublayer steam-liquid system, the model was developed by a simple energy balance of liquid sublayer evaporation as the vapor blanket tends to disturb the balance between the buoyancy force and the drag force exerted upon it. The model is compared with the well-known Biasi et al. correlation as well as the Atomic Energy of Canada Limited lookup table against 102 uniformly heated round tube CHF data and 34 nonuniformly heated round tube CHF data. The comparison shows that the model provides better accuracy and a reasonable agreement between the predicted values and experimental CHF data

  5. Anisotropy and buoyancy in nuclear turbulent heat transfer - critical assessment and needs for modelling

    International Nuclear Information System (INIS)

    Groetzbach, G.

    2007-12-01

    Computational Fluid Dynamics (CFD) programs have a wide application field in reactor technique, like to diverse flow types which have to be considered in Accelerator Driven nuclear reactor Systems (ADS). This requires turbulence models for the momentum and heat transfer with very different capabilities. The physical demands on the models are elaborated for selected transport mechanisms, the status quo of the modelling is discussed, and it is investigated which capabilities are offered by the market dominating commercial CFD codes. One topic of the discussion is on the already earlier achieved knowledge on the distinct anisotropy of the turbulent momentum and heat transport near walls. It is shown that this is relevant in channel flows with inhomogeneous wall conditions. The related consequences for the turbulence modelling are discussed. The second topic is the turbulent heat transport in buoyancy influenced flows. The only turbulence model for heat transfer which is available in the large commercial CFD-codes is based on the Reynolds analogy. This means, it is required to prescribe suitable turbulent Prandtl number distributions. There exist many correlations for channel flows, but they are seldom used in practical applications. Here, a correlation is deduced for the local turbulent Prandtl number which accounts for many parameters, like wall distance, molecular Prandtl number of the fluid, wall roughness and local shear stress, thermal wall condition, etc. so that it can be applied to most ADS typical heat transporting channel flows. The spatial dependence is discussed. It is shown that it is essential for reliable temperature calculations to get accurate turbulent Prandtl numbers especially near walls. If thermal wall functions are applied, then the correlation for the turbulent Prandtl number has to be consistent with the wall functions to avoid unphysical discretisation dependences. In using Direct Numerical Simulation (DNS) data for horizontal fluid layers it

  6. Effects of Liquid Metal Fin on Critical Heat Flux under IVR-ERVC Condition

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    The molten fuel is relocated to bottom of reactor vessel after core is damaged and not cooled continuously. In-vessel retention through external reactor vessel cooling (IVR-ERVC) is presented to terminate the progression of accidents by removing the decay heat. IVR-ERVC is suitable for small size reactors like AR-600, AP-1000. There is uncertainty for high power reactor like APR-1400 and CAP-1400. This uncertainty originates from the thermal margin between the CHF value and real heat flux on the reactor vessel under severe accidents. The main mechanism of heat removal on IVR-ERVC strategy is boiling on the outer wall of reactor vessel. The boiling heat transfer is limited due to the CHF phenomenon. There should be an enough margin for preventing the CHF in boiling heat transfer systems. The CHF tests for IVR-ERVC system were conducted to confirm or increase the thermal margin. The design of thermal insulator was changed to vent the vapor smoothly. Forming the coating layer on the vessel surface was proposed to enhance the CHF margin. The liquid metal was designed to flood the space around the reactor vessel. The liquid metal has high boiling point and superb thermal conductivity in comparison with the coolant. In this work, experimental tests were conducted to validate the CFD results about the IVR-ERVC system with liquid metal. The behavior of vapor was observed to predict the tendency of CHF increase with small-scaled facility to simulate the IVR-ERVC system.

  7. Multi-band description of the specific heat and thermodynamic critical field in MgB2 superconductor

    Science.gov (United States)

    Szcześniak, R.; Jarosik, M. W.; Tarasewicz, P.; Durajski, A. P.

    2018-05-01

    The thermodynamic properties of MgB2 superconductor can be explained using the multi-band models. In the present paper we have examined the experimental data available in literature and we have found out that it is possible to reproduce the measured values of the superconducting energy gaps, the thermodynamic critical magnetic field and specific heat jump within the framework of two-band Eliashberg formalism and appropriate defined free energy difference between superconducting and normal state. Moreover, we found that the obtained results differ significantly from the predictions of the conventional Bardeen-Cooper-Schrieffer theory.

  8. CSER 94-09: Implications of the heat anomaly in Tank 106-C to criticality safety

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, C.A.

    1994-10-01

    Water is periodically added to Tank C-106 to cool its waste. In March 1994 addition of water was temporarily discontinued to determine if the tank could be adequately cooled at a lower water level. Following an addition of water, a temperature fluctuation was observed on one of the thermocouple trees. This Criticality Safety Evaluation Report (CSER) explains why the anomalous temperature measurements could not have been caused by nuclear criticality. Waste in Tank C-106 was discharged from processing facilities under controls designed to ensure that the contents of the tank would remain well subcritical under all credible conditions. The observed temperature profile does not fit the profile expected from a criticality event. In addition, there has been no indication of any significant increase in the rate of water evaporation.

  9. Research on cooling of ultra high critical heat flux with external flow boiling of water. Challenge to achieve ultra high critical heat flux and improvement in estimation of critical heat flux. JAERI's nuclear research promotion program, H11-004 (Contract research)

    International Nuclear Information System (INIS)

    Monde, Masanori; Mitsutake, Yuichi; Ishida, Kenji; Hino, Ryutaro

    2003-03-01

    An ultra high critical heat flux (CHF) has been challenged with a highly subcooled water jet impinging on a small rectangular heated surface. Major objective of the study is to achieve an ultra high heat flux cooling as large as 100 MW/m 2 and to establish an accurate estimation method of the CHF. The experiments were carried out over the experimental range; a fixed jet diameter of 2 mm, jet velocity of 5 - 35 m/s, degree of subcooling of 80 - 170 K and system pressure of 0.1 - 1.0 MPa. The rectangular heated surface with a thin nickel foil of 0.03 - 0.3 mm in thickness, 5 and 10 mm in length, and 4 mm in width and heated by a direct current. Effects of thickness of heater wall, jet velocity and subcooling on the CHF were experimentally elucidated. The experimental results show that the CHF decreases about 50% as the heater thickness, namely heat capacity of heater decreases. Characteristics of the CHF with heater length of 10 mm are correlated within ±20% by the generalized correlation of subcooled CHF proposed by the authors. However, the CHF with the shorter heater length of 5 mm shows large deviation of -40% especially at lower subcooling and higher velocity. The maximum CHF of 212 MW/m 2 was achieved at the subcooling of 151 K, the jet velocity of 35 m/s and system pressure of 0.5 MPa. The maximum CHF under atmospheric pressure approaches to 48% of the ultimate maximum heat flux given by the assumptions that vapor molecules leave a liquid-vapor interface at the average speed of a Boltzman-Maxwellian gas and any molecules returning to the interface are not permitted. The ratio of the CHF and ultimate maximum heat flux was considerably enhanced from the existing record of 30%. This study can give the feasibility of ultra high heat flux removal facing in a development of components such as a diverter of a fusion reactor. (author)

  10. Compressibility and specific heats of heavier condensed rare gases near the liquid-vapour critical point

    International Nuclear Information System (INIS)

    March, N.H.

    2003-08-01

    Sarkisov (J. Chem. Phys. 119, 373, 2003) has recently discussed the structural behaviour of a simple fluid near the liquid-vapour critical point. His work, already compared with computer simulation studies, is here brought into direct contact for the heavier condensed rare gases Ar, Kr and Xe with (a) experiment and (b) earlier theoretical investigations. Directions for future studies then emerge. (author)

  11. Effects of Al{sub 2}O{sub 3} nanoparticles deposition on critical heat flux of R-123 in flow boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seok Bin; Bang, In Cheol [School of Mechanical and Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST), Ulsan (Korea, Republic of)

    2015-06-15

    In this study, R-123 flow boiling experiments were carried out to investigate the effects of nanoparticle deposition on heater surfaces on flow critical heat flux (CHF) and boiling heat transfer. It is known that CHF enhancement by nanoparticles results from porous structures that are very similar to layers of Chalk River unidentified deposit formed on nuclear fuel rod surfaces during the reactor operation period. Although previous studies have investigated the surface effects through surface modifications, most studies are limited to pool boiling conditions, and therefore, the effects of porous surfaces on flow boiling heat transfer are still unclear. In addition, there have been only few reports on suppression of wetting for decoupled approaches of reasoning. In this study, bare and Al{sub 2}O{sub 3} nanoparticle-coated surfaces were prepared for the study experiments. The CHF of each surface was measured with different mass fluxes of 1,600 kg/m{sup 2}s, 1,800 kg/m{sup 2}s, 2,100 kg/m{sup 2}s, 2,400 kg/m{sup 2}s, and 2,600 kg/m{sup 2}s. The nanoparticle-coated tube showed CHF enhancement up to 17% at a mass flux of 2,400 kg/m{sup 2}s compared with the bare tube. The factors for CHF enhancement are related to the enhanced rewetting process derived from capillary action through porous structures built-up by nanoparticles while suppressing relative wettability effects between two sample surfaces as a highly wettable R-123 refrigerant was used as a working fluid.

  12. Disappearance of criticality in branched-chain thermal explosion with heat loss

    International Nuclear Information System (INIS)

    Okoya, Samuel S.

    2003-09-01

    In the framework of the currently developed branched-chain thermal explosion theory, the equation governing leakage through a hole of a reaction vessel is given. The critical ignition, extinction and transition temperature excess, activation energy parameter and modified Semenov's number are estimated employing this equation. We calculated numerically and obtained analytically these non-dimensional parameters with and without initiation respectively. The similar solution for Semenov model appear as a limiting case of our solution. We also obtained the ignition times. (author)

  13. CFD simulation on critical heat flux of flow boiling in IVR-ERVC of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xiang, E-mail: zhangxiang3@snptc.com.cn [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Hu, Teng [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Chen, Deqi, E-mail: chendeqi@cqu.edu.cn [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China); Zhong, Yunke; Gao, Hong [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China)

    2016-08-01

    Highlights: • CFD simulation on CHF of boiling two-phase flow in ERVC is proposed. • CFD simulation result of CHF agrees well with that of experimental result. • The characteristics of boiling two-phase flow and boiling crisis are analyzed. - Abstract: The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as the local CHF does not exceed the local heat flux, the lower head of the pressure vessel can be cooled sufficiently to prevent from failure. In this paper, a CFD simulation is carried out to investigate the CHF of ERVC. This simulation is performed by a CFD code fluent couple with a boiling model by UDF (User-Defined Function). The experimental CHF of ERVC obtained by State Nuclear Power Technology Research and Development Center (SNPTRD) is used to validate this CFD simulation, and it is found that the simulation result agrees well with the experimental result. Based on the CFD simulation, detailed analysis focusing on the pressure distribution, velocity distribution, void fraction distribution, heating wall temperature distribution are proposed in this paper.

  14. Assessment of fluid-to-fluid modelling of critical heat flux in horizontal 37-element bundle flows

    International Nuclear Information System (INIS)

    Yang, S.K.

    2006-01-01

    Fluid-to-fluid modelling laws of critical heat flux (CHF) available in the literature were reviewed. The applicability of the fluid-to-fluid modelling laws was assessed using available data ranging from low to high mass fluxes in horizontal 37-element bundles simulating a CANDU fuel string. Correlations consisting of dimensionless similarity groups were derived using modelling fluid data (Freon-12) to predict water CHF data in horizontal 37-element bundles with uniform and non-uniform axial-heat flux distribution (AFD). The results showed that at mass fluxes higher than ∼4,000 kg/m 2 s (water equivalent value), the vertical fluid-to-fluid modelling laws of Ahmad (1973) and Katto (1979) predict water CHF in horizontal 37-element bundles with non-uniform AFD with average errors of 1.4% and 3.0% and RMS errors of 5.9% and 6.1%, respectively. The Francois and Berthoud (2003) fluid-to-fluid modelling law predicts CHF in non-uniformly heated 37-element bundles in the horizontal orientation with an average error of 0.6% and an RMS error of 10.4% over the available range of 2,000 to 6,200 kg/m 2 s. (author)

  15. Critical evaluation of analytical models for stochastic heating in dual-frequency capacitive discharges

    International Nuclear Information System (INIS)

    Sharma, S; Turner, M M

    2013-01-01

    Dual-frequency capacitive discharges are widespread in the semiconductor industry and are used, for example, in etching of semiconductor materials to manufacture microchips. In low-pressure dual radio-frequency capacitive discharges, stochastic heating is an important phenomenon. Recent theoretical work on this problem using several different approaches has produced results that are broadly in agreement insofar as scaling with the discharge parameters is concerned, but there remains some disagreement in detail concerning the absolute size of the effect for the case of dual-frequency capacitive discharges. In this work, we investigate the dependence of stochastic heating on various discharge parameters with the help of particle-in-cell (PIC) simulation. The dual-frequency analytical models are in fair agreement with PIC results for values of the low-frequency current density amplitude J lf (or dimensionless control parameter H lf ∼ 5) typical of many modern experiments. However, for higher values of J lf (or higher H lf ), new physical phenomena (like field reversal, reflection of ions, etc) appear and the simulation results deviate from existing dual-frequency analytical models. On the other hand, for lower J lf (or lower H lf ) again the simulation results deviate from analytical models. So this research work produces a relatively extensive set of simulation data that may be used to validate theories over a wide range of parameters. (paper)

  16. Premature and stable critical heat flux for downward flow in a narrow rectangular channel

    International Nuclear Information System (INIS)

    Lee, Juhyung; Chang, Soon Heung; Jeong, Yong Hoon; Jo, Daeseong

    2014-01-01

    It has been recommended that RRs and MTRs be designed to have sufficient margins for CHF and the onset of FI as well, since unstable flow could leads to premature CHF under very low wall heat flux in comparison to stable CHF. Even the fact and previous studies, however, the understanding of relationship among FI, premature CHF and stable CHF is not sufficient to date. In this regards, subcooled flow boiling in a vertical rectangular channel was experimentally investigated to enhance the understanding of the CHF and the effect of the two-phase flow instability on it under low pressure conditions, especially for downward flow which was adopted for Jordan Research and Training Reactor (JRTR) and Kijang research reactor (KJRR) to achieve easier fuel and irradiation rig loading. In this study, CHF for downward flow of water under low pressure in narrow rectangular channel was experimentally investigated. For conditions such as downward flow, narrow rectangular channel and low pressure, it has been deduced from literature that flow instability could largely influence on triggering CHF at lower heat flux, i. e. premature CHF. Total 54 CHF data, which includes premature and stable data was obtained for various fluid conditions and system configurations including inlet stiffness. The upper and lower boundaries of CHF were newly proposed based on the experiment

  17. Calculation of critical heat transfer in horizontal evaporator pipes in cooling systems of high-rise buildings

    Science.gov (United States)

    Aksenov, Andrey; Malysheva, Anna

    2018-03-01

    An exact calculation of the heat exchange of evaporative surfaces is possible only if the physical processes of hydrodynamics of two-phase flows are considered in detail. Especially this task is relevant for the design of refrigeration supply systems for high-rise buildings, where powerful refrigeration equipment and branched networks of refrigerants are used. On the basis of experimental studies and developed mathematical model of asymmetric dispersed-annular flow of steam-water flow in horizontal steam-generating pipes, a calculation formula has been obtained for determining the boundaries of the zone of improved heat transfer and the critical value of the heat flux density. A new theoretical approach to the solution of the problem of the flow structure of a two-phase flow is proposed. The applied method of dissipative characteristics of a two-phase flow in pipes and the principle of a minimum rate of entropy increase in stabilized flows made it possible to obtain formulas that directly reflect the influence of the viscous characteristics of the gas and liquid media on their distribution in the flow. The study showed a significant effect of gravitational forces on the nature of the phase distribution in the cross section of the evaporative tubes. At a mass velocity of a two-phase flow less than 700 kg / m2s, the volume content of the liquid phase near the upper outer generating lines of the tube is almost an order of magnitude lower than the lower one. The calculation of the heat transfer crisis in horizontal evaporative tubes is obtained. The calculated dependence is in good agreement with the experimental data of the author and a number of foreign researchers. The formula generalizes the experimental data for pipes with the diameter of 6-40 mm in the pressure of 2-7 MPa.

  18. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.

  19. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon

    2006-02-01

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation

  20. Humidity measurements in passive heat and moisture exchangers applications: a critical issue.

    Science.gov (United States)

    Dubini, G; Fumero, R

    2000-01-01

    A reliable, quantitative assessment of humidification performances of passive heat and moisture exchangers in mechanically-ventilated patients is still to be achieved, although relevant efforts have been made to date. One of the major problems to tackle consists in the difficulty of humidity measurements, both in vivo (during either anaesthesia or intensive care unit treatments) and in vitro set-ups. In this paper a review of the basic operation principles of humidity sensors as well as an analysis of their usage within in vivo and in vitro tests are presented. Particular attention is devoted to the limitations arising from the specific measurement set-up, as they may affect the results notably.

  1. A critical examination of the validity of simplified models for radiant heat transfer analysis.

    Science.gov (United States)

    Toor, J. S.; Viskanta, R.

    1972-01-01

    Examination of the directional effects of the simplified models by comparing the experimental data with the predictions based on simple and more detailed models for the radiation characteristics of surfaces. Analytical results indicate that the constant property diffuse and specular models do not yield the upper and lower bounds on local radiant heat flux. In general, the constant property specular analysis yields higher values of irradiation than the constant property diffuse analysis. A diffuse surface in the enclosure appears to destroy the effect of specularity of the other surfaces. Semigray and gray analyses predict the irradiation reasonably well provided that the directional properties and the specularity of the surfaces are taken into account. The uniform and nonuniform radiosity diffuse models are in satisfactory agreement with each other.

  2. Review of Critical Heat Flux Correlations for Upward Flow in a Vertical Thin Rectangular Channel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Gil Sik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    From the view point of safety, this type of fuel has higher resistance to earthquake and external impact. The cross section of coolant flow channel in the reactor core composed with the plate fuel is a thin rectangular shape. Thermal-hydraulic characteristics of this thin rectangular channel are different with those of general circular rod fuel bundle flow channel. Accordingly it could be thought that the CHF correlation in a thin rectangular channel is different with that in a circular channel, for which a large number of researches on CHF prediction have been carried out. The objective of this paper is to review previous researches on CHF in a thin rectangular channel, summarize the important conclusion and propose the new simple CHF correlation, which is based on the data set under high pressure and high flow rate condition. The researches on CHF in rectangular channel have been partially carried out according to the pressure, heated surface number, heated surface wettability effect, flow driving force and flow direction conditions. From the literature researches on CHF for upward flow in a vertical thin rectangular channel, some CHF prediction methods were reviewed and compared. There is no universal correlation which can predict CHF at all conditions, but generally, Katto empirical correlation is known to be useful at high pressure and high flow rate. The new simple correlation was developed from the restricted data set, the CHF prediction capacity of which is better than that of Katto. Even though the prediction consistency of the new simple correlation is lower, MAE and RMS error decreased quite. For the more development of the new simple CHF correlation, the more advanced regression analysis method and theoretical analysis should be studied in future.

  3. Review of Critical Heat Flux Correlations for Upward Flow in a Vertical Thin Rectangular Channel

    International Nuclear Information System (INIS)

    Choi, Gil Sik; Chang, Soon Heung

    2014-01-01

    From the view point of safety, this type of fuel has higher resistance to earthquake and external impact. The cross section of coolant flow channel in the reactor core composed with the plate fuel is a thin rectangular shape. Thermal-hydraulic characteristics of this thin rectangular channel are different with those of general circular rod fuel bundle flow channel. Accordingly it could be thought that the CHF correlation in a thin rectangular channel is different with that in a circular channel, for which a large number of researches on CHF prediction have been carried out. The objective of this paper is to review previous researches on CHF in a thin rectangular channel, summarize the important conclusion and propose the new simple CHF correlation, which is based on the data set under high pressure and high flow rate condition. The researches on CHF in rectangular channel have been partially carried out according to the pressure, heated surface number, heated surface wettability effect, flow driving force and flow direction conditions. From the literature researches on CHF for upward flow in a vertical thin rectangular channel, some CHF prediction methods were reviewed and compared. There is no universal correlation which can predict CHF at all conditions, but generally, Katto empirical correlation is known to be useful at high pressure and high flow rate. The new simple correlation was developed from the restricted data set, the CHF prediction capacity of which is better than that of Katto. Even though the prediction consistency of the new simple correlation is lower, MAE and RMS error decreased quite. For the more development of the new simple CHF correlation, the more advanced regression analysis method and theoretical analysis should be studied in future

  4. Flow visualization study of post-critical heat flux in inverted flow

    International Nuclear Information System (INIS)

    Babelli, I.; Revankar, S.T.; Ishii, M.

    1994-01-01

    A visual study of film boiling was carried out to determine the flow regime transition in the post-CHF region for a transient bottom reflooding of a hot transparent test section. The effect of test liquid subcooling and inlet velocity on flow transition as well as on the quench front propagation was investigated. The respective ranges for liquid velocity and subcooling were 1.8-26.8 cm/s, and 20-45 C, respectively. The test liquid was Freon 113 which was introduced into the bottom of the quartz test section whose walls were maintained well above the film boiling temperature of the test liquid, via a transparent heat transfer fluid. The flow regimes observed down stream of the upward moving quench front were the rough wavy, the agitated, and the dispersed droplet/ligaments in agreement with a steady state, two-phase core injection study carried on recently by one of the authors. A correlation for the flow regime transition between the inverted annular and the dispersed droplet/ligament flow patterns was developed. The correlation showed a marked dependence on the void fraction at the CHF location and hence on the flow regime encountered in the pre-CHF region. (orig.)

  5. The implementation of microstructural and heat treatment models to development of forming technology of critical aluminum-alloy parts

    Science.gov (United States)

    Biba, Nikolay; Alimov, Artem; Shitikov, Andrey; Stebunov, Sergei

    2018-05-01

    The demand for high performance and energy efficient transportation systems have boosted interest in lightweight design solutions. To achieve maximum weight reductions, it is not enough just to replace steel parts by their aluminium analogues, but it is necessary to change the entire concept of vehicle design. In this case we must develop methods for manufacturing a variety of critical parts with unusual and difficult to produce shapes. The mechanical properties of the material in these parts must also be optimised and tightly controlled to provide the best distribution within the part volume. The only way to achieve these goals is to implement technology development methods based on simulation of the entire manufacturing chain from preparing a billet through the forming operations and heat treatment of the product. The paper presents an approach to such technology development. The simulation of the technological chain starts with extruding a round billet. Depending on the extrusion process parameters, the billet can have different levels of material workout and variation of grain size throughout the volume. After extrusion, the billet gets formed into the required shape in a forging process. The main requirements at this stage are to get the near net shape of the product without defects and to provide proper configuration of grain flow that strengthens the product in the most critical direction. Then the product undergoes solution treatment, quenching and ageing. The simulation of all these stages are performed by QForm FEM code that provides thermo-mechanical coupled deformation of the material during extrusion and forging. To provide microstructure and heat treatment simulation, special subroutines has been developed by the authors. The proposed approach is illustrated by an industrial case study.

  6. A review of critical heat flux prediction technique and its application in CANDU reactor

    International Nuclear Information System (INIS)

    Park, Jee Won; Roh, Gyu Hong

    1997-09-01

    The CHF prediction method being used for CANDU reactor have been critically reviewed. The AECL's CHF prediction totally depends on the look-up table which has been developed from many CHF databank. These databanks include not only the water-cooled bundle-CHF data but also the freon-cooled bundle-CHF data. The CHF look-up tables have been developed by smoothing and interpolating (with some extrapolations) the experimental data to construct a practically useful CHF table. Therefore, the table look-up method has advantages of accuracy, consistency in a wide range of thermal-hydraulic parameters. It seems, however, that since the existing look-up table is constructed by many steps of modification of the original experimental data (e.g., the look-up table is constructed not only using the horizontal flow data but also the vertical flow data), one should be very careful when one try to generate a look-up table for other fuel geometries. In other words, a reliable look-up table can be constructed by performing experiments for new fuel geometry. Finally, it should be noted that the modifications to the original experimental data has simple form with many modification parameters for taking into account of different geometrical effects. This report presents the backbone and the validity of AECL CHF look-up table. (author). 22 refs., 2 tabs., 2 figs

  7. Critical heat flux of water in vertical round tubes at low-pressure and low-flow conditions

    International Nuclear Information System (INIS)

    Park, Jae-Wook; Kim, Hong-Chae; Beak, Won-Pil; Chang, Soon Heung

    1997-01-01

    A series of critical heat flux (CHF) tests have been performed to provide a reliable set of CHF data for water flow in vertical round tubes at low pressure and low flow (LPLF) conditions. The range of experimental conditions is as follows: diameter 8, 10 mm; heated length 0.5, 1 m; pressure 2-9 bar, mass flux 50-200 kg/m 2 s; inlet subcooling 350, 450 kJ/kg. The observed parametric trends are generally consistent with the previous understanding except for the effects of system pressure and tube diameter. The pressure effect is small but very complicated; existing CHF correlations do not represent this parametric trend properly. CHF increases with the increase in diameter at fixed exit conditions, contrary to the general understanding. The artificial neural networks are applied to the round tube CHF data base at LPLF (P = 110-1100 kPa, G = 0-500 kg/m 2 s) conditions. The trained backpropagation networks (BPNs) predict CHF better than any other CHF correlations. Parametric trends of CHF based on the BPN for fixed inlet conditions generally agree well with our experimental results. (author)

  8. Enhanced pool boiling critical heat flux induced by capillary wicking effect of a Cr-sputtered superhydrophilic surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Son, Hong Hyun; Seo, Gwang Hyeok; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    In light of boiling heat transfer, the smooth surface potentially reduces active nucleation of bubbles and rewetting of dry spots near the critical heat flux (CHF). This kind of process is highly likely to deteriorate the CHF. Thus, it is essential to produce appropriate microstructures on the surface for the enhancement of the CHF. In this study, to investigate the microstructural effect of thin film-fabricated surfaces on the pool boiling CHF, we controlled the surface roughness in a narrow range of 0.1-0.25 μm and its morphologies, in the form of micro-scratches using PVD sputtering technique. Specifically for DC magnetron sputtering, pure chromium (Cr) was selected as a target material owing to its high oxidation resistance. In order to analyze the CHF trend with changes in roughness, we introduced existing capillary wicking-based models because superhydrophilic characteristics of microstructures are highly related to the capillary wicking behaviors in micro-flow channels. After Cr sputtering under given conditions, the Cr-sputtered surfaces showed superhydrophilic characteristics and its capability became more enhanced with an increase of surface roughness. Judging from spreading behavior of a liquid droplet, the presence of micro-wicking channels, coupled with Cr nanostructures, effectively enhanced the advancing rate of drop base diameter. The CHF exhibited an increasing trend with increasing surface roughness. However, the enhancement ratio agreed poorly with the predictions of the roughness factor-based models, all of which originated from a conventional static force balance.

