Analysis of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Xiong Wenbin ◽  
Xie Qin ◽  
Li Huwei ◽  
Yang Sengai ◽  
Mao Huan ◽  
...  

The whole core model of China experimental fast reactor (CEFR) is established according to the parameters of China experimental fast reactor, which are given by technical publication from the International Atomic Energy Agency (IAEA-TECDOC-1531), and the physical parameters of CEFR are simulated with the Monte Carlo N-particle code (MCNP4a). The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design values, which successfully demonstrate the acceptable fidelity of the MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.

Author(s):  
Xiong Wen-bin ◽  
Cao Jian ◽  
Xie Qin ◽  
Wang Zhan-yong ◽  
Wang Chuang ◽  
...  

The whole core model of CEFR is established according to the parameters of China Experimental Fast Reactor which are given by IAEA-TECDOC-1531, and the physical parameters of CEFR are simulated with the MCNP4a program. The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design value, which successfully demonstrates the reliability of this MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.


Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  

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.


2020 ◽  
Vol 148 ◽  
pp. 107710
Author(s):  
Tuan Quoc Tran ◽  
Jiwon Choe ◽  
Xianan Du ◽  
Hyunsuk Lee ◽  
Deokjung Lee

2018 ◽  
Vol 32 (04) ◽  
pp. 1850031 ◽  
Author(s):  
T. Sahdane ◽  
A. Mhirech ◽  
L. Bahmad ◽  
B. Kabouchi

In this paper, we study the influence of the physical parameters on the thermal and magnetic properties of a nano-particle of bi-fullerene-like structure X[Formula: see text]-Y[Formula: see text] separated by nonmagnetic spherical surfaces, using the Metropolis Monte Carlo (MC) simulations. The coupling between the two spheres containing the spins [Formula: see text] = 3/2 and the spins S = 7/2 belongs to the core and shell, respectively. This study is based on the RKKY (Ruderman–Kittel–Kasuya–Yoshida) interaction type. To complete this work, we carry out the hysteresis cycles of the studied system for different nonmagnetic surface (NMS) numbers.


Physics World ◽  
2021 ◽  
Vol 34 (10) ◽  
pp. 12ii-12ii
Author(s):  
Michael Banks

The International Atomic Energy Agency (IAEA) is to conduct a safety review of the planned discharge of millions of tonnes of treated waste water from the Fukushima nuclear reactor into the ocean.


Author(s):  
Clark J. Radcliffe ◽  
Xian Li Huang

Abstract Sound and vibration transmission modeling methods are important to the design process for high quality automotive vehicles. Statistical Energy Analysis (SEA) is an emerging design tool for the automotive industry that was initially developed in the 1960’s to estimate root-mean-square sound and vibration levels in structures and interior spaces. Although developed to estimate statistical mean values, automotive design application of SEA needs the additional ability to predict statistical variances of the predicted mean values of sound and vibration. This analytical ability would allow analysis of vehicle sound and vibration response sensitivity to changes in vehicle design specifications and their statistical distributions. This paper will present an algorithm to extend the design application of the SEA method through prediction of the variances of RMS responses of vibro-acoustic automobile structures and interior spaces from variances in SEA automotive model physical parameters. The variance analysis is applied to both a simple, complete illustrative example and a more complex automotive vehicle example. Example variance results are verified through comparison with a Monte Carlo test of 2,000 SEA responses whose physical parameters were given Gaussian distributions with means at design values. Analytical predictions of the response statistics agree with the statistics generated by the Monte Carlo method but only require about 1/300 of the computational effort.


