Development of a porous body model for decay heat removal studies in a pool type sodium cooled fast reactor

2012 ◽  
Vol 4 (3) ◽  
pp. 202-216
Author(s):  
U. Partha Sarathy ◽  
T. Sundararajan ◽  
C. Balaji ◽  
K. Velusamy
Keyword(s):  
Fast Reactor ◽  
Porous Body ◽  
Heat Removal ◽  
Decay Heat ◽  
Body Model ◽  
Removal Studies ◽  
Pool Type

Decay heat removal in pool type fast reactor using passive systems

2012 ◽  
Vol 250 ◽  
pp. 480-499 ◽  
Author(s):  
U. Parthasarathy ◽  
T. Sundararajan ◽  
C. Balaji ◽  
K. Velusamy ◽  
P. Chellapandi ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Heat Removal ◽  
Passive Systems ◽  
Decay Heat ◽  
Pool Type

Porous Body Model Based Parametric Study for Sodium to Air Heat Exchanger Used in Fast Reactors

2015 ◽  
Vol 8 (1) ◽  
Author(s):  
S. P. Pathak ◽  
V. A. Suresh Kumar ◽  
I. B. Noushad ◽  
K. K. Rajan ◽  
K. Velusamy ◽  
...  
Keyword(s):  
Porous Body ◽  
Removal Rate ◽  
Heat Removal ◽  
Test Facility ◽  
Outlet Temperature ◽  
Finned Tubes ◽  
Decay Heat ◽  
Body Model ◽  
Fast Breeder Reactors ◽  
Breeder Reactors

Sodium to air heat exchangers (AHX) with finned tubes is used in fast breeder reactors for decay heat removal. The aim of decay heat removal is to maintain the fuel, clad, coolant, and structural temperatures within safety limits. To investigate the thermal hydraulic features of AHX, a robust porous body based computational fluid dynamics (CFD) model has been developed and validated against the experimental data obtained from a model AHX of 2 MW capacity in Steam Generator Test Facility at the Indira Gandhi Centre for Atomic Research, Kalpakkam. In the present paper, the developed porous body model is used to study the sodium and air temperature distribution and the influence of various parameters that affect the heat removal rate and sodium outlet temperature in full-size AHX used in the fast breeder reactors. The parameters include mass flow rates and inlet temperatures of sodium and air. The focus of the study has been to identify conditions that can pose the risk of sodium freezing.

Development and Basic Verification of Decay Heat Removal Analysis Code of Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Ping Song ◽  
Dalin Zhang ◽  
Tangtao Feng ◽  
Shibao Wang ◽  
Yapei Zhang ◽  
...  
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Heat Removal ◽  
Code Verification ◽  
Design And Optimization ◽  
Decay Heat ◽  
Removal Capacity ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Pool Type

As one of the generation IV reactors, pool-type Sodium-cooled Fast Reactors (SFRs) is attracting more and more attention. Loss of flow and heat sink accident is one of the most serious accidents for SFRs. Therefore, the decay heat removal capacity after emergency shutdown is of great importance. This paper has developed a one-dimensional code named Decay heat Removal Analysis Code for Sodium-cooled Fast Reactor (DRAC-SFR) to analyze the decay heat removal capacity. In the code, the decay heat removal system contains the primary loop, the intermediate loop and air circuit. The decay heat is removed out step by step with the above three loops. Many studies have been conducted on code verification. The international benchmark analysis of EBRII reactor is applied in the code verification. The calculation is compared with the experimental data and the results of DRAC-SFR agreed well with the experimental data. The comparison with the steady state of China Experimental Fast Reactor (CEFR) shows a good agreement with the design value. The errors of all the compared parameters are within 2%. What’s more, calculation is performed to analyze the characteristics of the decay heat removal capacity for CEFR. Thus, code verification shows that DRAC-SFR is proper to evaluate the decay heat removal capacity for SFRs and has the ability to provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

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.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

2021 ◽  
Author(s):  
Janos Bodi ◽  
Alexander Ponomarev ◽  
Evaldas Bubelis ◽  
Konstantin Mikityuk
Keyword(s):  
Power Plant ◽  
Fast Reactor ◽  
Heat Removal ◽  
Current Paper ◽  
Decay Heat ◽  
Sodium Fast Reactor ◽  
Station Blackout ◽  
Safety Assessments ◽  
The Individual

Abstract As part of the ESFR-SMART project, safety assessments are being conducted on the updated European Sodium Fast Reactor (ESFR) design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In the current paper, the analyzed accidental scenario is the so called Protected Station Blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and SIM-SFR system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRS). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project have been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

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.

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