scholarly journals One dimensional modelling of flow in a helium loop

2018 ◽  
Vol 168 ◽  
pp. 07007
Author(s):  
František Világi ◽  
Branislav Knížat ◽  
Marek Mlkvik ◽  
František Urban ◽  
Róbert Olšiak ◽  
...  
Keyword(s):  
Temperature Distribution ◽  
Numerical Model ◽  
Gaseous Medium ◽  
Nuclear Reactors ◽  
Natural Circulation ◽  
Heat Removal ◽  
Variable Density ◽  
Sufficient Accuracy ◽  
One Dimensional ◽  
Decay Heat

The natural circulation helium loop is a facility designed for decay heat removal from ALLEGRO fast nuclear reactors. The article deals with the observation of pressure and velocity relations during steady flow of helium. A one-dimensional numerical model of flow capable of determining the velocity with sufficient accuracy is presented in the article. The model describes the flow of highly compressed gaseous medium with variable density in direct pipelines with local resistances. It’s a hydraulic model, which means that the temperature distribution along the loop must be known. The article also includes the evaluation of local resistances in DHR and GFR, which significantly affects the resulting accuracy. The results from numerical model are compared with experiments.

scholarly journals Mathematical model of steady flow in a helium loop

2019 ◽  
Vol 20 (7) ◽  
pp. 704
Author(s):  
František Világi ◽  
Branislav Knížat ◽  
Marek Mlkvik ◽  
František Urban ◽  
Róbert Olšiak ◽  
...  
Keyword(s):  
Mathematical Model ◽  
Numerical Model ◽  
Steady Flow ◽  
Gaseous Medium ◽  
Natural Circulation ◽  
Variable Density ◽  
Experimental Facility ◽  
Flowing Medium ◽  
Current State ◽  
Natural Circulation Loop

The article describes the application of a mathematical model to a natural circulation loop. A set of measurements were conducted at the experimental facility. The pressure and velocity relations were observed during the steady flow of helium. The main goal was to create a numerical model of flow capable of determining the velocity of flowing medium. The model describes the flow of highly compressed gaseous medium with variable density in direct pipelines with local resistances. At the current state, the temperature values along the loop are taken as input to the model. The article also includes the evaluation of local resistances in DHR and GFR, which significantly affects the resulting accuracy. The results from a numerical model are compared with experiments.

Modeling and Experiments of a Passive Decay Heat Removal System for Advanced Nuclear Reactors

2017 ◽  
Author(s):  
Andrea Bersano ◽  
Mario De Salve ◽  
Cristina Bertani ◽  
Nicolò Falcone ◽  
Bruno Panella
Keyword(s):  
Scaling Laws ◽  
Nuclear Reactors ◽  
Natural Circulation ◽  
Heat Removal ◽  
Experimental Results ◽  
Passive Systems ◽  
Decay Heat ◽  
Safety Systems ◽  
Heat Removal System ◽  
Removal System

Within the field of research and development of innovative nuclear reactors, in particular Generation IV reactors and Small Modular Reactors (SMR), the design and the improvement of safety systems play a crucial role. Among all the safety systems high attention is dedicated to passive systems that do not need external energy to operate, with a very high reliability also in the case of station blackout, and which are largely used in evolutionary technology reactors. The aim of this work is the experimental and numerical analysis of a passive system that operates in natural circulation in order to study the mechanism and the efficiency of heat removal. The final goal is the development of a methodology that can be used to study this class of systems and to assess the thermal-hydraulic code RELAP5 for these specific applications. Starting from a commercial size system, which is the decay heat removal system of the experimental lead cooled reactor ALFRED, an experimental facility has been designed, built and tested with the aim of studying natural circulation in passive systems for nuclear applications. The facility has been simulated and optimized using the thermal-hydraulic code RELAP5-3D. During the experimental tests, temperatures and pressures are measured and the experimental results are compared with the ones predicted by the code. The results show that the system operates effectively, removing the given thermal power. The code can predict well the experimental results but high attention must be dedicated to the modeling of components where non-condensable gases are present (condenser pool and surrounding ambient). This facility will be also used to validate the scaling laws among systems that operate in natural circulation.

