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  • 1
    Online Resource
    Online Resource
    Hindawi Limited ; 2023
    In:  Science and Technology of Nuclear Installations Vol. 2023 ( 2023-3-7), p. 1-12
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2023 ( 2023-3-7), p. 1-12
    Abstract: Owing to pipe thinning, fatigue damage, and aging, pipes, valves, and devices installed in the primary and secondary systems of nuclear power plants may leak high-temperature/high-pressure reactor coolant. Thus, a system must be developed to determine if the leakage is exceeding the operating limit of the nuclear power plant, thereby mitigating any loss of life or economic loss in such cases. In this study, a validated numerical analysis method was established to initially simulate the leakage behavior and subsequently to evaluate the small amount of leakage in the compartment. For this purpose, a vapor-jet collision test in the compartment and a vapor-jet test in the pipe were performed; numerical analysis was conducted, and comparative analysis was performed to verify the validity of the established method. The evaluation results suggested that the proposed numerical analysis method could optimally simulate the flow characteristics of the steam jet. Notably, compared to the existing evaluation method, the proposed approach simulated a more detailed behavior of the jet formed at the leakage point. In future research, the results of this study (data) will be used to inform the design of the second phase of the leak-capture system and will be served as the foundation for a performance-optimization study on the capture system.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2023
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
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  • 2
    Online Resource
    Online Resource
    Hindawi Limited ; 2022
    In:  Science and Technology of Nuclear Installations Vol. 2022 ( 2022-1-30), p. 1-10
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2022 ( 2022-1-30), p. 1-10
    Abstract: A study on ONB (onset of nucleate boiling) in two vertical rectangular channels are experimentally conducted in a range of mass flux varying from 100 to 300 kg/(m2·s), inlet water temperature from 70 to 100°C, heat flux from 10 to 70 kW/m2, and local pressure of 0.145 MPa. The cross-section sizes are 1.8 mm ∗ 60 mm and 2.8 mm ∗ 60 mm, respectively. Three boiling incipience judgment methods have been used to locate ONB sites and found that Δ T ONB (the wall superheat at ONB site) increases with the decrease of inlet temperature and increases as mass flux increases. The results also indicate that although the bubble size and behaviors in the narrow channel are different from that in the nonnarrow channel at the ONB site, the heat transfer has not been influenced evidently. In addition, Δ T ONB in both channels can be predicted by the correlation proposed by Thom within the error range of ±30%.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2022
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 3
    Online Resource
    Online Resource
    Hindawi Limited ; 2022
    In:  Science and Technology of Nuclear Installations Vol. 2022 ( 2022-8-29), p. 1-17
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2022 ( 2022-8-29), p. 1-17
    Abstract: The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2022
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 4
    Online Resource
    Online Resource
    Hindawi Limited ; 2012
    In:  Science and Technology of Nuclear Installations Vol. 2012 ( 2012), p. 1-9
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2012 ( 2012), p. 1-9
    Abstract: The subcooled decompression under temperature gradient experiment performed by Takeda and Toda in 1979 has been reproduced using the in-house code WAHA version 3. The sudden blowdown of a pressurized water pipe under temperature gradient generates a travelling pressure wave that changes from decompression to compression, and vice versa, every time it reaches the two-phase region near the orifice break. The pressure wave amplitude and frequency are obtained at different locations of the pipe's length. The value of the wave period during the first 20 ms of the experiment seems to be correct but the pressure amplitude is overpredicted. The main three parameters that contribute to the pressure wave behavior are: the break orifice (critical flow model), the ambient pressure at the outlet, and the number of volumes used for the calculation. Recent studies using RELAP5 code have reproduced the early pressure wave (transient) of the same experiment reducing the discharge coefficient and the bubble diameter. In the present paper, the long-term pipe pressure, that is, 2 seconds after rupture, is used to estimate the break orifice that originates the pressure wave. The numerical stability of the WAHA code is clearly proven with the results using different Courant numbers.
