The Experts below are selected from a list of 216 Experts worldwide ranked by ideXlab platform

A. P. Sorokin - One of the best experts on this subject based on the ideXlab platform.

  • Fast Reactor: an experimental study of thermohydraulic processes in different operating regimes
    Thermal Engineering, 2017
    Co-Authors: A. N. Opanasenko, A. P. Sorokin, D. G. Zaryugin, A. A. Trufanov
    Abstract:

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the Fast Neutron Reactor (hereinafter, Fast Reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the Reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the Reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a Fast Reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the Reactor remains within acceptable limits and decay heat removal from the Reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the Reactor plant parameters on the similarity criteria basis.

  • Investigations of coolant stratification and temperature pulsations in nuclear power installations
    Thermal Engineering, 2013
    Co-Authors: D. G. Zaryugin, S. G. Kolyakin, A. N. Opanasenko, A. P. Sorokin
    Abstract:

    Specific features of stratified flows are considered taking Fast-Neutron Reactors as an example, the possibilities of scale-wise simulation of such flows are discussed, and some results obtained from experimental and numerical investigations of flows in mixing chambers and pipelines are presented. The mixing of sodium jets in the upper chamber at the outlet from the Fast-Neutron Reactor core is analyzed.

  • An experimental study of heat transfer in the core of a BREST-OD-300 Reactor with lead cooling on models
    Thermal Engineering, 2002
    Co-Authors: A. V. Zhukov, Yu. A. Kuzina, A. P. Sorokin, V. N. Leonov, V. P. Smirnov, A. G. Sila-novitskii
    Abstract:

    The results of thermohydraulic investigations for designing the experimental demonstration Fast Neutron Reactor BREST-OD-300 with lead cooling are considered.

Haojun Zhou - One of the best experts on this subject based on the ideXlab platform.

  • investigations of sample reactivity worth measurement in a Fast Neutron Reactor with the inverse kinetics method
    Annals of Nuclear Energy, 2018
    Co-Authors: Shumiao Wang, Haojun Zhou, Zhongxiong Bai, Xiaoqiang Fan, Yanpeng Yin
    Abstract:

    Abstract A new method for the measurement of sample reactivity worth in a Fast Neutron Reactor named the inverse kinetics method is proposed in the paper. The sample reactivity worth could be obtained by measuring the reactivity step change in the process of sample fetching and placing in the delayed critical Reactor. Compared with the traditional period method, the advantage is that the accuracy of Reactor reactivity control will not exert any influence on the uncertainty of reactivity worth measurement. The inverse kinetics method has been used to measure the reactivity worth of Ф20 mm × 9 mm Au, V and Be samples at the center of the upper surface of a highly enriched uranium Fast Neutron Reactor, and the results are 5.17¢, 4.40¢ and 5.90¢ respectively, which are consistent with those obtained with the period method. The standard uncertainty of measurement of the results is about 0.03¢, which achieves an obvious improvement compared with that (∼0.08¢) of the period method.

  • investigations of reactivity worth measurement in a Fast Neutron Reactor with the inverse kinetics method
    arXiv: Instrumentation and Detectors, 2017
    Co-Authors: Shumia Wang, Haojun Zhou, Zhongxiong Bai, Xiaoqiang Fan, Yanpeng Yin
    Abstract:

    A new method for the measurement of sample reactivity worth in a Fast Neutron Reactor named the inverse kinetics method is proposed in the paper. The sample reactivity worth could be obtained by measuring the reactivity step change in the process of sample fetching and placing in the delayed critical Reactor. Compared with the traditional period method, the advantage is that the accuracy of Reactor reactivity control will not exert any influence on the uncertainty of reactivity worth measurement. The inverse kinetics method has been used to measure the reactivity worth of {\phi}20mmx9mm Au, V and Be samples at the center of the upper surface of a highly enriched uranium Fast Neutron Reactor, and the results are 5.17, 4.40 and 5.90 respectively, which are consistent with those obtained with the period method. The standard uncertainty of measurement of the results is about 0.03, which achieves an obvious improvement compared with that (~0.08) of the period method.

