The Experts below are selected from a list of 297 Experts worldwide ranked by ideXlab platform
P L Kirillov - One of the best experts on this subject based on the ideXlab platform.
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Supercritical Water Cooled Reactors
Thermal Engineering, 2008Co-Authors: P L KirillovAbstract:The main advantages of implementation of Water-moderated Water-Cooled power Reactors with supercritical parameters in power engineering are shown.
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changing over to supercritical parameters the way to advance nuclear power stations with Water Cooled Reactors
Thermal Engineering, 2001Co-Authors: P L KirillovAbstract:One promising way to advance a nuclear power station (NPS) with Water-Cooled Reactors is considered. It allows us to considerably simplify the thermal scheme of the NPS, to exclude steam generators and other items of equipment, and, hence, to reduce the capital outlay and the time of the construction of the NPS.
S Koshizuka - One of the best experts on this subject based on the ideXlab platform.
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safety design principle of supercritical Water Cooled Reactors
2004Co-Authors: Y Ishiwatari, S KoshizukaAbstract:This paper summarizes safety of high temperature supercritical pressure light Water Cooled and moderated reactor (SCLWR-H) developed by Univ. Tokyo. Although SCLWR-H is a logical evolution of LWR, the cooling system and safety principle are different from it. SCLWR-H is once-through cooling cycle where all the coolant is led to the turbine. The safety design principle of once-through SCLWR-H is to keep core flow rate while keeping Water inventory is important in LWR. The safety system of SCLWR-H is similar to that of BWR since both Reactors are direct cycle. Cladding temperature is limited at transient instead of heat flux because heat transfer deterioration of single-phase supercritical Water is milder than DNB and dryout of subcritical Water. A set of safety analyses shows that SCLWR-H safety is maintained. Once-through cooling cycle, relatively small pressure change at supercritical pressure and heat conduction to many Water rods characterizes safety behavior of SCLWR-H. Depressurization induces core flow and increases core cooling because high-density Water passively flow from the RPV through the ADS. Such effect of depressurization is advantage in SCLWR-H safety. Decrease in Water inventory by depressurization is not a problem for SCLWR-H because Water inventory is not important safety factor. Another advantagemore » is that ATWS behavior is mild and better than LWR. It is mitigated without scram and depressurization. (authors)« less
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Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor
2002Co-Authors: S Koshizuka, Yuki Ishiwatari, Akifumi YamajiAbstract:The paper describes elements of design consideration of supercritical-pressure, light Water Cooled Reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the Reactors are 1. Compact reactor and turbines due to high specific enthalpy of supercritical Water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water Cooled Reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast reactor core due to small coolant flow rate, potentially easy shift to fast breeder reactor without changing coolant technology. 10.Potential ofmore » producing energy products such as hydrogen and high quality hydro carbons. (authors)« less
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startup thermal considerations for supercritical pressure light Water Cooled Reactors
Nuclear Technology, 2001Co-Authors: Toru Nakatsuka, S KoshizukaAbstract: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 inmore » the bypass line appears desirable.« less
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development of a loca analysis code for the supercritical pressure light Water Cooled Reactors
Annals of Nuclear Energy, 1998Co-Authors: S KoshizukaAbstract:Abstract The supercritical-pressure light Water Cooled Reactors aim at cost reduction by system simplification and higher thermal efficiency, and have flexibility for the fuel cycle due to technical feasibility for various neutron spectrum Reactors. Since loss-of-coolant accident (LOCA) behavior at supercritical pressure conditions cannot be analyzed with the existing codes for the current light Water Reactors, a LOCA analysis code for the supercritical-pressure light Water Cooled reactor is developed in this study. This code, which is named SCRELA, is composed of two parts: the blowdown and reflood analysis modules. The blowdown analysis module is designed based on the homogeneous equilibrium model. The reflood analysis module is modeled by the thermal equilibrium relative velocity model. SCRELA is validated by the REFLA-TRAC code, which is developed in the Japan Atomic Energy Research Institute based on TRAC-PF1. A large break LOCA of the thermal neutron spectrum reactor (SCLWR) is analyzed by SCRELA. The result shows that the peak clad temperature (PCT) is nearly 980°C about 60s after the break and the PCT position is quenched at 170s This means that PCT is sufficiently lower than the safety limit of 1260°C. In conclusion, the developed code shows the safety of SCLWR under the large break LOCA, and is expected to be applied to LOCA analysis of other types of the supercritical-pressure light Water Cooled Reactors.
