An Investigation of the Collapse and Surface Rewet in Film Boiling in Forced Vertical Flow

1975 ◽  
Vol 97 (2) ◽  
pp. 166-172 ◽  
Author(s):  
O. C. Iloeje ◽  
D. N. Plummer ◽  
W. M. Rohsenow ◽  
P. Griffith
Keyword(s):  
Nucleate Boiling ◽  
Boiling Temperature ◽  
Film Boiling ◽  
Test Surface ◽  
Loss Of Coolant Accident ◽  
Mass Fluxes ◽  
Loss Of Coolant ◽  
Scale Deposit ◽  
Transient Boiling ◽  
Scale Deposits

A transient boiling experiment has been run in such a way that one can acquire data in forced convection film, transition, and nucleate boiling regions for a specified pressure, quality, and mass flux. Transient boiling experiments were conducted at the Nuclear Energy Division of the General Electric Company for water in a 0.492 in. ID inconel X-750 tube at mass fluxes of 50,000, 100,000, and 250,000 lbm/hr-ft2, quality range of 30–100 percent and a pressure of 1000 psia. The reduced boiling curves for these data indicated temperature differences at burnout on the order of 100–200° F and temperature differences at the minimum ranging from 700 to 1100° F. These results (higher than in other experiments) are felt to be caused by scale deposit, axial conduction, and roughnesses on the test surface. Physical evidence indicates that the test surface became coated with an appreciable scale deposit when subjected to the initial temperatures in excess of 1500 °F in a steam atmosphere. It has been found (reference [1]) that BWR fuel will normally have scale deposit on the heat transfer surface and thus the qualitative effects of scale deposits in this report are expected to apply in BWR Loss-of-Coolant accident evaluation. An empirical correlation was developed for the data for minimum film boiling temperature differences. The correlation was based on Berenson’s minimum pool film boiling temperature difference correlation in order to provide a technique for extrapolating to different pressures.

Effect of Subcooling on the Quenching of a Simulated Nuclear Fuel Rod

2001 ◽  
Author(s):  
J. Sinha ◽  
L. E. Hochreiter ◽  
F. B. Cheung
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Reactor ◽  
Nucleate Boiling ◽  
Boiling Curve ◽  
Boiling Temperature ◽  
Film Boiling ◽  
Working Fluid ◽  
Filler Materials ◽  
Fuel Rod ◽  
The Right

Abstract An experimental study of the effect of liquid subcooling on the minimum film boiling temperature during quenching of a simulated nuclear fuel rod was performed under carefully controlled laboratory conditions. The rod was designed and fabricated with a proper combination of cladding and filler materials and was instrumented with embedded thermocouples distributed at various axial and radial locations. Quenching of the rods was done using distilled water with various degrees of subcooling as the working fluid. It was found that the rate of quenching depends strongly on the liquid subcooling. Multiple maxima were observed between film and nucleate boiling in the boiling curve for high subcoolings. Increasing subcooling had the effect of moving the entire boiling curve up and to the right, resulting in a considerably high minimum film boiling temperature. The strong effect of liquid subcooling on the quenching of a fuel rod should be properly accounted for in the study of reflood heat transfer in a nuclear reactor following a design-based accident.

scholarly journals Quenching Experiments with CrAl-coated Zircaloy Cladding in Reflooding Water Flows

Energies ◽  
2021 ◽  
Vol 14 (7) ◽  
pp. 1859
Author(s):  
Wang Kee In ◽  
Kwan Geun Lee
Keyword(s):  
Nuclear Power ◽  
Early Stage ◽  
Nucleate Boiling ◽  
Front Velocity ◽  
Light Water Reactors ◽  
Test Conditions ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Film Layer ◽  
Water Reactors

A quenching experiment is performed to investigate the heat transfer characteristics and cooling performance of CrAl-coated Zircaloy (Zr) cladding in a water flow. The CrAl-coated Zr cladding is one of the accident tolerant fuels for light water reactors. The uncoated Zr cladding is also used in this quenching experiment for comparison. This experiment simulates reflood quenching of fuel rod during loss of coolant accident (LOCA) in nuclear power plant. The test conditions were determined to represent the peak cladding temperature, the coolant subcooling and the reflood velocity in the event of LOCA. The flow visualization showed the film boiling during early stage of reflood quenching and the transition to nucleate boiling. The film layer decreases as the coolant subcooling increases and becomes wavy as the reflood velocity increases. The CrAl-coated Zr cladding showed more wavy and thinner film than the uncoated Zr cladding. The rewetting temperature increases as the initial wall temperature and/or the coolant subcooling increases. The quench front velocity increases significantly as the coolant subcooling increases. The reflood velocity has a negligible effect on rewetting temperature and quench front velocity.

scholarly journals Multiphysics Investigation on Coolant Thermohydraulic Conditions and Fuel Rod Behavior During a Loss-of-Coolant Accident

2021 ◽  
Vol 9 ◽  
Author(s):  
Meng Yu ◽  
Libo Qian ◽  
Wenzhong Zhou
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Void Fraction ◽  
Mechanical Performance ◽  
Nucleate Boiling ◽  
Physical Models ◽  
Coolant Temperature ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

