• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.92-102
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• 2017

The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1118-1124
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• 2023

Experimental study on the fluidelastic instability (FEI) of a curved tube bundle in single phase downward cross flow is investigated for the design qualification and analysis input preparation of helical coiled steam generator tubing. A 6×9 normal square curved tube array with equal and different vertical/horizontal pitch-to-diameter ratio was under-tested up to 6 m/s in term of gap flow velocity to measure the critical velocity for FEI. The critical velocity for FEI was measured at the turning point from the vibration amplitude plot along the gap flow velocity. Our test results were compared with straight tube results and published data in the design guideline. The applicability of the current design guidelines to a curved tube bundle is also assessed. We found that introducing frequency difference in a curved tube array increases the critical velocity for fluidelastic instability.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1742-1756
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• 2023

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a Large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high-temperature flow of approximately 390 ℃ was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high-temperature region. The phenomenological discussion in this paper helps understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.871-888
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• 2012

This paper aims at presenting the state of the art, the recent progress, and the perspective for the future, in the modelling of two-phase flow in the horizontal legs of a PWR. All phenomena relevant for safety analysis are listed first. The selection of the modelling approach for system codes is then discussed, including the number of fluids or fields, the space and time resolution, and the use of flow regime maps. The classical two-fluid six-equation one-pressure model as it is implemented in the CATHARE code is then presented and its properties are described. It is shown that the axial effects of gravity forces may be correctly taken into account even in the case of change of the cross section area or of the pipe orientation. It is also shown that it can predict both fluvial and torrential flow with a possible hydraulic jump. Since phase stratification plays a dominant role, the Kelvin-Helmholtz instability and the stability of bubbly flow regime are discussed. A transition criterion based on a stability analysis of shallow water waves may be used to predict the Kelvin-Helmholtz instability. Recent experimental data obtained in the METERO test facility are analysed to model the transition from a bubbly to stratified flow regime. Finally, perspectives for further improvement of the modelling are drawn including dynamic modelling of turbulence and interfacial area and multi-field models.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1387-1401
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• 2018

An aircraft impact (AI) on a nuclear power plant (NPP) is considered to be a beyond-design-basis event that draws considerable attention in the nuclear field. As some NPPs have already adopted the seismic isolation technology, and there are relevant standards to guide the application of this technology in future NPPs, a new challenge is that nuclear power engineers have to determine a reasonable method for performing AI analysis of base-isolated NPPs. Hence, dynamic influences of the seismic isolation on the vibration and structural damage characteristics of the base-isolated CPR1000 containment are studied under various aircraft loads. Unlike the seismic case, the impact energy of AI is directly impacting on the superstructure. Under the coupled influence of the seismic isolation and the various AI load, the flexible isolation layer weakens the constraint function of the foundation on the superstructure, the results show that the seismic isolation bearings will produce a large horizontal deformation if the AI load is large enough, the acceleration response at the base-mat will also be significantly affected by the different horizontal stiffness of the isolation bearing. These concerns require consideration during the design of the seismic isolation system.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.95-102
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• 2007

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1079-1092
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• 2020

A parametric study of several parameters relevant to design safety on the spent nuclear fuel (SNF) rod response under a drop accident is presented. In the view of the complexity of interactions between the independent safety-related parameters, a factorial design of experiment is employed as an efficient method to investigate the main effects and the interactions between them. A detailed single full-length fuel rod is used with consideration of post-irradiated fuel conditions under horizontal and vertical free-drops onto an unyielding surface using finite-element analysis. Critical drop heights and critical g-loads that yield the threshold plastic strain in the cladding are numerically estimated to evaluate the fuel rod structural resistance to impact load. The combinatory effects of four uncertain parameters (pellet-cladding interfacial bonding, material properties, spacer grid stiffness, rod internal pressure) and the interactions between them on the fuel rod response are investigated. The principal finding of this research showed that the effects of above-mentioned parameters on the load-carrying capacity of fuel rod are significantly different. This study could help to prioritize the importance of data in managing and studying the structural integrity of the SNF.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2454-2465
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• 2023

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1164-1171
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• 2020

T-junction structures play an important role in nuclear power plant systems. Research on liquid entrainment is mostly based on small-scale branch pipes (d/D ≤ 0.2) and attention paid to large-scale branch pipes (0.33 < d/D < 1) is insufficient. Accordingly, this study implements a series of experiments on the liquid entrainment of T-junction with different angles (32.2°,47.9°,62.3°,90°) through a large-scale branch (d/D = 0.675). The onset liquid entrainment is related to the gas phase Froude number Fr_g, the dimensionless gas chamber height h_b/d and the branch pipe angle 𝜃. As Fr_g increases, h_b/d also rises. With a constant h_b/d, the onset liquid entrainment changes from droplets entrainment by the gas phase to that by the rising liquid film. The steady-state liquid entrainment is related to w_3g, h/d and 𝜃. With constant w_3g and h/d, the branch quality grows as the branch angle increases. With a certain h/d, the branch quality increases, as the w_3g number increases.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3635-3642
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• 2021

Natural circulation systems (NCSs) are extensively applied in nuclear power plants because of their simplicity and inherent safety features. For some passive natural circulation systems in floating nuclear power plants (FNPPs), the ocean is commonly used as the heat sink. Condensation induced water hammer (CIWH) events may appear as the steam directly contacts the subcooled seawater, which seriously threatens the safe operation and integrity of the NCSs. Nevertheless, the research on the formation mechanisms of CIWH is insufficient, especially in NCSs. In this paper, the characteristics of flow rate and fluid temperature are emphatically analyzed. Then the formation types of CIWH are identified by visualization method. The experimental results reveal that due to the different size and formation periods of steam slugs, the flow rate presents continuous and irregular oscillation. The fluid in the horizontal hot pipe section near the water tank is always subcooled due to the reverse flow phenomenon. Moreover, the transition from stratified flow to slug flow can cause CIWH and enhance flow instability. Three types of formation mechanisms of CIWH, including the Kelvin-Helmholtz instability, the interaction of solitary wave and interface wave, and the pressure wave induced by CIWH, are obtained by identifying 67 CIWH events.