Simulation Calculation of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP

2017 ◽  
Author(s):  
Xiong Wen-bin ◽  
Cao Jian ◽  
Xie Qin ◽  
Wang Zhan-yong ◽  
Wang Chuang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Nuclear Safety ◽  
Core Model ◽  
Physical Parameters ◽  
Safety Review ◽  
The Core ◽  
Simulation Calculation ◽  
Calculation Results ◽  
Design Value ◽  
China Experimental Fast Reactor

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.

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 ◽  
...  
Keyword(s):  
Monte Carlo ◽  
International Atomic Energy Agency ◽  
Fast Reactor ◽  
Physical Parameters ◽  
Energy Agency ◽  
Safety Review ◽  
The Core ◽  
Calculation Results ◽  
Design Values ◽  
China Experimental Fast Reactor

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.

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.

Modeling of Reactivity Effects and Transient Behaviour of Large Sodium Fast Reactor

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk
Keyword(s):  
Fast Reactor ◽  
Core Model ◽  
Primary System ◽  
Control Rod ◽  
Transient Behaviour ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Reactivity Coefficients ◽  
Core Conditions ◽  
Point Kinetics

Abstract In the paper the reactivity characteristics of the core of the large sodium fast reactor Superphenix (SPX) were evaluated and compared with available experimental data. The analysis was performed using the TRACE system code modified for the fast reactor applications. The simplified core model was developed aiming to overcome the lack of detailed information on design and realistic core conditions. Point Kinetics neutronic model with all relevant reactivity feedbacks was used to calculate transient power. The paper focuses on challenging issue of modelling of the transient thermal responses of primary system structural elements resulting in reactivity feedbacks specific to such large fast reactor which cannot be neglected. For these effects, the model was equipped with dedicated heat structures to reproduce important feedbacks due to vessel wall, diagrid, strongback, control rod drive lines thermal expansion. Peculiarly, application of the model was considered for a whole range of core conditions from zero power to 100% nominal. The developed core model allowed reproducing satisfactorily the core reactivity balance between zero power at 180?C and full power conditions. Additionally, the reactivity coefficients k, g, h at three power levels were calculated and satisfactory agreement with experimental measurements was also observed. The study demonstrated feasibility of application of relatively simple model with adjusted parameters for analysis of different conditions of very complex system.

Loading Scheme Research of Small Modular Sodium Cooled Reactor in a Factory

2017 ◽  
Author(s):  
Hu Xiao ◽  
Chen Xiao Liang
Keyword(s):  
Neutron Source ◽  
Fast Reactor ◽  
Thermal Power ◽  
Reactor Vessel ◽  
Monte Carlo Code ◽  
Construction Time ◽  
Mcnp Simulation ◽  
The Core ◽  
China Experimental Fast Reactor ◽  
Loading Scheme

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.

Numerical Simulation of Safety Rod and its Drive Mechanism in Sodium-Cooled Fast Reactor During Scram Action

2018 ◽  
Author(s):  
Li Yan ◽  
Hu Wenjun ◽  
Ren Lixia
Keyword(s):  
Numerical Simulation ◽  
Fast Reactor ◽  
Simulation Method ◽  
Drive Mechanism ◽  
Pipe System ◽  
The Core ◽  
Hydraulic Method ◽  
Drop Time ◽  
China Experimental Fast Reactor ◽  
The Relationship

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.

Scales in K(ℝ) at the end of a weak gap

2008 ◽  
Vol 73 (2) ◽  
pp. 369-390 ◽  
Author(s):  
J. R. Steel
Keyword(s):  
Induction Hypothesis ◽  
The Other ◽  
Core Model ◽  
The Core ◽  
Woodin Cardinals ◽  
Other Hand

In this note we shall proveTheorem 0.1. Letbe a countably ω-iterable-mouse which satisfies AD, and [α, β] a weak gap of. Supposeis captured by mice with iteration strategies in ∣α. Let n be least such that ; then we have that believes that has the Scale Property.This complements the work of [5] on the construction of scales of minimal complexity on sets of reals in K(ℝ). Theorem 0.1 was proved there under the stronger hypothesis that all sets definable over are determined, although without the capturing hypothesis. (See [5, Theorem 4.14].) Unfortunately, this is more determinacy than would be available as an induction hypothesis in a core model induction. The capturing hypothesis, on the other hand, is available in such a situation. Since core model inductions are one of the principal applications of the construction of optimal scales, it is important to prove 0.1 as stated.Our proof will incorporate a number of ideas due to Woodin which figure prominently in the weak gap case of the core model induction. It relies also on the connection between scales and iteration strategies with the Dodd-Jensen property first discovered in [3]. Let be the pointclass at the beginning of the weak gap referred to in 0.1. In section 1, we use Woodin's ideas to construct a Γ-full a mouse having ω Woodin cardinals cofinal in its ordinals, together with an iteration strategy Σ which condenses well in the sense of [4, Def. 1.13]. In section 2, we construct the desired scale from and Σ.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

