Uncertainty Quantification of the RELAP5 Interfacial Friction Model in the Rod Bundle Geometry

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Ikuo Kinoshita ◽  
Toshihide Torige ◽  
Minoru Yamada
Keyword(s):  
Model Uncertainty ◽  
National Laboratory ◽  
Friction Model ◽  
Numerical Analyses ◽  
Test Facility ◽  
Interfacial Friction ◽  
Oak Ridge ◽  
Drift Flux Model ◽  
Drift Flux ◽  
Rod Bundle

Interfacial friction in the core affects the two-phase mixture level and the distribution of the dispersed gas phase during a small-break loss-of-coolant accident (LOCA). The RELAP5/MOD3.2 code uses the drift flux model to describe the interfacial friction force in vertical dispersed flow, and the Chexal–Lellouche drift flux correlation is used for the rod bundle geometry. In the present study, the RELAP5 model uncertainty was quantified for the bubbly–slug interfacial friction model in the rod bundle geometry by conducting numerical analyses of void profile tests in the Thermal Hydraulic Test Facility (THTF) of the Oak Ridge National Laboratory (ORNL). The model uncertainty parameter was defined as a multiplier for the interfacial friction coefficient. Numerical analyses were performed by adjusting the multiplier so that the predicted void fractions agreed with the measured test data. The resultant distribution of the multipliers represented the interfacial friction model uncertainty.

Uncertainty Quantification of the RELAP5 Interfacial Friction Model in the Rod Bundle Geometry

2014 ◽  
Author(s):  
Ikuo Kinoshita ◽  
Toshihide Torige ◽  
Minoru Yamada
Keyword(s):  
Model Uncertainty ◽  
Numerical Analyses ◽  
Test Facility ◽  
Interfacial Friction ◽  
Two Phase ◽  
Friction Coefficients ◽  
Oak Ridge ◽  
Rod Bundle ◽  
Separate Effect ◽  
Phase Mixture

An application of the Best Estimate Plus Uncertainty (BEPU) method is made to an analysis of the “Intentional depressurizaion of steam generator secondary side” which is an accident management procedure in a small-break loss-of-coolant accident (SBLOCA) with high pressure injection (HPI) system failure. RELAP5/MOD3.2 is used as the system analysis code. Interfacial friction in the core affects the two-phase mixture level and the distribution of the dispersed gas phase. This phenomenon is very important in terms of the influence its uncertainty has on the peak cladding temperature. The RELAP5/MOD3.2 code uses drift-velocity to describe the interfacial friction coefficients in vertical dispersed flow. The Chexal-Lellouche drift-flux correlation is used for the rod bundle geometry. In the present study, the RELAP5 model uncertainty was quantified regarding the interfacial friction coefficients in the rod bundle geometry by conducting numerical analyses of separate effect tests. As the separate effect tests, two-phase mixture level swell tests in the Thermal Hydraulic Test Facility (THTF) of the Oak Ridge National Laboratory (ORNL) were used. After considering applicability to the SBLOCA, tests were selected for which conditions of pressures and rod powers were similar to PWR plant conditions. A total of 55 data were used. The model uncertainty parameter was defined as a multiplier for the interfacial friction coefficient. Numerical analyses were performed by adjusting the multiplier so that the predicted void fractions agreed with the experimental measured data. The resultant distribution of the multipliers represented the model uncertainty. The mean, standard deviation, minimum and maximum values of this uncertainty distribution were 0.88, 0.55, 0.13 and 3.0, respectively.

One-dimensional drift-flux model for two-phase flow in pool rod bundle systems

2012 ◽  
Vol 40 ◽  
pp. 166-177 ◽  
Author(s):  
Shao-Wen Chen ◽  
Yang Liu ◽  
Takashi Hibiki ◽  
Mamoru Ishii ◽  
Yoshitaka Yoshida ◽  
...  
Keyword(s):  
Two Phase Flow ◽  
Phase Flow ◽  
Two Phase ◽  
One Dimensional ◽  
Drift Flux Model ◽  
Drift Flux ◽  
Rod Bundle ◽  
Flux Model

Drift-flux model for rod bundle geometry

2015 ◽  
Vol 83 ◽  
pp. 229-247 ◽  
Author(s):  
Tetsuhiro Ozaki ◽  
Takashi Hibiki
Keyword(s):  
Drift Flux Model ◽  
Drift Flux ◽  
Rod Bundle ◽  
Flux Model

scholarly journals Conceptual Design of a MEDE Treatment System for Sodium Bonded Fuel