  9. Two-phase flows and heat transfer within systems with ambient pressure above the thermodynamic critical pressure

    Science.gov (United States)

    Hendricks, R. C.; Braun, M. J.; Mullen, R. L.

    1986-01-01

    In systems where the design inlet and outlet pressure P sub amb are maintained above the thermodynamic critical pressure P sub c, it is often assumed that heat and mass transfer are governed by single-phase relations and that two-phase flows cannot occur. This simple rule of thumb is adequate in many low-power designs but is inadequate for high-performance turbomachines, boilers, and other systems where two-phase regions can exist even though P sub amb P sub c. Heat and mass transfer and rotordynamic-fluid-mechanic restoring forces depend on momentum differences, and those for a two-phase zone can differ significantly from those for a single-phase zone. By using a laminar, variable-property bearing code and a rotating boiler code, pressure and temperature surfaces were determined that illustrate nesting of a two-phase region within a supercritical pressure region. The method of corresponding states is applied to bearings with reasonable rapport.

  10. Prediction of the critical heat flux for saturated upward flow boiling water in vertical narrow rectangular channels

    International Nuclear Information System (INIS)

    Choi, Gil Sik; Chang, Soon Heung; Jeong, Yong Hoon

    2016-01-01

    A study, on the theoretical method to predict the critical heat flux (CHF) of saturated upward flow boiling water in vertical narrow rectangular channels, has been conducted. For the assessment of this CHF prediction method, 608 experimental data were selected from the previous researches, in which the heated sections were uniformly heated from both wide surfaces under the high pressure condition over 41 bar. For this purpose, representative previous liquid film dryout (LFD) models for circular channels were reviewed by using 6058 points from the KAIST CHF data bank. This shows that it is reasonable to define the initial condition of quality and entrainment fraction at onset of annular flow (OAF) as the transition to annular flow regime and the equilibrium value, respectively, and the prediction error of the LFD model is dependent on the accuracy of the constitutive equations of droplet deposition and entrainment. In the modified Levy model, the CHF data are predicted with standard deviation (SD) of 14.0% and root mean square error (RMSE) of 14.1%. Meanwhile, in the present LFD model, which is based on the constitutive equations developed by Okawa et al., the entire data are calculated with SD of 17.1% and RMSE of 17.3%. Because of its qualitative prediction trend and universal calculation convergence, the present model was finally selected as the best LFD model to predict the CHF for narrow rectangular channels. For the assessment of the present LFD model for narrow rectangular channels, effective 284 data were selected. By using the present LFD model, these data are predicted with RMSE of 22.9% with the dryout criterion of zero-liquid film flow, but RMSE of 18.7% with rivulet formation model. This shows that the prediction error of the present LFD model for narrow rectangular channels is similar with that for circular channels.

  11. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

    2017-06-15

    Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

  12. An improved Peltier effect-based instrument for critical temperature threshold measurement in cold- and heat-induced urticaria.

    Science.gov (United States)

    Magerl, M; Abajian, M; Krause, K; Altrichter, S; Siebenhaar, F; Church, M K

    2015-10-01

    Cold- and heat-induced urticaria are chronic physical urticaria conditions in which wheals, angioedema or both are evoked by skin exposure to cold and heat respectively. The diagnostic work up of both conditions should include skin provocation tests and accurate determination of critical temperature thresholds (CTT) for producing symptoms in order to be able to predict the potential risk that each individual patient faces and how this may be ameliorated by therapy. To develop and validate TempTest(®) 4, a simple and relatively inexpensive instrument for the accurate determination of CTT which may be used in clinical practice. TempTest(®) 4 has a single 2 mm wide 350 mm U-shaped Peltier element generating a temperature gradient from 4 °C to 44 °C along its length. Using a clear plastic guide placed over the skin after provocation, CTT values may be determined with an accuracy of ±1 °C. Here, TempTest(®) 4 was compared with its much more expensive predecessor, TempTest(®) 3, in inducing wheals in 30 cold urticaria patients. Both TempTest(®) 4 and TempTest(®) 3 induced wheals in all 30 patients between 8 ° and 28 °C. There was a highly significant (P < 0.0001) correlation between the instruments in the CTT values in individual patients. The TempTest(®) 4 is a simple, easy to use, licensed, commercially available and affordable instrument for the determination of CTTs in both cold- and heat-induced urticaria. © 2014 European Academy of Dermatology and Venereology.

  13. A 3-D CFD approach to the mechanistic prediction of forced convective critical heat flux at low quality

    International Nuclear Information System (INIS)

    Jean-Marie Le Corre; Cristina H Amon; Shi-Chune Yao

    2005-01-01

    Full text of publication follows: The prediction of the Critical Heat Flux (CHF) in a heat flux controlled boiling heat exchanger is important to assess the maximal thermal capability of the system. In the case of a nuclear reactor, CHF margin gain (using improved mixing vane grid design, for instance) can allow power up-rate and enhanced operating flexibility. In general, current nuclear core design procedures use quasi-1D approach to model the coolant thermal-hydraulic conditions within the fuel bundles coupled with fully empirical CHF prediction methods. In addition, several CHF mechanistic models have been developed in the past and coupled with 1D and quasi-1D thermal-hydraulic codes. These mechanistic models have demonstrated reasonable CHF prediction characteristics and, more remarkably, correct parametric trends over wide range of fluid conditions. However, since the phenomena leading to CHF are localized near the heater, models are needed to relate local quantities of interest to area-averaged quantities. As a consequence, large CHF prediction uncertainties may be introduced and 3D fluid characteristics (such as swirling flow) cannot be accounted properly. Therefore, a fully mechanistic approach to CHF prediction is, in general, not possible using the current approach. The development of CHF-enhanced fuel assembly designs requires the use of more advanced 3D coolant properties computations coupled with a CHF mechanistic modeling. In the present work, the commercial CFD code CFX-5 is used to compute 3D coolant conditions in a vertical heated tube with upward flow. Several CHF mechanistic models at low quality available in the literature are coupled with the CFD code by developing adequate models between local coolant properties and local parameters of interest to predict CHF. The prediction performances of these models are assessed using CHF databases available in the open literature and the 1995 CHF look-up table. Since CFD can reasonably capture 3D fluid

  14. Study on ex-vessel cooling of RPV (behavior of coalesced bubbles and trigger condition of critical heat flux on inclined plate)

    International Nuclear Information System (INIS)

    Ohtake, H.; Koizumi, Y.; Takano, K.I.

    2001-01-01

    The Ex-vessel cooling of Reactor-Pressure-Vessel in Light-Water-Reactor at the severe accident have been proposed for future nuclear reactors. The estimation of Critical-Heat-Flux on a downward-facing curvilinear surface, like a hemisphere, is important to the assessment of the cooling. In this study, the CHFs on inclined surfaces were examined experimentally focusing on orientation of the heating surface. In order to discuss detailed mechanism of the CHF, the behaviors of coalesced bubbles near the heating surface were investigated through visual observations. The critical heat flux obtained in the present experiments increased with the inclined angle over the present experimental range. The dependence of the inclined angle on the critical heat flux was q CHF,R-113 [q] = f (q 0.33 ) for the present experimental results. The effect of the surface orientation on the critical heat flux was roughly explained by using the simple analytical model based on the macro-layer model and Kelvin-Helmholtz instability. From visual observations for behavior of bubbles near the heating surface, whereas the coalesced bubble covered over the heating surface for the inclined angle of 0 degree, the coalesced bubble moved upward to avoid packing the bubble on the surface above 5 degree. As the inclined angle increased, the velocity of the coalesced bubble was high, the period covered the heater and the bubble length were small. The results suggested that the CHF was closely related to forming the coalesced bubble and the behavior of the bubble. (author)

  15. The effect of grid assembly mixing vanes on critical heat flux values and azimuthal location in fuel assemblies

    International Nuclear Information System (INIS)

    De Crecy, F.

    1994-01-01

    Critical heat flux (CHF) is one of the limiting phenomena for a PWR. It has been widely studied for years, but many facts are still not satisfactorily understood. This paper deals with the effect of the grid assembly mixing vanes on both the value of the CHF and the azimuthal location of the departure from nucleate boiling (DNB). A series of experimental studies was performed on electrically heated, 5x5 square pitched, vertical rod bundles. Two specific grid assembly designs were used: with and without mixing vanes. DNB was detected by eight thermocouples welded internally in each rod at the same level in order to determine the azimuthal location. The coolant was Freon-12 flowing upwards to simulate high pressure water (as defined by Stevens). Single-phase flow experiments were also conducted to measure the exit temperature field in order to obtain the mixing coefficients for subchannel analysis.The results show very clearly that the mixing vanes have a significant effect on both the DNB azimuthal location and the CHF value. - Without mixing vanes, DNB occurs mainly on the most central rod and preferentially at the azimuthal location facing the adjacent rod. - With mixing vanes, DNB can occur on any of the nine central rods and is distributed in an apparently random way around the rod. -The effect of the mixing vanes on CHF is dramatic and depends a great deal on the parameter range (pressure, local mass velocity and local quality). Generally speaking, CHF with mixing vanes is significantly higher than without mixing vanes, but this effect can be inverted in some cases.In order to understand this fact more clearly, it is necessary to perform detailed analysis of subchannel behavior. Indeed, the analyses show that the magnitude of this effect is closely related to the mixing coefficients used. These mixing coefficients, estimated from the single-phase flow experiments, are subject to large uncertainties in two-phase flow. ((orig.))

  16. Two-phase flow characteristic of inverted bubbly, slug, and annular flow in post-critical heat flux region

    International Nuclear Information System (INIS)

    Ishii, M.; Denten, J.P.

    1989-01-01

    Inverted annular flow can be visualized as a liquid jet-like core surrounded by a vapor annulus. While many analytical and experimental studies of heat transfer in this regime have been performed, there is very little understanding of the basic hydrodynamics of the post-critical heat flux (CHF) flow field. However, a recent experimental study was done that was able to successfully investigate the effects of various steady-state inlet flow parameters on the post-CHF hydrodynamics of the film boiling of a single phase liquid jet. This study was carried out by means of a visual photographic analysis of an idealized single phase core inverted annular flow initial geometry (single phase liquid jet core surrounded by a coaxial annulus of gas). In order to extend this study, a subsequent flow visualization of an idealized two-phase core inverted annular flow geometry (two-phase central jet core, surrounded by a coaxial annulus of gas) was carried out. The objective of this second experimental study was to investigate the effect of steady-state inlet, pre-CHF two-phase jet core parameters on the hydrodynamics of the post-CHF flow field. In actual film boiling situations, two-phase flows with net positive qualities at the CHF point are encountered. Thus, the focus of the present experimental study was on the inverted bubbly, slug, and annular flow fields in the post dryout film boiling region. Observed post dryout hydrodynamic behavior is reported. A correlation for the axial extent of the transition flow pattern between inverted annular and dispersed droplet flow (the agitated regime) is developed. It is shown to depend strongly on inlet jet core parameters and jet void fraction at the dryout point

  17. Development of classification and prediction methods of critical heat flux using fuzzy theory and artificial neural networks

    International Nuclear Information System (INIS)

    Moon, Sang Ki

    1995-02-01

    This thesis applies new information techniques, artificial neural networks, (ANNs) and fuzzy theory, to the investigation of the critical heat flux (CHF) phenomenon for water flow in vertical round tubes. The work performed are (a) classification and prediction of CHF based on fuzzy clustering and ANN, (b) prediction and parametric trends analysis of CHF using ANN with the introduction of dimensionless parameters, and (c) detection of CHF occurrence using fuzzy rule and spatiotemporal neural network (STN). Fuzzy clustering and ANN are used for classification and prediction of the CHF using primary system parameters. The fuzzy clustering classifies the experimental CHF data into a few data clusters (data groups) according to the data characteristics. After classification of the experimental data, the characteristics of the resulted clusters are discussed with emphasis on the distribution of the experimental conditions and physical mechanisms. The CHF data in each group are trained in an artificial neural network to predict the CHF. The artificial neural network adjusts the weight so as to minimize the prediction error within the corresponding cluster. Application of the proposed method to the KAIST CHF data bank shows good prediction capability of the CHF, better than other existing methods. Parametric trends of the CHF are analyzed by applying artificial neural networks to a CHF data base for water flow in uniformly heated vertical round tubes. The analyses are performed from three viewpoints, i.e., for fixed inlet conditions, for fixed exit conditions, and based on local conditions hypothesis. In order to remove the necessity of data classification, Katto and Groeneveld et al.'s dimensionless parameters are introduced in training the ANNs with the experimental CHF data. The trained ANNs predict the CHF better than any other conventional correlations, showing RMS error of 8.9%, 13.1%, and 19.3% for fixed inlet conditions, for fixed exit conditions, and for local

  18. Effects of a FeCrAl layer fabricated by sputtering process on pool boiling critical heat flux

    International Nuclear Information System (INIS)

    Seo, Gwang Hyeok; Son, Hong Hyun; Jeun, Gyoodong; Kim, Sung Joong

    2016-01-01

    The thermal safety margin of a FeCrAl-layered heater was investigated measuring pool boiling critical heat flux (CHF). Boiling experiments were conducted in a pool of deionized water at atmospheric pressure. For a comparison work, bare and FeCrAl-layered heater samples were prepared. The sputtering technique was employed to fabricate the FeCrAl layer. It was confirmed that the key sputtering parameters on the surface structure were substrate temperature and deposition time. As compared to the bare sample, surface wettability and roughness increased. Higher values of the surface roughness were observed at temperatures of 150degC and 600degC. The FeCrAl-layered heaters showed improved CHF up to ∼40%. The highest enhancement of 42% was observed for the heater sample fabricated at a substrate temperature of 150degC. With employing recent CHF models that incorporate the surface effects, it was evaluated that increased roughness at the micrometer scale mainly contributed to the CHF enhancement. Furthermore, visual observations showed at least 2 msec reduction in the rewetting times for the FeCrAl-layered heaters, and the improved CHF may be attributed to the suppressed hot dry spots due to the rewetting phenomena. (author)

  19. Critical heat flux prediction by using radial basis function and multilayer perceptron neural networks: A comparison study

    International Nuclear Information System (INIS)

    Vaziri, Nima; Hojabri, Alireza; Erfani, Ali; Monsefi, Mehrdad; Nilforooshan, Behnam

    2007-01-01

    Critical heat flux (CHF) is an important parameter for the design of nuclear reactors. Although many experimental and theoretical researches have been performed, there is not a single correlation to predict CHF because it is influenced by many parameters. These parameters are based on fixed inlet, local and fixed outlet conditions. Artificial neural networks (ANNs) have been applied to a wide variety of different areas such as prediction, approximation, modeling and classification. In this study, two types of neural networks, radial basis function (RBF) and multilayer perceptron (MLP), are trained with the experimental CHF data and their performances are compared. RBF predicts CHF with root mean square (RMS) errors of 0.24%, 7.9%, 0.16% and MLP predicts CHF with RMS errors of 1.29%, 8.31% and 2.71%, in fixed inlet conditions, local conditions and fixed outlet conditions, respectively. The results show that neural networks with RBF structure have superior performance in CHF data prediction over MLP neural networks. The parametric trends of CHF obtained by the trained ANNs are also evaluated and results reported

  20. Development of an artificial neural network to predict critical heat flux based on the look up tables

    Energy Technology Data Exchange (ETDEWEB)

    Terng, Nilton; Carajilescov, Pedro, E-mail: Nil.terng@gmail.com, E-mail: pedro.carajilescov@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais

    2015-07-01

    The critical heat flux (CHF) is one of the principal thermal hydraulic limits of PWR type nuclear reactors. The present work consists in the development of an artificial neural network (ANN) to estimate the CHF, based on Look Up Table CHF data, published by Groeneveld (2006). Three parameters were considered in the development of the ANN: the pressure in the range of 1 to 21 MPa, the mass flux in the range of 50 to 8000 kg m{sup -2} s{sup -1} and the thermodynamic quality in the range of - 0.5 to 0.9. The ANN model considered was a multi feed forward net, which have two feedforward ANN. The first one, called main neural network, is used to calculate the result of CHF, and the second, denominated spacenet, is responsible to modify the main neural network according to the input. Comparing the ANN predictions with the data of the Look Up Table, it was observed an average of the ratio of 0.993 and a root mean square error of 13.3%. With the developed ANN, a parametric study of CHF was performed to observe the influence of each parameter in the CHF. It was possible to note that the CHF decreases with the increase of pressure and thermodynamic quality, while CHF increases with the mass flow rate, as expected. However, some erratic trends were also observed which can be attributed to either unknown aspect of the CHF phenomenon or uncertainties in the data. (author)

  1. Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

    International Nuclear Information System (INIS)

    Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

    1995-01-01

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

  2. Nucleate pool boiling, film boiling and single-phase free convection at pressures up to the critical state. Part I: Integral heat transfer for horizontal copper cylinders

    Energy Technology Data Exchange (ETDEWEB)

    Gorenflo, Dieter; Baumhoegger, Elmar; Windmann, Thorsten; Herres, Gerhard [Institut fuer Energie- und Verfahrenstechnik, Universitaet Paderborn, Warburger Str. 100, D-33098 Paderborn (Germany)

    2010-11-15

    Transcritical working cycles for refrigerants have led to increased interest in heat transfer near the Critical State. In general, experimental results for this region differ significantly from those far from it because some fluid properties vary much more there than at a greater distance. In this paper, measurements for two-phase and single-phase free convective heat transfer from an electrically heated copper tube with 25 mm O.D. to refrigerant R125 are discussed for fluid states very close to the Critical Point and far from it. It is shown that heat transfer for film boiling slightly below and for free convection slightly above the critical pressure is very similar. The new - and also previous - experimental data for nucleate boiling, film boiling, and single-phase free convection are compared with calculated results between atmospheric and critical pressure. It can be concluded that the Principle of Corresponding States in its simplest form is very well suited to transfer the results to other refrigerants. In Part II, particular attention will be given to a minimum superheat for nucleate boiling and a maximum superheat for film boiling and single-phase free convection within the circumferential variation of the isobaric wall superheat on the lower parts of the tube. (author)

  3. Heat treatment effect on the strain dependence of the critical current for an internal-tin processed Nb3Sn strand

    International Nuclear Information System (INIS)

    Oh, Sangjun; Park, Soo-Hyeon; Lee, Chulhee; Choi, Heekyung; Kim, Keeman

    2010-01-01

    A comparative study on the effects of heat treatment, especially, the duration of the A15 reaction temperature plateau on the strain dependence of the critical current for an internal-tin processed Nb 3 Sn strand has been carried out. The strain dependence of the critical current is measured by a variable temperature Walter spiral probe that we have developed. It was shown that prolonged heat treatment can be a very effective way to improve the strain dependency. For a quantitative analysis, the measured data were analyzed with various proposed scaling laws: the scaling law based on strong-coupling theory, the modified deviatoric strain scaling law, and the interpolative scaling law. We found that there is a slight increase in the critical temperature and a substantial improvement in the maximum pinning force. The origin of improved strain dependency is further discussed.

  4. An experimental study of trans-critical CO2 water–water heat pump using compact tube-in-tube heat exchangers

    International Nuclear Information System (INIS)

    Jiang, Yuntao; Ma, Yitai; Li, Minxia; Fu, Lin

    2013-01-01

    Highlights: • Thermodynamic analyses of transcritical CO 2 cycle with and without IHX are provided. • A transcritical CO 2 heat pump system adopts compact tube-in-tube heat exchangers. • Experiment results of systems with and without IHX have been analyzed and compared. • IHX can improve the performance of the transcritical CO 2 heat pump system. - Abstract: A transcritical CO 2 water–water heat pump system is introduced in this study, which employs compact tube-in-tube evaporator and gas cooler. Its primary test standards and operating conditions are introduced. Under test conditions, experiments have been carried out with compression cycles with and without internal heat exchanger (IHX). Experiment results have been analyzed and compared, showing that IHX can improve the coefficient of performance of the system. The analyses are done mainly on the variations of outlet CO 2 temperature of the gas cooler, compressor discharge pressure, compressor lubricant temperature, hot water mass flow rate, etc. When the inlet water temperature of the gas cooler is 15 °C, 20 °C, 25 °C respectively, the hot water temperature ranges from 45 °C to 70 °C, the relative COP h (coefficient of performance when heating) change index (RCI COP ) of the heat pump system with IHX is about 3.5–8% higher than that without IHX. The relative capacity change index (RCI Q ) of the heat pump system with IHX is about 5–10% higher than that without IHX. Temperature of CO 2 increases at the outlet of the gas cooler when the outlet water temperature of the gas cooler increases. Lowering the outlet CO 2 temperature of the gas cooler is an important way to improve the performance of the system

  5. Critical role of heat shock protein 27 in bufalin-induced apoptosis in human osteosarcomas: a proteomic-based research.

    Directory of Open Access Journals (Sweden)

    Xian-biao Xie

    Full Text Available Bufalin is the primary component of the traditional Chinese herb "Chan Su". Evidence suggests that this compound possesses potent anti-tumor activities, although the exact molecular mechanism(s is unknown. Our previous study showed that bufalin inhibited growth of human osteosarcoma cell lines U2OS and U2OS/MTX300 in culture. Therefore, this study aims to further clarify the in vitro and in vivo anti-osteosarcoma effects of bufalin and its molecular mechanism of action. We found bufalin inhibited both methotrexate (MTX sensitive and resistant human osteosarcoma cell growth and induced G2/M arrest and apoptosis. Using a comparative proteomics approach, 24 differentially expressed proteins following bufalin treatment were identified. In particular, the level of an anti-apoptotic protein, heat shock protein 27 (Hsp27, decreased remarkably. The down-regulation of Hsp27 and alterations of its partner signaling molecules (the decrease in p-Akt, nuclear NF-κB p65, and co-immunoprecipitated cytochrome c/Hsp27 were validated. Hsp27 over-expression protected against bufalin-induced apoptosis, reversed the dephosphorylation of Akt and preserved the level of nuclear NF-κB p65 and co-immunoprecipitated Hsp27/cytochrome c. Moreover, bufalin inhibited MTX-resistant osteosarcoma xenograft growth, and a down-regulation of Hsp27 in vivo was observed. Taken together, bufalin exerted potent anti-osteosarcoma effects in vitro and in vivo, even in MTX resistant osteosarcoma cells. The down-regulation of Hsp27 played a critical role in bufalin-induced apoptosis in osteosarcoma cells. Bufalin may have merit to be a potential chemotherapeutic agent for osteosarcoma, particularly in MTX-resistant groups.

  6. Critical current, electro-mechanical properties and specific heat of bronze Nb{sub 3}Sn conductors

    Energy Technology Data Exchange (ETDEWEB)

    Uglietti, D.; Seeber, B.; Abacherli, V.; Flukiger, R. [Geneva Univ., Groupe Applique de Physique (GAP) (Switzerland); Wang, X.Y.; Junod, A.; Flukiger, R. [Geneva Univ., Dept. Phys. Mat. Condensee (DPMC) (Switzerland)

    2004-07-01

    The fabrication process leading to a Nb{sub 3}Sn wire by using the bronze route with 15.4 wt per cent of Sn is described. The critical current density, J{sub c}, is studied as a function of the applied magnetic field, B, up to 25 T; the uniaxial strain, {epsilon}, was measured up to 17. In the second part our device for measuring I{sub c}({epsilon}) is presented. The device is based on the concept of the Walters spring (WASP), which allows to measure long length wires (voltage taps distance up to 50 cm), up to 1000 A and to obtain an absolute measurement of the strain value. It is thus possible to measure the voltage-current relation of technical superconducting wires and tapes down to 0.01 {mu}V/cm, an important requirement for the characterisation in view of applications like NMR high field magnets which require persistent mode operation with high current densities. Finally specific heat measurements on Nb{sub 3}Sn wires prepared at GAP have allowed to determine for the first time the overall distribution of T{sub c} in the filaments. The onset of T{sub c} was observed at 17.2 K, the T{sub c} distribution being centred at 15.9 K. This analysis confirms the reduction of T{sub c} due to the Ti addition and the presence of a distribution of Sn in Nb{sub 3}Sn bronze wires. (authors)

  7. A study on the development of advanced models to predict the critical heat flux for water and liquid metals

    International Nuclear Information System (INIS)

    Lee, Yong Bum

    1994-02-01

    The critical heat flux (CHF) phenomenon in the two-phase convective flows has been an important issue in the fields of design and safety analysis of light water reactor (LWR) as well as sodium cooled liquid metal fast breeder reactor (LMFBR). Especially in the LWR application many physical aspects of the CHF phenomenon are understood and reliable correlations and mechanistic models to predict the CHF condition have been proposed. However, there are few correlations and models which are applicable to liquid metals. Compared with water, liquid metals show a divergent picture for boiling pattern. Therefore, the CHF conditions obtained from investigations with water cannot be applied to liquid metals. In this work a mechanistic model to predict the CHF of water and a correlation for liquid metals are developed. First, a mechanistic model to predict the CHF in flow boiling at low quality was developed based on the liquid sublayer dryout mechanism. In this approach the CHF is assumed to occur when a vapor blanket isolates the liquid sublayer from bulk liquid and then the liquid entering the sublayer falls short of balancing the rate of sublayer dryout by vaporization. Therefore, the vapor blanket velocity is the key parameter. In this work the vapor blanket velocity is theoretically determined based on mass, energy, and momentum balance and finally the mechanistic model to predict the CHF in flow boiling at low quality is developed. The accuracy of the present model is evaluated by comparing model predictions with the experimental data and tabular data of look-up tables. The predictions of the present model agree well with extensive CHF data. In the latter part a correlation to predict the CHF for liquid metals is developed based on the flow excursion mechanism. By using Baroczy two-phase frictional pressure drop correlation and Ledinegg instability criterion, the relationship between the CHF of liquid metals and the principal parameters is derived and finally the

  8. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    OpenAIRE

    H. P. Rahardjo; V. I. Sri Wardhani

    2017-01-01

    The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety an...

  9. The validation of science virtual test to assess 7th grade students’ critical thinking on matter and heat topic (SVT-MH)

    Science.gov (United States)

    Sya’bandari, Y.; Firman, H.; Rusyati, L.

    2018-05-01

    The method used in this research was descriptive research for profiling the validation of SVT-MH to measure students’ critical thinking on matter and heat topic in junior high school. The subject is junior high school students of 7th grade (13 years old) while science teacher and expert as the validators. The instruments that used as a tool to obtain the data are rubric expert judgment (content, media, education) and rubric of readability test. There are four steps to validate SVT-MH in 7th grade Junior High School. These steps are analysis of core competence and basic competence based on Curriculum 2013, expert judgment (content, media, education), readability test and trial test (limited and larger trial test). The instrument validation resulted 30 items that represent 8 elements and 21 sub-elements to measure students’ critical thinking based on Inch in matter and heat topic. The alpha Cronbach (α) is 0.642 which means that the instrument is sufficient to measure students’ critical thinking matter and heat topic.