  • measurements of effective delayed Neutron fraction in a Fast Neutron Reactor using the perturbation method
    Chinese Physics C, 2016
    Co-Authors: Haojun Zhou, Zhenghong Li, Yikang Pu
    Abstract:

    A perturbation method is proposed to obtain the effective delayed Neutron fraction β eff of a cylindrical highly enriched uranium Reactor. Based on reactivity measurements with and without a sample at a specified position using the positive period technique, the Reactor reactivity perturbation Δρ of the sample in β eff units is measured. Simulations of the perturbation experiments are performed using the MCNP program. The PERT card is used to provide the difference dk of effective Neutron multiplication factors with and without the sample inside the Reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation β eff = dk/Δρ is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the Reactor are measured and calculated. The average β eff value of the Reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for β eff can be used to evaluate the reliabilities of the delayed Neutron parameters. This work shows that the delayed Neutron data of 235U and 238U from G.R. Keepin's publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2.

  • measurements of effective delayed Neutron fraction in a Fast Neutron Reactor using the perturbation method
    arXiv: Instrumentation and Detectors, 2015
    Co-Authors: Haojun Zhou, Zhenghong Li, Yikang Pu
    Abstract:

    The perturbation method is proposed to obtain the effective delayed Neutron fraction (\b{eta}eff) of a cylindrical highly enriched uranium Reactor. Based on the reactivity measurements with and without a sample at a designable position using the positive periodic technique, the Reactor reactivity perturbation {\Delta}\r{ho} of the sample in \b{eta}eff units is measured. The simulation of the perturbation experiments are performed by MCNP program. The PERT card is used to provide the difference dk of effective Neutron multiplication factors with and without the sample inside the Reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation \b{eta}eff =dk/{\Delta}\r{ho} is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the Reactor are measured and calculated. The average \b{eta}eff value of the Reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for \b{eta}eff can be used to evaluate the reliabilities of the delayed Neutron parameters. This work showed that the delayed Neutron data of 235U and 238U from G.R.Keepin's publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2.

A. N. Opanasenko - One of the best experts on this subject based on the ideXlab platform.

  • Fast Reactor: an experimental study of thermohydraulic processes in different operating regimes
    Thermal Engineering, 2017
    Co-Authors: A. N. Opanasenko, A. P. Sorokin, D. G. Zaryugin, A. A. Trufanov
    Abstract:

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the Fast Neutron Reactor (hereinafter, Fast Reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the Reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the Reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a Fast Reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the Reactor remains within acceptable limits and decay heat removal from the Reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the Reactor plant parameters on the similarity criteria basis.

  • Investigations of coolant stratification and temperature pulsations in nuclear power installations
    Thermal Engineering, 2013
    Co-Authors: D. G. Zaryugin, S. G. Kolyakin, A. N. Opanasenko, A. P. Sorokin
    Abstract:

    Specific features of stratified flows are considered taking Fast-Neutron Reactors as an example, the possibilities of scale-wise simulation of such flows are discussed, and some results obtained from experimental and numerical investigations of flows in mixing chambers and pipelines are presented. The mixing of sodium jets in the upper chamber at the outlet from the Fast-Neutron Reactor core is analyzed.

D. G. Zaryugin - One of the best experts on this subject based on the ideXlab platform.

  • Fast Reactor: an experimental study of thermohydraulic processes in different operating regimes
    Thermal Engineering, 2017
    Co-Authors: A. N. Opanasenko, A. P. Sorokin, D. G. Zaryugin, A. A. Trufanov
    Abstract:

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the Fast Neutron Reactor (hereinafter, Fast Reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the Reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the Reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a Fast Reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the Reactor remains within acceptable limits and decay heat removal from the Reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the Reactor plant parameters on the similarity criteria basis.

  • Investigations of coolant stratification and temperature pulsations in nuclear power installations
    Thermal Engineering, 2013
    Co-Authors: D. G. Zaryugin, S. G. Kolyakin, A. N. Opanasenko, A. P. Sorokin
    Abstract:

    Specific features of stratified flows are considered taking Fast-Neutron Reactors as an example, the possibilities of scale-wise simulation of such flows are discussed, and some results obtained from experimental and numerical investigations of flows in mixing chambers and pipelines are presented. The mixing of sodium jets in the upper chamber at the outlet from the Fast-Neutron Reactor core is analyzed.

A. A. Trufanov - One of the best experts on this subject based on the ideXlab platform.

  • Fast Reactor: an experimental study of thermohydraulic processes in different operating regimes
    Thermal Engineering, 2017
    Co-Authors: A. N. Opanasenko, A. P. Sorokin, D. G. Zaryugin, A. A. Trufanov
    Abstract:

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the Fast Neutron Reactor (hereinafter, Fast Reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the Reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the Reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a Fast Reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the Reactor remains within acceptable limits and decay heat removal from the Reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the Reactor plant parameters on the similarity criteria basis.