Progressive Evolution of the Core of the Fast Breeder Test Reactor

2010 ◽  
Author(s):  
S. Varatharajan ◽  
K. V. Sureshkumar ◽  
K. V. Kasiviswanathan ◽  
G. Srinivasan
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Power Reactor ◽  
Test Bed ◽  
Fast Breeder Test Reactor ◽  
The Core ◽  
Linear Heat ◽  
Mox Fuel ◽  
Test Reactor ◽  
Carbide Fuel

The second stage of Indian nuclear programme envisages the deployment of fast reactors on a large scale for the effective use of India’s limited uranium reserves. The Fast Breeder Test Reactor (FBTR) at Kalpakkam is a loop type, sodium cooled fast reactor, meant as a test bed for the fuels and structural materials for the Indian fast reactor programme. The reactor was made critical with a unique high plutonium MK-I carbide fuel (70% PuC+30%UC). Being a unique untested fuel of its kind, it was decided to test it as a driver fuel, with conservative limits on Linear Heat Rating and burn-up, based on out-of-pile studies. FBTR went critical in Oct 1985 with a small core of 23 MK-I fuel subassemblies. The Linear Heat Rating and burn-up limits for the fuel were conservatively set at 250 W/cm & 25 GWd/t respectively. Based on out-of-pile simulation in 1994, it was possible to raise the LHR to 320 W/cm. It was decided that when the fuel reaches the target burn-up of 25 GWd/t, the MK-I core would be progressively replaced with a larger core of MK-II carbide fuel (55% PuC+45%UC). Induction of MK-II subassemblies was started in 1996. However, based on the Post-Irradiation Examination (PIE) of the MK-I fuel at 25, 50 & 100 GWd/t, it became possible to enhance the burn-up of the MK-I fuel to 155 GWd/t. More than 900 fuel pins of MK-I composition have reached 155 GWd/t without even a single failure and have been discharged. One subassembly (61 pins) was taken to 165 GWd/t on trial basis, without any clad failure. The core has been progressively enlarged, adding MK-I subassemblies to compensate for the burn-up loss of reactivity and replacement of discharged subassemblies. The induction of MK-II fuel was stopped in 2003. One test subassembly simulating the composition of the MOX fuel (29% PuO2) to be used in the 500 MWe Prototype Fast Breeder Reactor was loaded in 2003. It is undergoing irradiation at 450 W/cm, and has successfully seen a burn-up of 92.5 GWd/t. In 2006, it was proposed to test high Pu MOX fuel (44% PuO2), in order to validate the fabrication and fuel cycle processes developed for the power reactor MOX fuel. Eight MOX subassemblies were loaded in FBTR core in 2007. The current core has 27 MK-I, 13 MK-II, eight high Pu MOX and one power reactor MOX fuel subassemblies. The reactor power has been progressively increased from 10.5 MWt to 18.6 MWt, due to the progressive enlargement of the core. This paper presents the evolution of the core based on the progressive enhancement of the burn-up limit of the unique high Pu carbide fuel.

scholarly journals Characteristics of a gas-cooled fast reactor with minor actinide loading

2021 ◽  
Vol 8 (2) ◽  
pp. 1-9
Author(s):  
Hoai Nam Tran ◽  
Yasuyoshi Kato ◽  
Van Khanh Hoang ◽  
Sy Minh Tuan Hoang
Keyword(s):  
Fast Reactor ◽  
Supercritical Co2 ◽  
Cycle Length ◽  
Minor Actinide ◽  
Co2 Gas ◽  
Control Rod ◽  
The Core ◽  
Thermal Output ◽  
Long Life ◽  
Gas Turbine Cycle

This paper presents the neutronics characteristics of a prototype gas-cooled (supercritical CO2-cooled) fast reactor (GCFR) with minor actinide (MA) loading in the fuel. The GCFR core is designed with a thermal output of 600 MWt as a part of a direct supercritical CO2 (S-CO2) gas turbine cycle. Transmutation of MAs in the GCFR has been investigated for attaining low burnup reactivity swing and reducing long-life radioactive waste. Minor actinides are loaded uniformly in the fuel regions of the core. The burnup reactivity swing is minimized to 0.11% ∆k/kk’ over the cycle length of 10 years when the MA content is 6.0 wt%. The low burnup reactivity swing enables minimization of control rod operation during burnup. The MA transmutation rate is 42.2 kg/yr, which is equivalent to the production rates in 7 LWRs of the same electrical output.

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