The Measurement of the High Temperature Sodium Pipe in CEFR

2013 ◽  
Author(s):  
Huajin Yu ◽  
Long Tang ◽  
Min Qi ◽  
Yuanwu Ye
Keyword(s):  
High Temperature ◽  
Fast Reactor ◽  
Nuclear Reactor ◽  
Heat Removal ◽  
Measure Data ◽  
Secondary Circuit ◽  
Thermal Displacement ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Displacement Measurements

The main causes of invalidation of heat transfer piping are stress, thermal displacement and vibration in fast reactor. Strain and displacement measurements are of particular importance in the commissioning and running periods, which is to ensure safety of nuclear reactor. The defect and abnormity of sodium piping can be discovered betimes by the strain and displacement measurements, and the hidden problems of piping can be avoided. Based on the investigation of measurement on piping of power plant, the strain and thermal displacement measurements are carried into execution on the main cooling systems of secondary circuit and the accident residual heat removal system of China Experimental Fast Reactor(CEFR) during the commissioning and running periods. According to the measure data, the accurateness of pipe stress analysis is validated, and the security of systems can be safeguarded. This paper describes the strain and displacement measurement on the high-temperature sodium piping, including the measure method for the double-layer piping, the test of detection capability of high temperature strain gauges, displacement measure device, the analysis and adjustment method of measure data, and so on.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

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.

Thermal-Hydraulic Character Effect Factors Analysis for Passive Residual Heat Removal System of Sodium-Cooled Fast Reactor

2012 ◽  
Author(s):  
Xiaolong Sun ◽  
Minjun Peng ◽  
Genglei Xia
Keyword(s):  
Heat Exchanger ◽  
Fast Reactor ◽  
Heat Exchangers ◽  
Heat Removal ◽  
Factors Analysis ◽  
Higher Temperature ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Removal System ◽  
Residual Heat

A theoretical investigation was performed on the passive residual heat removal system (PRHRs) of China Experimental Fast Reactor (CEFR). The system contained two air heat exchangers, two independent heat exchangers and several pipes. By means of JTOPMERET code developed by GSE, the thermodynamic model of the system had been built. Based on the model, transient characteristics of the PRHRs have been discussed in detail. Through the calculation analysis, it is then indicated that the capacity of removing residual heat was affected by some factors. The faster speed of air in the air heat exchanger, the higher temperature and the larger heat transfer area of the air heat exchanger are favorable to the PRHRs. At the same time, the calculated parameters variation trends are reasonable according to the Experimental data from CEFR.

Analysis of Passive Residual Heat Removal System at Primary Side when Station Blackout Occurs

2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Author(s):  
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao
Keyword(s):  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Primary Circuit ◽  
Station Blackout ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.

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.

Preliminary Analysis of Unprotected Loss of Heat Sink in Small Long Life Gas Cooled Fast Reactor

2015 ◽  
Vol 751 ◽  
pp. 268-272
Author(s):  
Su'ud Zaki ◽  
Nuri Trianti ◽  
Rosidah M. Indah
Keyword(s):  
Fast Reactor ◽  
Heat Sink ◽  
Channel Model ◽  
Cooling System ◽  
Hot Spot ◽  
Natural Circulation ◽  
Heat Removal ◽  
Reducing Power ◽  
Secondary System ◽  
Heat Removal System

The failure of the secondary side in Gas Cooled Fast Reactor system, which may contain co-generation system, will cause loss of heat sink (LOHS) accident. In this study accident analysis of unprotected loss of heat sink due to the failure of the secondary cooling system has been investigated. The thermal hydraulic model include transient hot spot channel model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is also plays important role to limit the level of consequence during the accident. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate the failure of the secondary system by reducing power through reactivity feedback and remove the rest of heat through natural circulations based decay heat removal (RVACS system).

Transient Study on Sodium Heat Pipe in Passive Heat Removal System of Molten Salt Reactor

2013 ◽  
Author(s):  
Chenglong Wang ◽  
Suizheng Qiu ◽  
Wenxi Tian ◽  
Yingwei Wu ◽  
Guanghui Su
Keyword(s):  
High Temperature ◽  
Molten Salt ◽  
Heat Pipe ◽  
Heat Removal ◽  
Heat Pipes ◽  
Molten Salt Reactor ◽  
Sodium Heat Pipe ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

High temperature heat pipes are effective devices for heat transfer, which are characterized by remarkable advantages in conductivity, isothermality and passivity. It is of significance to apply heat pipes on new concept passive residual heat removal system (PRHRS) of molten salt reactor (MSR). In this paper, the transient performance of high temperature sodium heat pipe is simulated with numerical method in the case of MSR accident. The model of the heat pipe is composed of three conjugate heat transfers, i.e. the vapor space, wick structure and wall. Based on finite element method, the governing equations and boundary conditions are solved by using FORTRAN code to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results indicated that high temperature sodium heat pipe had a good operating characteristic and removed the residual heat of fuel salt rapidly under the accident of MSR.

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

2015 ◽  
Vol 52 (9) ◽  
pp. 1102-1121 ◽  
Author(s):  
Osamu Watanabe ◽  
Kazuhiro Oyama ◽  
Junji Endo ◽  
Norihiro Doda ◽  
Ayako Ono ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Evaluation Methodology ◽  
Decay Heat ◽  
Heat Removal System ◽  
Removal System

Allegro: The European Gas Fast Reactor Demonstrator Project

2009 ◽  
Author(s):  
Christian Poette ◽  
Vale´rie Brun-Magaud ◽  
Franck Morin ◽  
Jean-Franc¸ois Pignatel ◽  
Richard Stainsby ◽  
...  
Keyword(s):  
High Temperature ◽  
Fast Reactor ◽  
Heat Removal ◽  
Reactor System ◽  
Large Temperature ◽  
The Core ◽  
Intermediate Heat Exchanger ◽  
Current Design ◽  
Ceramic Fuel ◽  
High Temperature Components

In the Gas Fast Reactor development plan, ALLEGRO is the first necessary step towards the electricity generating prototype GFR. The ALLEGRO start of operation is planned by 2020. This needs to define all design options in 2010 and to start detailed design studies in 2013. ALLEGRO is a low power Gas Cooled Fast Reactor studied in the European framework. It is a loop type, non electricity generating reactor. Its power is about 80 MW. Several objectives are assigned to ALLEGRO. At first, it will demonstrate the viability of the GFR reactor system, no reactor of this type having been built in the past. Most of the GFR architecture, materials and components features are considered at reduced scale in ALLEGRO, excluding the energy conversion system. ALLEGRO will rely on the same safety options as the reactor system. In addition, the ALLEGRO core will allow the progressive qualification of the GFR ceramic fuel, with the possibility to load some ceramic carbide or nitride sub-assemblies in a first MOX core, with SiC/SiCf cladding and wrappers. When such unit test will be considered convincing enough, the diagrid and circuits are designed to accept full high temperature ceramic cores. The core neutrons can also be used to irradiate structural materials with fast neutron spectrum and in a large temperature range. The core can also include innovative irradiation fuel devices (samples or full bundles) for other reactor systems. Finally, branches on the main intermediate heat exchanger will allow the testing and validation of high temperature components and processes. The pre-conceptual design of ALLEGRO is shared between European partners through the GCFR 6th R&D Framework Program. After recalling the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: • Core design and neutron performances, • The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the MOX core, • Fuel handling principles and solutions, • System design and global reactor architecture which is largely influenced by the Decay Heat Removal strategy (DHR) for depressurized accidents.

Sign in / Sign up

Export Citation Format

Share Document