Sensitivity Analysis of Unsteady Corium Solidification in an Initially Emptied Horizontal Turbulent Pipe Flow

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Shawn Somers-Neal ◽  
Vien Nguyen ◽  
Edgar Matida ◽  
Vinh Tang ◽  
Tarik Kaya
Keyword(s):  
Reactor Core ◽  
Horizontal Tube ◽  
Sensitivity Study ◽  
Turbulent Pipe Flow ◽  
Bulk Temperature ◽  
Inlet Temperature ◽  
Melt Flow ◽  
Penetration Length ◽  
Horizontal Pipes ◽  
Melt Penetration

Abstract In a reactor core meltdown under postulated severe accidents, the molten material called corium could be ejected or relocated through existing vessel penetrations. There exists, however, a potential for plugging of melt flow due to its complete solidification providing the availability of an adequate heat sink. Simulations of the melt flow in a horizontal tube were carried out to conduct a sensitivity study on the effect of key parameters on the melt penetration distance and bulk temperature distribution of the corium. The Reynolds number was varied from 10,000 to 20,000, inlet temperature was varied from 2600 K to 3000 K, the corium thermal conductivity was varied from 10 W/m·K to 20 W/m·K, and the pipe diameter was varied from 0.0095 m to 0.019 m. In addition, a comparison was made with an analytical model based on a modified Epstein's model and a previous numerical model. The study provided insight into the lower bound, which was found to be 98 mm, and the upper bound was 258 mm when predicting the potential penetration length of corium in horizontal pipes.

Heat transfer characteristic modelling and the effect of operating conditions on re-cooled cycle for a scramjet

2011 ◽  
Vol 115 (1164) ◽  
pp. 83-90 ◽  
Author(s):  
W. Bao ◽  
J. Qin ◽  
W. X. Zhou
Keyword(s):  
Heat Transfer ◽  
Heat Sink ◽  
Transfer Characteristic ◽  
Cooling Capacity ◽  
Sensitivity Study ◽  
Performance Model ◽  
Operating Conditions ◽  
Inlet Temperature ◽  
Fuel Flow ◽  
Temperature And Pressure

Abstract A re-cooled cycle has been proposed for a regeneratively cooled scramjet to reduce the hydrogen fuel flow for cooling. Upon the completion of the first cooling, fuel can be used for secondary cooling by transferring the enthalpy from fuel to work. Fuel heat sink (cooling capacity) is thus repeatedly used and fuel heat sink is indirectly increased. Instead of carrying excess fuel for cooling or seeking for any new coolant, the cooling fuel flow is reduced, and fuel onboard is adequate to satisfy the cooling requirement for the whole hypersonic vehicle. A performance model considering flow and heat transfer is build. A model sensitivity study of inlet temperature and pressure reveals that, for given exterior heating condition and cooling panel size, fuel heat sink can be obviously increased at moderate inlet temperature and pressure. Simultaneously the low-temperature heat transfer deterioration and Mach number constrains can also be avoided.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

Experimental and Numerical Investigation of Solidification of Gallium in an Initially Emptied Horizontal Pipe Flow

2020 ◽  
Author(s):  
Shawn Somers-Neal ◽  
Alex Pegarkov ◽  
Edgar Matida ◽  
Vinh Tang ◽  
Tarik Kaya
Keyword(s):  
Numerical Model ◽  
Molten Metal ◽  
Reactor Core ◽  
Experimental Tests ◽  
Severe Accident ◽  
Penetration Length ◽  
Horizontal Pipe ◽  
Cooling Pipe ◽  
Severe Accidents ◽  
Core Meltdown

Abstract In a reactor core meltdown under postulated severe accidents, the molten material (corium) could be ejected or relocated through existing vessel penetrations (cooling pipe connections), thus potentially contaminating other locations in the power plant. There exists, however, a potential for plugging of melt flow due to its complete solidification, providing the availability of an adequate heat sink. Therefore, a numerical model was created to simulate the flow of molten metal through an initially empty horizontal pipe. The numerical model was verified using a previously developed analytical model and validated against experimental tests with gallium (low melting temperature) as a substitute for corium. The numerical model was able to predict the penetration length (length of distance travelled by the molten metal) after a complete blockage occurred with an average percent error range of 9%. Since the numerical model has been verified and validated, the model can be updated to predict the penetration length in the cooling pipe in case of a severe accident.