  10. Literature survey of heat transfer and hydraulic resistance of water, carbon dioxide, helium and other fluids at supercritical and near-critical pressures

    Energy Technology Data Exchange (ETDEWEB)

    Pioro, I.L.; Duffey, R.B

    2003-04-01

    This survey consists of 430 references, including 269 Russian publications and 161 Western publications devoted to the problems of heat transfer and hydraulic resistance of a fluid at near-critical and supercritical pressures. The objective of the literature survey is to compile and summarize findings in the area of heat transfer and hydraulic resistance at supercritical pressures for various fluids for the last fifty years published in the open Russian and Western literature. The analysis of the publications showed that the majority of the papers were devoted to the heat transfer of fluids at near-critical and supercritical pressures flowing inside a circular tube. Three major working fluids are involved: water, carbon dioxide, and helium. The main objective of these studies was the development and design of supercritical steam generators for power stations (utilizing water as a working fluid) in the 1950s, 1960s, and 1970s. Carbon dioxide was usually used as the modeling fluid due to lower values of the critical parameters. Helium, and sometimes carbon dioxide, were considered as possible working fluids in some special designs of nuclear reactors. (author)

  11. Replacing critical radiators to increase the potential to use low-temperature district heating – A case study of 4 Danish single-family houses from the 1930s

    International Nuclear Information System (INIS)

    Østergaard, Dorte Skaarup; Svendsen, Svend

    2016-01-01

    Low-temperature district heating is a promising technology for providing homes with energy-efficient heating in the future. However, it is of great importance to maintain thermal comfort in existing buildings when district heating temperatures are lowered. This case study evaluated the actual radiator sizes and heating demands in 4 existing Danish single-family houses from the 1930s. A year-long dynamic simulation was performed for each of the houses to evaluate the potential to lower the heating system temperatures. The results indicate that there is a large potential to use low-temperature district heating in existing single-family houses. In order to obtain the full potential of low-temperature district heating, critical radiators must be replaced. Based on a novel method, a total of nine radiators were identified to be critical to ensure thermal comfort and low return temperatures in two of the case-houses. If these radiators were replaced it would be possible to lower the average heating system temperatures to 50 °C/27 °C in all four houses. - Highlights: • Comparison of dynamically calculated heat demands and radiator sizes. • Method for identification and evaluation of critical radiators was tested. • Existing houses can be heated with low-temperature heating for most of the year. • Replacing critical radiators helps ensure comfort and low return temperatures.

  12. Investigation of the critical edge ion heat flux for L-H transitions in Alcator C-Mod and its dependence on B T

    Science.gov (United States)

    Schmidtmayr, M.; Hughes, J. W.; Ryter, F.; Wolfrum, E.; Cao, N.; Creely, A. J.; Howard, N.; Hubbard, A. E.; Lin, Y.; Reinke, M. L.; Rice, J. E.; Tolman, E. A.; Wukitch, S.; Ma, Y.; ASDEX Upgrade Team; Alcator C-Mod Team

    2018-05-01

    This paper presents investigations on the role of the edge ion heat flux for transitions from L-mode to H-mode in Alcator C-Mod. Previous results from the ASDEX Upgrade tokamak indicated that a critical value of edge ion heat flux per particle is needed for the transition. Analysis of C-Mod data confirms this result. The edge ion heat flux is indeed found to increase linearly with density at given magnetic field and plasma current. Furthermore, the Alcator C-Mod data indicate that the edge ion heat flux at the L-H transition also increases with magnetic field. Combining the data from Alcator C-Mod and ASDEX Upgrade yields a general expression for the edge ion heat flux at the L-H transition. These results are discussed from the point of view of the possible physics mechanism of the L-H transition. They are also compared to the L-H power threshold scaling and an extrapolation for ITER is given.

  13. Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.A.; Boyd, R.D.; Easor, J.R.; Gauster, W.B.; Gordon, J.D.; Mattas, R.F.; Morgan, G.D.; Ulrickson, M.A,; Watson, R.D.; Wolfer, W.G,

    1984-06-01

    A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted.

  14. Magnetic fusion energy plasma interactive and high heat flux components. Volume II. Technical assessment of the critical issues and problem areas in high heat flux materials and component development

    International Nuclear Information System (INIS)

    Abdou, M.A.; Boyd, R.D.; Easor, J.R.

    1984-06-01

    A technical assessment of the critical issues and problem areas for high heat flux materials and components (HHFMC) in magnetic fusion devices shows these problems to be of critical importance for the successful operation of near-term fusion experiments and for the feasibility and attractiveness of long-term fusion reactors. A number of subgroups were formed to assess the critical HHFMC issues along the following major lines: (1) source conditions, (2) systems integration, (3) materials and processes, (4) thermal hydraulics, (5) thermomechanical response, (6) electromagnetic response, (7) instrumentation and control, and (8) test facilities. The details of the technical assessment are presented in eight chapters. The primary technical issues and needs for each area are highlighted

  15. Highly precise (liquid + liquid) equilibrium and heat capacity measurements near the critical point for [Bmim][BF4] + 1H, 1H, 2H, 2H perfluoroctanol

    International Nuclear Information System (INIS)

    Pérez-Sánchez, G.; Troncoso, J.; Losada-Pérez, P.; Méndez-Castro, P.; Romaní, L.

    2013-01-01

    Highlights: • Highly precise liquid–liquid curves for [Bmim][BF 4 ] + perfluoroctanol are reported. • Critical behavior of heat capacity for the same system was also characterized. • In contrast to previous results, no coulombic/solvophobic crossover for coexistence curve diameter was found. • The system criticality shows characteristics both solvophobic and coulombic. -- Abstract: Liquid + liquid equilibrium of the system [Bmim][BF 4 ] + 1H, 1H, 2H, 2H perfluoroctanol using a highly precise methodology based on refractive index measurements was experimentally determined. In addition, isobaric heat capacity near the critical point was obtained. The performance of the new refractive index set-up was successfully checked against the coexistence curve of the system dimethyl carbonate + decane, since highly accurate data are available in the literature. The choice of [Bmim][BF 4 ] + 1H, 1H, 2H, 2H perfluoroctanol was motivated by a previous experimental work, whose results suggest that this system could present characteristics of both solvophobic and coulombic behavior, which are the two categories to which an ionic system can belong. Although this was previously observed for other ionic systems, this mixture presented a very striking feature: the diameter of the coexistence curve seemed to change its criticality in the studied temperature range, from solvophobic far away to coulombic close to the critical point. The results of this work reveal that, in fact, [Bmim][BF 4 ] + 1H, 1H, 2H, 2H perfluoroctanol presents characteristics of both solvophobic and coulombic criticality, but no evidence of the observed crossover over the experimental temperature range has been found

  16. Critical factors for profitable combined production of heat, power and biofuels; Kritiska faktorer foer loensam produktion i bioenergikombinat

    Energy Technology Data Exchange (ETDEWEB)

    Nohlgren, Ingrid; Gunnarsson, Emma; Lundqvist, Per; Stigander, Haakan; Widmark, Annika (AaF, Stockholm (Sweden))

    2012-02-15

    During the last 5-10 years, research and development efforts have been made in the field of polygeneration of heat and power with production of 'other green' products such as transport fuels or wood pellets. The driving force for heat and power producers is the potential of increased profitability through additional sales of heat. The driving force for wood pellet and some transport fuel producers is the potential of low cost process steam or heat. However, in the case of gasification based transport fuel production processes the situation is different. The process generates a surplus of heat, which can benefit from the proximity of a district heating net. In addition, some polygeneration combinations could provide other advantages such as more efficient raw material handling. Together with these driving forces, the EU renewable energy directive (which targets 10 % renewable energy use in the transport sector by 2020), shows that the market for production of renewable transport fuel is expanding. To refine Swedish biomass resources to more highly valuable products such as wood pellets or renewable transport fuels would maintain industry and employment opportunities within Sweden and at the same time fulfils the international and national climate targets. The overall aim with this project is to describe the factors which are crucial for the opportunity for profitable polygeneration of heat, power and wood pellets or renewable transport fuels and how these factors influence the location of such a plant within Sweden. The important factors can be categorized as: (1) Supply of raw material, (2) distribution of raw material and products, (3) Demand of products and (4) Integration between the different plants. In this project, only general aspects are described and should be seen as guidance for the industry (both energy and forest industry) which has an interest in polygeneration. The project gives an overview of different possibilities, opportunities and

  17. Replacing critical radiators to increase the potential to use low-temperature district heating – A case study of 4 Danish single-family houses from the 1930s

    DEFF Research Database (Denmark)

    Østergaard, Dorte Skaarup; Svendsen, Svend

    2016-01-01

    radiator sizes and heating demands in 4 existing Danish single-family houses from the 1930s. A year-long dynamic simulation was performed for each of the houses to evaluate the potential to lower the heating system temperatures. The results indicate that there is a large potential to use low......-temperature district heating in existing single-family houses. In order to obtain the full potential of low-temperature district heating, critical radiators must be replaced. Based on a novel method, a total of nine radiators were identified to be critical to ensure thermal comfort and low return temperatures in two...

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  19. Study on critical heat flux in narrow rectangular channel with repeated-rib roughness. 1. Experimental facility and preliminary experiments

    International Nuclear Information System (INIS)

    Kinoshita, Hidetaka; Terada, Atsuhiko; Kaminaga, Masanori; Hino, Ryutaro

    2001-10-01

    In the design of a spallation target system, the water cooling system, for example a proton beam window and a safety hull, is used with narrow channels, in order to remove high heat flux and prevent lowering of system performance by absorption of neutron. And in narrow channel, heat transfer enhancement using 2-D rib is considered for reduction the cost of cooling component and decrease inventory of water in the cooling system, that is, decrease of the amount of irradiated water. But few studies on CHF with rib have been carried out. Experimental and analytical studies with rib-roughened test section, in 10:1 ratio of pitch to height, are being carried out in order to clarify the CHF in rib-roughened channel. This paper presents the review of previous researches on heat transfer in channel with rib roughness, overview of the test facility and the preliminary experimental and analytical results. As a result, wall friction factors were about 3 times as large as that of smooth channel, and heat transfer coefficients are about 2 times as large as that of smooth channel. The obtained CHF was as same as previous mechanistic model by Sudo. (author)

  20. IR-thermography-based investigation of critical heat flux in subcooled flow boiling of water at atmospheric and high pressure conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bucci, Matteo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Seong, Jee H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Buongiorno, Jdacopo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Richenderfer, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kossolapov, A. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2017-11-01

    Here we report on MIT’s THM work in Q4 2016 and Q1 2017. The goal of this project is to design, construct and execute tests of flow boiling critical heat flux (CHF) at high-pressure using high-resolution and high-speed video and infrared (IR) thermometry, to generate unique data to inform the development of and validate mechanistic boiling heat transfer and CHF models. In FY2016, a new test section was designed and fabricated. Data was collected at atmospheric conditions at 10, 25 and 50 K subcoolings, and three mass fluxes, i.e. 500, 750 and 1000 kg/m2/s. Starting in Q4 2016 and continuing forward, new post-processing techniques have been developed to analyze the data collected. These new algorithms analyze the time-dependent temperature and heat flux distributions to calculate nucleation site density, nucleation frequency, growth and wait time, dry area fraction, and the complete heat flux partitioning. In Q1 2017 a new flow boiling loop was designed and constructed to support flow boiling tests up 10 bar pressure and 180 °C. Initial shakedown and testing has been completed. The flow loop and test section are now ready to begin high-pressure flow boiling testing.

  1. Power load limits of the WENDELSTEIN 7-X target elements-comparison of experimental results and design values for power loads up to the critical heat flux

    International Nuclear Information System (INIS)

    Greuner, H; Boeswirth, B; Boscary, J; Leuprecht, A; Plankensteiner, A

    2007-01-01

    The power load limits of the WENDELSTEIN7-X divertor target elements were experimentally evaluated with heat loads considerably exceeding the expected operating conditions. The water-cooled elements are designed for steady-state heat flux of 10 MW m -2 and to remove a power load up to 100 kW. The elements must allow a limited operation time at 12 MW m -2 steady-state and should not fail for short pulses of up to 15 MW m -2 for cooling conditions in the subcooled nucleate boiling regime. In the framework of the qualification phase, pre-series target elements were loaded up to 24 MW m -2 without loss of CFC tiles. A critical heat flux at the target of 31 MW m -2 was achieved. The paper discusses the results of the tests performed at the high heat flux test facility GLADIS. The experimental results compared to transient nonlinear fine element method (FEM) calculations confirm a high thermal safety margin of the target design sufficient for plasma operation in W7-X

  2. Changes in antioxidants are critical in determining cell responses to short- and long-term heat stress.

    Science.gov (United States)

    Sgobba, Alessandra; Paradiso, Annalisa; Dipierro, Silvio; De Gara, Laura; de Pinto, Maria Concetta

    2015-01-01

    Heat stress can have deleterious effects on plant growth by impairing several physiological processes. Plants have several defense mechanisms that enable them to cope with high temperatures. The synthesis and accumulation of heat shock proteins (HSPs), as well as the maintenance of an opportune redox balance play key roles in conferring thermotolerance to plants. In this study changes in redox parameters, the activity and/or expression of reactive oxygen species (ROS) scavenging enzymes and the expression of two HSPs were studied in tobacco Bright Yellow-2 (TBY-2) cells subjected to moderate short-term heat stress (SHS) and long-term heat stress (LHS). The results indicate that TBY-2 cells subjected to SHS suddenly and transiently enhance antioxidant systems, thus maintaining redox homeostasis and avoiding oxidative damage. The simultaneous increase in HSPs overcomes the SHS and maintains the metabolic functionality of cells. In contrast the exposure of cells to LHS significantly reduces cell growth and increases cell death. In the first phase of LHS, cells enhance antioxidant systems to prevent the formation of an oxidizing environment. Under prolonged heat stress, the antioxidant systems, and particularly the enzymatic ones, are inactivated. As a consequence, an increase in H2 O2 , lipid peroxidation and protein oxidation occurs. This establishment of oxidative stress could be responsible for the increased cell death. The rescue of cell growth and cell viability, observed when TBY-2 cells were pretreated with galactone-γ-lactone, the last precursor of ascorbate, and glutathione before exposure to LHS, highlights the crucial role of antioxidants in the acquisition of basal thermotolerance. © 2014 Scandinavian Plant Physiology Society.

  3. Universal fine structure of the specific heat at the critical λ-point for an ideal Bose gas in an arbitrary trap

    International Nuclear Information System (INIS)

    Tarasov, S V; Kocharovsky, Vl V; Kocharovsky, V V

    2014-01-01

    We analytically find the universal fine structure of the noted discontinuity in the value and/or derivative of the specific heat of an ideal Bose gas in an arbitrary trap in the whole critical region around the λ-point of the Bose–Einstein condensation. The result reveals a remarkable dependence of the λ-point structure on the trap's form and boundary conditions, even for a macroscopically large system. We suggest measuring this strong effect in the experiments with a controllable trap potential. (paper)

  4. Global Gene-Expression Analysis to Identify Differentially Expressed Genes Critical for the Heat Stress Response in Brassica rapa.

    Directory of Open Access Journals (Sweden)

    Xiangshu Dong

    Full Text Available Genome-wide dissection of the heat stress response (HSR is necessary to overcome problems in crop production caused by global warming. To identify HSR genes, we profiled gene expression in two Chinese cabbage inbred lines with different thermotolerances, Chiifu and Kenshin. Many genes exhibited >2-fold changes in expression upon exposure to 0.5- 4 h at 45°C (high temperature, HT: 5.2% (2,142 genes in Chiifu and 3.7% (1,535 genes in Kenshin. The most enriched GO (Gene Ontology items included 'response to heat', 'response to reactive oxygen species (ROS', 'response to temperature stimulus', 'response to abiotic stimulus', and 'MAPKKK cascade'. In both lines, the genes most highly induced by HT encoded small heat shock proteins (Hsps and heat shock factor (Hsf-like proteins such as HsfB2A (Bra029292, whereas high-molecular weight Hsps were constitutively expressed. Other upstream HSR components were also up-regulated: ROS-scavenging genes like glutathione peroxidase 2 (BrGPX2, Bra022853, protein kinases, and phosphatases. Among heat stress (HS marker genes in Arabidopsis, only exportin 1A (XPO1A (Bra008580, Bra006382 can be applied to B. rapa for basal thermotolerance (BT and short-term acquired thermotolerance (SAT gene. CYP707A3 (Bra025083, Bra021965, which is involved in the dehydration response in Arabidopsis, was associated with membrane leakage in both lines following HS. Although many transcription factors (TF genes, including DREB2A (Bra005852, were involved in HS tolerance in both lines, Bra024224 (MYB41 and Bra021735 (a bZIP/AIR1 [Anthocyanin-Impaired-Response-1] were specific to Kenshin. Several candidate TFs involved in thermotolerance were confirmed as HSR genes by real-time PCR, and these assignments were further supported by promoter analysis. Although some of our findings are similar to those obtained using other plant species, clear differences in Brassica rapa reveal a distinct HSR in this species. Our data could also provide a

  5. Pre-critical phenomena of two-flavor color superconductivity in heated quark matter. Diquark-pair fluctuations and non-Fermi liquid behavior

    International Nuclear Information System (INIS)

    Kitazawa, Masakiyo; Kunihiro, Teiji; Koide, Tomoi; Nemoto, Yukio

    2005-01-01

    We investigate the fluctuations of the diquark-pair field and their effects on observables above the critical temperature T c in two-flavor color superconductivity (CSC) at moderate density using a Nambu-Jona-Lasinio-type effective model of QCD. Because of the strong-coupling nature of the dynamics, the fluctuations of the pair field develop a collective mode, which has a prominent strength even well above T c . We show that the collective mode is actually the soft mode of CSC. We examine the effects of the pair fluctuations on the specific heat and the quark spectrum for T above but close to T c . We find that the specific heat exhibits singular behavior because of the pair fluctuations, in accordance with the general theory of second-order phase transitions. The quarks display a typical non-Fermi liquid behavior, owing to the coupling with the soft mode, leading to a pseudo-gap in the density of states of the quarks in the vicinity of the critical point. Some experimental implications of the precursory phenomena are also discussed. (author)

  6. Specific heat near the Lambda point in 4He and 3He- 4He mixtures: test of universality of the critical exponent and the amplitude ratio, and observation of the critical-tricritical crossover effect

    International Nuclear Information System (INIS)

    Takada, T.; Watanabe, T.

    1980-01-01

    The specific heat under saturated vapor pressure of pure 4 He and of six 3 He- 4 He mixtures up to X=0.545 was measured in the temperature range 3 x 10 -6 -2 K. The critical exponents α/sub phi/ and α'/sub phi/ along the path phi=phi/sub lambda/ are independent of X up to X=0.545, where phi(=μ 3 -μ 4 ) is the difference between chemical potentials. If we take account of higher order terms, the exponent α/sub phi/(=α'/sub phi/) and the amplitude ratio A/sub //A'are independent of X up to X=0.545. The values of α/sub phi/ and A/sub //A'/sub phi/ are -0.023 and 1.090, respectively. The critical-tricriticall crossover effect was observed for X=0.545 and the boundary of crossover region closest to the critical region was at theta/T/sub lambda/1(times)=10 -4 , where theta is the distance Vertical BarT-T/sub lambda/Vertical Bar along the path phi=phi/sub lambda/. This value is in good agreement with the estimated value by Riedel et al. But, remarkably, in the case of X=0.439 this effect was not observed

  7. A new flooding correlation development and its critical heat flux predictions under low air-water flow conditions in Savannah River Site assembly channels

    International Nuclear Information System (INIS)

    Lee, S.Y.

    1993-01-01

    The upper limit to countercurrent flow, namely, flooding, is important to analyze the reactor coolability during an emergency cooling system (ECS) phase as a result of a large-break loss-of-coolant accident (LOCA) such as a double-ended guillotine break in the Savannah River Site (SRS) reactor system. During normal operation, the reactor coolant system utilizes downward flow through concentric heated tubes with ribs, which subdivided each annular channel into four subchannels. In this paper, a new flooding correlation has been developed based on the analytical models and literature data for adiabatic, steady-state, one-dimensional, air-water flow to predict flooding phenomenon in the SRS reactor assembly channel, which may have a counter-current air-water flow pattern during the ECS phase. In addition, the correlation was benchmarked against the experimental data conducted under the Oak Ridge National Laboratory multislit channel, which is close to the SRS assembly geometry. Furthermore, the correlation has also been used as a constitutive relationship in a new two-component two-phase thermal-hydraulics code FLOWTRAN-TF, which has been developed for a detailed analysis of SRS reactor assembly behavior during LOCA scenarios. Finally, the flooding correlation was applied to the predictions of critical heat flux, and the results were compared with the data taken by the SRS heat transfer laboratory under a single annular channel with ribs and a multiannular prototypic test rig

  8. Flow visualization study of post critical heat flux region for inverted bubbly, slug and annular flow regimes

    International Nuclear Information System (INIS)

    Denten, J.G.; Ishii, M.

    1988-11-01

    A visual study of film boiling using still photographic and high- speed motion picture methods was carried out in order to analyze the post-CHF hydrodynamics for steady-state inlet pre-CHF two-phase flow regimes. Pre-CHF two-phase flow regimes were established by introducing Freon 113 liquid and nitrogen gas into a jet core injection nozzle. An idealized, post-CHF two-phase core initial flow geometry (cylindrical multiphase jet core surrounded by a coaxial annulus of gas) was established at the nozzle exit by introducing nitrogen gas into the annular gap between the jet nozzle two-phase effluent and the heated test section inlet. For the present study three basic post-CHF flow regimes have been observed: the rough wavy regime (inverted annular flow preliminary break down), the agitated regime (transition between inverted annular and dispersed droplet flow), and the dispersed ligament/droplet regime. For pre-CHF bubbly flow in the jet nozzle, the post-CHF flow (beginning from jet nozzle exit/heated test section inlet) consists of the rough wavy regime, followed by the agitated and then the dispersed ligament/droplet regime. In the same way, for pre-CHF slug flow in the jet core, the post-CHF flow is comprised of the agitated regime at the nozzle exit, followed by the dispersed regime. Pre-CHF annular jet core flow results in a small, depleted post-CHF agitated flow regime at the nozzle exit, immediately followed by the dispersed ligament/droplet regime. Observed post dryout hydrodynamic behavior is reported, with particular attention given to the transition flow pattern between inverted annular and dispersed droplet flow. 43 refs., 20 figs., 5 tabs

  9. The reduction of optimal heat treatment temperature and critical current density enhancement of ex situ processed MgB2 tapes using ball milled filling powder

    Science.gov (United States)

    Fujii, Hiroki; Iwanade, Akio; Kawada, Satoshi; Kitaguchi, Hitoshi

    2018-01-01

    The optimal heat treatment temperature (Topt) at which best performance in the critical current density (Jc) property at 4.2 K is obtained is influenced by the quality or reactivity of the filling powder in ex situ processed MgB2 tapes. Using a controlled fabrication process, the Topt decreases to 705-735 °C, which is lower than previously reported by more than 50 °C. The Topt decrease is effective to suppress both the decomposition of MgB2 and hence the formation of impurities such as MgB4, and the growth of crystallite size which decreases upper critical filed (Hc2). These bring about the Jc improvement and the Jc value at 4.2 K and 10 T reaches 250 A/mm2. The milling process also decreases the critical temperature (Tc) below 30 K. The milled powder is easily contaminated in air and thus, the Jc property of the contaminated tapes degrades severely. The contamination can raise the Topt by more than 50 °C, which is probably due to the increased sintering temperature required against contaminated surface layer around the grains acting as a barrier.

  10. Critical behavior of binary mixture of {x C6H5CN + (1 - x) CH3(CH2)12CH3}: Measurements of coexistence curves, turbidity, and heat capacity

    International Nuclear Information System (INIS)

    Yin Tianxiang; Lei Yuntao; Huang Meijun; Chen Zhiyun; Mao Chunfeng; An Xueqin; Shen Weiguo

    2011-01-01

    Research highlights: → Coexistence curve, turbidity and heat capacity of critical solution were measured. → Critical amplitudes were determined to test universal ratios. → Complete scaling theory was verified. → Monotonic critical crossover behavior was demonstrated. - Abstract: (Liquid + liquid) coexistence curve, turbidity, and isobaric heat capacity per unit volume for the critical solution of {benzonitrile + n-tetradecane} have been measured. The critical exponents β, ν, γ, and α and system-dependent critical amplitudes B, ξ 0 , χ 0 , and A ± , corresponding to the difference of the general density variable of two coexisting phases Δρ, the correlation length ξ, the osmotic compressibility χ, and the isobaric heat capacity per unit volume C p V -1 , have been deduced and were used to test some universal ratios. The behavior of the diameter of the coexistence curves showed good agreement with the complete scaling theory. The analysis of effective critical exponent β eff , which was well described by the crossover model proposed by Anisimov and Sengers, and effective critical exponent α eff indicated monotonic crossover phenomena from 3D-Ising behavior to mean-field one as the temperature departed from the critical point.

  11. IAEA coordinated research programme on heat transfer behavior and thermo-hydraulics code testing for super critical water cooled reactors

    International Nuclear Information System (INIS)

    Bilbao y Leon, Sama; Aksan, Nusret

    2009-01-01

    One of the key roles of the IAEA is to foster the collaboration among Member States on the development of advances in technology for advanced nuclear power plants. There is high international interest, both in developing and industrialized countries, in innovative supercritical water-cooled reactors (SCWRs), primarily because such concepts will achieve high thermal efficiencies (44-45%) and promise improved economic competitiveness utilizing and building upon the recent developments for highly efficient fossil power plants. The SCWR has been selected as one of the promising concepts for development by the Generation-IV International Forum. Following the advice of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA has recently started a Coordinated Research Programme (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The first Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna, Austria in July 2008. This paper summarizes the current status of the CRP, including the Integrated Research Plan and the general schedule for the CRP. (author)

  12. Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle

    International Nuclear Information System (INIS)

    Onder, Ebru Nihan; Leung, Laurence; Kim, Hung; Rao, Yanfei

    2009-01-01

    The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU 1 reactor fuel bundle. Based primarily on the CANFLEX 2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures

  13. Effects of Oxidation and fractal surface roughness on the wettability and critical heat flux of glass-peened zirconium alloy tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Nitheanandan, T.; Bullock, C.D.; Slater, L.F.; McRae, G.A.

    2003-05-01

    Glass-bead peening the outside surfaces of zirconium alloy tubes has been shown to increase the Critical Heat Flux (CHF) in pool boiling of water. The CHF is found to correlate with the fractal roughness of the metal tube surfaces. In this study on the effect of oxidation on glass-peened surfaces, test measurements for CHF, surface wettability and roughness have been evaluated using various glass-peened and oxidized zirconium alloy tubes. The results show that oxidation changes the solid-liquid contact angle (i.e., decreases wettability of the metal-oxide surface), but does not change the fractal surface roughness, appreciably. Thus, oxidation of the glass-peened surfaces of zirconium alloy tubes is not expected to degrade the CHF enhancement obtained by glass-bead peening. (author)

  14. Steady-state and transient studies on critical heat flux of a PWR 5 x 5 fuel element bundle with complex spacer wire geometry

    International Nuclear Information System (INIS)

    Fulfs, H.; Katsaounis, A.; Kreubig, M.; Minden, C. von; Orlowski, R.

