scholarly journals Benchmarking COMSOL Multiphysics Single-Subchannel Thermal-Hydraulic Analysis of a TRIGA Reactor with RELAP5 Results and Experimental Data

2019 ◽  
Vol 2019 ◽  
pp. 1-14
Author(s):  
Ahmed K. Alkaabi ◽  
Jeffrey C. King
Keyword(s):  
Experimental Data ◽  
Thermal Power ◽  
Comsol Multiphysics ◽  
Hydraulic Analysis ◽  
Triga Reactor ◽  
The Core ◽  
Training Research ◽  
Thermal Hydraulic Analysis ◽  
Nuclear Applications ◽  
Federal Center

COMSOL Multiphysics has been used to conduct thermal-hydraulic analysis in multiple nuclear applications. The aim of this study is to benchmark the prediction accuracy of COMSOL Multiphysics in performing thermal-hydraulic analysis of TRIGA (Training, Research, Isotopes, General Atomics) reactors such as the Geological Survey TRIGA Reactor (GSTR) by comparing its predictions with RELAP5 (a widely used code in nuclear thermal-hydraulic analysis) results and experimental data. The GSTR type is Mark I with a full thermal power of 1 MW, and it resides at the Denver Federal Center (DFC) in Colorado. The numerical investigation of the present work is carried out by developing single-subchannel thermal-hydraulic models of the GSTR utilizing RELAP5 and COMSOL codes. The models estimate the temperatures (fuel, outer clad, and coolant) and water flow patterns in the core as well as fuel element powers at which void starts to form within the coolant subchannels. Then, these models’ predictions are quantitatively evaluated and compared with the measured data.

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code CUPID

2010 ◽  
Author(s):  
H. K. Cho ◽  
B. J. Yun ◽  
I. K. Park ◽  
J. J. Jeong
Keyword(s):  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Physical Models ◽  
Injection System ◽  
Hydraulic Analysis ◽  
Two Phase ◽  
Rate Condition ◽  
The Core ◽  
Benchmark Calculation ◽  
Thermal Hydraulic Analysis

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.

scholarly journals THERMAL-HYDRAULIC ANALYSIS OF IRT-4M IN REACTIVITY INSERTION ACCIDENT AT VR-1 REACTOR

2016 ◽  
Vol 4 ◽  
pp. 22
Author(s):  
Filip Fejt
Keyword(s):  
Atmospheric Pressure ◽  
Room Temperature ◽  
Nuclear System ◽  
Hydraulic Analysis ◽  
Research Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Coolant Circulation ◽  
Thermal Hydraulic Analysis ◽  
Reactivity Insertion

The paper deals with thermal-hydraulic analysis during reactivity insertion accident, i.e. a step increase of nuclear system reactivity by 0.7 eff, at VR-1 Reactor. The reactor utilizes IRT-4M type of fuel assemblies, and even though these fuel assemblies are designed for an operation at the high-power research reactors, they might be also used for zero-power reactors. The thermal-hydraulic analyses must take into account several specific assumptions that are derived from VR-1 reactor specifications. The reactor does not require a forced water flow for a fuel cooling, the core is placed in an open vessel with atmospheric pressure, and amount of coolant water in the vessel is sufficient for providing the inlet water at room temperature for the whole event. Coolant circulation is expected to be formed only by natural convection.

scholarly journals Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions

2021 ◽  
Vol 327 ◽  
pp. 01013
Author(s):  
Svetlomir Mitkov ◽  
Ivan Spasov ◽  
Nikola Kolev
Keyword(s):  
Hydraulic Analysis ◽  
The Core ◽  
Fine Mesh ◽  
Core Boundary ◽  
Safety Parameters ◽  
Equipment Failures ◽  
Main Steam Line ◽  
Main Steam ◽  
Control Rods ◽  
Thermal Hydraulic Analysis

The objective of this paper is to analyze the ability of a VVER-1000 core and its control system to cope with a hypothetical main steam line break (MSLB) accident in case of multiple equipment failures. The study involves the use of advanced 3D core calculation models benchmarked and validated for reactivity accidents in preceding studies. A MSLB core boundary condition problem is solved on a coarse (nodal) mesh with the coupled COBAYA/CTF neutronic/thermal hydraulic codes. The core thermal-hydraulic boundary conditions are obtained from a preceding full-plant MSLB simulation. The assessment of the core safety parameters is supplemented by a fine-mesh (sub-channel) thermal-hydraulic analysis of the hottest assemblies with the CTF code using information from the 3D nodal COBAYA/CTF calculations. Thirteen variants of a pessimistic MSLB scenario are considered, each of them assuming a number of equipment failures aggravated by eight control rods stuck out of the core after scram at different locations in the overcooled sector. The results (within the limitations of the adopted modeling assumptions) show that the core safety parameters do not exceed the safety limits in the simulated aggravated reactivity accidents.

