Results 1 - 10 of 73
Results 1 - 10 of 73. Search took: 0.019 seconds
|Sort by: date | relevance|
[en] To meet the energy requirement for a remote and isolated region, a concept of supercritical CO2 (S-CO2) cooled micro modular reactor (MMR) has been developed, which is easy to transport by fully modularizing the nuclear power system. From the previous work, only the on-design performances of MMR were presented. However, since the nuclear system needs to meet high safety standard, the design can be finalized only after the evaluation of the nuclear system response under multiple transient conditions. Such transient conditions may include part load operation and postulated accidents. The system response can be evaluated only with a high fidelity computer simulation, since it is almost impossible to construct a nuclear facility to test at such conditions. The simulation is performed to guarantee its autonomous operability and safety. Unfortunately, a system analysis computer code available for the S-CO2 cycle analysis is yet to be developed for the MMR safety evaluation. Description of a S-CO2 system analysis code platform is first presented in this paper. After that, the developed code is validated with experimental data. The steady state of MMR is first modeled with the code, which is followed by simulations of part load operation and postulated accidents of MMR. All simulation results show that the current design of MMR has the ability to maintain its safety and structural integrity even under extreme conditions to protect the public from radiation hazard at all times.
[en] Highlights: • A thermal hydraulic transient analysis code for OFNPs is developed. • Ocean condition models are established by considering the effects of ship motions. • The developed code is verified by experimental data under rolling motion. • Effects of the coupled ship motion on natural circulation system are studied. - Abstract: Offshore floating nuclear power plants (OFNPs) can effectively solve the offshore energy supply problem in marine resource development and island construction. Affected by ocean waves and other ocean conditions, the OFNPs can generate different kinds of ship motions, which can oscillate the thermal hydraulic parameters and threaten the reactor safety. In the present study, ocean condition theoretical models are established by considering the effects of three basic movement forms (static inclining, linear motion and rotation motion) as well as the coupled ship motions. A thermal hydraulic transient analysis code for OFNPs is developed by adding ocean condition theoretical models into the RELAP5/SCDAPSIM/MOD3.4 code. The experimental data obtained by zero power loading experiment and single-phase natural circulation experiment under rolling motion are used to verify the ocean condition theoretical models as well as the code modification strategy. Results show that the flow fluctuation behaviors caused by rolling motion can be well simulated by the developed code. The calculation capability of modified RELAP5 code under static inclining and heaving motion is also verified by comparing with RETRAN-02/GRAV code. Besides, the effects of the coupled ship motions on natural circulation system are studied by the modified RELAP5 code. Compared with the basic movement forms, the coupled ship motions can cause greater flow fluctuation and obviously reduce the core flow rate, which means the influence of the coupled ship motions is necessary to be considered in the safety analysis of OFNPs.
[en] Instrument setpoints are affected by many sources of uncertainty. ANSI/ISA-S67.04.01-2000 () and ISA-RP67.04.02-2000 () provide a basis for establishing safety-related instrument setpoints under several uncertainty terms including a random drift. However, the current methodologies and the recommended practices for a setpoint drift analysis need to be modified in order to control plant specific setpoint drift of an instrumentation device, especially when there are many ties in the measurements. In this paper, we propose an extended procedure supplemented with non-parametric and graphic tools for managing the setpoint drift, based on the plant-specific as-found/as-left data. The proposed procedure has three major advantages. First, supplementing non-parametric statistical methods can handle non-standard measurement data even when they have many ties. Second, adopting statistical graphic tools facilitates identifying data characteristics. Third, applying statistical process control (SPC) techniques may provide plant staff with an intuitive way of managing the instrumentation setpoint drift for a long term.