Towards Development of Uncertainty Library for Nuclear Reactor Core Simulation

2018 ◽  
Author(s):  
Hany S. Abdel-Khalik ◽  
Dongli Huang ◽  
Ondrej Chvala ◽  
G. Ivan Maldonado
Keyword(s):  
Cross Sections ◽  
Degrees Of Freedom ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Sensitivity Study ◽  
Reactor Physics ◽  
Rigorous Approach ◽  
Uncertainty Space ◽  
Core Simulation

Uncertainty quantification is an indispensable analysis for nuclear reactor simulation as it provides a rigorous approach by which the credibility of the predictions can be assessed. Focusing on propagation of multi-group cross-sections, the major challenge lies in the enormous size of the uncertainty space. Earlier work has explored the use of the physics-guided coverage mapping (PCM) methodology to assess the quality of the assumptions typically employed to reduce the size of the uncertainty space. A reduced order modeling (ROM) approach has been further developed to identify the active degrees of freedom (DOFs) of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. In the current work, a sensitivity study, based on the PCM and ROM results, is applied to identify a suitable compressed representation of the uncertainty space to render feasible the quantification and prioritization of the various sources of uncertainties. While the proposed developments are general to any reactor physics computational sequence, the proposed approach is customized to the TRITON-NESTLE computational sequence, simulating the BWR lattice model and the core model, which will serve as a demonstrative tool for the implementation of the algorithms.

scholarly journals Numerical Simulation of Reciprocating Flow Forced Convection in Two-Dimensional Channels

2003 ◽  
Vol 125 (3) ◽  
pp. 403-412 ◽  
Author(s):  
Cuneyt Sert ◽  
Ali Beskok
Keyword(s):  
Boundary Conditions ◽  
Forced Convection ◽  
Temperature Fields ◽  
Bottom Surface ◽  
Bulk Temperature ◽  
Element Formulation ◽  
Two Dimensional ◽  
Penetration Length ◽  
Time Accuracy ◽  
Reciprocating Flow

Numerical simulations of laminar, forced convection heat transfer for reciprocating, two-dimensional channel flows are performed as a function of the penetration length, Womersley (α) and Prandtl (Pr) numbers. The numerical algorithm is based on a spectral element formulation, which enables high-order spatial resolution with exponential decay of discretization errors, and second-order time-accuracy. Uniform heat flux and constant temperature boundary conditions are imposed on certain regions of the top surface, while the bottom surface is kept insulated. Periodicity of velocity and temperature fields is imposed on the side boundaries, while the flow is driven by an oscillating pressure gradient. These sets of boundary conditions enable time-periodic solution of the problem. Instantaneous and time-averaged surface and bulk temperature distributions, and Nusselt number variations are presented. For high α flows, the temperature field is significantly affected by the Richardson’s annular effect. Overall, forced convection increases by increasing the penetration length, α and Pr. Corresponding steady-flow simulations are performed by matching the volumetric flowrate. For the limited parameter space investigated in this paper, steady unidirectional forced convection is more effective than the reciprocating flow forced convection.

