• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2236-2241
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• 2008

The surface heat flux of nuclear fuel rod is the most important factor which can affect safety of reactor and fuel. If fuel rod surface heat flux exceeds the CHF(${\underline{C}}ritical$ ${\underline{H}}eat$ ${\underline{F}}lux$), fuel can be damaged. In case of double cooled annular fuel, which is under developing, contains two coolant channels. Therefore, a generated heat in the fuel pellet can move to inner or outer channel and heat flow direction is decided by both sides heat resistance which varied by dimension and material property change which caused by temperature and irradiation. The new program(called DUO) was developed. For the calculation of surface heat flux, a both sides convection by inner/outer coolant, s gap temperature jump and conduction in the fuel are modeled. Especially, temperature and time dependent fuel dimension and material property change are considered during the iteration. A sample calculation result shows that the DUO program has sufficient performance for annular fuel thermal hydraulics design.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.330-338
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• 2023

In this study, impacts of an enhanced-moderation Fuel Assembly (FA) named Truly Optimized PWR (TOP) lattice, which is modified based on the standard 17 × 17 PWR FA, are investigated in a natural circulation Soluble-Boron-Free (SBF) Small Modular Reactor (SMR). Two different TOP lattice designs are considered for the analysis; one is with 1.26 cm pin pitch and 0.38 cm fuel pellet radius, and the other is with 1.40 cm pin pitch and 0.41 cm fuel pellet radius. The NuScale core design is utilized as the base model and assumed to be successfully converted to an SBF core. The analysis is performed following the primary coolant circulation loop, and the reactor is modelled as a single channel for thermal-hydraulic analyses. It is assumed that the ratio of the core pressure drop to the total system pressure drop is around 0.3. The results showed that the reactor power could be increased by 2.5% and 9.8% utilizing 1.26/0.38 cm and 1.40/0.41 cm TOP designs, respectively, under the identical coolant inlet and outlet temperatures as the constraints.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.255-258
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• 2002

Liquid rocket engines using liquefied natural gas (LNG) or methane as a fuel is known to have several good characteristics, such as high specific impulse compared to other hydrocarbon fuels, environment-friendly exhaust gas, low production cost, and re-usability with low soot generation in the cooling channel. In this study, experimental combustion chambers capable of using LNC and $CH_{4}$ are being researched through experimental firing tests, and within easy range of eyes' inspection, there are the periodical existence of soot or discoloration in the chamber wall surface. This result means that mixture ratio of oxidizer and fuel fluctuates periodically between outer-row injectors in the mixing head in the circumferential direction. Therefore, based on this phenomenon, the variation of mixture ratio near the chamber wall caused by the spill pattern of a shear coaxial injector was analyzed quantitatively and the thermal heat flux Into the cooling channel is modified. Then, the calculated and modified results are compared with the measured ones.

• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.248-262
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• 2004

Turbulence models are separately assessed for a three dimensional thermal-hydraulic analysis of the integral reactor SMART. Seven models (mixing length, k-l, standard $k-{\epsilon},\;k-{\epsilon}-f{\mu},\;k-{\epsilon}-v2$, RRSM, and ERRSM) are investigated for flat plate channel flow, rotating channel flow, and square sectioned U-bend duct flow. The results of these models are compared to the DNS data and experiment data. The results are assessed in terms of many aspects such as economical efficiency, accuracy, theorization, and applicability. The standard $k-{\epsilon}$ model (high Reynolds model), the $k-{\epsilon}-v2$ model, and the ERRSM (low Reynolds models) are selected from the assessment results. The standard $k-{\epsilon}$ model using small grid numbers predicts the channel flow with higher accuracy in comparison with the other eddy viscosity models in the logarithmic layer. The elliptic-relaxation type models, $k-{\epsilon}-v2$, and ERRSM have the advantage of application to complex geometries and show good prediction for near wall flows.

• Journal of the Korean Society of Propulsion Engineers
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• v.6 no.4
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• pp.25-35
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• 2002

A calculation procedure for designing LRE regenerative cooling system is introduced. In LRE thrust chamber, heat is transfered from the hot gas to the wall by convection and radiation, then conduction through the wall and finally convection to the liquid coolant. A cooling channel is designed on the basis of heat transfer rate calculated by using criterial method and integral method. The result is compared with existing Russian cooling channel design code. Also a design logic and quantitative effect of various parameters were introduced to help better understanding for those who is not familiar to LRE system.

• Journal of the Korean Society of Propulsion Engineers
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• v.24 no.3
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• pp.81-124
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• 2020

In liquid oxygen/kerosene liquid rocket engines, kerosene is not only a propellant but also plays a role as a coolant to protect the combustion chamber wall from 3,000 K or more combustion gas. Since kerosene is exposed to high temperature passing through cooling channels, it may undergo heat-related chemical reactions leading to precipitation of carbon-rich solids. Such kerosene's thermal and fluidic characteristic test data are essential for the regeneratively cooled combustion chamber design. In this paper, we investigated foreign studies related to regenerative cooling channel and kerosene. Starting with general information on hydrocarbon fuels including kerosene, we attempted to systematically organize sedimentary phenomena on cooling channel walls, their causes/research results, coking test equipments/prevention methods, etc.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.1
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• pp.60-67
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• 2016

A pressurized heavy water reactor (PHWR) designed to refuel in service produces the energy required by nuclear fission. The fuel channel consists of components such as a pressure tube which directly contacts the fuel and is a passage for the reactor coolant, a calandria tube which contacts the moderator and is rolled joint with calandria, and a spacer which is not to contact the pressure tube and a calandria tube. As the fuel channel is one of the most important equipments, it requires accurate and periodic inspections to assess the integrity of a reactor in accordance with CSA N285.4. A fuel channel inspection system is developed to inspect fuel channels during in-service inspection in Wolsong unit. In this paper, the results and considerations of a field test are presented in order to show the effectiveness of the developed fuel channel inspection system.

• Journal of the Korean Society of Propulsion Engineers
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• v.20 no.4
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• pp.94-103
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• 2016

A regeneratively-cooled channel in a liquid rocket engine is used to effectively cool a combustion chamber inner wall from hot combustion gas, and the heat transfer/pressure loss characteristics should be predicted in advance to design cooling channels. In the present research, five cooling channels with different geometric dimensions were designed and the channels were respectively manufactured using cutter and endmill. By changing coolant velocity and downstream pressure, the effects of manufacturing method, channel shape, and flow condition on pressure losses were experimentally investigated and the results were compared with the analytical results. At same channel shape and flow condition, the pressure loss in the channel machined by the cutter was lower than that by the endmill. It was also found that the pressure loss ratio between the experimental result and the analytical data changed with the channel shape and flow condition.