2008 ◽  
Author(s):  
Carl E. Baily ◽  
Karen A. Moore ◽  
Collin J. Knight ◽  
Peter B. Wells ◽  
Paul J. Petersen ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Department Of Energy ◽  
Test Facility ◽  
Evaporation Chamber ◽  
Oak Ridge ◽  
Uranium Fuel ◽  
Fuel Assemblies ◽  
Metal Fuel ◽  
Enriched Uranium

Unirradiated sodium bonded metal fuel and casting scrap material containing highly enriched uranium (HEU) is stored at the Materials and Fuels Complex (MFC) on the Idaho National Laboratory (INL). This material, which includes intact fuel assemblies and elements from the Fast Flux Test Facility (FFTF) and Experimental Breeder Reactor-II (EBR-II) reactors, as well as scrap material from the casting of these fuels, has no current use under the terminated reactor programs for both facilities. The Department of Energy (DOE), under the Sodium-Bonded Spent Nuclear Fuel Treatment Record of Decision (ROD), has determined that this material could be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for commercial nuclear reactors. A plan is being developed to prepare, package, and transfer this material to the DOE HEU Disposition Program Office (HDPO), located at the Y-12 National Security Complex in Oak Ridge, Tennessee. Disposition of the sodium bonded material will require separating the elemental sodium from the metallic uranium fuel. A sodium distillation process known as MEDE (Melt-Drain-Evaporate), will be used for the separation process. The casting scrap material needs to be sorted to remove any foreign material or fines that are not acceptable to the HDPO program. Once all elements have been cut and loaded into baskets, they are then loaded into an evaporation chamber as the first step in the MEDE process. The chamber will be sealed and the pressure reduced to approximately 200 mtorr. The chamber will then be heated as high as 650 °C, causing the sodium to melt and then vaporize. The vapor phase sodium will be driven into an outlet line where it is condensed and drained into a receiver vessel. Once the evaporation operation is complete, the system is de-energized and returned to atmospheric pressure. This paper describes the MEDE process as well as a general overview of the furnace systems, as necessary, to complete the MEDE process.

Creep Strength and Microstructure of AL20-25+Nb Alloy Sheets and Foils for Advanced Microturbine Recuperators

2006 ◽  
Author(s):  
Philip J. Maziasz ◽  
John P. Shingledecker ◽  
Neal D. Evans ◽  
Yukinori Yamamoto ◽  
Karren L. More ◽  
...  
Keyword(s):  
Phase I ◽  
Phase Ii ◽  
Creep Resistance ◽  
National Laboratory ◽  
Rupture Life ◽  
Cost Effective ◽  
Test Facility ◽  
Oak Ridge ◽  
Collaborative Program ◽  
Wide Range

The Oak Ridge National Laboratory (ORNL) and ATI Allegheny-Ludlum began a collaborative program in 2004 to produce a wide range of commercial sheets and foils of the new AL20-25+Nb stainless alloy, specifically designed for advanced microturbine recuperator applications. There is a need for cost-effective sheets/foils with more performance and reliability at 650–750°C than 347 stainless steel, particularly for larger 200–250 kW microturbines. Phase I of this collaborative program produced the sheets and foils needed for manufacturing brazed plated-fin (BPF) aircells, while Phase II provided foils for primary surface (PS) aircells, and modified processing to change the microstructure of sheets and foils for improved creep-resistance. Phase I sheets and foils of AL20-25+Nb have much more creep-resistance than 347 steel at 700–750°C, and foils are slightly stronger than HR120 and HR230. Preliminary results for Phase II show nearly double the creep-rupture life of sheets at 750°C/100 MPa, with the first foils tested approaching the creep resistance of alloy 625 foils. AL20-25+Nb alloy foils are also now being tested in the ORNL Recuperator Test Facility.

Investigation of the Structure Velocity in a 3x3 Rod Bundle Under Bubbly and Cap-Bubbly Flow Regimes

2020 ◽  
Author(s):  
Pei-Syuan Ruan ◽  
Shao-Wen Chen ◽  
Min-Song Lin ◽  
Jin-Der Lee ◽  
Jong-Rong Wang
Keyword(s):  
Two Phase Flow ◽  
Bubbly Flow ◽  
Flow Regimes ◽  
Phase Flow ◽  
Two Phase ◽  
Drift Flux Model ◽  
Drift Flux ◽  
Rod Bundle ◽  
Flux Model ◽  
Sudden Jump