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supercritical pressure light Water Cooled Reactors for economical nuclear power plants
Progress in Nuclear Energy, 1998Co-Authors: S KoshizukaAbstract:Design studies of supercritical-pressure light-Water-Cooled Reactors (SCLWRs) have been carried out to pursue drastic improvement of the economy of nuclear power generation. The core is Cooled by supercritical Water which is superheated without the phase change. The cooling system is a once-through type; the whole core flow is driven by the feedWater pumps and is directly led to the turbine. No recirculation line is necessary. Besides, steam separators and dryers are not needed. Water rods are used to enhance the moderation and to increase the flow velocity around the fuel rods. The radial peaking factor is satisfactorily reduced by controlling uranium enrichment and gadolinia concentration as well as Water rods. Flattening of the radial power distribution is important to enhance the thermal efficiency. This can be achieved by the coolant density feedback and the out-in refueling pattern. Orificing is also effective to enhance the thermal efficiency. The thermal efficiency is above 40% with stainless steel cladding. Plant control system and safety system are also designed. The core flow should be directly maintained due to the once-through direct cycle. Plant behaviors of large break LOCAs and loss of offsite power are analyzed. Safety criteria are satisfied in both cases. The feasibility of SCLWR is shown.
Mujid S Kazimi - One of the best experts on this subject based on the ideXlab platform.
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coupled neutronic and thermal hydraulic out of phase stability of supercritical Water Cooled Reactors
Nuclear Technology, 2008Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:As the last topic of a series of U.S. reference supercritical Water-Cooled reactor (SCWR) design stability studies, coupled neutronic-thermal-hydraulic out-of-phase stability is analyzed and compared with that of a typical boiling Water reactor (BWR). A modal expansion method based on A modes (reactivity modes) of the neutron kinetic equation is applied, and the first subcritical mode of the neutron dynamics model is coupled with the coolant thermal-hydraulic model. The out-of-phase oscillation of the SCWR is found to be dominated by the reactor thermal hydraulics, whereas the BWR is more sensitive to the coolant density reactivity coefficient because of much stronger neutronic coupling. It is also found that the SCWR stability is sensitive to the details of the core simulation model and the hottest channel dominates the stability. The BWR is less sensitive to the core simulation model since it has much stronger neutronic coupling that is controlled by the whole-core average properties. Power and flow rate sensitivity analysis of the out-of-phase stability was also conducted for both the SCWR and the BWR. The SCWR stability is found to be more sensitive to the operating parameters than the typical BWR. Although the Water rod heating can improve the SCWR out-of-phase stability, it cannot significantly improve the sensitivity feature.
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core wide in phase stability of supercritical Water Cooled Reactors ii comparison with boiling Water Reactors
Nuclear Technology, 2008Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:To compare the stability features of a supercritical Water-Cooled reactor (SCWR) design with that of a typical boiling Water reactor (BWR), a stability analysis model for a typical BWR has been developed in addition to an already-developed model for the SCWR as presented in a companion paper. The homogenous equilibrium two-phase flow model, which is adequate at high pressures, is applied to the BWR stability analysis. The reactor core is simulated by three channels according to the radial power distribution and the inlet orifice coefficients. Similar to the SCWR model, the neutronic kinetic equation is expanded based on A modes (reactivity modes). The model is evaluated based on the Peach Bottom Atomic Power Station stability test data, and the results agree well with the experiment. The SCWR is found to be less sensitive to the coolant density neutronic reactivity coefficient than the typical BWR, since most of the neutronic moderation function is provided by the Water rods, where the density variation is either zero (if the Water rods are insulated) or small (if the Water rods are not insulated). The BWR is found to be less sensitive to changes in power level than the SCWR but has the same sensitivity level to the flow rate as the SCWR. A stability envelope that combines the single-channel and in-phase stability modes is developed. The decay ratios for the SCWR together with those for the typical BWR and the new Economic Simplified Boiling Water Reactor at nominal operational conditions are shown in the map. The stability sensitivity to operating conditions is also shown in the map, by increasing the power to 120% of nominal value and decreasing the flow rate to 80% of nominal value. It is found that the SCWR is more sensitive to the single-channel stability compared to the core-wide in-phase stability for all cases.
