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.

Validation Experiment for the Cooling Capability of Passive Auxiliary Feedwater System (PAFS) With PASCAL Facility

2012 ◽  
Author(s):  
Byoung-Uhn Bae ◽  
Seok Kim ◽  
Yu-Sun Park ◽  
Bok-Deuk Kim ◽  
Kyoung-Ho Kang ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Heat Exchanger ◽  
System Analysis ◽  
Thermal Power ◽  
Heat Removal ◽  
Steady State Condition ◽  
Single Tube ◽  
Decay Heat ◽  
Validation Experiment

The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the APR+ (Advanced Power Reactor Plus) which is intended to completely replace the conventional active auxiliary feedwater system. It removes the decay heat by cooling down the secondary system of the SG using condensation heat exchanger installed in the Passive Condensation Cooling Tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop), was constructed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. It simulates a single tube of the passive condensation heat exchangers, a steam-supply line, a return-water line, and a PCCT with a reduced area, which is equivalent to 1/240 of the prototype according to a volumetric scaling methodology with a full height. The objective of the experiment is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a steady state condition of the thermal power. From the experiment, two-phase flow phenomena in the horizontal heat exchanger and PCCT were investigated and the cooling capability of the condensation heat exchanger was validated. Test results showed that the design of the condensation heat exchanger in PAFS could satisfy the requirement for heat removal rate of 540 kW per a single tube and the prevention of water hammer phenomenon inside the tube. It also proved that the operation of PAFS played an important role in cooling down the decay heat by natural convection without any active system. The present experimental results will contribute to improve the model of the condensation and boiling heat transfer, and also to provide the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS.

scholarly journals E203 Experimental Studies of Natural Circulation Decay Heat Removal in Japan Sodium Cooled Fast Reactor (JSFR)

2009 ◽  
Vol 2009.14 (0) ◽  
pp. 427-428
Author(s):  
Hideki KAMIDE ◽  
Hiroyuki MIYAKOSHI ◽  
Osamu WATANABE ◽  
Yuzuru EGUCHI ◽  
Tomonari KOGA
Keyword(s):  
Fast Reactor ◽  
Natural Circulation ◽  
Experimental Studies ◽  
Heat Removal ◽  
Decay Heat

scholarly journals Thermal Characteristics of High Temperature Naturally Circulating Helium Cooling Loop

2018 ◽  
Vol 68 (1) ◽  
pp. 1-10
Author(s):  
František Dzianik ◽  
Štefan Gužela ◽  
Eva Puškášová
Keyword(s):  
Heat Exchanger ◽  
Thermal Performance ◽  
Fast Reactor ◽  
Natural Circulation ◽  
Thermal Characteristics ◽  
Heat Removal ◽  
Operating Conditions ◽  
Reactor Model ◽  
Decay Heat ◽  
Thermal Calculations

Abstract The paper deals with the process properties in terms of the heat transfer, i.e. the thermal performance of the thermal-process units within a helium loop intended for the testing of the decay heat removal (DHR) from the model of the gas-cooled fast reactor (GFR). The system is characterised by a natural circulation of helium, as a coolant, and assume the steady operating conditions of the circulation. The helium loop consists of four main components: the model of the gas-cooled fast reactor, the model of the heat exchanger for the decay heat removal, hot piping branch and cold piping branch. Using the thermal calculations, the thermal performance of the heat exchanger model and the thermal performance of the gas-cooled fast reactor model are determined. The calculations have been done for several defined operating conditions which correspond to the different helium flow rates within the system.

Supercritical Natural Circulation Loop

2021 ◽  
pp. 338-369
Author(s):  
Tanuj Srivastava ◽  
Pranab Sutradhar ◽  
Milan Krishna Singha Sarkar ◽  
Dipankar Narayan Basu
Keyword(s):  
Nuclear Reactors ◽  
Natural Circulation ◽  
Point Of View ◽  
Circulation Loop ◽  
Working Fluid ◽  
System Behavior ◽  
Steady State Condition ◽  
Natural Circulation Loop ◽  
Stability Point ◽  
Future Direction

Supercritical natural circulation loop is a compelling technology for cooling of modern nuclear reactors, which promises enhanced thermal-hydraulic performance in a simple design. Being a new concept, related knowledge base is relatively thin and involves several conflicting theories and controversies. The chapter summarizes the observation till date, starting from the very fundamentals. The phenomenon of natural circulation under steady state condition and suitability of supercritical medium as working fluid are discussed in detail. Different methods of analyses, including analytical, simple 1-d numerical, and multidimensional computational codes, as well as experimental, are elucidated. A comprehensive discussion is presented about the effect of various geometric and operating parameters on the system behavior, from both thermal-hydraulic and stability point of view. Finally, a few recommendations are included about the operation of such loops and future direction of research.

Sign in / Sign up

Export Citation Format

Share Document