scholarly journals Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

2008 ◽  
Vol 2008 ◽  
pp. 1-11 ◽  
Author(s):  
Avinash J. Gaikwad ◽  
P. K. Vijayan ◽  
Sharad Bhartya ◽  
Kannan Iyer ◽  
Rajesh Kumar ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Transport ◽  
Nuclear Power ◽  
Natural Circulation ◽  
Heat Removal ◽  
Transport Systems ◽  
Critical Examination ◽  
Decay Heat ◽  
Station Blackout

Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT), SGs, and PDHRs) under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections). On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

S-CO2 Turbine Design for Decay Heat Removal System of Sodium Cooled Fast Reactor

2016 ◽  
Author(s):  
Seong Kuk Cho ◽  
Jekyoung Lee ◽  
Jeong Ik Lee ◽  
Jae Eun Cha
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Design Methodology ◽  
Nuclear Reactors ◽  
Heat Removal ◽  
Design Code ◽  
Decay Heat ◽  
Turbine Design ◽  
Heat Removal System ◽  
Removal System

A Sodium-cooled Fast Reactor (SFR) has receiving attention as one of the promising next generation nuclear reactors because it can recycle the spent nuclear fuel produced from the current commercial nuclear reactors and accomplish higher thermal efficiency than the current commercial nuclear reactors. However, after shutdown of the nuclear reactor core, the accumulated fission products of the SFR also decay and release heat via radiation within the reactor. To remove this residual heat, a decay heat removal system (DHRS) with supercritical CO2 (S-CO2) as the working fluid is suggested with a turbocharger system which achieves passive operational capability. However, for designing this system an improved S-CO2 turbine design methodology should be suggested because the existing methodology for designing the S-CO2 Brayton cycle has focused only on the compressor design near the critical point. To develop a S-CO2 turbine design methodology, the non-dimensional number based design and the 1D mean line design method were modified and suggested. The design methodology was implemented into the developed code and the code results were compared with existing turbine experimental data. The data were collected under air and S-CO2 environment. The developed code in this research showed a reasonable agreement with the experimental data. Finally using the design code, the turbocharger design for the suggested DHRS and prediction of the off design performance were carried out. As further works, more effort will be put it to expand the S-CO2 turbine test data for validating the design code and methodology.

Performance of Sodium Natural Circulation Loop for Passvie Decay Heat Removal

2008 ◽  
Author(s):  
Hae-Yong Jeong ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Yong-Bum Lee ◽  
Dohee Hahn ◽  
...  
Keyword(s):  
Fast Reactor ◽  
System Analysis ◽  
Natural Circulation ◽  
Heat Removal ◽  
Circulation Loop ◽  
Ambient Atmosphere ◽  
Steady State Condition ◽  
Decay Heat ◽  
Natural Circulation Loop ◽  
Test Loop

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Analysis of a Convection Loop for GFR Post-LOCA Decay Heat Removal

2004 ◽  
Author(s):  
Wesley C. Williams ◽  
Pavel Hejzlar ◽  
Pradip Saha
Keyword(s):  
Heat Transfer ◽  
Steady State ◽  
Flow Regime ◽  
Natural Circulation ◽  
Heat Removal ◽  
Computer Code ◽  
Steady State Analysis ◽  
Decay Heat ◽  
Cooling Loops ◽  
State Analysis

A computer code (LOCA-COLA) has been developed at MIT for steady state analysis of convective heat transfer loops. In this work, it is used to investigate an external convection loop for decay heat removal of a post-LOCA GFR. The major finding is that natural circulation cooling of the GFR is feasible under certain circumstances. Both helium and CO2 cooled system components are found to operate in the mixed convection regime, the effects of which are noticeable as heat transfer enhancement or degradation. It is found that CO2 outperforms helium under identical natural circulation conditions. Decay heat removal is found to have a quadratic dependence on pressure in the laminar flow regime and linear dependence in the turbulent flow regime. Other parametric studies have been performed as well. In conclusion, convection cooling loops are a credible means for GFR decay heat removal and LOCA-COLA is an effective tool for steady state analysis of cooling loops.

Sign in / Sign up

Export Citation Format

Share Document