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core wide in phase stability of supercritical Water Cooled Reactors i sensitivity to design and operating conditions
Nuclear Technology, 2008Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:Using a three-region supercritical Water flow model, the core-wide in-phase stability of the U.S. reference supercritical Water-Cooled reactor (SCWR) design is investigated. The reactor core is simulated as three channels according to the radial power distribution. A method based on λ modes (reactivity modes) expansion of neutronic kinetic equations is applied. A constant pressure drop boundary condition between the feedWater pump and the turbine control valve is assumed. Cases with and without Water rods heating are studied. It is found that the stability of the U.S. reference SCWR design is sensitive to the flow restrictions in the hot fluid or the steam line. As long as the restriction in the steam line is small, the design will be stable. A pressure loss coefficient of 0.25 is assumed for the exit valve on the steam line in this analysis. With this value, the SCWR is stable with a large margin. It is concluded that the presence of Water rods heating will reduce the stability margin and increase the flow rate sensitivity while maintaining the power sensitivity level. The decay ratios for the three density wave oscillation modes, i.e., single hot channel, coupled neutronic out-of-phase and in-phase, are compared at steady-state conditions. It is found that the single hot channel oscillation mode is the most limiting one in the absence of the Water rods heating, while the in-phase oscillation mode is most limiting in the presence of Water rods heating.
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hot channel stability of supercritical Water Cooled Reactors i steady state and sliding pressure startup
Nuclear Technology, 2007Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:The drastic change of fluid density in the reactor core of a supercritical Water-Cooled reactor (SCWR) gives rise to a concern about density-wave stability. Using a single-channel thermal-hydraulic model, stability boundary maps for the U.S. reference SCWR design have been constructed for both steady state and sliding pressure startup conditions. The supercritical Water flow in the reactor core has been simulated using a three-region model: a heavy fluid with constant density, a mixture of heavy fluid and light fluid similar to a homogeneous-equilibrium two-phase mixture, and a light fluid, which behaves like an ideal gas or superheated steam. Two important nondimensional numbers, namely, a pseudosubcooling number N pSUb and an expansion number N exp , have been identified for the supercritical region. The stability map in the supercritical region is then plotted in the plane made of these two numbers. The U.S. reference SCWR design operates in a stable region with a large margin. Sensitivity studies produced results consistent with the findings of the earlier research done for the subcritical two-phase flow. During the sliding pressure startup of the SCWR, a two-phase steam-Water mixture at subcritical pressure will appear in the reactor core. A nonhomogeneous (e.g., drift-flux) nonequilibrium two-phase flow model was applied. The characteristic equation was numerically integrated, and stability boundary maps were plotted on the traditional subcooling number versus phase change number (or Zuber number) plane. These maps have been used to develop a sliding pressure SCWR startup strategy avoiding thermal-hydraulic flow instabilities.