    1980-01-01

    The results will be described in exemplary presentations completely and concluding. The experimental examination of the steady state simularity of critical heat flux (CHF) in freon 12 and water at identical PWR-5 x 15-rod bundles will show that hot rod/hot channels position as well as CHF can be transformed from model to original fluid with good accuracy. The investigated mass flow and power transients (only in freon 12) point out a definite influence of initial and boundary conditions on CHF and CHF time delay at changing rates higher than 10 to 20%/s. On the contrary simulation of primary pump failure (LOFA) shows no or only small improvement in CHF behaviour while a coupled Scram prevents from reaching the boiling crisis. (orig.) [de

  15. Can Temperate-Water Immersion Effectively Reduce Rectal Temperature in Exertional Heat Stroke? A Critically Appraised Topic.

    Science.gov (United States)

    Truxton, Tyler T; Miller, Kevin C

    2017-09-01

    Clinical Scenario: Exertional heat stroke (EHS) is a medical emergency which, if left untreated, can result in death. The standard of care for EHS patients includes confirmation of hyperthermia via rectal temperature (T rec ) and then immediate cold-water immersion (CWI). While CWI is the fastest way to reduce T rec , it may be difficult to lower and maintain water bath temperature in the recommended ranges (1.7°C-15°C [35°F-59°F]) because of limited access to ice and/or the bath being exposed to high ambient temperatures for long periods of time. Determining if T rec cooling rates are acceptable (ie, >0.08°C/min) when significantly hyperthermic humans are immersed in temperate water (ie, ≥20°C [68°F]) has applications for how EHS patients are treated in the field. Are T rec cooling rates acceptable (≥0.08°C/min) when significantly hyperthermic humans are immersed in temperate water? T rec cooling rates of hyperthermic humans immersed in temperate water (≥20°C [68°F]) ranged from 0.06°C/min to 0.19°C/min. The average T rec cooling rate for all examined studies was 0.11±0.06°C/min. Clinical Bottom Line: Temperature water immersion (TWI) provides acceptable (ie, >0.08°C/min) T rec cooling rates for hyperthermic humans post-exercise. However, CWI cooling rates are higher and should be used if feasible (eg, access to ice, shaded treatment areas). Strength of Recommendation: The majority of evidence (eg, Level 2 studies with PEDro scores ≥5) suggests TWI provides acceptable, though not ideal, T rec cooling. If possible, CWI should be used instead of TWI in EHS scenarios.

  16. Two-phase flow characteristic of inverted bubbly, slug and annular flow in post-critical heat flux region

    International Nuclear Information System (INIS)

    Ishii, M.; Denten, J.P.

    1988-01-01

    Inverted annular flow can be visualized as a liquid jet-like core surrounded by a vapor annulus. While many analytical and experimental studies of heat transfer in this regime have been performed, there is very little understanding of the basic hydrodynamics of the post-CHF flow field. However, a recent experimental study was done that was able to successfully investigate the effects of various steady-state inlet flow parameters on the post-CHF hydrodynamics of the film boiling of a single phase liquid jet. This study was carried out by means of a visual photographic analysis of an idealized single phase core inverted annular flow initial geometry (single phase liquid jet core surrounded by a coaxial annulus of gas). In order to extend this study, a subsequent flow visualization of an idealized two-phase core inverted annular flow geometry (two-phase central jet core, surrounded by a coaxial annulus of gas) was carried out. The objective of this second experimental study was to investigate the effect of steady-state inlet, pre-CHF two-phase jet core parameters on the hydrodynamics of the post-CHF flow field. In actual film boiling situations, two-phase flows with net positive qualities at the CHF point are encountered. Thus, the focus of the present experimental study was on the inverted bubbly, slug, and annular flow fields in the post dryout film boiling region. Observed post dryout hydrodynamic behavior is reported. A correlation for the axial extent of the transition flow pattern between inverted annular and dispersed droplet flow (the agitated regime) is developed. It is shown to depend strongly on inlet jet core parameters and jet void fraction at the dryout point. 45 refs., 9 figs., 4 tabs

  17. Dryout-type critical heat flux in vertical upward annular flow: effects of entrainment rate, initial entrained fraction and diameter

    Science.gov (United States)

    Wu, Zan; Wadekar, Vishwas; Wang, Chenglong; Sunden, Bengt

    2018-01-01

    This study aims to reveal the effects of liquid entrainment, initial entrained fraction and tube diameter on liquid film dryout in vertical upward annular flow for flow boiling. Entrainment and deposition rates of droplets were included in mass conservation equations to estimate the local liquid film mass flux in annular flow, and the critical vapor quality at dryout conditions. Different entrainment rate correlations were evaluated using flow boiling data of water and organic liquids including n-pentane, iso-octane and R134a. Effect of the initial entrained fraction (IEF) at the churn-to-annular flow transition was also investigated. A transitional Boiling number was proposed to separate the IEF-sensitive region at high Boiling numbers and the IEF-insensitive region at low Boiling numbers. Besides, the diameter effect on dryout vapor quality was studied. The dryout vapor quality increases with decreasing tube diameter. It needs to be pointed out that the dryout characteristics of submillimeter channels might be different because of different mechanisms of dryout, i.e., drying of liquid film underneath long vapor slugs and flow boiling instabilities.

  18. Study of critical free-area ratio during the snow-melting process on pavement using low-temperature heating fluids

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Huajun [School of Energy and Environment Engineering, Hebei University of Technology, Tianjin 300401 (China); Chen, Zhihao [Faculty of Engineering, Yokohama National University, Hodogaya, Yokohama 240-8501 (Japan)

    2009-01-15

    Critical free-area ratio (CFR) is an interesting phenomenon during the snow-melting process on pavement using low-temperature heating fluids such as geothermal tail water and industrial waste water. This paper is performed to further investigate the mechanism of CFR and its influencing factors. A simplified theoretical model is presented to describe the heat and mass transfer process on pavement. Especially the variation of thermal properties and the capillary effect of snow layer are considered. Numerical computation shows that the above theoretical model is effective for the prediction of CFR during the snow-melting process. Furthermore, the mechanism of CFR is clarified in detail. CFR is independent of the layout of hydronic pipes, the fluid temperature, the idling time, and weather conditions. It is both the non-uniform temperature distribution and complicated porous structure of snow layer that lead to the occurrence of CFR. Besides, the influences of operation parameters including the fluid temperature, the idling time, the pipe spacing and buried depths on snow melting are analyzed, which are helpful for the next optimal design of snow-melting system. (author)

  19. Study of critical free-area ratio during the snow-melting process on pavement using low-temperature heating fluids

    Energy Technology Data Exchange (ETDEWEB)

    Wang Huajun [School of Energy and Environment Engineering, Hebei University of Technology, Tianjin 300401 (China)], E-mail: huajunwang@126.com; Chen Zhihao [Faculty of Engineering, Yokohama National University, Hodogaya, Yokohama 240-8501 (Japan)

    2009-01-15

    Critical free-area ratio (CFR) is an interesting phenomenon during the snow-melting process on pavement using low-temperature heating fluids such as geothermal tail water and industrial waste water. This paper is performed to further investigate the mechanism of CFR and its influencing factors. A simplified theoretical model is presented to describe the heat and mass transfer process on pavement. Especially the variation of thermal properties and the capillary effect of snow layer are considered. Numerical computation shows that the above theoretical model is effective for the prediction of CFR during the snow-melting process. Furthermore, the mechanism of CFR is clarified in detail. CFR is independent of the layout of hydronic pipes, the fluid temperature, the idling time, and weather conditions. It is both the non-uniform temperature distribution and complicated porous structure of snow layer that lead to the occurrence of CFR. Besides, the influences of operation parameters including the fluid temperature, the idling time, the pipe spacing and buried depths on snow melting are analyzed, which are helpful for the next optimal design of snow-melting system.

  20. Critical energy in the cyclotron heating of ions in a mirror machine; Energia critica en el calentamiento ciclotronico de los iones en una maquina espejo

    Energy Technology Data Exchange (ETDEWEB)

    Gutierrez T, C.; Hernandez A, O. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2002-07-01

    The problem of heating in the plasma sources where the geometry of the magnetic field forms a magnetic mirror as it is the case of the Ecr sources type, for maintaining the reload, it continues being an actual important problem. There are two methods for the analysis of this problem. The first of these methods is the stochastic mechanism of a particle where it is considered the existence of three characteristic frequencies as the cyclotron frequency, the electromagnetic field frequency and the transit frequency. The second method is that related with the non linear interaction of waves where the collective effects of the particles are the most important. In this work, in the Hamiltonian formalism, the stochastic mechanism in the cyclotron heating is analysed. It is considered the particular case of a plasma source with an external magnetic field, type mirror where a TE{sub 11} electromagnetic wave is injected. The critical energy in the resonance mixing is calculated by the Poincare mapping method. The heterogeneity of the magnetic field is analysed. (Author)

  1. Study of critical free-area ratio during the snow-melting process on pavement using low-temperature heating fluids

    International Nuclear Information System (INIS)

    Wang Huajun; Chen Zhihao

    2009-01-01

    Critical free-area ratio (CFR) is an interesting phenomenon during the snow-melting process on pavement using low-temperature heating fluids such as geothermal tail water and industrial waste water. This paper is performed to further investigate the mechanism of CFR and its influencing factors. A simplified theoretical model is presented to describe the heat and mass transfer process on pavement. Especially the variation of thermal properties and the capillary effect of snow layer are considered. Numerical computation shows that the above theoretical model is effective for the prediction of CFR during the snow-melting process. Furthermore, the mechanism of CFR is clarified in detail. CFR is independent of the layout of hydronic pipes, the fluid temperature, the idling time, and weather conditions. It is both the non-uniform temperature distribution and complicated porous structure of snow layer that lead to the occurrence of CFR. Besides, the influences of operation parameters including the fluid temperature, the idling time, the pipe spacing and buried depths on snow melting are analyzed, which are helpful for the next optimal design of snow-melting system

  2. Introducing the concept of critical Fo in batch heat processing Introduzindo o conceito de Fo crítico no processamento térmico em batelada

    Directory of Open Access Journals (Sweden)

    Homero Ferracini Gumerato

    2009-12-01

    Full Text Available The determination of the sterilization value for low acid foods in retorts includes a critical evaluation of the factory's facilities and utilities, validation of the heat processing equipment (by heat distribution assays, and finally heat penetration assays with the product. The intensity of the heat process applied to the food can be expressed by the Fo value (sterilization value, in minutes, at a reference temperature of 121.1 °C, and a thermal index, z, of 10 °C, for Clostridium botulinum spores. For safety reasons, the lowest value for Fo is frequently adopted, being obtained in heat penetration assays as indicative of the minimum process intensity applied. This lowest Fo value should always be higher than the minimum Fo recommended for the food in question. However, the use of the Fo value for the coldest can fail to statistically explain all the practical occurrences in food heat treatment processes. Thus, as a result of intense experimental work, we aimed to develop a new focus to determine the lowest Fo value, which we renamed the critical Fo. The critical Fo is based on a statistical model for the interpretation of the results of heat penetration assays in packages, and it depends not only on the Fo values found at the coldest point of the package and the coldest point of the equipment, but also on the size of the batch of packages processed in the retort, the total processing time in the retort, and the time between CIPs of the retort. In the present study, we tried to explore the results of physical measurements used in the validation of food heat processes. Three examples of calculations were prepared to illustrate the methodology developed and to introduce the concept of critical Fo for the processing of canned food.A determinação do valor de esterilização de alimentos de baixa acidez em autoclaves compreende uma minuciosa avaliação das instalações e utilidades da fábrica, uma validação do equipamento de processo t

  3. An application of liquid sublayer dryout mechanism to the prediction of critical heat flux under low pressure and low velocity conditions in round tubes

    International Nuclear Information System (INIS)

    Lee, Kwang-Won; Yang, Jae-Young; Baik, Se-Jin

    1997-01-01

    Based on several experimental evidences for nucleate boiling in annular film and the existence of residual liquid film flow rate at the critical heat flux (CHF) location, the liquid sublayer dryout (LSD) mechanism under annular film is firstly introduced to evaluate the CHF data at low pressure and low velocity (LPLV) conditions, which would not be predicted by a normal annular film dryout (AFD) model. In this study, the CHF occurrence due to annular film separation or breaking down is phenomenologically modelled by applying the LSD mechanism to this situation. In this LSD mechanism, the liquid sublayer thickness, the incoming liquid velocity to the liquid sublayer, and the axial distance from the onset of annular flow to the CHF location are used as the phenomena-controlling parameters. From the model validation on the 1406 CHF data points ranging over P = 0.1 - 2 MPa, G = 4 - 499 kg/m 2 s, L/D = 4 - 402, most of CHF data (more than 1000 points) are predicted within ±30% error bounds by the LSD mechanism. However, some calculation results that critical qualities are less than 0.4 are considerably overestimated by this mechanism. These overpredictions seem to be caused by inadequate CHF mechanism classification criteria and an insufficient consideration of the flow instability effect on CHF. Further studies for a new classification criterion screening the CHF data affected by flow instabilities and a new bubble detachment model for LPLV conditions are needed to improve the model accuracy. (author)

  4. Critical behavior of binary mixture of {(1 − x) C6H5CN + x CH3(CH2)9CH3}: Measurements of coexistence curves, turbidity, and heat capacity

    International Nuclear Information System (INIS)

    Yin Tianxiang; Lei Yuntao; Mao Chunfeng; Chen Zhiyun; An Xueqin; Shen Weiguo

    2012-01-01

    Highlights: ► Coexistence curve, isobaric heat capacity and turbidity measurements have been reported. ► Asymmetry of the coexistence curves has been analyzed by the complete scaling theory. ► Heat capacity has been shown to be important in describing the asymmetric criticality. ► Universal amplitude ratios have been tested. - Abstract: (Liquid + liquid) coexistence curve, turbidity, and isobaric heat capacity per unit volume for the critical solution of {benzonitrile + n-undecane} have been measured. The critical exponents β, ν, γ, and α have been deduced, which were found to be consistent with the theoretic predictions. Meanwhile, the experimental data have also been analyzed to obtain the system-dependent critical amplitudes B, ξ 0 , χ 0 , A ± , and D corresponding to the difference of the general density variable of two coexisting phases Δρ, the correlation length ξ, the osmotic compressibility χ, the isobaric heat capacity per unit volume C p V −1 , and the first term of correction-to-scaling for the isobaric heat capacity per unit volume, which were used to test some universal ratios. It was found that the coexistence curve may be well described by the crossover model proposed by Gutkowski et al. The critical-fluctuation induced contribution to the background heat capacity B cr was obtained and used to analyze the asymmetric behavior of the diameter of the coexistence curve. The result indicated that the asymmetry of the coexistence curve can be well described by the complete scaling theory proposed by Anisimov et al., and the heat capacity does make a significant contribution to this asymmetric behavior.

  5. Critical behaviour of binary mixture of {xC6H5CN + (1 - x)CH3(CH2)7CH3}: Measurements of coexistence curves, light scattering, and heat capacity

    International Nuclear Information System (INIS)

    Lei Yuntao; Chen Zhiyun; Wang Nong; Mao Chunfeng; An Xueqin; Shen Weiguo

    2010-01-01

    Liquid + liquid coexistence, light scattering, and isobaric heat capacity per unit volume for the critical solutions of (benzonitrile + n-nonane) have been measured. The critical exponents relating to the coexistence curve β, the osmotic compressibility γ, the correlation length ν, and the heat capacity α have been deduced and the values are consistent with the 3D-Ising values in the range close to the critical point. The experimental results of the liquid + liquid coexistence were analyzed to examine the Wegner correction terms and the behaviour of the diameter of the coexistence curves. The light scattering data were well described by the crossover model proposed by Anisimov and Sengers, and showed a tendency of monotonic crossover of the critical exponents γ and ν from the 3D-Ising values to the mean-field values as the temperature departures from the critical point. From calorimetric measurements, the amplitude A ± and the critical background B cr of the heat capacity in the critical region have been deduced and some universal ratios are tested.

  6. Numerical investigations on the effect of the axial interval between intensifying spacer grids on the critical heat flux value for fuel assemblies with non-uniform axial power distribution

    International Nuclear Information System (INIS)

    Kireeva, D.; Oleksyuk, D.

    2015-01-01

    In this paper a number of numerical studies on intensifying heat exchange conducted by NRC 'Kurchatov Institute' are presented. A standardised heat exchange intensifying spacer grid (UDRI) can be installed at any height along the fuel assembly (FA) heat-generating section. When installed at the bottom of a fuel assembly, the UDRI facilitates intensive coolant mixing; the UDRI mounted at the top of a FA provides better mixing and the enhancement in heat exchange. The application of the heat exchange intensifying spacer grids results in better flattening of the coolant parameters along the cross-section and higher critical heat flux ratio. The investigations were carried out by means of numerical code SC-INT using mesh generation that have been specially designed by NRC 'Kurchatov Institute' to perform calculations for fuel assemblies equipped with the intensifying spacer grids. The effect of the axial interval between UDRI grids on the critical heat flux value for two typical axial power shapes has been investigated. The derived optimal solutions for the positioning of intensifying grids are also presented

  7. An utilization of liquid sublayer dryout mechanism in predicting critical heat flux under low pressure and low velocity conditions in round tubes

    International Nuclear Information System (INIS)

    Lee, Kwang-Won; Baik, Se-Jin; Ro, Tae-Sun

    2000-01-01

    From a theoretical assessment of extensive critical heat flux (CHF) data under low pressure and low velocity (LPLV) conditions, it was found out that lots of CHF data would not be well predicted by a normal annular film dryout (AFD) mechanism, although their flow patterns were identified as annular-mist flow. To predict these CHF data, a liquid sublayer dryout (LSD) mechanism has been newly utilized in developing the mechanistic CHF model based on each identified CHF mechanism. This mechanism postulates that the CHF occurrence is caused by dryout of the thin liquid sublayer resulting from the annular film separation or breaking down due to nucleate boiling in annular film or hydrodynamic fluctuation. In principle, this mechanism well supports the experimental evidence of residual film flow rate at the CHF location, which can not be explained by the AFD mechanism. For a comparative assessment of each mechanism, the CHF model based on the LSD mechanism is developed together with that based on the AFD mechanism. The validation of these models is performed on the 1406 CHF data points ranging over P=0.1-2 MPa, G=4-499 kg m -2 s -1 , L/D=4-402. This model validation shows that 1055 and 231 CHF data are predicted within ±30 error bound by the LSD mechanism and the AFD mechanism, respectively. However, some CHF data whose critical qualities are <0.4 or whose tube length-to-diameter ratios are <70 are considerably overestimated by the CHF model based on the LSD mechanism. These overestimations seem to be caused by an inadequate CHF mechanism classification and an insufficient consideration of the flow instability effect on CHF. Further studies for a new classification criterion screening the CHF data affected by flow instabilities as well as a new bubble detachment model for LPLV conditions, are needed to improve the model accuracy.

  8. High-energy components of 'designer gasoline and designer diesel fuel' I. Heat capacities, enthalpy increments, vapor pressures, critical properties, and derived thermodynamic functions for bicyclopentyl between the T=(10 and 600) K

    International Nuclear Information System (INIS)

    Chirico, R.D.; Steele, W.V.

    2004-01-01

    Measurements leading to the calculation of the standard thermodynamic properties for gaseous bicyclopentyl (Chemicals Abstracts registry number [1636-39-1]) are reported. Experimental methods include adiabatic heat-capacity calorimetry, comparative ebulliometry, and differential-scanning calorimetry (d.s.c.). The critical temperature was determined by d.s.c. and the critical pressure and critical density were estimated. Standard molar entropies, standard molar enthalpies, and standard molar Gibbs free energies of formation are reported at selected temperatures between T=(298.15 and 600) K. Formation properties were calculated with a literature value for the enthalpy of combustion in the liquid phase. All results are compared with available literature values

  9. The effect of the advanced drift-flux model of ASSERT-PV on critical heat flux, flow and void distributions in CANDU bundle subchannels

    International Nuclear Information System (INIS)

    Hammouda, N.; Rao, Y.F.

    2017-01-01

    Highlights: • Presentation of the “advanced” drift-flux model of the subchannel code ASSERT-PV. • Study the effect of the drift-flux model of ASSERT on CHF and flow distribution. • Quantify model component effects with flow, quality and dryout power measurements. - Abstract: This paper studies the effect of the drift flux model of the subchannel code ASSERT-PV on critical heat flux (CHF), void fraction and flow distribution across fuel bundles. Numerical experiments and comparison against measurements were performed to examine the trends and relative behaviour of the different components of the model under various flow conditions. The drift flux model of ASSERT-PV is composed of three components: (a) the lateral component or diversion cross-flow, caused by pressure difference between connected subchannels, (b) the turbulent diffusion component or the turbulent mixing through gaps of subchannels, caused by instantaneous turbulent fluctuations or flow oscillations, and (c) the void drift component that occurs due to the two-phase tendency toward a preferred distribution. This study shows that the drift flux model has a significant impact on CHF, void fraction and flow distribution predictions. The lateral component of the drift flux model has a stronger effect on CHF predictions than the axial component, especially for horizontal flow. Predictions of CHF, void fraction and flow distributions are most sensitive to the turbulent diffusion component of the model, followed by the void drift component. Buoyancy drift can be significant, but it does not have as much influence on CHF and flow distribution as the turbulent diffusion and void drift.

  10. Numerical prediction of critical heat flux in nuclear fuel rod bundles with advanced three-fluid multidimensional porous media based model

    International Nuclear Information System (INIS)

    Zoran Stosic; Vladimir Stevanovic

    2005-01-01

    Full text of publication follows: The modern design of nuclear fuel rod bundles for Boiling Water Reactors (BWRs) is characterised with increased number of rods in the bundle, introduced part-length fuel rods and a water channel positioned along the bundle asymmetrically in regard to the centre of the bundle cross section. Such design causes significant spatial differences of volumetric heat flux, steam void fraction distribution, mass flux rate and other thermal-hydraulic parameters important for efficient cooling of nuclear fuel rods during normal steady-state and transient conditions. The prediction of the Critical Heat Flux (CHF) under these complex thermal-hydraulic conditions is of the prime importance for the safe and economic BWR operation. An efficient numerical method for the CHF prediction is developed based on the porous medium concept and multi-fluid two-phase flow models. Fuel rod bundle is observed as a porous medium with a two-phase flow through it. Coolant flow from the bundle entrance to the exit is characterised with the subsequent change of one-phase and several two-phase flow patterns. One fluid (one-phase) model is used for the prediction of liquid heating up in the bundle entrance region. Two-fluid modelling approach is applied to the bubbly and churn-turbulent vapour and liquid flows. Three-fluid modelling approach is applied to the annular flow pattern: liquid film on the rods wall, steam flow and droplets entrained in the steam stream. Every fluid stream in applied multi-fluid models is described with the mass, momentum and energy balance equations. Closure laws for the prediction of interfacial transfer processes are stated with the special emphasis on the prediction of the steam-water interface drag force, through the interface drag coefficient, and droplets entrainment and deposition rates for three-fluid annular flow model. The model implies non-equilibrium thermal and flow conditions. A new mechanistic approach for the CHF prediction

  11. Plasma heating

    International Nuclear Information System (INIS)

    Wilhelm, R.

    1989-01-01

    Successful plasma heating is essential in present fusion experiments, for the demonstration of DpT burn in future devices and finally for the fusion reactor itself. This paper discusses the common heating systems with respect to their present performance and their applicability to future fusion devices. The comparative discussion is oriented to the various function of heating, which are: - plasma heating to fusion-relevant parameters and to ignition in future machines, -non-inductive, steady-pstate current drive, - plasma profile control, -neutral gas breakdown and plasma build-up. In view of these different functions, the potential of neutral beam injection (NBI) and the various schemes of wave heating (ECRH, LH, ICRH and Alven wave heating) is analyzed in more detail. The analysis includes assessments of the present physical and technical state of these heating methods, and makes suggestions for future developments and about outstanding problems. Specific attention is given to the still critical problem of efficient current drive, especially with respect to further extrapolation towards an economically operating tokamak reactor. Remarks on issues such as reliability, maintenance and economy conclude this comparative overview on plasma heating systems. (author). 43 refs.; 13 figs.; 3 tabs

  12. Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry; Badania teoretyczne i eksperymentalne kryzysu wrzenia w warunkach wrzenia przechlodzonego w przeplywie w kanale pionowym

    Energy Technology Data Exchange (ETDEWEB)

    Staron, E [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1997-12-31

    Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs.

  13. Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry; Badania teoretyczne i eksperymentalne kryzysu wrzenia w warunkach wrzenia przechlodzonego w przeplywie w kanale pionowym

    Energy Technology Data Exchange (ETDEWEB)

    Staron, E. [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1996-12-31

    Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs.

  14. Structure, resistivity, critical field, specific-heat jump at Tc, Meissner effect, a.c. and d.c. Susceptibility of the high-temperature superconductor La2-xSrxCuO4

    International Nuclear Information System (INIS)

    Decroux, M.; Junod, A.; Bezinge, A.

    1987-01-01

    The temperature dependence of the resistivity, the magnetic properties and the specific heat were investigated on sintered samples of La 1.85 Sr 0.15 CuO 4 having zero resistance below 35 K. The crystal structure at 300K (tetragonal K 2 NiF 4 -type) was refined from X-ray powder diffraction data. The d.c. susceptibility shows no indication for the existence of localized Cu 2+ moments. The observation of a 60% Meissner effect and a smeared jump at T c in the specific-heat curve prove the intrinsic character of this superconducting state. The amplitude of this jump is compatible with the DOS estimated from the Pauli susceptibility. With a critical magnetic field slope dH c2 /dT| Tc = - 2.5 T/K, the orbital critical field is expected to be of the order of 64 T

  15. The Effects of Channel Curvature and Protrusion Height on Nucleate Boiling and the Critical Heat Flux of a Simulated Electronic Chip

    Science.gov (United States)

    1994-05-01

    parameters and geometry factor. 57 3.2 Laminar sublayer and buffer layer thicknesses for geometry of Mudawar and Maddox.ŝ 68 3.3 Correlation constants...transfer from simulated electronic chip heat sources that are flush with the flow channel wall. Mudawar and Maddox2" have studied enhanced surfaces...bias error was not estimated; however, the percentage of heat loss measured compares with that previously reported by Mudawar and Maddox19 for a

  16. Transient heat transfer characteristics of liquid helium

    International Nuclear Information System (INIS)

    Tsukamoto, Osami

    1976-01-01

    The transient heat transfer characteristics of liquid helium are investigated. The critical burnout heat fluxes for pulsive heating are measured, and empirical relations between the critical burnout heat flux and the length of the heat pulse are given. The burnout is detected by observing the super-to-normal transition of the temperature sensor which is a thin lead film prepared on the heated surface by vacuum evaporation. The mechanism of boiling heat transfer for pulsive heating is discussed, and theoretical relations between the critical burnout heat flux and the length of the heat pulse are derived. The empirical data satisfy the theoretical relations fairly well. (auth.)

