Neutronic simulation of China Experimental Fast Reactor start-up tests. Part I: SARAX code deterministic calculation

Measuring of Sodium Void Reactivity Effect (SVRE), one of the most important tests in China Experimental Fast Reactor (CEFR) physical start-up, is described in the paper, including test method, test results and evaluation of test results. The results met to the test requirement and the sign base of CEFR TIB Accident Special Inspect System. The calculation analysis of CEFR SVRE test has been completed, which provides data support before the test and verifies the reliability of the calculation systems after the test. The technology for analysis and measuring of SVRE in sodium-cooled fast reactor has been accumulated through the research of this test.

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MONTE CARLO SIMULATION OF NEUTRONICS START-UP TESTS AT CHINA EXPERIMENTAL FAST REACTOR (CEFR)

EPJ Web of Conferences ◽

10.1051/epjconf/202124710008 ◽

2021 ◽

Vol 247 ◽

pp. 10008

Author(s):

Jiwon Choe ◽

Chirayu Batra ◽

Vladimir Kriventsev ◽

Deokjung Lee

Keyword(s):

Monte Carlo ◽

Fast Reactor ◽

Impact Analysis ◽

Physical Models ◽

Monte Carlo Code ◽

Control Rod ◽

Preliminary Results ◽

Research Activities ◽

Start Up ◽

China Experimental Fast Reactor

China Experimental Fast Reactor (CEFR) is a small size sodium-cooled fast reactor (SFR) with a high neutron leakage core fueled by uranium oxide. The CEFR core with 20 MW(e) power reached its first criticality in July 2010, and several start-up tests were conducted from 2010 to 2011. The China Institute of Atomic Energy (CIAE) proposed to release some of the neutronics start-up test data for the IAEA benchmark within the scope of the IAEA’s coordinated research activities through the coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”, launched in 2018. This benchmark aims to perform validation and verification of the physical models and the neutronics simulation codes by comparing calculation results against collected experimental data. The six physics start-up tests considered for this CRP include evaluation of the criticality, control rod worth, void reactivity, temperature coefficient, swap reactivity, and foil irradiation. Twenty-nine participating research organizations are performing independent blind calculations during the first phase of the project. As a part of this coordinated research, IAEA performed neutronics calculations using Monte Carlo code SERPENT. Two kinds of 3D core models, homogenous and heterogeneous, were calculated using SERPENT, with ENDF/B-VII.0 continuous energy library. Preliminary results with a reasonably good estimation of criticality, as well as theoretically sound results of other five test cases, are available. The paper will discuss the core modelling assumptions, challenges and key findings of modelling a dense SFR core, preliminary results of the first phase of the CRP, heterogeneity impact analysis between homogenous core models and heterogeneous core models and future work to be performed as a part of this four-year project.

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

Volume 9: Student Paper Competition ◽

10.1115/icone26-81626 ◽

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.

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Time dependent start-up thermal analysis of a Super Fast Reactor

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2013.04.014 ◽

2013 ◽

Vol 263 ◽

pp. 129-137 ◽

Cited By ~ 4

Author(s):

Sutanto ◽

Yoshiaki Oka

Keyword(s):

Thermal Analysis ◽

Fast Reactor ◽

Time Dependent ◽

Start Up

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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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