1997 ◽  
Vol 119 (4) ◽  
pp. 629-634 ◽  
Author(s):  
C. J. Radcliffe ◽  
X. L. Huang

Sound and vibration transmission modeling methods are important to the design process for high quality automotive vehicles. Statistical Energy Analysis (SEA) is an emerging design tool for the automotive industry that was initially developed in the 1960’s to estimate root-mean-square sound and vibration levels in structures and interior spaces. Although developed to estimate statistical mean values, automotive design application of SEA needs the additional ability to predict statistical variances of the predicted mean values of sound and vibration. This analytical ability would allow analysis of vehicle sound and vibration response sensitivity to changes in vehicle design specifications and their statistical distributions. This paper will present an algorithm to extend the design application of the SEA method through prediction of the variances of RMS. responses of vibro-acoustic automobile structures and interior spaces from variances in SEA automotive model physical parameters. The variance analysis is applied to both a simple, complete illustrative example and a more complex automotive vehicle example. Example variance results are verified through comparison with a Monte Carlo test of 2,000 SEA responses whose physical parameters were given Gaussian distributions with means at design values. Analytical predictions of the response statistics agree with the statistics generated by the Monte Carlo method but only require about 1/300 of the computational effort.


Author(s):  
Xiong Wen-bin ◽  
Yan Xiu-ping ◽  
Wang Bo ◽  
Wang Zhan-yong ◽  
Bie Ye-wang ◽  
...  

This study investigates the reactor core physical properties of the AP1000, which applies the MCNP4a program to model the AP1000 reactor core with the parameters and data from the DCD (19th Edition) of the AP1000 Nuclear Power Plant, which has been submitted to the NRC. The model is applied to calculate and verify the physical parameters of AP1000 core design. The results match well with the design values in the DCD of the AP1000 nuclear power plant. The design values have been calculated by the KENO-Va program, which proves the correctness of the MCNP model. The model will be improved and applied for safety review and verification analysis of AP1000 nuclear power plant in the future.


Author(s):  
Hu Xiao ◽  
Chen Xiao Liang

An innovative small modular sodium cooled fast reactor called S1 is designed by China Institute of Atomic Energy (CIAE). As an encapsulated nuclear source with thermal power of 3MW, S1 is characterized by small volume, light weight, high safety and sound reliability. The S1 systems adopt modularization, by which the core will be loaded in a factory and filled with sodium, then shipped to the assembled onsite, thus the construction time for S1 can be substantially reduced. In this paper, the Monte Carlo code MCNP is used to calculate the loading scheme of S1. Considering factory’s special conditions for loading and active chemical property of sodium, a special loading pattern is adopted for S1 — loading fuel first, then sodium filling. Ensure that the neutron source and detectors are well matched during loading, and detector counting rate is no less than 2cps when only neutron source but no fuel exists in the core. Three positions where the 252Cf neutron source is placed are studied in this paper: (1) at the center of the core; (2) at the middle of outside core plane; (3) at the bottom of outside the reactor vessel. Through MCNP simulation calculations and comparison of large resulting data, it finds the neutron source should be reasonably placed at the bottom of the reactor vessel where 252Cf strength is 105 s−1 neutrons, and the ex-core detectors are distributed symmetrically at the center of outside core plane; the most befitting moderator material of detector surface is methacrylate-C5H8O3. In this paper, 1/N extrapolation method is used during loading and kinds of loading schemes have been studied with reference to the principles of China experimental fast reactor (CEFR) and regulations of relevant research reactors, and 5-batch loading scheme is finally chosen as the optimal loading scheme. S1 is prepared for sodium filling at 250 °C. It shows that neutron flux variation of core can be more reliably monitored when the ex-core detectors are placed about 120cm away from the center core through MCNP simulation calculation. Such arrangement can also meet the monitoring requirements for loading and sodium filling.


Author(s):  
Li Yan ◽  
Hu Wenjun ◽  
Ren Lixia

Safety rod and its drive mechanism is one of the shutdown systems in sodium-cooled fast reactor, which must be quickly inserted into the core to achieve emergency shutdown in the event of an accident. Therefore, it is necessary to study the falling process of safety rod. In this paper, the numerical simulation method is used to analyze the falling process of safety rod and its drive mechanism in China Experimental Fast Reactor. According to the flow path of the safety rod and its drive mechanism, the pipe system hydraulic method is used to model the safety rod and its drive mechanism and calculate the hydraulic force of the safety rod and its drive mechanism during the falling process. The relationship between time, displacement, velocity and acceleration is presented. The drop time of safety rod is calculated, which is compared with the experimental results. The factors that affect the drop process are analyzed and a sensitivity analysis is presented.


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