Sensitivity Study of the South Texas Project Power Plant Steady-State Simulations Using RELAP5-3D Coupled With Dakota

2012 ◽  
Author(s):  
O. A. Rodriguez ◽  
R. Vaghetto ◽  
Y. A. Hassan
Keyword(s):  
Steady State ◽  
Power Plant ◽  
Uncertainty Quantification ◽  
National Laboratory ◽  
Sandia National Laboratory ◽  
Sensitivity Study ◽  
South Texas ◽  
Normal Operation ◽  
Inlet Temperature ◽  
The South

A RELAP5-3D input deck of the South Texas Project (STP) power plant was created in order to study the thermal-hydraulic behavior of the plant during normal operation (steady-state) and during a Loss of Coolant Accident (LOCA). It is important to study the sensitivity of selected output parameters such as the total coolant mass flow rate, the peak clad temperature, the secondary pressure, as a function of specific input parameters (reactor nominal power, vessel inlet temperature, steam generators primary side heat transfer coefficient, primary pressure etc.) in order to identify the variables that play a role in the uncertainty of the thermal-hydraulic calculations. RELAP5-3D, one of the most used best estimate thermal-hydraulic system codes, was coupled with DAKOTA, developed by Sandia National Laboratory for Uncertainty Quantification and Sensitivity Analysis in order to simplify the simulation process and the analysis of the results. In the present paper, the results of the sensitivity study for selected output parameters of the steady-state simulations are presented. The coupled software was validated by repeating one set of simulations using the RELAP5-3D standalone version and by analyzing the simulation results with respect of the physical expectations and behavior of the power plant. The thermal-hydraulic parameters of interest for future uncertainty quantification calculations were identified.

Analysis of Thermal-Hydraulic Fluctuations in Trillo NPP With CTF/PARCS v. 2.7 Coupled Code

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
A. Abarca ◽  
R. Miró ◽  
G. Verdú ◽  
J. A. Bermejo
Keyword(s):  
Reactor Core ◽  
Low Frequency ◽  
Protection System ◽  
Inlet Temperature ◽  
Frequency Noise ◽  
Pressurized Water ◽  
The Core ◽  
Coupled Codes ◽  
Water Reactors ◽  
New Research

The low-frequency noises are fluctuations in the neutron flux density, in the low-frequency range up to 4 Hz, which generate noise in the neutron instrumentation and could affect the limitation and protection system of the reactor core. Some European pressurized water reactors (PWRs) experienced the effect of low-frequency noise, opening a new research line for the verification of the neutron-kinetics/thermal-hydraulic coupled codes. A CTF/PARCS v. 2.7 simulation study to verify whether periodical fluctuations in the core inlet temperature could activate the core protection system has been done, obtaining the frequency spectrum of the power oscillation amplitudes.

scholarly journals Uncertainty and Sensitivity Studies with TRACE-SUSA and TRACE-DAKOTA by Means of Transient BFBT Data

2013 ◽  
Vol 2013 ◽  
pp. 1-9 ◽  
Author(s):  
Wadim Jaeger ◽  
Victor Hugo Sánchez Espinoza ◽  
Francisco Javier Montero Mayorga ◽  
Cesar Queral
Keyword(s):  
Pressure Drop ◽  
Void Fraction ◽  
Sensitivity Study ◽  
Hydraulic Diameter ◽  
Inlet Temperature ◽  
Uncertain Parameters ◽  
Wall Roughness ◽  
Model Uncertainties ◽  
Void Fractions ◽  
Sensitivity Studies

In the present paper, an uncertainty and sensitivity study is performed for transient void fraction and pressure drop measurements. Two transients have been selected from the NUPEC BFBT database. The first one is a turbine trip without bypass and the second one is a trip of a recirculation pump. TRACE (version 5.0 patch 2) is used for the thermohydraulic study and SUSA and DAKOTA are used for the quantification of the model uncertainties and the evaluation of the sensitivities. As uncertain parameters geometrical values, hydraulic diameter, and wall roughness are considered while mass flow rate, power, pressure, and inlet subcooling (inlet temperature) are chosen as boundary and input conditions. Since these parameters change with time, it is expected that the importance of them on pressure drop and void fraction will change, too. The results show that the pressure drop is mostly sensitive to geometrical variations like the hydraulic diameter and the form loss coefficient of the spacer grid. For low void fractions, the parameter of the highest importance is the inlet temperature/subcooling while at higher void fraction the power is also of importance.

