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Study on thermal hydraulic characteristics under startup of SCWR
Abstract
To investigate the startup characteristics of the supercritical pressure water-cooled reactor (SCWR) system, a complete startup system model of the SCWR was established with the analysis code SCTRAN, based on the high-performance light water reactor (HPLWR) steam cycle and SCWR circulation startup loop. The correctness of the model was verified in comparison with the steady-state parameters of the steam cycle of the HPLWR. A sliding pressure startup procedure with a circulation loop that employs a control system was analyzed, and the transient performances of the core, steam drum, turbine, reheaters, steam extraction and heaters at each stage were obtained. The calculation results showed that the startup sequence and startup thermal parameters could fully meet the expectations: the system starts up stably and the core remains in the single phase; the inlet steam of the turbine stays superheated; the core inlet temperature can reach 280 °C after the adoption of the high-pressure and low-pressure heaters; and the inlet pressure of the high-pressure turbine can be kept constant. During the startup process, the maximum cladding surface temperature (MCST) remains below the limit of 650 °C, suggesting that the entire startup procedure is safe and reliable.
This book of proceedings compiles the articles of the European Nuclear Young Generation Forum 2021. The event was hosted by the Spanish Young Generation Network and took place in Tarragona. Eventually, more than 150 professionals attended the physical event, whereas another 100 followed the forum from home due to the COVID19 pandemic restrictions. The motto of the Technical Program of ENYGF21 was “It is time to look ahead”, a claim of the vital role to play for the young generations in dealing with all the challenges that nuclear will face this decade. This motto shaped our technical program and the content of the book.
Supercritical Water-cooled Reactor (SCWR) is the only water-cooled reactor proposed for the Generation IV nuclear system. The dynamics of its steam temperature are strongly influenced by the reactor power and also with a high degree of nonlinearity. It is difficult to control using the traditional PI control only. In this paper, moving boundary method is adopted to construct the dynamic model of the steam temperature in a Canadian SCWR. The design of the feedforward controller and fuzzy adaptive PI controller is based on the dynamic characteristics of the steam temperature. Through numerical simulation, it is found that the feedforward control can reduce the effect of the reactor power on the steam temperature. The control parameters can be tuned online using fuzzy adaptive control and the control performance can be improved. Steam temperature can promptly stabilize and the variation is reduced using the designed fuzzy control system. It is concluded that the control performance is much better than that with PI controller and the control requirements of Canadian SCWR are satisfied. © 2016, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.
The turbine of SCWR is supercritical and impulse which rotors are made of materials maturely used in the turbines of supercritical thermal power plants. The turbine's seal steam is from clean steam source. In order to realize the sliding pressure startup, the thermal system includes a startup and shut down system which supply sufficient flow for the reactor in startup and shut down phase. This paper describes the turbine and thermal system of SCWR.
Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain an understanding of the conceptual design elements and specific analysis methods for supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference for engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology. © Springer Science+Business Media, LLC 2010. All rights reserved.
A 1000 MW supercritical water-cooled reactor (SCWR) design concept CSR1000 with independent intellectual property has been developed based on the existing technologies of ABWR, PWR and SCWR. In order to simplify the structural design and obtain more uniform moderation, the CSR1000 standard fuel assembly cluster is built of 4 small size square fuel assemblies, and each of them has only one water rod. Furthermore, a two-pass coolant flow scheme is adopted to increase the coolant flow stability and core averaged outlet temperature. The core consists of 157 fuel assembly clusters, and the out-in loading pattern is employed. After core conceptual design evaluation, these key parameters of the core such as cycle length, average discharge burnup, core coolant averaged outlet temperature, maximum fuel cladding temperature and maximum linear power density have been presented.
This paper presents a phenomenologically based model for subcooled nucleate boiling. It is shown that the current state of understanding is such that mechanistic models cannot be uniquely determined; however, comparisons of the mechanistic model presented with existing data indicate good agreement.