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hot channel stability of supercritical Water Cooled Reactors ii effect of Water rod heating and comparison with bwr stability
Nuclear Technology, 2007Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:The single hot-channel thermal-hydraulic stability model is expanded to investigate the effects of heat transport from fuel rods and to Water rods on supercritical Water-Cooled reactor (SCWR) stability. Furthermore, the stability margin of the SCWR is compared with that of a typical boiling Water reactor (BWR) by conducting a sensitivity study on operating conditions. The fuel thermal-dynamic effect is studied by coupling a lumped-parameter fuel model with the three-region coolant thermal-hydraulics model. It is found that the fuel heat capacity would dampen the oscillations in the coolant channel and therefore increase the stability of the system. Also, heating of the Water rods, which could be allowed in the core, would improve single-channel stability. The stability sensitivity to power and flow rate conditions is analyzed for the U.S. reference SCWR design and compared with a typical BWR. The SCWR is found to be more sensitive to power and flow rate changes than the typical BWR. The Water rod heating cannot significantly improve this sensitivity feature of the SCWR stability. The traditional stability measure of oscillation amplitude decay ratio does not capture the extent to which a stability margin exists in a particular design of the SCWR. The robustness of stability should be ascertained by examining accommodation of the potential variation and/or uncertainty about the nominal conditions.
Jiyun Zhao - One of the best experts on this subject based on the ideXlab platform.
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Supercritical Water heat transfer for nuclear reactor applications: A review
Annals of Nuclear Energy, 2016Co-Authors: Mohammad Mizanur Rahman, Mehrdad Shahmohammadi Beni, Ho Choi Hei, Ji Dongxu, Weidong He, Jiyun ZhaoAbstract:Understanding of the heat transfer characteristics of supercritical Water is one of the most important issues in Supercritical Water-Cooled Reactors (SCWRs) development. The main objective of the present study is to perform literature survey on the supercritical Water heat transfer researches in order to provide the references for the SCWR researchers. Both the experimental and numerical studies related to the supercritical Water heat transfer, especially for nuclear reactor applications, are reviewed. It is found that the majority researches are focusing on the supercritical Water flow inside circular tubes in which first order closure models assuming isotropic behavior of turbulence are being used. However, for flow channels with different geometries such as sub-channel of fuel assembly, anisotropic behavior of turbulence and secondary flows are observed. Therefore, additional researches are needed for supercritical Water heat transfer under SCWR operating conditions.
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coupled neutronic and thermal hydraulic out of phase stability of supercritical Water Cooled Reactors
Nuclear Technology, 2008Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:As the last topic of a series of U.S. reference supercritical Water-Cooled reactor (SCWR) design stability studies, coupled neutronic-thermal-hydraulic out-of-phase stability is analyzed and compared with that of a typical boiling Water reactor (BWR). A modal expansion method based on A modes (reactivity modes) of the neutron kinetic equation is applied, and the first subcritical mode of the neutron dynamics model is coupled with the coolant thermal-hydraulic model. The out-of-phase oscillation of the SCWR is found to be dominated by the reactor thermal hydraulics, whereas the BWR is more sensitive to the coolant density reactivity coefficient because of much stronger neutronic coupling. It is also found that the SCWR stability is sensitive to the details of the core simulation model and the hottest channel dominates the stability. The BWR is less sensitive to the core simulation model since it has much stronger neutronic coupling that is controlled by the whole-core average properties. Power and flow rate sensitivity analysis of the out-of-phase stability was also conducted for both the SCWR and the BWR. The SCWR stability is found to be more sensitive to the operating parameters than the typical BWR. Although the Water rod heating can improve the SCWR out-of-phase stability, it cannot significantly improve the sensitivity feature.
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core wide in phase stability of supercritical Water Cooled Reactors ii comparison with boiling Water Reactors
Nuclear Technology, 2008Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:To compare the stability features of a supercritical Water-Cooled reactor (SCWR) design with that of a typical boiling Water reactor (BWR), a stability analysis model for a typical BWR has been developed in addition to an already-developed model for the SCWR as presented in a companion paper. The homogenous equilibrium two-phase flow model, which is adequate at high pressures, is applied to the BWR stability analysis. The reactor core is simulated by three channels according to the radial power distribution and the inlet orifice coefficients. Similar to the SCWR model, the neutronic kinetic equation is expanded based on A modes (reactivity modes). The model is evaluated based on the Peach Bottom Atomic Power Station stability test data, and the results agree well with the experiment. The SCWR is found to be less sensitive to the coolant density neutronic reactivity coefficient than the typical BWR, since most of the neutronic moderation function is provided by the Water rods, where the density variation is either zero (if the Water rods are insulated) or small (if the Water rods are not insulated). The BWR is found to be less sensitive to changes in power level than the SCWR but has the same sensitivity level to the flow rate as the SCWR. A stability envelope that combines the single-channel and in-phase stability modes is developed. The decay ratios for the SCWR together with those for the typical BWR and the new Economic Simplified Boiling Water Reactor at nominal operational conditions are shown in the map. The stability sensitivity to operating conditions is also shown in the map, by increasing the power to 120% of nominal value and decreasing the flow rate to 80% of nominal value. It is found that the SCWR is more sensitive to the single-channel stability compared to the core-wide in-phase stability for all cases.