    Type of Medium: Online Resource
    ISSN: 1687-6075 , 1687-6083
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2012
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 5
    Online Resource
    Online Resource
    Hindawi Limited ; 2021
    In:  Science and Technology of Nuclear Installations Vol. 2021 ( 2021-3-13), p. 1-6
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2021 ( 2021-3-13), p. 1-6
    Abstract: In response to a station blackout accident similar to the Fukushima nuclear accident, China’s Generation III nuclear power HPR1000 designed and developed a passive residual heat removal system connected to the secondary side of the steam generator. Based on the two-phase natural circulation principle, the system is designed to bring out long-term core residual heat after an accident to ensure that the reactor is in a safe state. The steady-state characteristic test and transient start and run test of the PRS were carried out on the integrated experiment bench named ESPRIT. The experiment results show that the PRS can establish natural circulation and discharge residual heat of the first loop. China’s Fuqing no. 5 nuclear power plant completed the installation of the PRS in September 2019 and carried out commissioning work in October. This debugging is the first real-world debugging of the new design. This paper introduces the design process of the PRS debugging scheme.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2021
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 6
    Online Resource
    Online Resource
    Hindawi Limited ; 2023
    In:  Science and Technology of Nuclear Installations Vol. 2023 ( 2023-7-10), p. 1-15
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2023 ( 2023-7-10), p. 1-15
    Abstract: The neutronics and thermal-hydraulics (N/TH) coupling behavior analysis is a key issue for nuclear power plant design and safety analysis. Due to the high-dimensional partial differential equations (PDEs) derived from the N/TH system, it is usually time consuming to solve such a large-scale nonlinear equation by the traditional numerical solution method of PDEs. To solve this problem, this work develops a reduced order model based on the proper orthogonal decomposition (POD) and artificial neural networks (ANNs) to simulate the N/TH coupling system. In detail, the POD method is used to extract the POD modes and corresponding coefficients from a set of full-order model results under different boundary conditions. Then, the backpropagation neural network (BPNN) is utilized to map the relationship between the boundary conditions and POD coefficients. Therefore, the physical fields under the new boundary conditions could be calculated by the predicated POD coefficients from ANN and POD modes from snapshot. In order to assess the performance of an ANN-POD-based reduced order method, a simplified pressurized water reactor model under different inlet coolant temperatures and inlet coolant velocities is utilized. The results show that the new reduced order model can accurately predict the distribution of the physical fields, as well as the effective multiplication factor in the N/TH coupling nuclear system, whose relative errors are within 1%.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2023
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 7
    Online Resource
    Online Resource
    Hindawi Limited ; 2009
    In:  Science and Technology of Nuclear Installations Vol. 2009 ( 2009), p. 1-1
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2009 ( 2009), p. 1-1
    Type of Medium: Online Resource
    ISSN: 1687-6075 , 1687-6083
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2009
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 8
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2020 ( 2020-11-2), p. 1-15
    Abstract: The accurate prediction of the neutronic and thermal-hydraulic coupling system transient behavior is important in nuclear reactor safety analysis, where a large-scale nonlinear coupling system with strong stiffness should be solved efficiently. In order to reduce the stiffness and huge computational cost in the coupling system, the high-performance numerical techniques for solving delayed neutron precursor equation are a key issue. In this work, a new precursor integral method with an exponential approximation is proposed and compared with widely used Taylor approximation-based precursor integral methods. The truncation errors of exponential approximation and Taylor approximation are analyzed and compared. Moreover, a time control technique is put forward which is based on flux exponential approximation. The procedure is tested in a 2D neutron kinetic benchmark and a simplified high-temperature gas-cooled reactor-pebble bed module (HTR-PM) multiphysics problem utilizing the efficient Jacobian-free Newton–Krylov method. Results show that selecting appropriate flux approximation in the precursor integral method can improve the efficiency and precision compared with the traditional method. The computation time is reduced to one-ninth in the HTR-PM model under the same accuracy when applying the exponential integral method with the time adaptive technique.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2020
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 9
    Online Resource
    Online Resource
    Hindawi Limited ; 2022
    In:  Science and Technology of Nuclear Installations Vol. 2022 ( 2022-5-19), p. 1-14
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2022 ( 2022-5-19), p. 1-14
    Abstract: In order to grasp the distribution characteristics of the pressure field on the inner wall of the volute casing of an atypical open impeller centrifugal pump, the instantaneous pressure at different operating conditions was experimentally measured under four operating rotational speeds to obtain the distribution characteristics of the average static pressure field in the volute casing of this pump model. The pressure pulsation amplitude and pressure pulsation intensity were also analyzed at different rotational speed cases, and the standard deviation analysis was performed. The results showed that the instantaneous pressure pulsation on the inner wall of the volute casing strongly fluctuates during the pump operating, and the closer to the volute casing outlet, the more intense the pressure pulsation was. After increasing the pump shaft speed, the fluctuation amplitude gradually decreased. The pressure pulsation on the wall of tip clearance is more intense than that on the inner wall of the volute shell. The intensity of the pressure pulsation on the wall of tip clearance decreases with the increase of the rotational speed, and the higher the speed, the less intense the pressure pulsation.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2022
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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  • 10
    Online Resource
    Online Resource
    Hindawi Limited ; 2021
    In:  Science and Technology of Nuclear Installations Vol. 2021 ( 2021-4-24), p. 1-10
    In: Science and Technology of Nuclear Installations, Hindawi Limited, Vol. 2021 ( 2021-4-24), p. 1-10
    Abstract: The granular flow is one of the principal issues for the design of pebble bed reactors. Particularly, the clogging phenomenon raises an important issue for pebble bed reactors. In this paper, we conduct experiments and discrete particle simulation of two-dimensional discharge granular flow from a conical hopper, to study the effect of the particle bed height h and hopper angle α on the clogging phenomenon. In general, the clogging probability J increases with height h and starts to saturate when h is larger than a critical value. The experimental result trends are supported by discrete simulations. To understand the underlying physical mechanism, we conduct discrete particle simulations for various h values, focusing on the following parameters: the statistical averaging of the volume fraction, velocity, and contact pressure of particles near the aperture during the discharge. We found that, among all relevant variables, the contact pressure of particles is the main cause of the increasement of J when h increases. An exponential law between the pebble bed h and clogging probability J has been established based on these observations and Janssen model. As for hopper angle α , J shows an almost constant behavior for any rise in α followed by a sudden regression at α = 75 ° . Surprisingly, the effect of α is most obvious for intermediate values of h , where we observe a sharp increasement of clogging probability. The same trend is observed in the two-dimensional discrete simulation results.
    Type of Medium: Online Resource
    ISSN: 1687-6083 , 1687-6075
    Language: English
    Publisher: Hindawi Limited
    Publication Date: 2021
    detail.hit.zdb_id: 2397941-0
    SSG: 3,6
    Location Call Number Limitation Availability
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