The complete set of eigenvalues and eigenfunctions for the classical Graetz problem is presented and the solution is extended to cover arbitrary wall temperature or heat-flux variations.
The modification to ATHLET-SC code is introduced in this paper, which realizes the simulation of trans-critical transients using two-phase model. With the modified code, the thermal-hydraulic dynamic behavior of the mixed SCWR core during the startup process is simulated. The startup process is similar to the design of SCLWR-H sliding pressure startup. The results show that maximum temperature of cladding-surface does not exceed 650°C in the whole startup process, and the sudden change of water properties in the trans-critical transients will not cause harmful influence to the heat transfer of the fuel cladding.
The published data on heat transfer in turbulent flow of air, water or superheated steam in tube bundles are reduced to a single set and are assessed from a single point of view by a navel technique. New and reliable recommendations for heat transfer calculations were thereby obtained.
Safety analysis is one of the chief difficulties during the research and design of SCWRs. Currently, the development of the SCWR safety analysis code is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs, has been developed based on RETRAN-02, the best estimate code used for safety analysis of Light Water Reactors. The ability of SCTRAN to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with the APROS and RELAP5-3D codes, respectively. The results show that the SCTRAN code developed in this study is capable of performing safety analysis for SCWRs, and the results are reliable. SCTRAN developed in this study is of great value and engineering significance, and could be utilized in the Chinese SCWR research and development.
The startup system of a supercritical pressure light water cooled fast reactor (Super FR) is studied by time dependent thermal-hydraulic analysis. The plant analysis code is developed based on an innovative upward flow pattern in all the assemblies of the Super FR. A recirculation system consisting of a steam drum, a circulation pump, and a heat exchanger is used for the startup. Detailed procedures are performed and the maximum cladding surface temperature (MCST) at rated power, 640 °C, is used as the criterion. Firstly a small constant nuclear power is used for rising the core feed water temperature to be 280 °C through the recirculation system. Secondly, pressurization is done in the recirculation system from atmospheric to operating pressure, 25 MPa, by raising the power. Thirdly, line-switching from recirculation mode to once-through direct-cycle is performed while turbines are started by supercritical steam at supercritical pressure. Finally the power is raised to be 100% of power followed by raising the flow rate. During pressurization the heat flux margin is large due to low power used for pressurization and the MCST is much lower than the criterion. The MCST is not sensitive to the inlet temperature, the flow rate, and the gap volume of the core because of high flow to power ratio. Smaller dimension of steam drum can be used for pressurization stably. The MCST satisfies the criterion both during subcritical pressure and during power-raising.
A model has been developed to calculate the heat transfer coefficients from the fuel rods to the steam-droplet mixture typical of Boiling Water Reactors under Emergency Core Cooling System (ECCS) operation conditions during a postulated loss-of-coolant accident. The model includes the heat transfer by convection to the vapor, the radiation from the surfaces to both the water droplets and the vapor, and the effects of droplet evaporation. The combined convection and radiation heat transfer coefficient can be evaluated with respect to the characteristic droplet size. Calculations of the heat transfer coefficient based on the droplet sizes obtained from the existing literature are consistent with those determined empirically from the Full-Length-Emergency-Cooling-Heat-Transfer (FLECHT) program. The present model can also be used to assess the effects of geometrical distortions (or deviations from nominal dimensions) on the heat transfer to the cooling medium in a rod bundle.
Supercritical-pressure light water-cooled reactors (SCRs) are innovative systems aimed at high efficiency and cost reduction. The once-through direct-cycle plant system is the leading system of fossil-fired power plants (FPPs). Estimates of the coolability and necessary sizes of the SCR startup systems, sequences, and required equipment for startup are investigated with reference to supercritical FPPs. There are two types of supercritical boilers. One is a constant pressure boiler, and the other is a variable pressure boiler.
First, startup of the constant pressure boiler is examined. The reactor starts at a supercritical pressure. A startup bypass system consisting of a flash tank and pressure-reducing valves is required. Second, startup of the variable pressure boiler is investigated. The reactor starts at a subcritical pressure, and the pressure increases with the load. A steam-water separator and a drain tank are required for startup.
