Development of a thermal hydraulic analysis code for gas-cooled reactors with annular fuels

Han, Kyu-Hyun; Seo, Kyoungwoo; Hwang, Dae-Hyun; Chang, Soon Heung

doi:10.1016/j.nucengdes.2005.07.006

Cited by 14 publications

(7 citation statements)

References 6 publications

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“…Where, ρ αρ v + (1 − α)ρ l is the mixture density in unit of kg/ m 3 , u → is the velocity vector of fluid through flow area dS in unit of Feng and Hejzlar (2007) America Thermal hydraulic analysis of annular fuel subchannel DUO_THERM Yang et al (2009) Korea Research on temperature and heat flux distribution of annular fuel MATRA-THAF Han and Chang (2003) Korea Thermal hydraulic analysis of annular fuel subchannel MATRA-GCR Han et al (2006) Korea Subchannel analysis for HTGR SAAF Diao and Ji (2015) China Thermal hydraulic analysis of annular fuel subchannel ATHAS-AF Zhu (2012) China Thermal hydraulic analysis of annular fuel subchannel SACAF Xia et al (2019) China Thermal hydraulic analysis of annular fuel subchannel m/s, n → denotes the flow direction (if the fluid flows from the channels to another channel, n → 1, otherwise, n → -1).…”

Section: Mass Conservation Equationmentioning

confidence: 99%

“…Annular fuel has advantages of shorter heat transfer path and larger heat transfer area than traditional solid rod fuel. It has been considered as a novel fuel design for different types of nuclear reactor concepts, such as liquid cooled fast reactor (Rowinski and White, 2015), CANDU (Nava Dominguez and Rao, 2020), gas cooled reactor (Han et al, 2006) and water cooled reactor (Shin and Chun, 2012;Kwon and Kim, 2019).…”

Section: Introductionmentioning

confidence: 99%

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Researches on Special Thermal Hydraulic Phenomena of Annular Fuel Assembly by Sub-Channel Analysis Code-NACAF

Zhang

Shan

et al. 2021

Front. Energy Res.

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Dual cooled annular fuel is a novel fuel design, which has the potential to improve the reactor power density while maintaining or improving its safety margin. The effects of tight-lattice geometry, fuel burnup, fuel expansion, coolant channel blockage on the thermal hydraulic performance of annular fuel is studied to illustrate its special features in this paper. A sub-channel analysis code named NACAF, which includes empirical constitutive models in consideration of tight-lattice effects on prediction of pressure drop, critical heat flux and turbulent mixing, channel blockage model, heat conduction model for dual surface cooling condition, coolant flowrate distribution between inner and outer channel, is developed for annular fuel assembly or core analysis based on homogenous fluid model. Validation work is carried out with comparing NACAF results with analytical solutions, as well as numerical results of existing sub-channel code for annular fuel, such as VIPRE-01 and TAFIX. Comparison results demonstrates NACAF’s prediction error is acceptable and it has the ability to simulate thermal hydraulic performance of annular fuels or annular fuel bundles. Based on the developed and verified NACAF, the special thermal hydraulic phenomena of annular fuel are studied to clarify the features of annular fuel.

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Section: Mass Conservation Equationmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Researches on Special Thermal Hydraulic Phenomena of Annular Fuel Assembly by Sub-Channel Analysis Code-NACAF

Zhang

Shan

et al. 2021

Front. Energy Res.

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“…Hence, in 2001 M. S. Kazimi [1] proposed annular fuel rods concept for water reactors and the investigation has shown better performance than the default solid fuel pin design. The feasibility study was conducted by the United States of America in case of water and sodium cooled reactors applications [2][3][4], while South Korea has conducted evaluation for water and gas-cooled reactors [5][6][7][8][9][10]. It was found that the temperature distribution is much lower in annular fuel element, while pressure drop along the channel can be decreased if the design of fuel assembly is optimized.…”

Section: Introductionmentioning

confidence: 99%

Innovative model of annular fuel design for lead-cooled fast reactors

Rowinski

White

Zhao

2015

Progress in Nuclear Energy

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“…Hence, Kazimi et al [92] proposed the annular fuel rods will show better performance than the default solid fuel pin design. The United States of America conducted independent feasibility studies for water and Sodium-cooled Fast Reactors (SFRs) [93][94][95], while South Korea evaluationed for water and gas-cooled reactors [86,[96][97][98][99][100]. The temperature distribution is much less in annular fuel elements, while the pressure drop along the channel can be minimized by optimizing the fuel assembly geometry.…”

Section: Introductionmentioning

confidence: 99%

“…SuperCritical Water-cooled Reactors (SCWRs) [101][102][103][104]. (iii) Gas-cooled Fast Reactors (GFRs) [48,49,97].…”

Section: Introductionmentioning

confidence: 99%

Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

Rowinski¹

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Analytical methods and Computational Fluid Dynamics (CFD) examined the safety and performance of advanced nuclear reactors. Thermal and pressure drop evaluations of an innovative annular fuel in the European lead-cooled System ELSY fast reactor were conducted. Annular fuel has superior cooling capacity, and for a square fuel lattice, the pressure drop can be lower than the reference value for standard solid fuel. Consequently, when annular fuel rods replace solid fuel rods of the same dimensions the power rating is upgraded. While the hexagonal lattice cools more effectively, the pressure drop is higher than the standard fuel assembly and incompatible with power augmentation. Numerical investigations of supercritical flow of both carbon dioxide and water in a vertical tube under non-uniform heat flux applied at the wall demonstrated that current correlations for the Heat Transfer Coefficient (HTC) at low enthalpy values are accurate to ±15%. At high enthalpy, only the general trend is replicated due to large property value differences near the wall and in the bulk flow. In these circumstances, the thermal conductivity, specific heat, density and viscosity are 5-8 times lower after the transition from sub-to super-critical conditions. Moreover, the efficiency of different HTC correlation models is coolant dependant, e.g. the Swenson* formula proves superior for carbon dioxide, whereas the Ornatsky model** provides better agreement with water, where a lower maximum wall temperature was obtained when a non-uniform heat flux was applied, as compared to the reference case of uniform heat flux. For supercritical water flow inside a 2  2 fuel rod bundle, non-uniform heat flux increases wall temperature beyond the reference uniform flux. Also, wall temperature peaks occur only at the gaps between the fuel rods.

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Development of a thermal hydraulic analysis code for gas-cooled reactors with annular fuels

Cited by 14 publications

References 6 publications

Researches on Special Thermal Hydraulic Phenomena of Annular Fuel Assembly by Sub-Channel Analysis Code-NACAF

Researches on Special Thermal Hydraulic Phenomena of Annular Fuel Assembly by Sub-Channel Analysis Code-NACAF

Innovative model of annular fuel design for lead-cooled fast reactors

Computational fluid dynamics of advanced nuclear reactor cores and its contribution to safety analysis

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