  17. Effect of variable thermal conductivity and specific heat capacity on the calculation of the critical metal hydride thickness for Ti1.1CrMn

    DEFF Research Database (Denmark)

    Mazzucco, Andrea; Rokni, Masoud

    2014-01-01

    model is applied to the metal hydride system, with Ti 1.1 CrMn as the absorbing alloy, to predict the weight fraction of absorbed hydrogen and solid bed temperat ure . Dependencies of thermal conductivity and specific heat capacity upon pressure and hydrogen content respectively , are accounted for...

  18. Magnetic fusion energy plasma interactive and high heat flux components. Volume I. Technical assessment of the critical issues and problem areas in the plasma materials interaction field

    International Nuclear Information System (INIS)

    Conn, R.W.; Gauster, W.B.; Heifetz, D.; Marmar, E.; Wilson, K.L.

    1984-01-01

    A technical assessment of the critical issues and problem areas in the field of plasma materials interactions (PMI) in magnetic fusion devices shows these problems to be central for near-term experiments, for intermediate-range reactor devices including D-T burning physics experiments, and for long-term reactor machines. Critical technical issues are ones central to understanding and successful operation of existing and near-term experiments/reactors or devices of great importance for the long run, i.e., ones which will require an extensive, long-term development effort and thus should receive attention now. Four subgroups were formed to assess the critical PMI issues along four major lines: (1) PMI and plasma confinement physics experiments; (2) plasma-edge modelling and theory; (3) surface physics; and (4) materials technology for in-vessel components and the first wall. The report which follows is divided into four major sections, one for each of these topics

  19. A critical evaluation of the upper ocean heat budget in the Climate Forecast System Reanalysis data for the south central equatorial Pacific

    Energy Technology Data Exchange (ETDEWEB)

    Liu Hailong; Liu Xiangcui [State Key Laboratory of Atmospheric Sciences and Geophysical Fluid Dynamics, Institute of Atmospheric Physics, Chinese Academy of Sciences, Beijing (China); Zhang Minghua [Institute for Terrestrial and Planetary Atmospheres, Stony Brook University, State University of New York, Stony Brook, NY (United States); Lin Wuyin, E-mail: lhl@lasg.iap.ac.cn [Atmospheric Sciences Division, Brookhaven National Laboratory, Upton, NY (United States)

    2011-07-15

    Coupled ocean-atmospheric models suffer from the common bias of a spurious rain belt south of the central equatorial Pacific throughout the year. Observational constraints on key processes responsible for this bias are scarce. The recently available reanalysis from a coupled model system for the National Centers for Environmental Prediction (NCEP) Climate Forecast System Reanalysis (CFSR) data is a potential benchmark for climate models in this region. Its suitability for model evaluation and validation, however, needs to be established. This paper examines the mixed layer heat budget and the ocean surface currents-key factors for the sea surface temperature control in the double Inter-Tropical Convergence Zone in the central Pacific-from 5 deg. S to 10 deg. S and 170 deg. E to 150 deg. W. Two independent approaches are used. The first approach is through comparison of CFSR data with collocated station observations from field experiments; the second is through the residual analysis of the heat budget of the mixed layer. We show that the CFSR overestimates the net surface flux in this region by 23 W m{sup -2}. The overestimated net surface flux is mainly due to an even larger overestimation of shortwave radiation by 44 W m{sup -2}, which is compensated by a surface latent heat flux overestimated by 14 W m{sup -2}. However, the quality of surface currents and the associated oceanic heat transport in CFSR are not compromised by the surface flux biases, and they agree with the best available estimates. The uncertainties of the observational data from field experiments are also briefly discussed in the present study.

  20. Critical mass flux for flaming ignition of dead, dry wood as a function of external radiant heat flux and oxidizer flow velocity

    Science.gov (United States)

    Sara McAllister; Mark Finney; Jack Cohen

    2010-01-01

    Extreme weather often contributes to crown fires, where the fire spreads from one tree crown to the next as a series of piloted ignitions. An important aspect in predicting crown fires is understanding the ignition of fuel particles. The ignition criterion considered in this work is the critical mass flux criterion – that a sufficient amount of pyrolysis gases must be...

  1. Cyro Power and Heat Transfer

    National Research Council Canada - National Science Library

    Chow, L

    1998-01-01

    .... The heat generated from a 9x9-heater array was removed by liquid nitrogen pool boiling. The orientation and space limitation of the array were varied to explore their effects on the critical heat flux (CHF) value...

  2. CriticalEd

    DEFF Research Database (Denmark)

    Kjellberg, Caspar Mølholt; Meredith, David

    2014-01-01

    . Since the comments are not input sequentially, with regard to position, but in arbitrary order, this list must be sorted by copy/pasting the rows into place—an error-prone and time-consuming process. Scholars who produce critical editions typically use off-the-shelf music notation software......The best text method is commonly applied among music scholars engaged in producing critical editions. In this method, a comment list is compiled, consisting of variant readings and editorial emendations. This list is maintained by inserting the comments into a document as the changes are made......, consisting of a Sibelius plug-in, a cross-platform application, called CriticalEd, and a REST-based solution, which handles data storage/retrieval. A prototype has been tested at the Danish Centre for Music Publication, and the results suggest that the system could greatly improve the efficiency...

  3. Critical sizes and critical characteristics of nanoclusters, nanostructures and nanomaterials

    International Nuclear Information System (INIS)

    Suzdalev, I.P.

    2005-01-01

    Full text: Critical sizes and characteristics of nanoclusters and nanostructures are introduced as the parameters of nanosystems and nanomaterials. The next critical characteristics are considered: atomic and electronic 'magic number', critical size of cluster nucleation, critical size of melting-freezing of cluster, critical size of quantum (laser) radiation, critical sizes for the single electron conductivity, critical energy and magnetic field for the magnetic tunneling, critical cluster sizes for the giant magnetic resistance, critical size of the first order magnetic phase transition. The critical characteristics are estimated by thermodynamic approaches, by Moessbauer spectroscopy, AFM, heat capacity, SQUID magnetometry and other technique, The influence of cluster-cluster interactions, cluster-matrix interactions and cluster defects on cluster atomic dynamics, cluster melting, cluster critical sizes, Curie or Neel points and the character of magnetic phase transitions were investigated. The applications of critical size and critical characteristic parameters for the nanomaterial characterization are considered

  4. Investigation of inter-grain critical current density in Bi2Sr2CaCu2O8+δ superconducting wires and its relationship with the heat treatment protocol

    Science.gov (United States)

    Pallecchi, I.; Leveratto, A.; Braccini, V.; Zunino, V.; Malagoli, A.

    2017-09-01

    In this work we investigate the effect of each different heat treatment stage in the fabrication of Bi2Sr2CaCu2O8+δ superconducting wires on intra-grain and inter-grain superconducting properties. We measure magnetic critical temperature T c values and transport critical current density J c at temperatures from 4 K to 40 K and in fields up to 7 T. From an analysis of the temperature dependence of the self-field critical current density J c(T) that takes into account weak link behavior and the proximity effect, we study grain boundary (GB) transparency to supercurrents; we also establish a relationship between GB oxygenation in the different steps of the fabrication process and GB transparency to supercurrents. We find that GB oxygenation starts in the first crystallization stage, but it becomes complete in the plateau at 836 °C and in slow cooling stages and is further enhanced in the prolonged post-annealing step. Such oxygenation makes GBs more conductive, thus improving the inter-grain J c value and temperature dependence. On the other hand, from inspection of the T c values in the framework of the phase diagram dome, we find that grains are already oxygenated in the crystallization step up to the optimal doping, while successive slow cooling and post-annealing treatments further enhance the degree of overdoping, especially if carried out in oxygen atmosphere rather than in air.

  5. Heat transfer II essentials

    CERN Document Server

    REA, The Editors of

    1988-01-01

    REA's Essentials provide quick and easy access to critical information in a variety of different fields, ranging from the most basic to the most advanced. As its name implies, these concise, comprehensive study guides summarize the essentials of the field covered. Essentials are helpful when preparing for exams, doing homework and will remain a lasting reference source for students, teachers, and professionals. Heat Transfer II reviews correlations for forced convection, free convection, heat exchangers, radiation heat transfer, and boiling and condensation.

  6. Heat pumps: heat recovery

    Energy Technology Data Exchange (ETDEWEB)

    Pielke, R

    1976-01-01

    The author firstly explains in a general manner the functioning of the heat pump. Following a brief look at the future heat demand and the possibilities of covering it, the various methods of obtaining energy (making use of solar energy, ground heat, and others) and the practical applications (office heating, swimming pool heating etc.) are explained. The author still sees considerable difficulties in using the heat pump at present on a large scale. Firstly there is not enough maintenance personnel available, secondly the electricity supply undertakings cannot provide the necessary electricity on a wide basis without considerable investments. Other possibilities to save energy or to use waste energy are at present easier and more economical to realize. Recuperative and regenerative systems are described.

  7. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems.

  8. Magnetic Fusion Energy Plasma Interactive and High Heat Flux Components: Volume 5, Technical assessment of critical issues in the steady state operation of fusion confinement devices

    International Nuclear Information System (INIS)

    1988-01-01

    Critical issues for the steady state operation of plasma confinement devices exist in both the physics and technology fields of fusion research. Due to the wide range and number of these issues, this technical assessment has focused on the crucial issues associated with the plasma physics and the plasma interactive components. The document provides information on the problem areas that affect the design and operation of a steady state ETR or ITER type confinement device. It discusses both tokamaks and alternative concepts, and provides a survey of existing and planned confinement machines and laboratory facilities that can address the identified issues. A universal definition of steady state operation is difficult to obtain. From a physics point of view, steady state is generally achieved when the time derivatives approach zero and the operation time greatly exceeds the characteristic time constants of the device. Steady state operation for materials depends on whether thermal stress, creep, fatigue, radiation damage, or power removal are being discussed. For erosion issues, the fluence and availability of the machine for continuous operation are important, assuming that transient events such as disruptions do not limit the component lifetimes. The panel suggests, in general terms, that steady state requires plasma operation from 100 to 1000 seconds and an availability of more than a few percent, which is similar to the expectations for an ETR type device. The assessment of critical issues for steady state operation is divided into four sections: physics issues; technology issues; issues in alternative concepts; and devices and laboratory facilities that can address these problems

  9. Structure, resistivity, critical field, specific-heat jump at Tc, Meissner effect, a.c. and d.c. susceptibility of the high-Tc superconductor YBa2Cu3O7

    International Nuclear Information System (INIS)

    Junod, A.; Bezinge, A.; Graf, T.

    1987-01-01

    YBa 2 Cu 3 O 7 superconductors with inductive transitions as narrows as 0.45 K above 90 K were synthetized. Samples were characterized by thermogravimetry, differential thermal analysis, X-ray and neutron diffraction. The structure is characterized by a two-dimensional Cu-O network with square-pyramidal and square-planar coordinated Cu atoms. Results show a clear metallic behaviour of the resistivity. An orbital critical field as high as 300 T is extrapolated. Meissner flux expulsion up to 40% is observed. Small amounts of magnetic Cu 2+ ions are correlated with the presence of the impurity phase BaCuO 2 . The Pauli susceptibility and the specific-heat jump at T c are consistent with γ ≅ 2mJ/(K 2 gat) (9mJ/(K 2 mole-Cu)), neglecting all renormalizations

  10. Energy versus economic effectiveness in CHP (combined heat and power) applications: Investigation on the critical role of commodities price, taxation and power grid mix efficiency

    International Nuclear Information System (INIS)

    Comodi, Gabriele; Rossi, Mosè

    2016-01-01

    Starting from PES (primary energy saving) and CSR (cost saving ratio) definitions the work pinpoints a “grey area” in which CHP (combined heat and power – cogeneration) units can operate with profit and negative PES. In this case, CHP can be profitably operated with lower efficiency with respect to separate production of electrical and thermal energy. The work defines the R-index as the ratio between the cost of fuel and electricity. The optimal value of R-index for which CHP units operate with both environmental benefit (PES > 0) and economic profitability (CSR > 0) is the reference value of electrical efficiency, η_e_l_-_r_e_f, of separate production (national power grid mix). As a consequence, optimal R-index varies from Country to Country. The work demonstrates that the value of R corresponds to the minimum value of electrical efficiency for which any power generator operates with profit. The paper demonstrates that, with regard to the profitability of cogeneration, the ratio between the cost of commodities is more important than their absolute value so that different taxation of each commodity can be a good leverage for energy policy makers to promote high efficiency cogeneration, even in the absence of an incentive mechanism. The final part of the study presents an analysis on micro-CHP technologies payback times for different European Countries. - Highlights: • Investigation of the grey area where CHP profitably operates also with negative PES. • Study starts from definition of primary energy saving PES and cost saving ratio CSR. • Definition of the R-index as the ratio between the cost of fuel and electricity. • The optimal value of R for which the “grey area” disappears is R = η_e_l_-_r_e_f. • R is also the value of η_e_l for which any electric generator profitably operates.

  11. Thermo-fluid-dynamics of natural convection around a heated vertical plate with a critical assessment of the standard similarity theory

    Science.gov (United States)

    Guha, Abhijit; Nayek, Subhajit

    2017-10-01

    A compulsory element of all textbooks on natural convection has been a detailed similarity analysis for laminar natural convection on a heated semi-infinite vertical plate and a routinely used boundary condition for such analysis is u = 0 at x = 0. The same boundary condition continues to be assumed in related theoretical analyses, even in recent publications. The present work examines the consequence of this long-held assumption, which appears to have never been questioned in the literature, on the fluid dynamics and heat transfer characteristics. The assessment has been made here by solving the Navier-Stokes equations numerically with two boundary conditions—one with constrained velocity at x = 0 to mimic the similarity analysis and the other with no such constraints simulating the case of a heated vertical plate in an infinite expanse of the quiescent fluid medium. It is found that the fluid flow field given by the similarity theory is drastically different from that given by the computational fluid dynamics (CFD) simulations with unconstrained velocity. This also reflects on the Nusselt number, the prediction of the CFD simulations with unconstrained velocity being quite close to the experimentally measured values at all Grashof and Prandtl numbers (this is the first time theoretically computed values of the average Nusselt number N u ¯ are found to be so close to the experimental values). The difference of the Nusselt number (Δ N u ¯ ) predicted by the similarity theory and that by the CFD simulations (as well as the measured values), both computed with a high degree of precision, can be very significant, particularly at low Grashof numbers and at Prandtl numbers far removed from unity. Computations show that within the range of investigations (104 ≤ GrL ≤ 108, 0.01 ≤ Pr ≤ 100), the maximum value of Δ N u ¯ may be of the order 50%. Thus, for quantitative predictions, the available theory (i.e., similarity analysis) can be rather inadequate. With

  12. FTR europia gamma heating

    International Nuclear Information System (INIS)

    Ward, J.T. Jr.

    1975-01-01

    Calculated and experimental gamma heating rates of europia in the Engineering Mockup Critical Assembly (EMC) were correlated. A calculated to experimental (C/E) ratio of 1.086 was established in validating the theoretical approach and computational technique applied in the calculations. Gamma heat deposition rates in the FTR with Eu 2 O 3 control absorbers were determined from three-dimensional calculations. Maximum gamma heating was found to occur near the tip of a half-inserted row 5 control rod assembly--12.8 watts/gm of europia. Gamma heating profiles were established for a single half-inserted europia absorber assembly. Local heat peaking was found not to alter significantly heating rates computed in the FTR core model, where larger mesh interval sizes precluded examination of spatially-limited heating gradients. These computations provide the basis for thermal-hydraulic analyses to ascertain temperature profiles in the FTR under europia control

  13. Heat pipe heat storage performance

    Energy Technology Data Exchange (ETDEWEB)

    Caruso, A; Pasquetti, R [Univ. de Provence, Marseille (FR). Inst. Universitaire des Systemes; Grakovich, L P; Vasiliev, L L [A.V. Luikov Heat and Mass Transfer Inst. of the BSSR, Academy of Sciences, Minsk (BY)

    1989-01-01

    Heat storage offers essential thermal energy saving for heating. A ground heat store equipped with heat pipes connecting it with a heat source and to the user is considered in this paper. It has been shown that such a heat exchanging system along with a batch energy source meets, to a considerable extent, house heating requirements. (author).

  14. Heat transfer

    Indian Academy of Sciences (India)

    First page Back Continue Last page Overview Graphics. Heat transfer. Heat conduction in solid slab. Convective heat transfer. Non-linear temperature. variation due to flow. HEAT FLUX AT SURFACE. conduction/diffusion.

  15. Heat Waves

    Science.gov (United States)

    Heat Waves Dangers we face during periods of very high temperatures include: Heat cramps: These are muscular pains and spasms due ... that the body is having trouble with the heat. If a heat wave is predicted or happening… - ...

  16. Heat Islands

    Science.gov (United States)

    EPA's Heat Island Effect Site provides information on heat islands, their impacts, mitigation strategies, related research, a directory of heat island reduction initiatives in U.S. communities, and EPA's Heat Island Reduction Program.

  17. Magnetic heating in the sun

    International Nuclear Information System (INIS)

    Chiuderi, C.

    1981-01-01

    The observational evidence for magnetic heating in the solar corona is presented. The possible ways of investigating theoretically the nature of the heating processes are critically discussed. Merits and disadvantages of the basic mechanisms so far proposed are reviewed. Finally, a preliminary application of the magnetic heating concept to stellar coronae is presented. (orig.)

  18. Heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, E L; Eisenmann, G; Hahne, E [Stuttgart Univ. (TH) (F.R. Germany). Inst. fuer Thermodynamik und Waermetechnik

    1976-04-01

    A survey is presented on publications on design, heat transfer, form factors, free convection, evaporation processes, cooling towers, condensation, annular gap, cross-flowed cylinders, axial flow through a bundle of tubes, roughnesses, convective heat transfer, loss of pressure, radiative heat transfer, finned surfaces, spiral heat exchangers, curved pipes, regeneraters, heat pipes, heat carriers, scaling, heat recovery systems, materials selection, strength calculation, control, instabilities, automation of circuits, operational problems and optimization.

  19. The critical ionization velocity

    International Nuclear Information System (INIS)

    Raadu, M.A.

    1980-06-01

    The critical ionization velocity effect was first proposed in the context of space plasmas. This effect occurs for a neutral gas moving through a magnetized plasma and leads to rapid ionization and braking of the relative motion when a marginal velocity, 'the critical velocity', is exceeded. Laboratory experiments have clearly established the significance of the critical velocity and have provided evidence for an underlying mechanism which relies on the combined action of electron impact ionization and a collective plasma interaction heating electrons. There is experimental support for such a mechanism based on the heating of electrons by the modified two-stream instability as part of a feedback process. Several applications to space plasmas have been proposed and the possibility of space experiments has been discussed. (author)

  20. Thinking Critically about Critical Thinking

    Science.gov (United States)

    Mulnix, Jennifer Wilson

    2012-01-01

    As a philosophy professor, one of my central goals is to teach students to think critically. However, one difficulty with determining whether critical thinking can be taught, or even measured, is that there is widespread disagreement over what critical thinking actually is. Here, I reflect on several conceptions of critical thinking, subjecting…

  1. Heat Stress

    Science.gov (United States)

    ... Publications and Products Programs Contact NIOSH NIOSH HEAT STRESS Recommend on Facebook Tweet Share Compartir OSHA-NIOSH ... hot environments may be at risk of heat stress. Exposure to extreme heat can result in occupational ...

  2. Effect of Al content on critical CTOD properties in heat affected zone of C-Mn microalloyed steel. Teitanso teigokin koyosetsu netsu eikyobu no genkai CTOD tokusei ni oyobosu Al ganyuryo no eikyo

    Energy Technology Data Exchange (ETDEWEB)

    Fukada, Y.; Komizo, Y. (Sumitomo Metal Industries Ltd., Osaka (Japan))

    1992-08-05

    Two types of molten alloys specimen with a base of 0.10%C-0.20%Si-1.40%Mn-0.01%Ti system and varied Al content, were studied. The critical crack tip opening displacement(CTOD) properties in heat effected zone(HAZ) of extreme low Al content steel was stable at. extremely low temperature and there was no formation of M-A. Fine ferrite has a texture of [alpha] main body and it has been thought that the the change in the CTOD properties with the variation in Al content has been due to the difference in the texture. In case of Al content steel plate, the interfacial energy has been decreased due to excessive carbon concentration at [gamma] /[alpha] interface, M-A formation has been easier by the suppression of [alpha] transformation. In case of extremely low Al content steel plate, [alpha] transformation has been promoted and cementite deposition has been estimated from a small amount. of left [gamma] of extremely high carbon concentration. As for SH-CCT diagram of extremely low Al content steel plate, compare to Al content steel plate, [alpha] noze has shifted toward shorter time and formation of [alpha] has been easier within the normal welding cooling rate, and microstructures of [alpha] texture have formed in HAZ. 21 refs., 12 figs., 1 tab.

  3. Experimental study of heat exchange coefficients, critical heat flux and charge losses, using water-steam mixtures in turbulent flow in a vertical tube; Etude experimentale des coefficients d'echanges thermiques, des flux de chaleur critiques et des pertes de charge avec des melanges eau-vapeur en ecoulement turbulent dans un tube vertical

    Energy Technology Data Exchange (ETDEWEB)

    Perroud, P; De La Harpe, A; Rebiere, J [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1960-12-15

    Two stainless steel tubes were used (with diameters of 5 and 10 mm, lengths 400 and 600 mm respectively), heated electrically (50 Hz). The mixture flows from top to bottom. The work was carried out mainly on mixtures of high concentration (x > 0.1), at pressures between 50 and 60 kg/cm{sup 2}, flowing as a liquid film on the walls of the tube with droplets suspended in the central current of steam. By analysis of the heat transfer laws the exchange mechanisms were established, and the conditions under which the critical heat flux may be exceeded without danger of actual burnout were determined. In this way high output concentrations (x{sub s} > 0.9) may be obtained. An attempt has been made to find out to what extent existing correlation formulae can be used to account for the phenomena observed. It is shown that those dealing with exchange coefficients can only be applied in a first approximation in cases where exchange by convection is preponderant, and only below the critical flux. The formulae proposed by WAPD and CISE do not give a satisfactory estimation of the critical heat flux, and the essential reasons for this inadequacy are explained. Lastly, the Martinelli and Nelson method may be used to an approximation of 30 per cent for the calculation of charge losses. (author) [French] On a utilise deux tubes en acier inox (avec des diametres de 5 et 10 mm, et des longueurs respectives 400 et 600 mm) chauffes electriquement (50 Hz). Le melange s'ecoule de haut en bas. Les etudes ont porte plus specialement sur des melanges de titres eleves (x > 0,1) a des pressions comprises entre 60 et 90 kg/cm{sup 2} dont l'ecoulement se fait avec film liquide annulaire et gouttelettes en suspension dans le coeur de vapeur. Par l'analyse des lois de transfert de chaleur, on a precise les mecanismes d'echanges et l'on a d'autre part determine dans quelles conditions le flux de chaleur critique peut etre depasse sans danger de 'burnout' proprement dit. On peut ainsi obtenir des

  4. Critical Care

    Science.gov (United States)

    Critical care helps people with life-threatening injuries and illnesses. It might treat problems such as complications from surgery, ... attention by a team of specially-trained health care providers. Critical care usually takes place in an ...

  5. Heat pumps

    CERN Document Server

    Macmichael, DBA

    1988-01-01

    A fully revised and extended account of the design, manufacture and use of heat pumps in both industrial and domestic applications. Topics covered include a detailed description of the various heat pump cycles, the components of a heat pump system - drive, compressor, heat exchangers etc., and the more practical considerations to be taken into account in their selection.

  6. How Critical Is Critical Thinking?

    Science.gov (United States)

    Shaw, Ryan D.

    2014-01-01

    Recent educational discourse is full of references to the value of critical thinking as a 21st-century skill. In music education, critical thinking has been discussed in relation to problem solving and music listening, and some researchers suggest that training in critical thinking can improve students' responses to music. But what exactly is…

  7. Critical Jostling

    Directory of Open Access Journals (Sweden)

    Pippin Barr

    2016-11-01

    Full Text Available Games can serve a critical function in many different ways, from serious games about real world subjects to self-reflexive commentaries on the nature of games themselves. In this essay we discuss critical possibilities stemming from the area of critical design, and more specifically Carl DiSalvo’s adversarial design and its concept of reconfiguring the remainder. To illustrate such an approach, we present the design and outcomes of two games, Jostle Bastard and Jostle Parent. We show how the games specifically engage with two previous games, Hotline Miami and Octodad: Dadliest Catch, reconfiguring elements of those games to create interactive critical experiences and extensions of the source material. Through the presentation of specific design concerns and decisions, we provide a grounded illustration of a particular critical function of videogames and hope to highlight this form as another valuable approach in the larger area of videogame criticism.

  8. Critical Proximity

    OpenAIRE

    Simon, Jane

    2010-01-01

    This essay considers how written language frames visual objects. Drawing on Michel Foucault’s response to Raymond Roussel’s obsessive description, the essay proposes a model of criticism where description might press up against its objects. This critical closeness is then mapped across the conceptual art practice and art criticism of Ian Burn. Burn attends to the differences between seeing and reading, and considers the conditions which frame how we look at images, including how w...

  9. Criticality Model

    International Nuclear Information System (INIS)

    Alsaed, A.

    2004-01-01

    The ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, ''Models'', in that they are procedural, rather than mathematical. These model reports document the detailed methodology necessary to implement the approach presented in the Disposal Criticality Analysis Methodology Topical Report and provide calculations utilizing the methodology. Thus, the governing procedure for this type of report is AP-3.12Q, ''Design Calculations and Analyses''. The ''Criticality Model'' is of this latter type, providing a process evaluating the criticality potential of in-package and external configurations. The purpose of this analysis is to layout the process for calculating the criticality potential for various in-package and external configurations and to calculate lower-bound tolerance limit (LBTL) values and determine range of applicability (ROA) parameters. The LBTL calculations and the ROA determinations are performed using selected benchmark experiments that are applicable to various waste forms and various in-package and external configurations. The waste forms considered in this calculation are pressurized water reactor (PWR), boiling water reactor (BWR), Fast Flux Test Facility (FFTF), Training Research Isotope General Atomic (TRIGA), Enrico Fermi, Shippingport pressurized water reactor, Shippingport light water breeder reactor (LWBR), N-Reactor, Melt and Dilute, and Fort Saint Vrain Reactor spent nuclear fuel (SNF). The scope of this analysis is to document the criticality computational method. The criticality

  10. Heat pumps

    CERN Document Server

    Brodowicz, Kazimierz; Wyszynski, M L; Wyszynski

    2013-01-01

    Heat pumps and related technology are in widespread use in industrial processes and installations. This book presents a unified, comprehensive and systematic treatment of the design and operation of both compression and sorption heat pumps. Heat pump thermodynamics, the choice of working fluid and the characteristics of low temperature heat sources and their application to heat pumps are covered in detail.Economic aspects are discussed and the extensive use of the exergy concept in evaluating performance of heat pumps is a unique feature of the book. The thermodynamic and chemical properties o

  11. Advances in heat transfer enhancement

    CERN Document Server

    Saha, Sujoy Kumar; Sundén, Bengt; Wu, Zan

    2016-01-01

    This Brief addresses the phenomena of heat transfer enhancement. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to three other monographs including “Critical Heat Flux in Flow Boiling in Microchannels,” this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.