This article simulates the multiphysics coolant thermohydraulic conditions and fuel performance of a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA). In the coolant channel of a PWR, the coolant undergoes a series of different boiling regimes along the axial direction. At the inlet of the coolant channel, heat exchange between the cladding wall and coolant is based on single-phase forced convection. As the coolant flow distance increases, the boiling regime gradually converts to nucleate boiling. When a LOCA occurs, on the one hand, the coolant flux and coolant pressure decrease sharply; on the other hand, the heat flux at the cladding wall decreases relatively slowly. They both contribute to a swift increase in coolant temperature. As a consequence, a boiling crisis may occur as critical heat flux (CHF) decreases. In this article, the void fraction along the length of coolant channel in a reactor and mechanical performance of Zr cladding enwrapping UO2 fuel are investigated by establishing a fully coupled multiphysics model based on the CAMPUS code. Physical models of coolant boiling regimes are implemented into the CAMPUS code by adopting different heat transfer models and void fraction models. Physical properties of the coolant are implemented into the CAMPUS code using curve-fitting results. All physical models and parameters related to solid heat transfer are implemented into the CAMPUS code with a 2D axisymmetric geometry. The modeling results help enhance our understanding of void fraction along the length of the coolant channel and mechanical performance of Zr cladding enwrapping UO2 fuel under different coolant pressure and mass flux conditions during a LOCA.

Modeling Cladding-Coolant Heat Transfer of High-Burnup Fuel During RIA

2006 ◽  
Author(s):  
Wenfeng Liu ◽  
Mujid S. Kazimi
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Thermal Properties ◽  
Heat Conduction ◽  
Nucleate Boiling ◽  
Boiling Temperature ◽  
Film Boiling ◽  
Solid Contact ◽  
Boiling Regime ◽  
High Burnup

This paper describes a model for the cladding-coolant heat transfer of high burnup fuel during a Reactivity Initiated Accident (RIA) which is implemented in the fuel performance code FRAPTRAN 1.2. The minimum stable film boiling temperature, affected by the subcooling and the clad oxidation, is modeled by a modified Henry correlation. This accounts for the effects of thermal properties of the cladding surface on the transient temperature drop during liquid-solid contact. The transition boiling regime is described as the interpolation of the heat flux between two anchor points on the boiling curve: the Critical Heat Flux (CHF) and minimum stable film boiling. The CHF correlation is based on the Zuber hydrodynamic model multiplied by a subcooling factor. Frederking correlation is chosen to model the film boiling regime. The heat conduction through the oxide layer of the cladding surface of high burnup fuel is calculated by solving heat conduction equations with thermal properties of zirconia taken from MATPRO. This model is validated in the FRAPTRAN code for test cases of both high burnup and fresh test fuel rods including the burnup level (0–56 MW d/kg), peak fuel enthalpy deposit (70–190 cal/g), degree of subcooling (0–80 °C), and extent of oxidation (0–25 micron). The modified code demonstrates the capability of differentiating between the departure from nucleate boiling (DNB) and none-DNB cases. The predicted peak cladding temperature (PCT) and duration of DNB achieves generally good agreement with the experimental data. It is found that the cladding surface oxidation of high burnup fuel causes an early rewetting of cladding or suppresses DNB due to two factors: 1) Thick zirconia layer may delay the heat conducted to the surface while keeping the surface heat transfer in the most effective nucleate boiling regime. 2) The transient liquid-solid contact resulting from vapor breaking down would cause a lower interface temperature for an oxidized surface, essentially raises the minimum stable film boiling temperature.

scholarly journals A Study of Prediction Model Improvement for Air-Oxidation Breakaway in a Postulated Spent Nuclear Fuel Pool Complete Loss of Coolant Accident

2021 ◽  
Vol 13 (3) ◽  
pp. 1442
Author(s):  
Sanggil Park ◽  
Jaeyoung Lee ◽  
Min Bum Park
Keyword(s):  
Kinetic Model ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Complete Loss ◽  
Air Oxidation ◽  
Loss Of Coolant Accident ◽  
Spent Nuclear Fuel Pool ◽  
Loss Of Coolant ◽  
Model Derivation ◽  
Model Improvement

The temperature of zirconium alloy cladding on the postulated spent nuclear fuel pool complete loss of coolant accident is abruptly increased at a certain time and the cladding is almost fully oxidized to weak ZrO2 in the air. This abrupt temperature escalation phenomenon induced by the air-oxidation breakaway is called a zirconium fire. Although an air-oxidation breakaway kinetic model correlated between time and temperature has been implemented in the MELCOR code, it is likely to bring about unexpected large errors because of many limitations of model derivation. This study suggests an improved time–temperature correlated kinetic model using the Johnson–Mehl equation. It is based on that the air-oxidation breakaway is initiated by the phase transformation from the tetragonal to monoclinic ZrO2 at the oxide–metal interface in the cladding. This new model equation is also evaluated with the Zry-4 air-oxidation literature data. This equation resulted in the almost similar air-oxidation breakaway timing to the actual experimental data at 800 °C. However, at 1000 °C, it showed an error of about 8 min. This could be inferred from the influence of the ZrN phase change due to the nitrogen existing in air.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

2021 ◽  
Vol 134 ◽  
pp. 103648
Author(s):  
Katarzyna Skolik ◽  
Chris Allison ◽  
Judith Hohorst ◽  
Mateusz Malicki ◽  
Marina Perez-Ferragut ◽  
...  
Keyword(s):  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant
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