Neutronic Analysis for VVER-440 Type Reactor Using PARCS Code

2018 ◽  
Author(s):  
Luca Ratti ◽  
Guido Mazzini ◽  
Marek Ruščák ◽  
Valerio Giusti
Keyword(s):  
Czech Republic ◽  
Cross Sections ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Models ◽  
Reactor Core ◽  
Sensitivity Study ◽  
The Czech Republic ◽  
State Office ◽  
Scale Models

The Czech Republic National Radiation Protection Institute (SURO) provides technical support to the Czech Republic State Office for Nuclear Safety, providing safety analysis and reviewing of the technical documentations for Nuclear Power Plants (NPPs). For this reason, several computational models created in SURO were prepared using different codes as tools to simulate and investigate the design base and beyond design base accidents scenarios. This paper focuses on the creation of SCALE and PARCS neutronic models for a proper analysis of the VVER-440 reactor analysis. In particular, SCALE models of the VVER-440 fuel assemblies have been created in order to produce collapsed and homogenized cross sections necessary for the study with PARCS of the whole VVER-440 reactor core. The sensitivity study of the suitable energy threshold to be adopted for the preparation with SCALE of collapsed two energy-group homogenized cross sections is also discussed. Finally, the results obtained with PARCS core model are compared with those reported in the VVER-440 Final Safety Report.

Lumped Capacitance and 3D CFD Conjugate Heat Transfer Modelling of an Automotive Turbocharger

2015 ◽  
Author(s):  
R. Burke ◽  
C. Copeland ◽  
T. Duda ◽  
M. A. Reyes-Belmonte
Keyword(s):  
Heat Transfer ◽  
High Speed ◽  
Weak Link ◽  
Three Dimensional ◽  
Sensitivity Study ◽  
Heat Fluxes ◽  
Strategy Development ◽  
Inlet Temperature ◽  
Compression Process ◽  
The Impact

One dimensional wave-action engine models have become an essential tool within engine development including stages of component selection, understanding system interactions and control strategy development. Simple turbocharger models are seen as a weak link in the accuracy of these simulation tools and advanced models have been proposed to account for phenomena including heat transfer. In order to run within a full engine code, these models are necessarily simple in structure yet are required to describe a highly complex 3D problem. This paper aims to assess the validity of one of the key assumptions in simple heat transfer models, namely, that the heat transfer between the compressor casing and intake air occurs only after the compression process. Initially a sensitivity study was conducted on a simple lumped capacity thermal model of a turbocharger. A new partition parameter was introduced αA, which divides the internal wetted area of the compressor housing into pre and post compression. The sensitivity of heat fluxes to αA was quantified with respect to the sensitivity to turbine inlet temperature (TIT). At low speeds, the TIT was the dominant effect on compressor efficiency whereas at high speed αA had a similar influence to TIT. However, modelling of the conduction within the compressor housing using an additional thermal resistance caused changes in heat flows of less than 10%. Three dimensional CFD analysis was undertaken using a number of cases approximating different values of αA. It was seen that when considering a case similar to αA=0, meaning that heat transfer on the compressor side is considered to occur only after the compression process, significant temperature could build up in the impeller area of the compressor housing, indicating the importance of the pre-compression heat path. The 3D simulation was used to estimate a realistic value for αA which was suggested to be between 0.15 and 0.3. Using a value of this magnitude in the lumped capacitance model showed that at low speed there would be less than 1% point effect on apparent efficiency which would be negligible compared to the 8% point seen as a result of TIT. In contrast, at high speeds, the impact of αA was similar to that of TIT, both leading to approximately 1% point apparent efficiency error.

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