Evaluation of China Experimental Fast Reactor Startup Tests – Final Report

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.

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LARGE FAST REACTOR DESIGN STUDY. Final Report

10.2172/4048452 ◽

1964 ◽

Author(s):

◽

Keyword(s):

Fast Reactor ◽

Reactor Design ◽

Final Report ◽

Design Study

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Neutronic simulation of China experimental fast reactor start-up tests- part II: MCS code Monte Carlo calculation

Annals of Nuclear Energy ◽

10.1016/j.anucene.2020.107710 ◽

2020 ◽

Vol 148 ◽

pp. 107710

Author(s):

Tuan Quoc Tran ◽

Jiwon Choe ◽

Xianan Du ◽

Hyunsuk Lee ◽

Deokjung Lee

Keyword(s):

Monte Carlo ◽

Fast Reactor ◽

Monte Carlo Calculation ◽

Start Up ◽

China Experimental Fast Reactor

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Probabilistic Distribution of Parameters Applied to PFM Analysis for Piping in China Experimental Fast Reactor

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-16520 ◽

2013 ◽

Author(s):

Min Qi ◽

Yueying Wang ◽

Jia Liu

Keyword(s):

Fracture Mechanics ◽

Fast Reactor ◽

Safety Assessment ◽

Small Sample Size ◽

Research Work ◽

Assessment Method ◽

Small Sample ◽

Probabilistic Distribution ◽

Structure Dimension ◽

China Experimental Fast Reactor

The safety assessment method based on probabilistic fracture mechanics (PFM) is well applied to pressure vessel and piping. The PFM analysis is more reasonable and reliable than determinate fracture mechanics (DFM) method. In PFM analysis, the uncertainty of main assessment parameters, such as loads, material character parameters, structure dimension and defect sizes are considered to be random, and the probabilistic distribution of these parameters are determined with the theory of probability statistics. Related to the practical engineering of China experimental fast reactor (CEFR), this paper has done some research work on the parameters probabilistic distribution, and a method was given to determine the optimum fitting probabilistic distribution function of parameters applied to PFM analysis for piping in the small sample size. The work of this paper makes the foundation of the further probabilistic safety assessment of CEFR piping.

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Measurement of Reaction Rate Distribution and Neutron Spectrum in China Experimental Fast Reactor

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-16413 ◽

2013 ◽

Author(s):

Xiaoliang Chen ◽

Zhendong Fan ◽

Xiaoxian Chen ◽

Dingsheng Hu

Keyword(s):

Low Power ◽

Neutron Spectrum ◽

Fast Reactor ◽

Reaction Rate ◽

Gamma Ray ◽

Activation Method ◽

Activity Measurement ◽

Axial Distribution ◽

Rate Distribution ◽

China Experimental Fast Reactor

China Experimental Fast Reactor (CEFR) has completed physics start-up tests in 2010 and connected the grid on 40%FP in 2011. The reaction rate distribution, neutron spectrum are some important parameters for CEFR neutron field. In order to measure these parameters some low power irradiation tests using foil activation method have been done in CEFR core. Two kinds of special irradiation test subassemblies have been developed and fabricated for irradiation in CEFR core. And a digital high purity Germanium gamma-ray spectrometer system has been established for foil activity measurement. After dozens of low power irradiation tests in CEFR core, the radial and axial distribution of 235U and 238U fission reaction rate have been measured. The distribution of 238U capture reaction rate in CEFR core was also obtained in these tests. The experimental values of reaction rate are according with the calculation values well. Neutron spectrum was measured by means of multifoil activation method. And a neutron spectrum adjusting code was also compiled to determine the neutron spectrum.

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A Study and Development of Primary Circuit Simulation Code for China Experimental Fast Reactor

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-15790 ◽

2013 ◽

Author(s):

Jin Wang ◽

Donghui Zhang ◽

Wenjun Hu ◽

Lixia Ren

Keyword(s):

Fast Reactor ◽

Nuclear Energy ◽

Circuit Simulation ◽

Thermal Power ◽

Physical Models ◽

Primary Circuit ◽

Cooling Channel ◽

Simulation Code ◽

Wide Range ◽

China Experimental Fast Reactor

A fast reactor is one of recommended candidates of Generation IV nuclear energy systems, which would meet wide requirements such as sustainability, safety and economics for nuclear energy development. To be the China’s first fast reactor, China Experimental Fast Reactor (CEFR) typical technical options are following: 65 MW thermal power and 20 MW electric power, three circuits of sodium-sodium-water, integrated pool type structure for the primary circuit. To establish modular simulation system for sodium fast reactor, the code which simulated the thermal-hydraulic behavior of primary circuit was developed. The physical models include reactor core, reactor vessel cooling channel, pumps, protection vessel, intermediate heat exchangers, ionization chamber cooling channel, cold sodium pool, hot sodium pool, inlet plenum, and pipes, etc. The code could compute coolant pressures, flow rates, and temperatures in the primary circuit. This module was designed for analysis of a wide range of transients. Although based on CEFR, it can treat an arbitrary arrangement of components.

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