  12. Critical Review

    DEFF Research Database (Denmark)

    Rosenbaum, Ralph K.; Olsen, Stig Irving

    2018-01-01

    Manipulation and mistakes in LCA studies are as old as the tool itself, and so is its critical review. Besides preventing misuse and unsupported claims, critical review may also help identifying mistakes and more justifiable assumptions as well as generally improve the quality of a study. It thus...... supports the robustness of an LCA and increases trust in its results and conclusions. The focus of this chapter is on understanding what a critical review is, how the international standards define it, what its main elements are, and what reviewer qualifications are required. It is not the objective...... of this chapter to learn how to conduct a critical review, neither from a reviewer nor from a practitioner perspective. The foundation of this chapter and the basis for any critical review of LCA studies are the International Standards ISO 14040:2006, ISO 14044:2006 and ISO TS 14071:2014....

  13. Heat transfer

    International Nuclear Information System (INIS)

    Saad, M.A.

    1985-01-01

    Heat transfer takes place between material systems as a result of a temperature difference. The transmission process involves energy conversions governed by the first and second laws of thermodynamics. The heat transfer proceeds from a high-temperature region to a low-temperature region, and because of the finite thermal potential, there is an increase in entropy. Thermodynamics, however, is concerned with equilibrium states, which includes thermal equilibrium, irrespective of the time necessary to attain these equilibrium states. But heat transfer is a result of thermal nonequilibrium conditions, therefore, the laws of thermodynamics alone cannot describe completely the heat transfer process. In practice, most engineering problems are concerned with the rate of heat transfer rather than the quantity of heat being transferred. Resort then is directed to the particular laws governing the transfer of heat. There are three distinct modes of heat transfer: conduction, convection, and radiation. Although these modes are discussed separately, all three types may occur simultaneously

  14. Beam induced RF heating

    CERN Document Server

    Salvant, B; Arduini, G; Assmann, R; Baglin, V; Barnes, M J; Bartmann, W; Baudrenghien, P; Berrig, O; Bracco, C; Bravin, E; Bregliozzi, G; Bruce, R; Bertarelli, A; Carra, F; Cattenoz, G; Caspers, F; Claudet, S; Day, H; Garlasche, M; Gentini, L; Goddard, B; Grudiev, A; Henrist, B; Jones, R; Kononenko, O; Lanza, G; Lari, L; Mastoridis, T; Mertens, V; Métral, E; Mounet, N; Muller, J E; Nosych, A A; Nougaret, J L; Persichelli, S; Piguiet, A M; Redaelli, S; Roncarolo, F; Rumolo, G; Salvachua, B; Sapinski, M; Schmidt, R; Shaposhnikova, E; Tavian, L; Timmins, M; Uythoven, J; Vidal, A; Wenninger, J; Wollmann, D; Zerlauth, M

    2012-01-01

    After the 2011 run, actions were put in place during the 2011/2012 winter stop to limit beam induced radio frequency (RF) heating of LHC components. However, some components could not be changed during this short stop and continued to represent a limitation throughout 2012. In addition, the stored beam intensity increased in 2012 and the temperature of certain components became critical. In this contribution, the beam induced heating limitations for 2012 and the expected beam induced heating limitations for the restart after the Long Shutdown 1 (LS1) will be compiled. The expected consequences of running with 25 ns or 50 ns bunch spacing will be detailed, as well as the consequences of running with shorter bunch length. Finally, actions on hardware or beam parameters to monitor and mitigate the impact of beam induced heating to LHC operation after LS1 will be discussed.

  15. Heat exchanger

    Science.gov (United States)

    Daman, Ernest L.; McCallister, Robert A.

    1979-01-01

    A heat exchanger is provided having first and second fluid chambers for passing primary and secondary fluids. The chambers are spaced apart and have heat pipes extending from inside one chamber to inside the other chamber. A third chamber is provided for passing a purge fluid, and the heat pipe portion between the first and second chambers lies within the third chamber.

  16. Carbon Dioxide Absorption Heat Pump

    Science.gov (United States)

    Jones, Jack A. (Inventor)

    2002-01-01

    A carbon dioxide absorption heat pump cycle is disclosed using a high pressure stage and a super-critical cooling stage to provide a non-toxic system. Using carbon dioxide gas as the working fluid in the system, the present invention desorbs the CO2 from an absorbent and cools the gas in the super-critical state to deliver heat thereby. The cooled CO2 gas is then expanded thereby providing cooling and is returned to an absorber for further cycling. Strategic use of heat exchangers can increase the efficiency and performance of the system.

  17. Heat pipe

    International Nuclear Information System (INIS)

    Triggs, G.W.; Lightowlers, R.J.; Robinson, D.; Rice, G.

    1986-01-01

    A heat pipe for use in stabilising a specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor, comprises an evaporator section, a condenser section, an adiabatic section therebetween, and a gas reservoir, and contains a vapourisable substance such as sodium. The heat pipe further includes a three layer wick structure comprising an outer relatively fine mesh layer, a coarse intermediate layer and a fine mesh inner layer for promoting unimpeded return of condensate to the evaporation section of the heat pipe while enhancing heat transfer with the heat pipe wall and reducing entrainment of the condensate by the upwardly rising vapour. (author)

  18. Critical Arts

    African Journals Online (AJOL)

    both formal and informal) in culture and social theory. CRITICAL ARTS aims to challenge and ... Book Review: Brian McNair, An Introduction to Political Communication (3rd edition), London: Routledge, 2003, ISBN 0415307082, 272pp. Phil Joffe ...

  19. Critical Proximity

    Directory of Open Access Journals (Sweden)

    Jane Simon

    2010-09-01

    Full Text Available This essay considers how written language frames visual objects. Drawing on Michel Foucault’s response to Raymond Roussel’s obsessive description, the essay proposes a model of criticism where description might press up against its objects. This critical closeness is then mapped across the conceptual art practice and art criticism of Ian Burn. Burn attends to the differences between seeing and reading, and considers the conditions which frame how we look at images, including how we look at, and through words. The essay goes on to consider Meaghan Morris’s writing on Lynn Silverman’s photographs. Both Morris and Burn offer an alternative to a parasitic model of criticism and enact a patient way of looking across and through visual landscapes.

  20. Critical proximity

    Directory of Open Access Journals (Sweden)

    Simon, Jane

    2010-01-01

    Full Text Available This essay considers how written language frames visual objects. Drawing on Michel Foucault’s response to Raymond Roussel’s obsessive description, the essay proposes a model of criticism where description might press up against its objects. This critical closeness is then mapped across the conceptual art practice and art criticism of Ian Burn. Burn attends to the differences between seeing and reading, and considers the conditions which frame how we look at images, including how we look at, and through words. The essay goes on to consider Meaghan Morris’s writing on Lynn Silverman’s photographs. Both Morris and Burn offer an alternative to a parasitic model of criticism and enact a patient way of looking across and through visual landscapes.

  1. Criticality safety

    International Nuclear Information System (INIS)

    Walker, G.

    1983-01-01

    When a sufficient quantity of fissile material is brought together a self-sustaining neutron chain reaction will be started in it and will continue until some change occurs in the fissile material to stop the chain reaction. The quantity of fissile material required is the 'Critical Mass'. This is not a fixed quantity even for a given type of fissile material but varies between quite wide limits depending on a number of factors. In a nuclear reactor the critical mass of fissile material is assembled under well-defined condition to produce a controllable chain reaction. The same materials have to be handled outside the reactor in all stages of fuel element manufacture, storage, transport and irradiated fuel reprocessing. At any stage it is possible (at least in principle) to assemble a critical mass and thus initiate an accidental and uncontrollable chain reaction. Avoiding this is what criticality safety is all about. A system is just critical when the rate of production of neutrons balances the rate of loss either by escape or by absorption. The factors affecting criticality are, therefore, those which effect neutron production and loss. The principal ones are:- type of nuclide and enrichment (or isotopic composition), moderation, reflection, concentration (density), shape and interaction. Each factor is considered in detail. (author)

  2. Autoclave nuclear criticality safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

    1991-12-31

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

  3. Critical Vidders

    DEFF Research Database (Denmark)

    Svegaard, Robin Sebastian Kaszmarczyk

    2015-01-01

    This article will introduce and take a look at a specific subset of the fan created remix videos known as vids, namely those that deal with feminist based critique of media. Through examples, it will show how fans construct and present their critique, and finally broach the topic of the critical ...

  4. Geothermal heat pump

    International Nuclear Information System (INIS)

    Bruno, R.; Tinti, F.

    2009-01-01

    In recent years, for several types of buildings and users, the choice of conditioning by heat pump and low enthalpy geothermal reservoir has been increasing in the Italian market. In fact, such systems are efficient in terms of energy and consumption, they can perform, even at the same time, both functions, heating and cooling and they are environmentally friendly, because they do not produce local emissions. This article will introduce the technology and will focus on critical points of a geothermal field design, from actual practice, to future perspectives for the geo exchanger improvement. Finally, the article presents a best practice case in Bologna district, with an economic analysis showing the convenience of a geothermal heat pump. Conclusions of the real benefits of these plants can be drawn: compared to a non-negligible initial cost, the investment has a pay-back period almost always acceptable, usually less than 10 years. [it

  5. Heat exchanger

    International Nuclear Information System (INIS)

    Leigh, D.G.

    1976-01-01

    The arrangement described relates particularly to heat exchangers for use in fast reactor power plants, in which heat is extracted from the reactor core by primary liquid metal coolant and is then transferred to secondary liquid metal coolant by means of intermediate heat exchangers. One of the main requirements of such a system, if used in a pool type fast reactor, is that the pressure drop on the primary coolant side must be kept to a minimum consistent with the maintenance of a limited dynamic head in the pool vessel. The intermediate heat exchanger must also be compact enough to be accommodated in the reactor vessel, and the heat exchanger tubes must be available for inspection and the detection and plugging of leaks. If, however, the heat exchanger is located outside the reactor vessel, as in the case of a loop system reactor, a higher pressure drop on the primary coolant side is acceptable, and space restriction is less severe. An object of the arrangement described is to provide a method of heat exchange and a heat exchanger to meet these problems. A further object is to provide a method that ensures that excessive temperature variations are not imposed on welded tube joints by sudden changes in the primary coolant flow path. Full constructional details are given. (U.K.)

  6. Direct Heat

    Energy Technology Data Exchange (ETDEWEB)

    Lienau, P J

    1990-01-01

    Potential resources and applications of earth heat in the form of geothermal energy are large. United States direct uses amount to 2,100 MWt thermal and worldwide 8,850 MWt above a reference temperature of 35 degrees Celsius. Space and district heating are the major direct uses of geothermal energy. Equipment employed in direct use projects is of standard manufacture and includes downhole and circulation pumps, transmission and distribution pipelines, heat exchangers and convectors, heat pumps and chillers. Direct uses of earth heat discussed are space and district heating, greenhouse heating and fish farming, process and industrial applications. The economic feasibility of direct use projects is governed by site specific factors such as location of user and resource, resource quality, system load factor and load density, as well as financing. Examples are presented of district heating in Klamath Falls, and Elko. Further developments of direct uses of geothermal energy will depend on matching user needs to the resource, and improving load factors and load density.

  7. Heat Stroke

    DEFF Research Database (Denmark)

    Mørch, Sofie Søndergaard; Andersen, Johnny Dohn Holmgren; Bestle, Morten Heiberg

    2017-01-01

    not diagnosed until several days after admittance; hence treatment with cooling was delayed. Both patients were admitted to the intensive care unit, where they were treated with an external cooling device and received treatment for complications. Both cases ended fatally. As global warming continues, more heat......Heat stroke is an acute, life-threatening emergency characterized clinically by elevated body temperature and central nervous system dysfunction. Early recognition and treatment including aggressive cooling and management of life-threatening systemic complications are essential to reduce morbidity...... and mortality. This case report describes two Danish patients diagnosed with heat stroke syndrome during a heat wave in the summer of 2014. Both patients were morbidly obese and had several predisposing illnesses. However since heat stroke is a rare condition in areas with temperate climate, they were...

  8. Heat transfer from two-side heated helical channels

    International Nuclear Information System (INIS)

    Shimonis, V.; Ragaishis, V.; Poshkas, P.

    1995-01-01

    Experimental results are presented on the heat transfer from two-side heated helical channels to gas (air) flows. The study covered six configurations and wide ranges of geometrical (D/h=5.5 to 84.2) and performance (Re=10 3 to 2*10 5 ) parameters. Under the influence of Re and of the channel curvature, the heat transfer from both the convex and the concave surfaces for two-side heating (q w1 ≅ q w2 ) is augmented by 20-30% over one-side heating. Improved relations to predict the critical values of Reynolds Re cr1 and Re cr2 are suggested. They enable more exact predictions of the heat transfer from convex surface in transient flows for one-side heating. The relation for annular channels is suggested for the turbulent heat transfer from the convex and concave surfaces of two-side heated helical channels. It can be adapted by introducing earlier expresions for one-side heated helical channels. (author). 6 refs., 2 tabs., 3 figs

  9. Open heat exchanger for improved heat efficiency in geothermal spas

    Energy Technology Data Exchange (ETDEWEB)

    Nasrabady, S.J.; Palsson, H.; Saevarsdottir, G.A.

    2008-09-15

    Hot spas and Jacuzzis are popular in Iceland due to the abundance of reasonably prized geothermal heat available. However the water from the district heating system is too warm to be admitted directly into the spa. For safety reasons the water is mixed with cold water, in order to reduce temperature from about 80 deg C down to 45 deg C, which leads to wasting a large quantity of heat. Therefore a design is suggested here that enables the feeding of geothermal water directly into the spa, omitting the step of mixing it with cold water. The idea is to employ an open heat exchanger that transfers heat from the geothermal water to the bulk water in the spa, before letting it mix with the spa water. A case study was done for one particular spa. Heat load was calculated and measured when the spa was in use, and when it was unused. A design is suggested employing a circular double-plate which is to be placed at the bottom of the spa. This unit will function as an open heat exchanger feeding district heating water into the spa. Free convection takes place at the upper side of the upper plate and forced convection below the upper plate. Heat transfer coefficient for both was calculated. Using results from calculations, temperature distribution at critical parts of spa and plate was modeled. Results are reasonable and promising for a good design that may considerably reduce the energy expenses for a continuously heated geothermal spa

  10. Critical reading and critical thinking Critical reading and critical thinking

    Directory of Open Access Journals (Sweden)

    Loni Kreis Taglieber

    2008-04-01

    Full Text Available The purpose of this paper is to provide, for L1 and L2 reading and writing teachers, a brief overview of the literature about critical reading and higher level thinking skills. The teaching of these skills is still neglected in some language classes in Brazil, be it in L1 or in L2 classes. Thus, this paper may also serve as a resource guide for L1 and/or L2 reading and writing teachers who want to incorporate critical reading and thinking into their classes. In modern society, even in everyday life people frequently need to deal with complicated public and political issues, make decisions, and solve problems. In order to do this efficiently and effectively, citizens must be able to evaluate critically what they see, hear, and read. Also, with the huge amount of printed material available in all areas in this age of “information explosion” it is easy to feel overwhelmed. But often the information piled up on people’s desks and in their minds is of no use due to the enormous amount of it. The purpose of this paper is to provide, for L1 and L2 reading and writing teachers, a brief overview of the literature about critical reading and higher level thinking skills. The teaching of these skills is still neglected in some language classes in Brazil, be it in L1 or in L2 classes. Thus, this paper may also serve as a resource guide for L1 and/or L2 reading and writing teachers who want to incorporate critical reading and thinking into their classes. In modern society, even in everyday life people frequently need to deal with complicated public and political issues, make decisions, and solve problems. In order to do this efficiently and effectively, citizens must be able to evaluate critically what they see, hear, and read. Also, with the huge amount of printed material available in all areas in this age of “information explosion” it is easy to feel overwhelmed. But often the information piled up on people’s desks and in their minds is of

  11. Criticality accident:

    International Nuclear Information System (INIS)

    Canavese, Susana I.

    2000-01-01

    A criticality accident occurred at 10:35 on September 30, 1999. It occurred in a precipitation tank in a Conversion Test Building at the JCO Tokai Works site in Tokaimura (Tokai Village) in the Ibaraki Prefecture of Japan. STA provisionally rated this accident a 4 on the seven-level, logarithmic International Nuclear Event Scale (INES). The September 30, 1999 criticality accident at the JCO Tokai Works Site in Tokaimura, Japan in described in preliminary, technical detail. Information is based on preliminary presentations to technical groups by Japanese scientists and spokespersons, translations by technical and non-technical persons of technical web postings by various nuclear authorities, and English-language non-technical reports from various news media and nuclear-interest groups. (author)

  12. Critical dynamics

    International Nuclear Information System (INIS)

    Dekker, H.

    1980-01-01

    It is shown how to solve the master equation for a Markov process including a critical point by means of successive approximations in terms of a small parameter. A critical point occurs if, by adjusting an externally controlled quantity, the system shows a transition from normal monostable to bistable behaviour. The fundamental idea of the theory is to separate the master equation into its proper irreducible part and a corrective remainder. The irreducible or zeroth order stochastic approximation will be a relatively simple Fokker-Planck equation that contains the essential features of the process. Once the solution of this irreducible equation is known, the higher order corrections in the original master equation can be incorporated in a systematic manner. (Auth.)

  13. Critical scattering

    International Nuclear Information System (INIS)

    Stirling, W.G.; Perry, S.C.

    1996-01-01

    We outline the theoretical and experimental background to neutron scattering studies of critical phenomena at magnetic and structural phase transitions. The displacive phase transition of SrTiO 3 is discussed, along with examples from recent work on magnetic materials from the rare-earth (Ho, Dy) and actinide (NpAs, NpSb, USb) classes. The impact of synchrotron X-ray scattering is discussed in conclusion. (author) 13 figs., 18 refs

  14. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1994-01-01

    It is approximately 10 years since the Third Edition of Heat Pipes was published and the text is now established as the standard work on the subject. This new edition has been extensively updated, with revisions to most chapters. The introduction of new working fluids and extended life test data have been taken into account in chapter 3. A number of new types of heat pipes have become popular, and others have proved less effective. This is reflected in the contents of chapter 5. Heat pipes are employed in a wide range of applications, including electronics cooling, diecasting and injection mo

  15. Heat conduction

    International Nuclear Information System (INIS)

    Grigull, U.; Sandner, H.

    1984-01-01

    Included are discussions of rates of heat transfer by conduction, the effects of varying and changing properties, thermal explosions, distributed heat sources, moving heat sources, and non-steady three-dimensional conduction processes. Throughout, the importance of thinking both numerically and symbolically is stressed, as this is essential to the development of the intuitive understanding of numerical values needed for successful designing. Extensive tables of thermophysical properties, including thermal conductivity and diffusivity, are presented. Also included are exact and approximate solutions to many of the problems that arise in practical situations

  16. Critical factors to bioenergy implementation

    International Nuclear Information System (INIS)

    Roos, A.; Hektor, B.; Rakos, C.

    1999-01-01

    Barriers to bioenergy technology implementation have received increased attention in recent years. This paper contributes to the identification and analysis of barriers and drivers behind bioenergy market growth, here labelled c ritical factors . It presents a framework for the analysis of both existing and projected bioenergy market potential, using economic concepts and models from transaction cost theory and industrial organization. The framework can be used for assessments of the potential for market growth of different bioenergy systems by decision makers in administration and industry. The following critical factors are identified: Integration with other economic activity, Scale effects on bioenergy markets, Competition in bioenergy markets, Competition with other business, National policy, Local policy and local opinion. The framework is demonstrated with five cases of real bioenergy markets: Pellet residential heating in USA, bioenergy power in USA, pellet residential heating in Sweden, biomass district heating in Sweden, and biomass district heating in Austria. Different applications of the framework are discussed

  17. District heating

    International Nuclear Information System (INIS)

    Hansen, L.

    1993-01-01

    The environmental risks and uncertainties of a high-energy future are disturbing and give rise to several reservations concerning the use of fossil fuels. A number of technologies will help to reduce atmospheric pollution. In Denmark special importance is attached to the following: Energy conservation. Efficient energy conversion. Renewable energy sources. District heating, combined production of heat and power. Many agree that district heating (DH), produced by the traditional heat-only plant, and combined heat and power (CHP) have enormous potential when considering thermal efficiency and lowered environmental impacts: The basic technology of each is proven, it would be relatively simple to satisfy a substantial part of the energy demand, and their high efficiencies mean reduced pollution including greenhouse gas emissions. This is especially important in high population density areas - the obviously preferred sites for such energy generation. Compared with individual heating DH can provide a community with an operationally efficient and most often also an economically competitive heat supply. This is particularly true under the circumstances where the DH system is supplied from CHP plants. Their use results in very substantial improvements in overall efficiency. Further environmental improvements arise from the reduced air pollution obtainable in reasonably large CHP plants equipped with flue gas cleaning to remove particles, sulphur dioxide, and nitrogen acids. As a consequence of these considerations, DH plays an important role in fulfilling the space and water heating demand in many countries. This is especially the case in Denmark where this technology is utilised to a very great extent. Indeed, DH is one of the reasons why Denmark has relatively good air quality in the cities. (au)

  18. Critical Mass

    CERN Multimedia

    AUTHOR|(CDS)2070299

    2017-01-01

    Critical Mass is a cycling event typically held on the last Friday of every month; its purpose is not usually formalized beyond the direct action of meeting at a set location and time and traveling as a group through city or town streets on bikes. The event originated in 1992 in San Francisco; by the end of 2003, the event was being held in over 300 cities around the world. At CERN it is held once a year in conjunction with the national Swiss campaing "Bike to work".

  19. Heat Rejection from a Variable Conductance Heat Pipe Radiator Panel

    Science.gov (United States)

    Jaworske, D. A.; Gibson, M. A.; Hervol, D. S.

    2012-01-01

    A titanium-water heat pipe radiator having an innovative proprietary evaporator configuration was evaluated in a large vacuum chamber equipped with liquid nitrogen cooled cold walls. The radiator was manufactured by Advanced Cooling Technologies, Inc. (ACT), Lancaster, PA, and delivered as part of a Small Business Innovative Research effort. The radiator panel consisted of five titanium-water heat pipes operating as thermosyphons, sandwiched between two polymer matrix composite face sheets. The five variable conductance heat pipes were purposely charged with a small amount of non-condensable gas to control heat flow through the condenser. Heat rejection was evaluated over a wide range of inlet water temperature and flow conditions, and heat rejection was calculated in real-time utilizing a data acquisition system programmed with the Stefan-Boltzmann equation. Thermography through an infra-red transparent window identified heat flow across the panel. Under nominal operation, a maximum heat rejection value of over 2200 Watts was identified. The thermal vacuum evaluation of heat rejection provided critical information on understanding the radiator s performance, and in steady state and transient scenarios provided useful information for validating current thermal models in support of the Fission Power Systems Project.

  20. Investigation Status of Heat Exchange while Boiling Hydrocarbon Fuel

    Directory of Open Access Journals (Sweden)

    D. S. Obukhov

    2006-01-01

    Full Text Available The paper contains analysis of heat exchange investigations while boiling hydrocarbon fuel. The obtained data are within the limits of the S.S. Kutateladze dependence proposed in 1939. Heat exchange at non-stationary heat release has not been investigated. The data for hydrocarbon fuel with respect to critical density of heat flow are not available even for stationary conditions.

  1. Heating experiments of JT-60

    International Nuclear Information System (INIS)

    1987-01-01

    In JT-60, after the finish of the first stage Joule experiment, the heating facilities were installed, and the heating experiment was started in August, 1986. As to neutral beam injection, the beam injection experiment at the maximum rating 20 MW carried out, and also as to RF, the injection experiment up to 1.4 MW was carried out in both ion cyclotron and low band hybrid waves. The results worthy of special mention in the heating experiment were the success in the current drive up to 1.7 MA at maximum using low band hybrid waves and the improvement of plasma confinement characteristics obtained by the compound heating of NBI and RF. In this paper, the main results of these heating experiments and their significance are explained. The JT-60 is the testing facilities for attaining the critical plasma condition by additionally heating the plasma which is generated by Joule electric discharge with NBI and RF heatings. The experimental operation cycle of the JT-60 consists of the unit cycle of two weeks, and the number of days in operation is nine days. The temperature of heated plasma rose to 70 million deg C in the 20 MW NBI heating. Hereafter, the improvement of confinement time by increasing the stored energy of plasma is attempted. (Kako, I.)

  2. Dictionary criticism

    DEFF Research Database (Denmark)

    Nielsen, Sandro

    2018-01-01

    Dictionary criticism is part of the lexicographical universe and reviewing of electronic and printed dictionaries is not an exercise in linguistics or in subject fields but an exercise in lexicography. It does not follow from this that dictionary reviews should not be based on a linguistic approach......, but that the linguistic approach is only one of several approaches to dictionary reviewing. Similarly, the linguistic and factual competences of reviewers should not be relegated to an insignificant position in the review process. Moreover, reviewers should define the object of their reviews, the dictionary, as a complex...... information tool with several components and in terms of significant lexicographical features: lexicographical functions, data and structures. This emphasises the fact that dictionaries are much more than mere vessels of linguistic categories, namely lexicographical tools that have been developed to fulfil...

  3. Review: heat pipe heat exchangers at IROST

    OpenAIRE

    E. Azad

    2012-01-01

    The use of the heat pipe as a component in a heat recovery device has gained worldwide acceptance. Heat pipes are passive, highly reliable and offer high heat transfer rates. This study summarizes the investigation of different types of heat pipe heat recovery systems (HPHRSs). The studies are classified on the basis of the type of the HPHRS. This research is based on 30 years of experience on heat pipe and heat recovery systems that are presented in this study. Copyright , Oxford University ...

  4. Heat pipes and heat pipe exchangers for heat recovery systems

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, L L; Grakovich, L P; Kiselev, V G; Kurustalev, D K; Matveev, Yu

    1984-01-01

    Heat pipes and heat pipe exchangers are of great importance in power engineering as a means of recovering waste heat of industrial enterprises, solar energy, geothermal waters and deep soil. Heat pipes are highly effective heat transfer units for transferring thermal energy over large distance (tens of meters) with low temperature drops. Their heat transfer characteristics and reliable working for more than 10-15 yr permit the design of new systems with higher heat engineering parameters.

  5. Unwanted heat

    International Nuclear Information System (INIS)

    Benka, M.