[en] Highlights: ► Natural circulation is an essential heat removal mechanism for nuclear power plants. ► A compressible 3-D CFD model is proposed to study a natural circulation loop. ► This CFD model is validated with the existing experimental data. ► Measured data include time histories of Tmax and temperature difference across the heater. ► Prediction of Ress vs Gr/NG show good agreement with measurement. - Abstract: Natural circulation is the most important heat removal mechanism for passive safety systems in new-design nuclear power plants. It is important to investigate the flow and heat transfer characteristics related to the natural circulation mechanism. In this paper, a compressible three-dimensional (3-D) CFD model is proposed to investigate these phenomena in a natural circulation loop. The flow and heat transfer behaviors in a loop can be reasonably captured by the present model, which includes the secondary flow within the elbow and the thermal stratification in the horizontal pipe, etc. This model is also validated with the existing experimental data. Compared with the measured results including time histories of local maximum fluid temperature and temperature difference across the heater, and the relationship between Ress and Grm/NG, the present predicted results show good agreement. These comparisons reveal that the present CFD methodology can be applied in simulating the thermal–hydraulic characteristics related to the natural circulation mechanism in confidence
[en] Highlights: ► New drag law in AIAD model was implemented in a CFD code to simulate the flows in nuclear reactor. ► The problems include the CCFL, hydraulic jump and pressurized thermal shock (PTS). ► The model is able to distinguish the local flow morphologies in frame of the Euler–Euler. ► CFD calculations agree well with the experimental data. - Abstract: This paper presents different CFD-simulations on flows which are relevant for nuclear reactor safety using a new modeling approach for the interfacial drag at free surfaces. The developed drag coefficient model was implemented together with the Algebraic Interfacial Area Density (AIAD) model () into the three-dimensional (3-D) computational fluid dynamics (CFD) code ANSYS-CFX. The applications considered include the prediction of counter-current flow limitations (CCFL) in a PWR hot leg, the development of hydraulic jump during the air–water co-current flow in a horizontal channel, and pressurized thermal shock (PTS) phenomena in a PWR cold leg and downcomer. For the modeling of these tasks, an Euler–Euler approach was used. This approach allows the use of different models depending on the local morphology. In the frame of an Euler–Euler simulation, the local morphology of the phases has to be considered in the drag model. To demonstrate the feasibility of the present approach, the computed main parameters of each case were compared with experimental data. It is shown that the CFD calculations agree well with the experimental data. This indicates that the AIAD model combined with new drag force modeling is a promising way to simulate the phenomena in frame of the Euler–Euler approach. Moreover the further validation of the model by including mass transfer effects should be carried out.
[en] Highlights: • Based on the experimental data, empirical gradient change was set as the quantified ONB criterion. • The effect of mass flux and inlet temperature on ONB was provided. • Existing ONB prediction correlations were evaluated based on the experimental data. • A new non-dimensional empirical correlation to predict ONB was developed based on the experimental data. - Abstract: Bubble nucleation itself is less important safety issue for nuclear reactor, but it can easily lead to critical thermal-hydraulic events such as OFI (Onset of Fluid Instability) or CHF (Critical Heat Flux) when a research reactor operates under atmospheric conditions. Thus, the ONB (Onset of Nucleate Boiling) margin for normal operation in research reactor is recommended. In the IAEA-TECDOC-233 report (IAEA, 1980), the ONB margin for a research reactor is recommended as well. Although the ONB margin in a research reactor is emphasized for such reasons, only a few experiments have been performed for downward flow direction in a narrow, rectangular channel. In addition, several existing ONB prediction correlations are arguably applicable to the flow boiling condition in the narrow rectangular channel because most of them are developed based on Hsu’s model, which was developed in the pool boiling cases. In the study, ONB experiments for various inlet temperature conditions and mass flux conditions were performed with increasing heat flux step by step. Based on experimental data, the effect of inlet temperature and mass flux on the wall superheat and heat flux at ONB was investigated. In addition, existing ONB prediction correlations were evaluated for predicting wall superheat and heat flux at ONB based on the experimental data. A new ONB prediction correlation was then developed for better-evaluation and was compared with other correlations.