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core wide in phase stability of supercritical Water Cooled Reactors i sensitivity to design and operating conditions
Nuclear Technology, 2008Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:Using a three-region supercritical Water flow model, the core-wide in-phase stability of the U.S. reference supercritical Water-Cooled reactor (SCWR) design is investigated. The reactor core is simulated as three channels according to the radial power distribution. A method based on λ modes (reactivity modes) expansion of neutronic kinetic equations is applied. A constant pressure drop boundary condition between the feedWater pump and the turbine control valve is assumed. Cases with and without Water rods heating are studied. It is found that the stability of the U.S. reference SCWR design is sensitive to the flow restrictions in the hot fluid or the steam line. As long as the restriction in the steam line is small, the design will be stable. A pressure loss coefficient of 0.25 is assumed for the exit valve on the steam line in this analysis. With this value, the SCWR is stable with a large margin. It is concluded that the presence of Water rods heating will reduce the stability margin and increase the flow rate sensitivity while maintaining the power sensitivity level. The decay ratios for the three density wave oscillation modes, i.e., single hot channel, coupled neutronic out-of-phase and in-phase, are compared at steady-state conditions. It is found that the single hot channel oscillation mode is the most limiting one in the absence of the Water rods heating, while the in-phase oscillation mode is most limiting in the presence of Water rods heating.
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hot channel stability of supercritical Water Cooled Reactors i steady state and sliding pressure startup
Nuclear Technology, 2007Co-Authors: Jiyun Zhao, Pradip Saha, Mujid S KazimiAbstract:The drastic change of fluid density in the reactor core of a supercritical Water-Cooled reactor (SCWR) gives rise to a concern about density-wave stability. Using a single-channel thermal-hydraulic model, stability boundary maps for the U.S. reference SCWR design have been constructed for both steady state and sliding pressure startup conditions. The supercritical Water flow in the reactor core has been simulated using a three-region model: a heavy fluid with constant density, a mixture of heavy fluid and light fluid similar to a homogeneous-equilibrium two-phase mixture, and a light fluid, which behaves like an ideal gas or superheated steam. Two important nondimensional numbers, namely, a pseudosubcooling number N pSUb and an expansion number N exp , have been identified for the supercritical region. The stability map in the supercritical region is then plotted in the plane made of these two numbers. The U.S. reference SCWR design operates in a stable region with a large margin. Sensitivity studies produced results consistent with the findings of the earlier research done for the subcritical two-phase flow. During the sliding pressure startup of the SCWR, a two-phase steam-Water mixture at subcritical pressure will appear in the reactor core. A nonhomogeneous (e.g., drift-flux) nonequilibrium two-phase flow model was applied. The characteristic equation was numerically integrated, and stability boundary maps were plotted on the traditional subcooling number versus phase change number (or Zuber number) plane. These maps have been used to develop a sliding pressure SCWR startup strategy avoiding thermal-hydraulic flow instabilities.
S L Solovev - One of the best experts on this subject based on the ideXlab platform.
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the use of supercritical parameters of a coolant a promising path to development of nuclear power plant Water Cooled Reactors in the 21st century
Thermal Engineering, 2006Co-Authors: B A Gabaraev, V N Smolin, S L SolovevAbstract:A modern concept of the development of Water-Cooled Reactors of nuclear power plants (NPPs) is considered. Data on the design of NPPs with supercritical-parameters coolant and the results of experimental studies are presented.