The results of computer calculations show that with both constant pressure and variable pressure startup, the peak cladding temperature does not exceed the operating limit through startup, and both startup sequences are feasible. The sizes of the components required for the startup systems are assessed. To simplify the plant system and to reduce the component size, variable pressure startup with steam separators in the bypass line appears desirable.
The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.
A simple dimensionless correlation for predicting heat-transfer coefficients during film condensation inside pipes is presented. It has been verified by comparison with a wide variety of experimental data. These include fluids water, R-11, R-12, R-22, R-113, methanol, ethanol, benzene, toluene, and trichloroethylene condensing in horizontal, vertical, and inclined pipes of diameters ranging from 7 to 40mm. One data set for condensation inside an annulus has also been analyzed. The range of parameters covered includes reduced pressures from 0.002 to 0.44, saturation temperatures from 21 to 310°C, vapor velocities from 3 to 300m/s, vapor qualities from 0 to 100%, mass flux 39000-758 000kg/m2 h, heat flux from 158 to 1 893000W/m2, all liquid Reynolds numbers from 100 to 63 000, and liquid Prandtl numbers from 1 to 13. The mean deviation for the 474 data points analyzed was found to be 15.4%.
Taylor-Helmholtz Hydrodynamic Instability and its significance with
regard to film boiling heat transfer from a horizontal surface is discussed. It
is shown that near the minimum film-boiling heat flux, the bubble spacing and
growth rate is determined by Taylor Instability neglecting the effect of fluid
depth and viscosity. Utilizing a simplified geometrical model, an analytical
expression for the heat-transfer coefficient near the minimum in film pool
boiling from a honizontal surface was derived. Combining this equation with the
available correlation for the minimum heat flux yields an analytical equation for
the temperature difference at the minimum, which defines the location of the
minimum point. The above equations agree with the available experimental
measurements made on n-pentane and carbon tetracliloride within plus or minus 10
per cent. (auth)
A phenomenological model for heat transfer in the post-critical-heat flux regime was developed on the basis of additive vapor and liquid transport mechanisms at the hot wall. The vapor heat transfer was derived from momentum-transfer analogy. A semi-empirical correlation for thermodynamic non-equilibrium was used to predict actual-qualities and superheated vapor temperatures. The liquid heat transfer was modeled as a transient three step process, involving liquid superheating at the hot wall, bubble nucleation, growth, and disruption of liquid mass, and evaporation of the residual liquid film. The area fraction of wall cooled by liquid was taken to decrease exponentially with increasing wall superheat. The overall correlation permits prediction of local heat flux, actual quality, and vapor temperature using values of local pressure, mass flow rate, equilibrium quality, and wall temperature as input information. In comparison with data for water boiling in tubes, from eight different experimental sources, the average deviation between calculated and experimentally derived values were 3.7% in actual-quality (for 2186 data points), and 14.2% in heat flux (for 3641 data points, at pressures up to 2200 psia).
An additive mechanism of micro- and macroconvective heat transfer was formulated to represent boiling heat transfer with net vapor generation to saturated, nonmetallic fluids in convective flow. Interaction of the two mechanisms was taken into account by means of two dimensionless functions: an effective two-phase Reynolds number function, F, and a bubble-growth suppression function, S. F was obtained as a function of the Martinelli parameter by both empirical correlation of heat transfer data and a momentum-analogy analysis. S was obtained as an empirical function of the two-phase Reynolds number. The correlation was tested with available data for water and organic fluids. The average deviation between calculated and measured boiling coefficients for all (over 600) data points from ten experimental cases was ± 12%.
Analytical expressions for bubble radii and growth rates derived by the authors are applied in an analysis of surface boiling at high heat transfer rates. It is shown that the product of bubble radius and radial velocity is a constant, independent of the bubble radius. This circumstance permits the formulation of a Reynolds number for the flow in the thin superheated liquid layer adjacent to the heating surface. The result of the analysis is then applied to maximal heat transfer rates in pool boiling.