scholarly journals Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

2011 ◽  
Vol 26 (1) ◽  
pp. 45-49 ◽  
Author(s):  
Atta Muhammad ◽  
Masood Iqbal ◽  
Tayyab Mahmood
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Safety Margin ◽  
Nucleate Boiling ◽  
Hydraulic Analysis ◽  
The Core ◽  
Computer Codes ◽  
Neutronic Parameters ◽  
Thermal Hydraulic Analysis

The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

scholarly journals Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Kien-Cuong Nguyen ◽  
Vinh-Vinh Le ◽  
Ton-Nghiem Huynh ◽  
Ba-Vien Luong ◽  
Nhi-Dien Nguyen
Keyword(s):  
Research Reactor ◽  
Nuclear Research ◽  
Nucleate Boiling ◽  
Fuel Cladding ◽  
Hydraulic Analysis ◽  
The Core ◽  
Nuclear Research Reactor ◽  
Calculation Results ◽  
Enriched Uranium ◽  
Thermal Hydraulic Analysis

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Steady-State Thermal-Hydraulic Analysis of the Oregon State University TRIGA Reactor Using RELAP5-3D

2009 ◽  
Vol 162 (3) ◽  
pp. 261-274 ◽  
Author(s):  
W. R. Marcum ◽  
B. G. Woods ◽  
M. R. Hartman ◽  
S. R. Reese ◽  
T. S. Palmer ◽  
...  
Keyword(s):  
Steady State ◽  
State University ◽  
Hydraulic Analysis ◽  
Triga Reactor ◽  
Oregon State University ◽  
Thermal Hydraulic Analysis

Assessment of the TRACE Code for the He-Cooled Systems Simulation Capability Against Some He-FUS3 Experimental Measurements

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Guido Mazzini ◽  
Miloš Kynčl ◽  
Marek Ruščák ◽  
Miroslav Hrehor ◽  
Alis Musa ◽  
...  
Keyword(s):  
Experimental Data ◽  
High Temperature ◽  
Fast Reactor ◽  
Research Center ◽  
Experimental Measurements ◽  
Water Coolant ◽  
Hydraulic Analysis ◽  
Collaborative Project ◽  
Systems Simulation ◽  
Thermal Hydraulic Analysis

This paper focuses on the TRACE code assessment for helium-cooled systems thermal-hydraulic analysis. In the frame of the GoFastR (gas cooled fast reactor) European Collaborative Project, ENEA has offered some selected experimental data for the organization of a benchmark exercise aimed at the validation of the system and CFD codes for the gas reactor transient analyses. One of the Research Center Řež teams participated in it with a CFD code application. Now, the experimental data are used in order to assess the TRACE code for the ongoing high-temperature helium loop (HTHL-2) licensing process. The results of the TRACE calculations agreed very well with the experimental measurements (often within the experimental uncertainties) data provided by the He-FUS3 facility, indicating that the code, despite developed for water coolant applications, if an appropriately tuned input is adopted, it can also be suitable for reasonably accurate gas technology thermo-hydraulic simulations.

Thermal-Hydraulic Analysis of a Passive Containment Cooling System Using Dynamic Liquid Film Model in CFD Code GASFLOW-MPI

2016 ◽  
Author(s):  
Jianjun Xiao ◽  
John R. Travis ◽  
Maurizio Bottoni
Keyword(s):  
Experimental Data ◽  
Liquid Film ◽  
Large Scale ◽  
Cooling System ◽  
Water Film ◽  
Hydraulic Analysis ◽  
Integral Test ◽  
Film Model ◽  
Water Reactors ◽  
Thermal Hydraulic Analysis

Falling water film modeling is crucial for the thermal hydraulic analysis of passive containment cooling system in advanced light water reactors. A dynamic liquid film model has been developed in the 3-D parallel CFD code GASFLOW-MPI to track the water film transport over the external surface of the steel containment. The gas phase, both inside and outside of the containment, the steel containment and the falling water film are coupled with each other through the essential source terms of inter-phase transport. The model has been verified by the Nusselt solution. Since very few useful experimental data can be found in the open literature, we compared the numerical results of GASFLOW-MPI and COMMIX code which is considered to be reliable because the code has been validated by the experimental data of the Westinghouse small/large scale integral test facilities. A passive containment cooling system has been simulated using the dynamic film model in GASFLOW-MPI. Good agreement was obtained when compared to the COMMIX results, regarding the water film thickness, velocity and temperature. The effect of mesh sensitivity on the heat and mass transfer needs further study. Further work is required concerning the heat transfer, evaporation and boiling at the interface of water film and gas mixtures in the annular space. Local film dry-out model will be implemented in the GASFLOW-MPI. In general, the calculation results using the dynamic film model positively demonstrated the capability of CFD code GASFLOW-MPI in simulation of passive containment cooling system.

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