    2006-01-01

    The number of small heating plants using biomass is growing. According to TREND's information, Hrinovska energeticka, is the only one that controls the whole supplier chain in cooperation with its parent company in Bratislava. Starting with the collection and processing of wood chips by burning, heat production and heat distribution to the end user. This gives the company better control over costs and consequently its own prices. Last year, the engineering company, Hrinovske storjarne, decided to focus only on its core business and sold its heating plant, Hrinovske tepelne hospodarstvo, to Intech Slovakia and changed the company name to Hrinovska energeticka. Local companies and inhabitants were concerned that the new owner would increase prices. But the company publicly declared and kept promises that the heat price for households would remain at 500 Slovak crowns/gigajoule (13.33 EUR/gigajoule ), one of the lowest prices in Slovakia. This year the prices increased slightly to 570 Slovak crowns (15.2 EUR). 'We needed - even at the cost of lower profit - to satisfy our customers so that we would not lose them. We used this time for transition to biomass. This will allow us to freeze our prices in the coming years,' explained the statutory representative of the company, Ivan Dudak. (authors)

  6. Heat Pipes

    Science.gov (United States)

    1990-01-01

    Bobs Candies, Inc. produces some 24 million pounds of candy a year, much of it 'Christmas candy.' To meet Christmas demand, it must produce year-round. Thousands of cases of candy must be stored a good part of the year in two huge warehouses. The candy is very sensitive to temperature. The warehouses must be maintained at temperatures of 78-80 degrees Fahrenheit with relative humidities of 38- 42 percent. Such precise climate control of enormous buildings can be very expensive. In 1985, energy costs for the single warehouse ran to more than 57,000 for the year. NASA and the Florida Solar Energy Center (FSEC) were adapting heat pipe technology to control humidity in building environments. The heat pipes handle the jobs of precooling and reheating without using energy. The company contacted a FSEC systems engineer and from that contact eventually emerged a cooperative test project to install a heat pipe system at Bobs' warehouses, operate it for a period of time to determine accurately the cost benefits, and gather data applicable to development of future heat pipe systems. Installation was completed in mid-1987 and data collection is still in progress. In 1989, total energy cost for two warehouses, with the heat pipes complementing the air conditioning system was 28,706, and that figures out to a cost reduction.

  7. Monopole heat

    International Nuclear Information System (INIS)

    Turner, M.S.

    1983-01-01

    Upper bounds on the flux of monopoles incident on the Earth with velocity -5 c(10 16 GeV m -1 ) and on the flux of monopoles incident on Jupiter with velocity -3 c(10 16 GeV m -1 ), are derived. Monopoles moving this slowly lose sufficient energy to be stopped, and then catalyse nucleon decay, releasing heat. The limits are obtained by requiring the rate of energy release from nucleon decay to be less than the measured amount of heat flowing out from the surface of the planet. (U.K.)

  8. Heat exchanger

    International Nuclear Information System (INIS)

    Drury, C.R.

    1988-01-01

    A heat exchanger having primary and secondary conduits in heat-exchanging relationship is described comprising: at least one serpentine tube having parallel sections connected by reverse bends, the serpentine tube constituting one of the conduits; a group of open-ended tubes disposed adjacent to the parallel sections, the open-ended tubes constituting the other of the conduits, and forming a continuous mass of contacting tubes extending between and surrounding the serpentine tube sections; and means securing the mass of tubes together to form a predetermined cross-section of the entirety of the mass of open-ended tubes and tube sections

  9. Heat Convection

    Science.gov (United States)

    Jiji, Latif M.

    Professor Jiji's broad teaching experience lead him to select the topics for this book to provide a firm foundation for convection heat transfer with emphasis on fundamentals, physical phenomena, and mathematical modelling of a wide range of engineering applications. Reflecting recent developments, this textbook is the first to include an introduction to the challenging topic of microchannels. The strong pedagogic potential of Heat Convection is enhanced by the follow ing ancillary materials: (1) Power Point lectures, (2) Problem Solutions, (3) Homework Facilitator, and, (4) Summary of Sections and Chapters.

  10. Combination solar photovoltaic heat engine energy converter

    Science.gov (United States)

    Chubb, Donald L.

    1987-01-01

    A combination solar photovoltaic heat engine converter is proposed. Such a system is suitable for either terrestrial or space power applications. The combination system has a higher efficiency than either the photovoltaic array or the heat engine alone can attain. Advantages in concentrator and radiator area and receiver mass of the photovoltaic heat engine system over a heat-engine-only system are estimated. A mass and area comparison between the proposed space station organic Rankine power system and a combination PV-heat engine system is made. The critical problem for the proposed converter is the necessity for high temperature photovoltaic array operation. Estimates of the required photovoltaic temperature are presented.

  11. Renewable Heating And Cooling

    Science.gov (United States)

    Renewable heating and cooling is a set of alternative resources and technologies that can be used in place of conventional heating and cooling technologies for common applications such as water heating, space heating, space cooling and process heat.

  12. Critical evaluation of the experiments and mathematical models for the determination of fission product release from the spherical fuel elements in cases of core heating accidents in modular HTR's

    International Nuclear Information System (INIS)

    Bailly, H.W.

    1987-01-01

    In this work, the thermal behaviour of modular reactors in cases of core heating accidents and the physical phenomena relevant for a release of radioactive materials from HTR fuel elements are explained as far as is necessary for understanding the work. The present mathematical models by which the release of radioactive materials from HTR fuel elements due to diffusion or breaking particles in cases of core heating accidents are also described, examined and evaluated with regard to their applicability to module reactors. The experiments used to verify the mathematical models are also evaluated. The mathematical models are in nearly all cases computer programs, which describe the complicated process of releasing radioactive materials quantitative mathematically. One should point out that these models are constantly being developed further, in line with the increasing amount of knowledge. To conclude the work, proposals are made for improving the certainty of information from experiments and mathematical models to determine the release behaviour of modular reactors. (orig./GL) [de

  13. Heat exchanger

    Science.gov (United States)

    Wolowodiuk, Walter

    1976-01-06

    A heat exchanger of the straight tube type in which different rates of thermal expansion between the straight tubes and the supply pipes furnishing fluid to those tubes do not result in tube failures. The supply pipes each contain a section which is of helical configuration.

  14. Heat exchangers

    International Nuclear Information System (INIS)

    1975-01-01

    The tubes of a heat exchanger tube bank have a portion thereof formed in the shape of a helix, of effective radius equal to the tube radius and the space between two adjacent tubes, to tangentially contact the straight sections of the tubes immediately adjacent thereto and thereby provide support, maintain the spacing and account for differential thermal expansion thereof

  15. Heat exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Harada, F; Yanagida, T; Fujie, K; Futawatari, H

    1975-04-30

    The purpose of this construction is the improvement of heat transfer in finned tube heat exchangers, and therefore the improvement of its efficiency or its output per unit volume. This is achieved by preventing the formation of flow boundary layers in gaseous fluid. This effect always occurs on flow of smooth adjacent laminae, and especially if these have pipes carrying liquid passing through them; it worsens the heat transfer of such a boundary layer considerably compared to that in the turbulent range. The fins, which have several rows of heat exchange tubes passing through them, are fixed at a small spacing on theses tubes. The fins have slots cut in them by pressing or punching, where the pressed-out material remains as a web, which runs parallel to the level of the fin and at a small distance from it. These webs and slots are arranged radially around every tube hole, e.g. 6 in number. For a suitable small tube spacing, two adjacent tubes opposite each other have one common slot. Many variants of such slot arrangements are illustrated.

  16. Heat exchanger

    International Nuclear Information System (INIS)

    Wolowodiuk, W.

    1976-01-01

    A heat exchanger of the straight tube type is described in which different rates of thermal expansion between the straight tubes and the supply pipes furnishing fluid to those tubes do not result in tube failures. The supply pipes each contain a section which is of helical configuration

  17. Critical point measurement of some polycyclic aromatic hydrocarbons

    International Nuclear Information System (INIS)

    Nikitin, Eugene D.; Popov, Alexander P.

    2015-01-01

    Highlights: • Critical properties of five polycyclic aromatic hydrocarbons were measured. • These hydrocarbons decompose at near-critical temperatures. • Pulse-heating method with short residence times was used. - Abstract: The critical temperatures and the critical pressures of five polycyclic aromatic compounds, namely, acenaphthene, fluorene, anthracene, phenanthrene, and pyrene have been measured. All the compounds studied decompose at near-critical temperatures. A pulse-heating technique applicable to measuring the critical properties of thermally unstable compounds has been used. The times from the beginning of a heating pulse to the moment of reaching the critical temperature were from (0.06 to 0.85) ms. The short residence times provide little degradation of the substances in the course of the experiments. The experimental critical parameters of the polycyclic aromatic compounds have been compared with those estimated by five predictive methods. The acentric factors of polycyclic aromatic compounds studied have been calculated

  18. Peeling mechanism of tomato under infrared heating

    Science.gov (United States)

    Critical behaviors of peeling tomatoes using infrared heat are thermally induced peel loosening and subsequent cracking. However, the mechanism of peel loosening and cracking due to infrared heating remains unclear. This study aimed at investigating the mechanism of peeling tomatoes under infrared h...

  19. Heat pipe heat exchangers in heat recovery systems

    Energy Technology Data Exchange (ETDEWEB)

    Stulc, P; Vasiliev, L L; Kiseljev, V G; Matvejev, Ju N

    1985-01-01

    The results of combined research and development activities of the National Research Institute for Machine Design, Prague, C.S.S.R. and the Institute for Heat and Mass Transfer, Minsk, U.S.S.R. concerning intensification heat pipes used in heat pipe heat exchangers are presented. This sort of research has been occasioned by increased interest in heat power economy trying to utilise waste heat produced by various technological processes. The developed heat pipes are deployed in construction of air-air, gas-air or gas-gas heat recovery exchangers in the field of air-engineering and air-conditioning. (author).

  20. Heat exchanger

    International Nuclear Information System (INIS)

    Bennett, J.C.

    1975-01-01

    A heat exchanger such as forms, for example, part of a power steam boiler is made up of a number of tubes that may be arranged in many different ways, and it is necessary that the tubes be properly supported. The means by which the tubes are secured must be as simple as possible so as to facilitate construction and must be able to continue to function effectively under the varying operating conditions to which the heat exchanger is subject. The arrangement described is designed to meet these requirements, in an improved way. The tubes are secured to a member extending past several tubes and abutment means are provided. At least some of the abutment means comprise two abutment pieces and a wedge secured to the supporting member, that acts on these pieces to maintain the engagement. (U.K.)

  1. Toward a Heat Recovery Chimney

    Directory of Open Access Journals (Sweden)

    Min Pan

    2011-11-01

    Full Text Available The worldwide population increase and subsequent surge in energy demand leads electricity producers to increase supply in an attempt to generate larger profit margins. However, with Global Climate Change becoming a greater focus in engineering, it is critical for energy to be converted in as environmentally benign a way as possible. There are different sustainable methods to meet the energy demand. However, the focus of this research is in the area of Waste Heat Recovery. The waste heat stored in the exiting condenser cooling water is delivered to the air flow through a water-air cross flow heat exchanger. A converging thermal chimney structure is then applied to increase the velocity of the airflow. The accelerated air can be used to turn on the turbine-generator installed on the top the thermal chimney so that electricity can be generated. This system is effective in generating electricity from otherwise wasted heat.

  2. Low grade waste heat recovery using heat pumps and power cycles

    International Nuclear Information System (INIS)

    Bor, D.M. van de; Infante Ferreira, C.A.; Kiss, Anton A.

    2015-01-01

    Thermal energy represents a large part of the global energy usage and about 43% of this energy is used for industrial applications. Large amounts are lost via exhaust gases, liquid streams and cooling water while the share of low temperature waste heat is the largest. Heat pumps upgrading waste heat to process heat and cooling and power cycles converting waste heat to electricity can make a strong impact in the related industries. The potential of several alternative technologies, either for the upgrading of low temperature waste heat such as compression-resorption, vapor compression and trans-critical heat pumps, or for the conversion of this waste heat by using organic Rankine, Kalina and trilateral cycle engines, are investigated with regards to energetic and economic performance by making use of thermodynamic models. This study focuses on temperature levels of 45–60 °C as at this temperature range large amounts of heat are rejected to the environment but also investigates the temperature levels for which power cycles become competitive. The heat pumps deliver 2.5–11 times more energy value than the power cycles in this low temperature range at equal waste heat input. Heat engines become competitive with heat pumps at waste heat temperatures at 100 °C and above. - Highlights: • Application of heat pump technology for heating and cooling. • Compression resorption heat pumps operating with large glides approaching 100 K. • Compression-resorption heat pumps with wet compression. • Potential to convert Industrial waste heat to power or high grade heat. • Comparison between low temperature power cycles and heat pumps

  3. Heat of mixing and morphological stability

    Science.gov (United States)

    Nandapurkar, P.; Poirier, D. R.

    1988-01-01

    A mathematical model, which incorporates heat of mixing in the energy balance, has been developed to analyze the morphological stability of a planar solid-liquid interface during the directional solidification of a binary alloy. It is observed that the stability behavior is almost that predicted by the analysis of Mullins and Sekerka (1963) at low growth velocities, while deviations in the critical concentration of about 20-25 percent are observed under rapid solidification conditions for certain systems. The calculations indicate that a positive heat of mixing makes the planar interface more unstable, whereas a negative heat of mixing makes it more stable, in terms of the critical concentration.

  4. Heating networks and domestic central heating systems

    Energy Technology Data Exchange (ETDEWEB)

    Kamler, W; Wasilewski, W

    1976-08-01

    This is a comprehensive survey of the 26 contributions from 8 European countries submitted to the 3rd International District Heating Conference in Warsaw held on the subject 'Heating Networks and Domestic Central Heating Systems'. The contributions are grouped according to 8 groups of subjects: (1) heat carriers and their parameters; (2) system of heating networks; (3) calculation and optimization of heating networks; (4) construction of heating networks; (5) operation control and automation; (6) operational problems; (7) corrosion problems; and (8) methods of heat accounting.

  5. Hydride heat pump with heat regenerator

    Science.gov (United States)

    Jones, Jack A. (Inventor)

    1991-01-01

    A regenerative hydride heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system. A series of at least four canisters containing a lower temperature performing hydride and a series of at least four canisters containing a higher temperature performing hydride is provided. Each canister contains a heat conductive passageway through which a heat transfer fluid is circulated so that sensible heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

  6. Heating systems for heating subsurface formations

    Science.gov (United States)

    Nguyen, Scott Vinh [Houston, TX; Vinegar, Harold J [Bellaire, TX

    2011-04-26

    Methods and systems for heating a subsurface formation are described herein. A heating system for a subsurface formation includes a sealed conduit positioned in an opening in the formation and a heat source. The sealed conduit includes a heat transfer fluid. The heat source provides heat to a portion of the sealed conduit to change phase of the heat transfer fluid from a liquid to a vapor. The vapor in the sealed conduit rises in the sealed conduit, condenses to transfer heat to the formation and returns to the conduit portion as a liquid.

  7. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1982-01-01

    A comprehensive, up-to-date coverage of the theory, design and manufacture of heat pipes and their applications. This latest edition has been thoroughly revised, up-dated and expanded to give an in-depth coverage of the new developments in the field. Significant new material has been added to all the chapters and the applications section has been totally rewritten to ensure that topical and important applications are appropriately emphasised. The bibliography has been considerably enlarged to incorporate much valuable new information. Thus readers of the previous edition, which has established

  8. Heat exchanger

    International Nuclear Information System (INIS)

    Dostatni, A.W.; Dostatni, Michel.

    1976-01-01

    In the main patent, a description was given of a heat exchanger with an exchange surface in preformed sheet metal designed for the high pressure and temperature service particularly encountered in nuclear pressurized water reactors and which is characterised by the fact that it is composed of at least one exchanger bundle sealed in a containment, the said bundle or bundles being composed of numerous juxtaposed individual compartments whose exchange faces are built of preformed sheet metal. The present addendun certificate concerns shapes of bundles and their positioning methods in the exchanger containment enabling its compactness to be increased [fr

  9. Critical properties of some aliphatic symmetrical ethers

    International Nuclear Information System (INIS)

    Nikitin, Eugene D.; Popov, Alexander P.; Bogatishcheva, Nataliya S.

    2014-01-01

    Highlights: • Critical properties of simple aliphatic ethers were measured. • The ethers decompose at near-critical temperatures. • Pulse-heating method with short residence times was used. -- Abstract: The critical temperatures T c and the critical pressures p c of dihexyl, dioctyl, and didecyl ethers have been measured. According to the measurements, the coordinates of the critical points are T c = (665 ± 7) K, p c = (1.44 ± 0.04) MPa for dihexyl ether, T c = (723 ± 7) K, p c = (1.19 ± 0.04) MPa for dioctyl ether, and T c = (768 ± 8) K, p c = (1.03 ± 0.03) MPa for didecyl ether. All the ethers studied degrade chemically at near-critical temperatures. A pulse-heating method applicable to measuring the critical properties of thermally unstable compounds has been used. The times from the beginning of a heating pulse to the moment of reaching the critical temperature were from 0.06 to 0.46 ms. The short residence times provide little decomposition of the substances in the course of the experiments. The critical properties of the ethers investigated in this work have been discussed together with those of methyl to butyl ethers. The experimental critical constants of the ethers have been compared with those estimated by the group-contribution methods of Wilson and Jasperson and Marrero and Gani. The Wilson/Jasperson method provides a better estimation of the critical temperatures and pressures of simple aliphatic ethers in comparison with the Marrero/Gani method if reliable normal boiling temperatures are used in the method of Wilson and Jasperson

  10. Regenerative Hydride Heat Pump

    Science.gov (United States)

    Jones, Jack A.

    1992-01-01

    Hydride heat pump features regenerative heating and single circulation loop. Counterflow heat exchangers accommodate different temperatures of FeTi and LaNi4.7Al0.3 subloops. Heating scheme increases efficiency.

  11. Low temperature nuclear heat

    Energy Technology Data Exchange (ETDEWEB)

    Kotakorpi, J.; Tarjanne, R. [comps.

    1977-08-01

    The meeting was concerned with the use of low grade nuclear heat for district heating, desalination, process heat, and agriculture and aquaculture. The sessions covered applications and demand, heat sources, and economics.

  12. Radiofrequency plasma heating: proceedings

    International Nuclear Information System (INIS)

    Swenson, D.G.

    1985-01-01

    The conference proceedings include sessions on Alfven Wave Heating, ICRF Heating and Current Drive, Lower Hybrid Heating and Current Drive, and ECRF Heating. Questions of confinement, diagnostics, instabilities and technology are considered. Individual papers are cataloged separately

  13. A state-of-the-art review on hybrid heat pipe latent heat storage systems

    International Nuclear Information System (INIS)

    Naghavi, M.S.; Ong, K.S.; Mehrali, M.; Badruddin, I.A.; Metselaar, H.S.C.

    2015-01-01

    The main advantage of latent heat thermal energy storage systems is the capability to store a large quantity of thermal energy in an isothermal process by changing phase from solid to liquid, while the most important weakness of these systems is low thermal conductivity that leads to unsuitable charging/discharging rates. Heat pipes are used in many applications – as one of the most efficient heat exchanger devices – to amplify the charging/discharging processes rate and are used to transfer heat from a source to the storage or from the storage to a sink. This review presents and critically discusses previous investigations and analysis on the incorporation of heat pipe devices into latent heat thermal energy storage with heat pipe devices. This paper categorizes different applications and configurations such as low/high temperature solar, heat exchanger and cooling systems, analytical approaches and effective parameters on the performance of hybrid HP–LHTES systems.

  14. High field laser heated solenoids

    International Nuclear Information System (INIS)

    Hoffman, A.L.

    1979-01-01

    A 10 kJ pulsed CO 2 laser and 3.8 cm bore, 15 T, 8 μs rise time, 1-m long fast solenoid facility has been constructed to demonstrate the feasibility of using long wavelength lasers to heat magnetically confined plasmas. The most critical physics requirement is the necessity of creating and maintaining an on-axis electron density minimum to trap the axially directed laser beam. Satisfaction of this requirement has been demonstrated by heating 1.5 Torr deuterium fill plasmas in 2.7 cm bore plasma tubes to line energies of approximately 1 kJ/m. (Auth.)

  15. Heat switch technology for cryogenic thermal management

    Science.gov (United States)

    Shu, Q. S.; Demko, J. A.; E Fesmire, J.

    2017-12-01

    Systematic review is given of development of novel heat switches at cryogenic temperatures that alternatively provide high thermal connection or ideal thermal isolation to the cold mass. These cryogenic heat switches are widely applied in a variety of unique superconducting systems and critical space applications. The following types of heat switch devices are discussed: 1) magnetic levitation suspension, 2) shape memory alloys, 3) differential thermal expansion, 4) helium or hydrogen gap-gap, 5) superconducting, 6) piezoelectric, 7) cryogenic diode, 8) magneto-resistive, and 9) mechanical demountable connections. Advantages and limitations of different cryogenic heat switches are examined along with the outlook for future thermal management solutions in materials and cryogenic designs.

  16. Integration of Heat Exchangers with Thermoelectric Modules

    DEFF Research Database (Denmark)

    Rezaniakolaei, Alireza

    2017-01-01

    processes wherein the critical system components such as the TEG module and the heat exchangers are thermally coupled. The optimization techniques of the TEG systems coupled with the heat transfer through the system using a maximum efficiency-power map for waste heat recovery applications offer maximum...... thermally interdependent in the system designs. This chapter studies the effect of the heat exchangers design on system performance, and discusses the challenges through accurate analyses techniques while introducing proper cooling technologies. Proper design of a TEG system involves design optimization...

  17. Supercritical heat transfer phenomena in nuclear system

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Kim, Moo Hwan; Anderson, Mark H.; Corradini, Michael L.

    2005-01-01

    A supercritical water (SCW) power cycle has been considered as one of the viable candidates for advanced fission reactor designs. However, the dramatic variation of thermo-physical properties with a modest change of temperature near the pseudo-critical point make existing heat transfer correlations such as the Dittus-Boelter correlation not suitably accurate to calculate the heat transfer in supercritical fluid. Several other correlations have also been suggested but none of them are able to predict the heat transfer over a parameter range, needed for reactor thermal-hydraulics simulation and design. This has prompted additional research to understand the characteristic of supercritical fluid heat transfer

  18. Split heat pipe heat recovery system

    OpenAIRE

    E. Azad

    2008-01-01

    This paper describes a theoretical analysis of a split heat pipe heat recovery system. The analysis is based on an Effectiveness-NTU approach to deduce its heat transfer characteristics. In this study the variation of overall effectiveness of heat recovery with the number of transfer units are presented. Copyright , Manchester University Press.

  19. Segmented heat exchanger

    Science.gov (United States)

    Baldwin, Darryl Dean; Willi, Martin Leo; Fiveland, Scott Byron; Timmons, Kristine Ann

    2010-12-14

    A segmented heat exchanger system for transferring heat energy from an exhaust fluid to a working fluid. The heat exchanger system may include a first heat exchanger for receiving incoming working fluid and the exhaust fluid. The working fluid and exhaust fluid may travel through at least a portion of the first heat exchanger in a parallel flow configuration. In addition, the heat exchanger system may include a second heat exchanger for receiving working fluid from the first heat exchanger and exhaust fluid from a third heat exchanger. The working fluid and exhaust fluid may travel through at least a portion of the second heat exchanger in a counter flow configuration. Furthermore, the heat exchanger system may include a third heat exchanger for receiving working fluid from the second heat exchanger and exhaust fluid from the first heat exchanger. The working fluid and exhaust fluid may travel through at least a portion of the third heat exchanger in a parallel flow configuration.

  20. Dual source heat pump

    Science.gov (United States)

    Ecker, Amir L.; Pietsch, Joseph A.

    1982-01-01

    What is disclosed is a heat pump apparatus for conditioning a fluid characterized by a fluid handler and path for circulating the fluid in heat exchange relationship with a refrigerant fluid; at least two refrigerant heat exchangers, one for effecting heat exchange with the fluid and a second for effecting heat exchange between refrigerant and a heat exchange fluid and the ambient air; a compressor for efficiently compressing the refrigerant; at least one throttling valve for throttling liquid refrigerant; a refrigerant circuit; refrigerant; a source of heat exchange fluid; heat exchange fluid circulating device and heat exchange fluid circuit for circulating the heat exchange fluid in heat exchange relationship with the refrigerant; and valves or switches for selecting the heat exchangers and direction of flow of the refrigerant therethrough for selecting a particular mode of operation. The heat exchange fluid provides energy for defrosting the second heat exchanger when operating in the air source mode and also provides a alternate source of heat.

  1. Heat pipes in modern heat exchangers

    International Nuclear Information System (INIS)

    Vasiliev, Leonard L.

    2005-01-01

    Heat pipes are very flexible systems with regard to effective thermal control. They can easily be implemented as heat exchangers inside sorption and vapour-compression heat pumps, refrigerators and other types of heat transfer devices. Their heat transfer coefficient in the evaporator and condenser zones is 10 3 -10 5 W/m 2 K, heat pipe thermal resistance is 0.01-0.03 K/W, therefore leading to smaller area and mass of heat exchangers. Miniature and micro heat pipes are welcomed for electronic components cooling and space two-phase thermal control systems. Loop heat pipes, pulsating heat pipes and sorption heat pipes are the novelty for modern heat exchangers. Heat pipe air preheaters are used in thermal power plants to preheat the secondary-primary air required for combustion of fuel in the boiler using the energy available in exhaust gases. Heat pipe solar collectors are promising for domestic use. This paper reviews mainly heat pipe developments in the Former Soviet Union Countries. Some new results obtained in USA and Europe are also included

  2. Phantom black holes and critical phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Azreg-Aïnou, Mustapha [Engineering Faculty, Başkent University, Bağlıca Campus, Ankara (Turkey); Marques, Glauber T. [Universidade Federal Rural da Amazônia ICIBE-LASIC, Av. Presidente Tancredo Neves 2501, CEP 66077-901—Belém/PA (Brazil); Rodrigues, Manuel E., E-mail: azreg@baskent.edu.tr, E-mail: gtadaiesky@hotmail.com, E-mail: esialg@gmail.com [Faculdade de Ciências Exatas e Tecnologia, Universidade Federal do Pará, Campus Universitário de Abaetetuba, CEP 68440-000, Abaetetuba, Pará (Brazil)

    2014-07-01

    We consider the two classes cosh and sinh of normal and phantom black holes of Einstein-Maxwell-dilaton theory. The thermodynamics of these holes is characterized by heat capacities that may have both signs depending on the parameters of the theory. Leaving aside the normal Reissner-Nordström black hole, it is shown that only some phantom black holes of both classes exhibit critical phenomena. The two classes share a nonextremality, but special, critical point where the transition is continuous and the heat capacity, at constant charge, changes sign with an infinite discontinuity. This point yields a classification scheme for critical points. It is concluded that the two unstable and stable phases coexist on one side of the criticality state and disappear on the other side, that is, there is no configuration where only one phase exists. The sinh class has an extremality critical point where the entropy diverges. The transition from extremality to nonextremality with the charge held constant is accompanied by a loss of mass and an increase in the temperature. A special case of this transition is when the hole is isolated (microcanonical ensemble), it will evolve by emission of energy, which results in a decrease of its mass, to the final state of minimum mass and vanishing heat capacity. The Ehrenfest scheme of classification is inaccurate in this case but the generalized one due to Hilfer leads to conclude that the transition is of order less than unity. Fluctuations near criticality are also investigated.