CHF look-up tables are used widely for the prediction of the critical heat flux (CHF). The CHF look-up table is basically a normalized data bank for a vertical 8 mm water-cooled tube. The 2006 CHF look-up table is based on a database containing more than 30,000 data points and provides CHF values at 24 pressures, 20 mass fluxes, and 23 qualities, covering the full range of conditions of practical interest. In addition, the 2006 CHF look-up table addresses several concerns with respect to previous CHF look-up tables raised in the literature. The major improvements of the 2006 CHF look-up table are:•An enhanced quality of the database (improved screening procedures, removal of clearly identified outliers and duplicate data).•An increased number of data in the database (an addition of 33 recent data sets).•A significantly improved prediction of CHF in the subcooled region and the limiting quality region.•An increased number of pressure and mass flux intervals (thus increasing the CHF entries by 20% compared to the 1995 CHF look-up table).•An improved smoothness of the look-up table (the smoothness was quantified by a smoothness index).A discussion of the impact of these changes on the prediction accuracy and table smoothness is presented. The 2006 CHF look-up table is characterized by a significant improvement in accuracy and smoothness.
The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density.
An improved look-up table for film-boiling heat-transfer coefficients has been derived for steam–water flow inside vertical tubes. Compared to earlier versions of the look-up table, the following improvements were made:•The database has been expanded significantly. The present database contains 77,234 film-boiling data points obtained from 36 sources.•The upper limit of the thermodynamic quality range was increased from 1.2 to 2.0. The wider range was needed as non-equilibrium effects at low flows can extend well beyond the point where the thermodynamic quality equals unity.•The surface heat flux has been replaced by the surface temperature as an independent parameter.•The new look-up table is based only on fully developed film-boiling data.•The table entries at flow conditions for which no data are available is based on the best of five different film-boiling prediction methods.The new film-boiling look-up table predicts the database for fully developed film-boiling data with an overall rms error in heat-transfer coefficient of 10.56% and an average error of 1.71%. A comparison of the prediction accuracy of the look-up table with other leading film-boiling prediction methods shows that the look-up table results in a significant improvement in prediction accuracy.
An experimental investigation of turbulent natural convection in air is described. The results of this study show good agreement with early investigations and remarkable agreement with the analytical correlation of Bayley [4]: for Rayleigh numbers up to 1012.Extensive measurements of the temperature field indicate a good similarity in the temperature profiles when compared on the basis of the natural coordinate ty. The use of power law temperature profiles is shown to be undesirable for the case of turbulent natural convection.RésuméUne recherche expérimentale de la convection naturelle turbulente dans l'air est décrite. Les résultats de cette étude montrent un bon accord avec des recherches antérieures et un accord remarquable avec la relation théorique de Bayley: (4) pour des nombres de Rayleigh allant jusqu'a 1012. Des mesures complètes du champ de température indiquent une bonne similitude pour les profils de température lorsqu'on les compare en utilisant la coordonnée physique y. L'emploi de profils de températures en puissance n'est pas désirable dans le cas de la convection naturelle turbulente.ZusammenfassungEs wird eine experimentelle Untersuchung der turbulenten freien Konvektion für Luft beschrieben. Die Ergebnisse dieser Arbeit zeigten gute Übereinstimmung mit früheren Untersuch-ungen und bemerkenswerte Übereinstimmung mit der analytischen Korrelation von Bayley [4] für Rayleigh-Zahlen bis 1012. Umfangreiche Messungen des Temperaturfeldes zeigen gute Ähnlichkeit der Temperaturprofile wenn sie auf Grund der natürlichen Koordinate y verglichen werden. Die Verwendung von Temperaturprofilen die einem Potenzgesetz gehorchen erweist sich für den Fall turbulenter freier Konvektion als ungünstig.
Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Mechanical Engineering, 1960. Vita. Includes bibliographical references (leaves 107-110). by John C. Chato. Ph.D.
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