Abstract This paper presents the experimental results and analyses of the structure velocity of air-water two-phase flow in a 3 × 3 rod bundle channel. A total of 56 flow conditions were tested and investigated for rod-gap, sub-channel, rod-wall and global regions of rod bundle geometry. The experimental tests were carried out under bubbly and cap-bubbly flow regimes with superficial gas and liquid velocities of 0–1 m/s and 1–1.7 m/s, respectively. The conductivity probes were set at different heights to measure the global and local void fractions. The structure velocity of air-water two-phase flow is the average bubble velocity calculated by the method in this study. The structure velocity were determined by utilizing the cross-correlation technique to analyze the time lags of the bubbles passing through the conductivity probes. The results of this study indicated that the structure velocity may increase with increasing superficial gas and liquid velocities. In low superficial gas velocity region, the structure velocity may first slightly increase and follow by a sudden jump which appear in most regions. After the sudden jump, the structure velocity may keep increasing mildly. The present structure velocities have been compared with the area-averaged gas velocities predicted by the drift flux model, and it appears that most structure velocities show a good agreement with the averaged gas velocities from the drift flux model after the jump.

Validation of Selected Cesar Friction Models of the ASTECV21 Code Based on Moby Dick Experiments

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
I. Gómez-García-Toraño ◽  
L. Laborde
Keyword(s):  
Momentum Equation ◽  
Test Facility ◽  
Primary Circuit ◽  
Correct Prediction ◽  
Moby Dick ◽  
Drift Flux Model ◽  
Drift Flux ◽  
Mass Fluxes ◽  
Mass Fractions ◽  
Friction Models

In the event of a loss of integrity of the main coolant line, a large mass and energy release from the primary circuit to the containment is to be expected. The temporal evolution of such depressurization is mainly governed by the critical flow, whose correct prediction requires, in first place, a correct description of the different friction terms. Within this work, selected friction models of the CESAR module of the Accident Source Term Evaluation Code (ASTEC) V2.1 integral code are validated against data from the Moby Dick test facility. Simulations are launched using two different numerical schemes: on the one hand, the classical five equation (drift flux) approach, with one momentum conservation equation for an average fluid plus one algebraic equation on the drift between the gas and the liquid; on the other hand, the recently implemented six equation approach, where two differential equations are used to obtain the phase velocities. The main findings are listed hereafter: The use of five equations provides an adequate description of the pressure loss as long as the mass fluxes remain below 1.24 kg/cm2 s and the gas mass fractions below 5.93 × 10 − 4. Beyond those conditions, the hypotheses of the drift flux model are exceeded and the use of an additional momentum equation is required. The use of an additional momentum equation leads to a better agreement with the experimental data for a wider range of mass fluxes and gas mass fractions. However, the qualitative prediction for high gas mass fractions still shows some deviations due to the decrease of the regular friction term at the end of the test section.

scholarly journals The use of acoustic inversion to estimate the bubble size distribution in pipelines

2012 ◽  
Vol 468 (2145) ◽  
pp. 2461-2484 ◽  
Author(s):  
Timothy G. Leighton ◽  
Kyungmin Baik ◽  
Jian Jiang
Keyword(s):  
Size Distribution ◽  
Bubble Size ◽  
National Laboratory ◽  
Free Field ◽  
Travelling Wave ◽  
Bubble Size Distribution ◽  
Optical Measurements ◽  
Test Facility ◽  
Oak Ridge ◽  
Acoustic Inversion

The most popular technique for estimating the gas bubble size distribution (BSD) in liquids is through the inversion of measured attenuation and/or sound speed of a travelling wave. The model inherent in such inversions never exactly matches the conditions of the measurement, and the size of the resulting error (which could well be small in quasi-free field conditions) cannot be quantified if only a free field code exists. Users may be unaware of errors because, with sufficient regularization, such inversions can always be made to produce an answer, the accuracy of which is unknown unless independent (e.g. optical) measurements are made. This study was commissioned to assess the size of this error for the mercury-filled steel pipelines of the target test facility (TTF) of the spallation neutron source at Oak Ridge National Laboratory, TN, USA. Large errors in estimating the BSD (greater than 1000% overcounts/undercounts) are predicted. A new inversion technique appropriate for pipelines such as TTF gives good BSD estimations if the frequency range is sufficiently broad. However, it also shows that implementation of the planned reduction in frequency bandwidth for the TTF bubble sensor would make even this inversion insufficient to obtain an accurate BSD in TTF.

Drift-flux model in a sub-channel of rod bundle geometry

2009 ◽  
Vol 52 (13-14) ◽  
pp. 3032-3041 ◽  
Author(s):  
J. Enrique Julia ◽  
Takashi Hibiki ◽  
Mamoru Ishii ◽  
Byong-Jo Yun ◽  
Goon-Cherl Park
Keyword(s):  
Drift Flux Model ◽  
Drift Flux ◽  
Rod Bundle ◽  
Flux Model
Sign in / Sign up

Export Citation Format

Share Document