[en] Highlights: ► A safety analysis code is developed for the molten salt reactors. ► The validation of the safety analysis code is done with the MSRE experimental data. ► The MSRE experiment and theoretical results do not show any oscillations with time. ► The unprotected loss of heat sink (ULOHS) is performed on the MOSART. ► The combination of ULOHS and ULOF is performed on the MOSART. - Abstract: The molten salt reactor (MSR) is one of the Generation IV reactors. The fuel is dissolved in the carrier salt and circulates in the loop. The technologies are different from that in the solid-fuel reactors. In this work, the attention is focused on development of safety analysis tool for the MSRs. A single channel model, a heat transfer model, a heat sink model and a liquid-fuel point kinetic model that takes into account the effect of circulation are employed. The validation of this code is done with the experimental data of pump coastdown and startup in MSRE. The unprotected loss of heat sink (ULOHS), combination of ULOHS and unprotected loss of flow (ULOF) are performed on the Molten Salt Actinide Recycler and Transmuter (MOSART). This work aims to study temperature fluctuation, corresponding power change, effect of flow delayed neutron precursors and temperature reactivity feedback in transient accident to examine the inherent safety design of MOSART. The transient results reveal that the large negative temperature feedback coefficients guarantee MOSART inherent safety and the range of temperature is within the safety margin in case of combination of accidental events
[en] Highlights: • A coupling algorithm was developed by combining DEM with a multi-fluid model. • This method was validated by perform the simulations of gas–solid fluidized beds. • Agreement was obtained between the simulation results and experimental data. - Abstract: Gas–solid fluidization is not only an essential phenomenon in many areas of industry, but is also used to understand particle behavior in a number of research fields. For the safety analysis of core disruptive accidents in liquid-metal fast reactors, a hybrid method is developed by combining the discrete element method with a fluid-dynamics model of the reactor safety analysis code SIMMER-III to reasonably simulate particle transient behavior, as well as the occurring thermal-hydraulic phenomena. As a preliminary validation procedure, the developed hybrid method is applied to simulations of gas–solid two-phase flows. In this study, numerical simulations of two typical gas–solid fluidized bed systems are performed. The particles in the beds are porous alumina of 70 μm diameter and glass of 530 μm diameter, which belong to Geldart groups A and B, respectively. The reasonable agreement between our simulation results and experimental data from the literature demonstrates the fundamental validity of the present simulation method for multiphase flows with large amounts of solid particles
[en] For effective reduction of occupational radiation exposure in a nuclear power plant, it is necessary to identify repetitive high radiation jobs during maintenance and refueling operation and comprehensively assess them. An integrated framework for effective reduction of occupational radiation exposure is proposed in this study. The framework consists of three parts; data collection, statistical analysis, and ALARA findings. A PC-based database program, INSTORE, is used for data collection and reduction, and the Rank Sum Method is used in identifying high radiation jobs. As a case study, the data accumulated in Kori Units 3 and 4 have been analyzed. The results of this study show that the radiation job classifications of SG related work have much effect on annual ORE collective dose in Kori Units 3 and 4. As an example of ALARA findings, hence, the improvements for the radiation job classifications of SG related work are summarized
[en] Highlights: •Heat transfer experiment of supercritical water in a rod bundle was performed. •Circumferential wall temperature and heat transfer coefficient were obtained. •Extensive heat transfer correlations were assessed against the test data. -- Abstract: The heat transfer coefficient of supercritical water in a rod bundle is essential for the fuel design of the Supercritical Water-Cooled Reactor (SCWR). Although numerous correlations have been proposed over the past few decades to predict the heat transfer coefficient, the conclusions are inconsistent due to the limited experimental data obtained in fuel bundle. In the present paper, 20 correlations were assessed against the experimental data obtained in a tight 2 × 2 rod bundle. Circumferential maximum wall temperature and minimum heat transfer coefficient were selected as the benchmark data to get a conservative conclusion in support of the fuel design and safety analysis. The assessments showed that the performances of these correlations vary greatly depending on the mass flux and heat flux. Most correlations give reasonable Nusselt numbers at normal and enhanced heat transfer regimes, but become worse when the heat transfer is impaired. Comparison of these correlations against the total of 714 data indicated that the correlation proposed by Chen-Fang is the best with an average error of −0.44% and a standard deviation of 6.4%. All of the experimental Nusselt numbers were successfully predicted within ±20% error band.