• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.274-284
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• 1992

A 1/15-scale CREARE experiment, which simulates the thermal-hydraulic behavior in the reactor pressure vessel of a PWR during a hypothetical Loss Of Coolant Accident, has been analyzed using CATHARE code for the associated model assessment to represent the phenomenon. The key parameters examined in the CREARE experiment were known as ECC water injection rate. ECC water subcooling, system pressure, and steam flow rate coming out from the core bottom. The present CATHARE simulation, however, has been mainly focused on qualitative analysis of a countercurrent flow in the downcomer. The discrepancy of the simulation results with the experimental data is considered arising primarily from an inadequate numerical representation as well as an interfacial friction model. Accordingly it is suggested from the sensitivity studies that either multidimensional approach or further examination of momentum equations at a junction near a volume element in CATHARE be necessary in order to represent the phenomenon more realistically.

• Korean Chemical Engineering Research
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• v.48 no.2
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• pp.198-204
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• 2010

Solids flow characteristics have been determined in a pressurized solids recycle systems of silica sand particles for the application in a pressurized CFB(PCFB). The solids recycle system is composed of a downcomer(0.10 m i.d. 2.25 m high) and a loop-seal(0.10 m i.d.). The silica sand($d_p=240{\mu}m$, ${\rho}_s=2582kg/m^3$) particles were transported at room temperature and system pressure($P_{sys}$) up to 0.71 MPa using air. Solids mass flux($G_s$) increases with increasing system pressure at constant aeration rate. Pressure gradient, solids velocity and actual gas velocity increase with increasing $P_{sys}$ at constant aeration rate. The Pressure drop number($\Phi$) on pressure gradient in downcomer has been correlated with Transportation number(Tr). Pressure drop across the loop-seal increases with increasing of $G_s$ irrespective of variation of $P_{sys}$. The obtained $G_s$ and Transportation number(Tr) have been correlated with the experimental variables.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.315-326
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• 2001

Direct ECC bypass phenomena that occur in a reactor vessel downcomer with a Direct Vessel Injection (DVI) system during the reflood phase of a Large Break Loss-of-Coolant Accident (LBLOCA) are experimentally investigated using a transparent l/7.5 scaled down test facility of the Upper Plenum Test Facility (UPTF). A series of separate effect tests are peformed in order to investigate the mechanisms of direct ECC bypass and to find out its scaling parameters. Various flow regimes and phasic distribution in downcomer are identified and mapped, and the fraction of direct ECC bypass is measured under a wide range of air and water injection conditions. From the counterpart test of the UPTF Test 21-D, the dimensionless gas velocity ( $j^{*}$$_{g,eff}$) is derived experimentally, which is believed to be a major scaling parameter for the fraction of direct ECC bypass. And it is found out that the direct ECC bypass is greatly affected by the spreading width of ECC water film and the geometric configuration of the downcomer.r.

• Journal of ILASS-Korea
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• v.8 no.1
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• pp.23-28
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• 2003

When a cold HPSI (High Pressure Safety Injection) fluid associated with a design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena may arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated collant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

• Journal of Mechanical Science and Technology
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• v.17 no.9
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• pp.1380-1387
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• 2003

When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.

• Journal of Environmental Science International
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• v.5 no.1
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• pp.93-102
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• 1996

Hydrodynamic characteristics such as gas holdup, liquid circulation velocity and bed expansion in a hexagonal inverse fluidized bed were investigated using air-water system by changing the ratio ($A_d$/$A_r$) of cross-sectional area between the riser and the downcomer, the liquid level($H_1$/H), and the superficial gas velocity($U_g$). The gas holdup and the liquid circulation velocity were steadily increased with the superficial gas velocity increasing, but at high superficial gas velocity, some of gas bubbles were carried over to a downcomer and circulated through the column. When the superficial gas velocity was high, the $A_d$/$A_r$ ratio in the range of 1 to 2.4 did not affect the liquid circulation velocity, but the maximum bed expansion was obtained at $A_d$/$A_r$ ratio of 1.25. The liquid circulation velocity was expressed as a model equation below with variables of the cross-sectional area ratio($A_d$/$A_r$) between riser to downcomer, the liquid level($H_1$/H), the superficial gas velocity($U_g$), the sparser height[(H-$H_s$)/H], and the draft Plate level($H_b$/H). $U_{ld}$ = 11.62U_g^{0.75}$${(\frac{H_1}{H})}^{10.30}$${(\frac{A_d}{A_r})}^{-0.52}$${(\frac({H-H_s}{H})}^{0.91}$${(\frac{H_b}{H})}^{0.13}$

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.200-208
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• 1986

The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.435-436
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.259.2-259
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.144-144
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• 2002