  3. Geothermal Heat Pump Benchmarking Report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1997-01-17

    A benchmarking study was conducted on behalf of the Department of Energy to determine the critical factors in successful utility geothermal heat pump programs. A Successful program is one that has achieved significant market penetration. Successfully marketing geothermal heat pumps has presented some major challenges to the utility industry. However, select utilities have developed programs that generate significant GHP sales. This benchmarking study concludes that there are three factors critical to the success of utility GHP marking programs: (1) Top management marketing commitment; (2) An understanding of the fundamentals of marketing and business development; and (3) An aggressive competitive posture. To generate significant GHP sales, competitive market forces must by used. However, because utilities have functioned only in a regulated arena, these companies and their leaders are unschooled in competitive business practices. Therefore, a lack of experience coupled with an intrinsically non-competitive culture yields an industry environment that impedes the generation of significant GHP sales in many, but not all, utilities.

  4. Heat tranfer decrease during water boiling in a tube for the heat flux step distribution by the tube length

    International Nuclear Information System (INIS)

    Remizov, O.V.; Sergeev, V.V.; Yurkov, Yu.I.

    1983-01-01

    The effect of the heat flux distribution along the circular tube length on supercritical convective heat transfer at parameters typical for steam generators heated by liquid metal is studied. The effect of conditions in a under- and a supercritical zones of a vertical tube with independently heated lower and upper sections on supercritical convective heat transfer is studied on a water circulation loop at 9.8-17.7 MPa pressure and 330-1000 kg/m 2 s mass velocities. The experimental heat fluxes varied within the following limits: at the upper section from 0 to 474 kW/m 2 , at the lower section from 190 to 590 kW/m 2 . Analysis of the obtained data shows that when heat flux changes in the supercritical zone rewetting of the heated surface and simultaneous existence of two critical zones are observed. The effect of heat flux in the supercritical zone on convective heat transfer is ambiguous: the heat flux growth up to 60-100 kW/m 2 leads to increasing minimum values of the heat transfer factor in the supercritical zone, and a further heat flux growth - to their reduction. The conclusion is made that the value of heat flux in the undercritical zone affects convective heat transfer in the supercritical zone mainly through changing the value of critical vapour content

  5. Heat and mass transfer in particulate suspensions

    CERN Document Server

    Michaelides, Efstathios E (Stathis)

    2013-01-01

    Heat and Mass Transfer in Particulate Suspensions is a critical review of the subject of heat and mass transfer related to particulate Suspensions, which include both fluid-particles and fluid-droplet Suspensions. Fundamentals, recent advances and industrial applications are examined. The subject of particulate heat and mass transfer is currently driven by two significant applications: energy transformations –primarily combustion – and heat transfer equipment. The first includes particle and droplet combustion processes in engineering Suspensions as diverse as the Fluidized Bed Reactors (FBR’s) and Internal Combustion Engines (ICE’s). On the heat transfer side, cooling with nanofluids, which include nanoparticles, has attracted a great deal of attention in the last decade both from the fundamental and the applied side and has produced several scientific publications. A monograph that combines the fundamentals of heat transfer with particulates as well as the modern applications of the subject would be...

  6. Heat pipes for temperature control

    International Nuclear Information System (INIS)

    Groll, M.

    1978-01-01

    Heat pipes have known for years as effective constructional elements for temperature control. With the aid of special techniques (gas, liquid, steam, and voltage control), special operating characteristics can be obtained, e.g. variable heat conduction or diode behaviour. Their main field of application is in spacecraft technology and in nuclear technology in the isothermalisation of irradiation capsules. The different control techniques are presented and critically evaluated on the basis of characteristic properties like heat transfer capacity, volume and mass requirements, complexity of structure and production, reliability, and temperature control characteristics. Advantages and shortcomings of the different concepts are derived and compared. The state of the art of these control techniques is established on the basis of four development levels. Finally, the necessity and direction of further R + D activities are discussed, and suggestions are made for further work. (orig./HP) [de

  7. Heat pipe heat exchanger for heat recovery in air conditioning

    Energy Technology Data Exchange (ETDEWEB)

    Abd El-Baky, Mostafa A.; Mohamed, Mousa M. [Mechanical Power Engineering Department, Faculty of Engineering, Minufiya University, Shebin El-Kom (Egypt)

    2007-03-15

    The heat pipe heat exchangers are used in heat recovery applications to cool the incoming fresh air in air conditioning applications. Two streams of fresh and return air have been connected with heat pipe heat exchanger to investigate the thermal performance and effectiveness of heat recovery system. Ratios of mass flow rate between return and fresh air of 1, 1.5 and 2.3 have been adapted to validate the heat transfer and the temperature change of fresh air. Fresh air inlet temperature of 32-40{sup o}C has been controlled, while the inlet return air temperature is kept constant at about 26{sup o}C. The results showed that the temperature changes of fresh and return air are increased with the increase of inlet temperature of fresh air. The effectiveness and heat transfer for both evaporator and condenser sections are also increased to about 48%, when the inlet fresh air temperature is increased to 40{sup o}C. The effect of mass flow rate ratio on effectiveness is positive for evaporator side and negative for condenser side. The enthalpy ratio between the heat recovery and conventional air mixing is increased to about 85% with increasing fresh air inlet temperature. The optimum effectiveness of heat pipe heat exchanger is estimated and compared with the present experimental data. The results showed that the effectiveness is close to the optimum effectiveness at fresh air inlet temperature near the fluid operating temperature of heat pipes. (author)

  8. Nonazeotropic Heat Pump

    Science.gov (United States)

    Ealker, David H.; Deming, Glenn

    1991-01-01

    Heat pump collects heat from water circulating in heat-rejection loop, raises temperature of collected heat, and transfers collected heat to water in separate pipe. Includes sealed motor/compressor with cooling coils, evaporator, and condenser, all mounted in outer housing. Gradients of temperature in evaporator and condenser increase heat-transfer efficiency of vapor-compression cycle. Intended to recover relatively-low-temperature waste heat and use it to make hot water.

  9. Heat transfer: Pittsburgh 1987

    International Nuclear Information System (INIS)

    Lyczkowski, R.W.

    1987-01-01

    This book contains papers divided among the following sections: Process Heat Transfer; Thermal Hydraulics and Phase Change Phenomena; Analysis of Multicomponent Multiphase Flow and Heat Transfer; Heat Transfer in Advanced Reactors; General Heat Transfer in Solar Energy; Numerical Simulation of Multiphase Flow and Heat Transfer; High Temperature Heat Transfer; Heat Transfer Aspects of Severe Reactor Accidents; Hazardous Waste On-Site Disposal; and General Papers

  10. Industrial waste heat for district heating

    International Nuclear Information System (INIS)

    Heitner, K.L.; Brooks, P.P.

    1982-01-01

    Presents 2 bounding evaluations of industrial waste heat availability. Surveys waste heat from 29 major industry groups at the 2-digit level in Standard Industrial Codes (SIC). Explains that waste heat availability in each industry was related to regional product sales, in order to estimate regional waste heat availability. Evaluates 4 selected industries at the 4-digit SIC level. Finds that industrial waste heat represents a significant energy resource in several urban areas, including Chicago and Los Angeles, where it could supply all of these areas residential heating and cooling load. Points out that there is a strong need to evaluate the available waste heat for more industries at the 4-digit level. Urges further studies to identify other useful industrial waste heat sources as well as potential waste heat users

  11. Automation of heating system with heat pump

    OpenAIRE

    Ferdin, Gašper

    2016-01-01

    Because of high prices of energy, we are upgrading our heating systems with newer, more fuel efficient heating devices. Each new device has its own control system, which operates independently from other devices in a heating system. With a relatively low investment costs in automation, we can group devices in one central control system and increase the energy efficiency of a heating system. In this project, we show how to connect an oil furnace, a sanitary heat pump, solar panels and a heat p...

  12. Nuclear criticality safety guide

    International Nuclear Information System (INIS)

    Pruvost, N.L.; Paxton, H.C.

    1996-09-01

    This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators

  13. Nuclear criticality safety guide

    Energy Technology Data Exchange (ETDEWEB)

    Pruvost, N.L.; Paxton, H.C. [eds.

    1996-09-01

    This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

  14. A Review of Wettability Effect on Boiling Heat Transfer Enhancement

    International Nuclear Information System (INIS)

    Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong

    2012-01-01

    Critical heat flux (CHF) and nucleate boiling heat transfer coefficient (NBHTC) are the key parameters characterizing pool boiling heat transfer. These variables are complicatedly related to thermal-hydraulic parameters of surface wettability, nucleation site density, bubble departure diameter and frequency, to mention a few. In essence, wettability effect on pool boiling heat transfer has been a major fuel to enhance the CHF. Often, however, the improved wettability effect hinders the nucleate boiling. Thus a comprehensive review of such wettability effect may enlighten a further study in this boiling heat transfer area. Phan et al. described surface wettability effects on boiling heat transfer

  15. Critical Curriculum Theory and Slow Ecopedagogical Activism

    Science.gov (United States)

    Payne, Phillip G.

    2015-01-01

    Enacting a critical environmental education curriculum theory with 8- to 9-year-old children in 1978 is now "restoried" in a "history of the present/future" like "case study" for prosecuting five interrelated problems confronting progress in environmental education and its research. They are: the intense heat of the…

  16. Intermittent heating of buildings

    Energy Technology Data Exchange (ETDEWEB)

    Kohonen, K

    1983-02-01

    Conditions for intermittent heating of buildings are considered both theoretically and experimentally. Thermal behaviour of buildings adn rooms in intermittent heating is simulated by a program based on the convective heat balance equation and by simplified RC-models. The preheat times and the heating energy savings compared with continuous heating are presented for typical lightweight, mediumweight and heavyweight classroom and office modules. Formulaes for estimating the oversizing of the radiator network, the maximum heat output of heat exchangers in district heating and the efficiency of heating boilers in intermittent heating are presented. The preheat times and heating energy savings with different heating control systems are determined also experimentally in eight existing buildings. In addition some principles for the planning and application of intermittent heating systems are suggested.

  17. Heat pump technology

    CERN Document Server

    Von Cube, Hans Ludwig; Goodall, E G A

    2013-01-01

    Heat Pump Technology discusses the history, underlying concepts, usage, and advancements in the use of heat pumps. The book covers topics such as the applications and types of heat pumps; thermodynamic principles involved in heat pumps such as internal energy, enthalpy, and exergy; and natural heat sources and energy storage. Also discussed are topics such as the importance of the heat pump in the energy industry; heat pump designs and systems; the development of heat pumps over time; and examples of practical everyday uses of heat pumps. The text is recommended for those who would like to kno

  18. Cryogenic heat transfer

    CERN Document Server

    Barron, Randall F

    2016-01-01

    Cryogenic Heat Transfer, Second Edition continues to address specific heat transfer problems that occur in the cryogenic temperature range where there are distinct differences from conventional heat transfer problems. This updated version examines the use of computer-aided design in cryogenic engineering and emphasizes commonly used computer programs to address modern cryogenic heat transfer problems. It introduces additional topics in cryogenic heat transfer that include latent heat expressions; lumped-capacity transient heat transfer; thermal stresses; Laplace transform solutions; oscillating flow heat transfer, and computer-aided heat exchanger design. It also includes new examples and homework problems throughout the book, and provides ample references for further study.

  19. Future heat supply of our cities. Heating by waste heat

    Energy Technology Data Exchange (ETDEWEB)

    Brachetti, H E [Stadtwerke Hannover A.G. (Germany, F.R.); Technische Univ. Hannover (Germany, F.R.))

    1976-08-01

    The energy-price crisis resulted in structural changes of the complete energy supply and reactivated the question of energy management with respect to the optimum solution of meeting the energy requirements for space heating. Condensation power plants are increasingly replaced by thermal stations, the waste heat of which is used as so-called district heat. Thermal power stations must be situated close to urban areas. The problem of emission of harmful materials can partly be overcome by high-level emission. The main subject of the article, however, is the problem of conducting and distributing the heat. The building costs of heat pipeline systems and the requirements to be met by heat pipelines such as strength, heat insulation and protection against humidity and ground water are investigated.

  20. Heat-Related Illnesses

    Science.gov (United States)

    ... Share this! EmergencyCareForYou » Emergency 101 » Heat-Related Illnesses Heat-Related Illnesses Dr. Glenn Mitchell , Emergency physician at ... about heat cramps and heat stroke and exhaustion. Heat Cramps Symptoms include muscle spasms, usually in the ...

  1. Heat-Related Illnesses

    Medline Plus

    Full Text Available ... Share this! EmergencyCareForYou » Emergency 101 » Heat-Related Illnesses Heat-Related Illnesses Dr. Glenn Mitchell , Emergency physician at ... about heat cramps and heat stroke and exhaustion. Heat Cramps Symptoms include muscle spasms, usually in the ...

  2. Heat transfer in heterogeneous propellant combustion systems

    International Nuclear Information System (INIS)

    Brewster, M.Q.

    1992-01-01

    This paper reports that heat transfer plays an important role in several critical areas of heterogeneous, solid-propellant combustion systems. These areas include heat feedback to the propellant surface, heat transfer between burning aluminum droplets and their surroundings, heat transfer to internal insulation systems, and heat transfer to aft-end equipment. Gas conduction dominates heat feedback to the propellant surface in conventional ammonium perchlorate (AP) composite propellants, although particle radiative feedback also plays a significant role in combustion of metalized propellants. Particle radiation plays a dominant role in heat transfer to internal insulation, compared with that of convection. However, conduction by impingement of burning aluminum particles, which has not been extensively studied, may also be significant. Radiative heat loss plays an important role in determining the burning rate of molten aluminum particles due to a highly luminous, oxide particle-laden, detached flame envelope. Radiation by aluminum oxide smoke particles also plays a dominant role in heat transfer from the exhaust plume to aft-end equipment. Uncertainties in aluminum oxide particle-size distribution and optical properties still make it difficult to predict radiative plume heat transfer accurately from first principles

  3. Absorption heat pump system

    Science.gov (United States)

    Grossman, G.

    1982-06-16

    The efficiency of an absorption heat pump system is improved by conducting liquid from a second stage evaporator thereof to an auxiliary heat exchanger positioned downstream of a primary heat exchanger in the desorber of the system.

  4. Heat Related Illnesses

    National Research Council Canada - National Science Library

    Carter, R; Cheuvront, S. N; Sawka, M. N

    2006-01-01

    .... The risk of serious heat illness can be markedly reduced by implementing a variety of countermeasures, including becoming acclimated to the heat, managing heat stress exposure, and maintaining hydration...

  5. Heat Roadmap Europe 1

    DEFF Research Database (Denmark)

    Connolly, David; Mathiesen, Brian Vad; Østergaard, Poul Alberg

    2012-01-01

    Heat Roadmap Europe (Pre-study 1) investigates the role of district heating in the EU27 energy system by mapping local conditions across Europe, identifying the potential for district heating expansion, and subsequently simulating the potential resource in an hourly model of the EU27 energy system....... In 2010, approximately 12% of the space heating demand in Europe is met by district heating, but in this study four alternative scenarios are considered for the EU27 energy system: 1. 2010 with 30% district heating 2. 2010 with 50% district heating 3. 2030 with 30% district heating 4. 2050 with 50......% district heating These scenarios are investigated in two steps. Firstly, district heating replaces individual boilers by converting condensing power plants to combined heat and power plants (CHP) to illustrate how district heating improves the overall efficiency of the energy system. In the second step...

  6. Multidimensional Heat Conduction

    DEFF Research Database (Denmark)

    Rode, Carsten

    1998-01-01

    Analytical theory of multidimensional heat conduction. General heat conduction equation in three dimensions. Steay state, analytical solutions. The Laplace equation. Method of separation of variables. Principle of superposition. Shape factors. Transient, multidimensional heat conduction....

  7. Oscillating heat pipes

    CERN Document Server

    Ma, Hongbin

    2015-01-01

    This book presents the fundamental fluid flow and heat transfer principles occurring in oscillating heat pipes and also provides updated developments and recent innovations in research and applications of heat pipes. Starting with fundamental presentation of heat pipes, the focus is on oscillating motions and its heat transfer enhancement in a two-phase heat transfer system. The book covers thermodynamic analysis, interfacial phenomenon, thin film evaporation,  theoretical models of oscillating motion and heat transfer of single phase and two-phase flows, primary  factors affecting oscillating motions and heat transfer,  neutron imaging study of oscillating motions in an oscillating heat pipes, and nanofluid’s effect on the heat transfer performance in oscillating heat pipes.  The importance of thermally-excited oscillating motion combined with phase change heat transfer to a wide variety of applications is emphasized. This book is an essential resource and learning tool for senior undergraduate, gradua...

  8. Heating in toroidal plasmas

    International Nuclear Information System (INIS)

    Knoepfel, H.; Mazzitelli, G.

    1984-01-01

    The article is a rather detailed report on the highlights in the area of the ''Heating in toroidal plasmas'', as derived from the presentations and discussions at the international symposium with the same name, held in Rome, March 1984. The symposium covered both the physics (experiments and theory) and technology of toroidal fusion plasma heating. Both large fusion devices (either already in operation or near completion) requiring auxiliary heating systems at the level of tens of megawatts, as well as physics of their heating processes and their induced side effects (as studied on smaller devices), received attention. Substantial progress was reported on the broad front of auxiliary plasma heating and Ohmic heating. The presentation of the main conclusions of the symposium is divided under the following topics: neutral-beam heating, Alfven wave heating, ion cyclotron heating, lower hybrid heating, RF current drive, electron cyclotron heating, Ohmic heating and special contributions

  9. Rhenium: a rare metal critical in modern transportation

    Science.gov (United States)

    John, David A.

    2015-01-01

    Rhenium is a silvery-white, metallic element with an extremely high melting point (3,180 degrees Celsius) and a heat-stable crystalline structure, making it exceptionally resistant to heat and wear. Since the late 1980s, rhenium has been critical for superalloys used in turbine blades and in catalysts used to produce lead-free gasoline.

  10. Design of a cavity heat pipe receiver experiment

    Science.gov (United States)

    Schneider, Michael G.; Brege, Mark H.; Greenlee, William J.

    1992-01-01

    A cavity heat pipe experiment has been designed to test the critical issues involved with incorporating thermal energy storage canisters into a heat pipe. The experiment is a replication of the operation of a heat receiver for a Brayton solar dynamic power cycle. The heat receiver is composed of a cylindrical receptor surface and an annular heat pipe with thermal energy storage canisters and gaseous working fluid heat exchanger tubes surrounding it. Hardware for the cavity heat pipe experiment will consist of a sector of the heat pipe, complete with gas tube and thermal energy storage canisters. Thermal cycling tests will be performed on the heat pipe sector to simulate the normal energy charge/discharge cycle of the receiver in a spacecraft application.

  11. Regenerative adsorbent heat pump

    Science.gov (United States)

    Jones, Jack A. (Inventor)

    1991-01-01

    A regenerative adsorbent heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system and at least a portion of the heat of adsorption. A series of at least four compressors containing an adsorbent is provided. A large amount of heat is transferred from compressor to compressor so that heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

  12. Heat Roadmap Europe

    DEFF Research Database (Denmark)

    David, Andrei; Mathiesen, Brian Vad; Averfalk, Helge

    2017-01-01

    The Heat Roadmap Europe (HRE) studies estimated a potential increase of the district heating (DH) share to 50% of the entire heat demand by 2050, with approximately 25–30% of it being supplied using large-scale electric heat pumps. This study builds on this potential and aims to document that suc......The Heat Roadmap Europe (HRE) studies estimated a potential increase of the district heating (DH) share to 50% of the entire heat demand by 2050, with approximately 25–30% of it being supplied using large-scale electric heat pumps. This study builds on this potential and aims to document...

  13. Prevention of criticality accidents

    International Nuclear Information System (INIS)

    Canavese, S.I.

    1982-01-01

    These notes used in the postgraduate course on Radiological Protection and Nuclear Safety discuss macro-and microscopic nuclear constants for fissile materials systems. Critical systems: their definition; criteria to analyze the critical state; determination of the critical size; analysis of practical problems about prevention of criticality. Safety of isolated units and of sets of units. Application of standards. Conception of facilities from the criticality control view point. (author) [es

  14. Burnout heat flux in natural flow boiling

    International Nuclear Information System (INIS)

    Helal, M.M.; Darwish, M.A.; Mahmoud, S.I.

    1978-01-01

    Twenty runs of experiments were conducted to determine the critical heat flux for natural flow boiling with water flowing upwards through annuli of centrally heated stainless steel tube. The test section has concentric heated tube of 14mm diameter and heated lengthes of 15 and 25 cm. The outside surface of the annulus was formed by various glass tubes of 17.25, 20 and 25.9mm diameter. System pressure is atmospheric. Inlet subcooling varied from 18 to 5 0 C. Obtained critical heat flux varied from 24.46 to 62.9 watts/cm 2 . A number of parameters having dominant influence on the critical heat flux and hydrodynamic instability (flow and pressure oscillations) preceeding the burnout have been studied. These parameters are mass flow rate, mass velocity, throttling, channel geometry (diameters ratio, length to diameter ratio, and test section length), and inlet subcooling. Flow regimes before and at the moments of burnout were observed, discussed, and compared with the existing physical model of burnout

  15. Waveguide circuit for LHRF heating in 'JT-60'

    International Nuclear Information System (INIS)

    Uehara, Kazuya; Saegusa, Mikio; Mizuno, Takenori; Sano, Keigo; Hara, Mitsuru; Oishi, Isamu; Kanai, Takao.

    1985-01-01

    As the heating method for attaining the critical condition in the critical plasma experiment apparatus 'JT-60' in the Japan Atomic Energy Research Institute, in addition to Joule heating, as the second heating method, neutral beam injection heating and high frequency heating have been adopted. For this high frequency heating, several tens to 200 MHz band of ICRF heating, several hundreds MHz to several GHz band of LHRF heating and several tens to 200 GHz band of ECR heating were considered, and in the JT-60, 100 MHz band (ICRF) and 2 GHz band (LHRF) have been adopted. Furukawa Electric Co., Ltd. has engaged in the development and manufacture of the waveguides of transmission system used for this high frequency heating through NEC Corp. This high frequency heating is to heat plasma by injecting high frequency radio waves into plasma proper, and reaches 10 MW for the whole high frequency heating. The system efficiently transmitting the radio waves of large power from a Klystron as a high frequency source to the JT-60 is the transmission system. The outline of the waveguides of the 2 GHz band transmission system and the individual performance of respective waveguides are reported. (Kako, I.)

  16. Heat pumps in district heating networks

    DEFF Research Database (Denmark)

    Ommen, Torben Schmidt; Markussen, Wiebke Brix; Elmegaard, Brian

    constraints limit the power plants. Efficient heat pumps can be used to decouple the constraints of electricity and heat production, while maintaining the high energy efficiency needed to match the politically agreed carbon emission goals. The requirements in terms of COP, location, capacity and economy...... and strategic planning in the energy sector. The paper presents a case study of optimal implementation of heat pumps in the present energy system of the Copenhagen area. By introduction of the correct capacity of heat pumps, a 1,6 % reduction in fuel consumption for electricity and heat production can...

  17. Desolvation of polymers by ultrafast heating: Influence of hydrophilicity

    Science.gov (United States)

    Sun, Si Neng; Urbassek, Herbert M.

    2010-10-01

    Using molecular-dynamics simulation, we investigate the consequences of ultrafast laser-induced heating of a small water droplet containing a solvated polymer. Two polymers are studied: polyethylene as an example of a hydrophobic, and polyketone as an example of a hydrophilic polymer. In both cases, when the droplet is heated below the critical temperature of water, strong water evaporation is started, but the polymer remains in contact with a central water cluster. However, upon heating beyond the critical temperature, the hydrophilic polyethylene becomes completely desolvated, while polyketone still remains solvated. We analyze this behavior in terms of the intermolecular interactions and of the expansion dynamics of the heated droplet.

  18. Criticality safety (prospect of study in NUCEF)

    International Nuclear Information System (INIS)

    Itagaki, Masafumi

    1996-01-01

    Experimental studies of criticality safety are under way using STACY and TRACY in NUCEF. Collection of fundamental data on criticality in a solution system is undergoing with STACY to confirm that the likelihood of criticality safety in the system constructed on the assumption of apparatuses in a reprocessing plant is enough large. Whereas some experiments simulating criticality accidents in a reprocessing plant using TRACY were designed to investigate the behaviors of fuel solution and radioactive matters in order to clarify whether it is possible to safely shut them in the facility even if a critical accident occurs. Both STACY and TRACY reached the criticality in 1995. Up to now a series of criticality experiments have been done using STACY with a core tank φ60 cm and the first periodical examination is now under way. On the other hand, we have a plan using TRACY to investigate the behaviors of nuclear heat solution at a criticality accident, and the releasing, transfer and deposition of radioactive materials. After reaching the criticality for the first, the performance verification test has been conducted. The full-scale study using TRACY is planned to begin in the second half of 1996. (M.N.)

  19. Solar heat storages in district heating networks

    Energy Technology Data Exchange (ETDEWEB)

    Ellehauge, K. (Ellehauge og Kildemoes, AArhus (DK)); Engberg Pedersen, T. (COWI A/S, Kgs. Lyngby (DK))

    2007-07-15

    This report gives information on the work carried out and the results obtained in Denmark on storages for large solar heating plants in district heating networks. Especially in Denmark the share of district heating has increased to a large percentage. In 1981 around 33% of all dwellings in DK were connected to a district heating network, while the percentage in 2006 was about 60% (in total 1.5 mio. dwellings). In the report storage types for short term storage and long term storages are described. Short term storages are done as steel tanks and is well established technology widely used in district heating networks. Long term storages are experimental and used in connection with solar heating. A number of solar heating plants have been established with either short term or long term storages showing economy competitive with normal energy sources. Since, in the majority of the Danish district heating networks the heat is produced in co-generation plants, i.e. plants producing both electricity and heat for the network, special attention has been put on the use of solar energy in combination with co-generation. Part of this report describes that in the liberalized electricity market central solar heating plants can also be advantageous in combination with co-generation plants. (au)

  20. Heat pipes for ground heating and cooling

    Energy Technology Data Exchange (ETDEWEB)

    Vasiliev, L L

    1988-01-01

    Different versions of heat pipe ground heating and cooling devices are considered. Solar energy, biomass, ground stored energy, recovered heat of industrial enterprises and ambient cold air are used as energy and cold sources. Heat pipe utilization of air in winter makes it possible to design accumulators of cold and ensures deep freezing of ground in order to increase its mechanical strength when building roadways through the swamps and ponds in Siberia. Long-term underground heat storage systems are considered, in which the solar and biomass energy is accumulated and then transferred to heat dwellings and greenhouses, as well as to remove snow from roadways with the help of heat pipes and solar collectors.