Void Reactivity Effect Research on Liquid Metal Cooled Fast Reactor

2017 ◽  
Author(s):  
Yang Lyu ◽  
Xiao Liang
Keyword(s):  
Liquid Metal ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Reactor Core ◽  
Nuclear Industry ◽  
Fourth Generation ◽  
Reactivity Effect ◽  
The Core ◽  
Reactivity Coefficient

In the fourth generation of advanced nuclear power systems, the liquid metal cooled fast reactor plays a more and more important role, such as SFR, LFR and ADS system with LBE coolant. Void reactivity effect means bubbles produced in the core area will induce the change of reactivity. And this reactivity will affect the safety of the reactor. Through investigation and comparison of several liquid metal cooled fast reactors in the nuclear industry, this paper studies bubbles in different positions and partial voiding of the active zone inside the core and fuel assemblies with Monte Carlo core physics calculation method and then concludes the main influencing factors of void reactivity coefficient. The results can provide reference for the follow-up research and development of new type liquid metal fast reactor core design.

scholarly journals Concept of a Fast Breeder Reactor To Transmute MA and LLFP

2021 ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Radioactive Waste ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Fission Products ◽  
Fast Breeder Reactor ◽  
Nuclear Power Reactor ◽  
The Core ◽  
Effective Use

Abstract The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MA: 237 Np, 241 Am, 243 Am, etc.) and long-lived fission products (LLFP: 129 I, 99 Tc, 79 Se, etc.). The purpose of this study was to show a system that can simultaneously achieve the breeding of fissile materials and the transmutation of MA and LLFP in one fast reactor. Transmutation was carried out by loading innovative Duplex type MA fuel in the core region and LLFP containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of about 1.1 was obtained, and MA and LLFP achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237 Np, 14.1%/y for 241 Am, 9.9%/y for 243 Am, 1.6%/y for 129 I, 0.75%/y for 99 Tc, and 4%/y for 79 Se. By simultaneously breeding fissile materials and transmuting MA and LLFP in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

Preliminary Core Design of the Long Life S-CO2 Cooled Fast Reactor With 300WMth

2014 ◽  
Author(s):  
Baolin Liu ◽  
Hongchun Wu ◽  
Youqi Zheng ◽  
Liangzhi Cao ◽  
Xianbao Yuan
Keyword(s):  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Single Channel ◽  
Reactor Core ◽  
Fuel Temperature ◽  
The Core ◽  
Long Life ◽  
High Conversion Ratio ◽  
Thermal Hydraulic Analysis

Gas cooled fast reactors are one of the Generation 4 nuclear power plants with hard neutron spectrum and high conversion ratio. In the study a long life Supercritical CO2 (S-CO2) cooled fast reactor core design with 300 MWth is presented. Physical calculation was carried out based on Dragon and CITATION, and thermal hydraulic analysis was performed based on the single channel code. The MOX fuel was utilized in the core design, and the tube-in-duct (TID) assembly was chosen for its excellent characteristics. According to the physical and thermal hydraulic coupling calculation, the reactor in the study can be operated with 300MWth for 20Ys without shuffling or refueling. Through the core life power peaking was kept relatively low, and the fuel temperature was kept below the 1800 degree centigrade.

scholarly journals Direct Numerical Simulation of the Flow Through a Randomly Packed Pebble Bed

2020 ◽  
Vol 142 (4) ◽  
Author(s):  
Mustafa Alper Yildiz ◽  
Gerrit Botha ◽  
Haomin Yuan ◽  
Elia Merzari ◽  
Richard C. Kurwitz ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Direct Numerical Simulation ◽  
Power Systems ◽  
Nuclear Power ◽  
Flow Behavior ◽  
Reactor Core ◽  
Reynolds Numbers ◽  
Fourth Generation ◽  
Flow Through ◽  
Flow Domain

Abstract The proposition for molten salt and high-temperature gas-cooled reactors has increased the focus on the dynamics and physics in randomly packed pebble beds. Research is being conducted on the validity of these designs as a possible contestant for the fourth-generation nuclear power systems. A detailed understanding of the coolant flow behavior is required in order to ensure proper cooling of the reactor core during normal and accident conditions. In order to increase the understanding of the flow through these complex geometries and enhance the accuracy of lower-fidelity modeling, high-fidelity approaches such as direct numerical simulation (DNS) can be utilized. Nek5000, a spectral-element computational fluid dynamics (CFD) code, was used to develop DNS fluid flow data. The flow domain consisted of 147 pebbles enclosed by a bounding wall. In the work presented, the Reynolds numbers ranged from 430 to 1050 based on the pebble diameter and inlet velocity. Characteristics of the flow domain such as volume averaged porosity, axial porosity, and radial porosity were studied and compared with correlations available in the literature. Friction factors from the DNS results for all Reynolds numbers were compared with correlations in the literature. The first- and second-order statistics show good agreement with the available experimental data. Turbulence length scales were analyzed in the flow. Reynolds stress anisotropy was characterized by utilizing invariant analysis. Overall, the results of the analysis in this study provide deeper understanding of the flow behavior and the effect of the wall in packed beds.

scholarly journals Concept of a fast breeder reactor to transmute MAs and LLFPs

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Toshio Wakabayashi
Keyword(s):  
Radioactive Waste ◽  
Power Systems ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Fission Products ◽  
Fast Breeder Reactor ◽  
Nuclear Power Reactor ◽  
The Core ◽  
Effective Use

AbstractThe long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

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.

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.

Structural Confinement and Safety Design of a Fast Reactor Core

1981 ◽  
Vol 103 (2) ◽  
pp. 289-294
Author(s):  
F. D. Ju ◽  
J. G. Bennett
Keyword(s):  
Fast Reactor ◽  
Lateral Pressure ◽  
Numerical Solutions ◽  
Homogeneous Medium ◽  
Safety Margin ◽  
Reactor Core ◽  
Gas Pressure ◽  
Equivalent Stiffness ◽  
Friction Forces ◽  
The Core

In certain fast-reactor designs, the core is an assemblage of a large number of containers of long, hexagonal, hollow cylinders mounted vertically. These so-called “hex-cans” sit individually on coolant nozzles held down by their own weight, and are held as a group laterally at two levels by two constraint rings. At operating temperature, the rings bear on the hex-can assembly because of differences in thermal expansion. The compression of the rings on the hex-can assembly serves to prevent lifting of the can individually or in groups because of any accidental buildup of gas pressure. In the analysis, it is observed that the large number of hexcans and the distribution of the temperature field are such that the cross section of the reactor core can be treated as in a locally uniform dilatational field. An approximate equation was developed relating the plane deformation of a hollow hex cylinder to the global lateral pressure. The parameters are the material constitution and the thickness index (the ratio of the interior and the exterior cross-flat dimensions). The effective range of the equation covers the thickness ratio from zero to the stability limit when the wall becomes too thin resulting in buckling under the lateral pressure. The design equation is exact for zero thickness index. For hollow hex cylinders, numerical solutions were also obtained by the finite element method as a comparison. For a thickness index of 0.9 to 0.95, the difference is less than 0.1 percent. The cylinder constitutive equation is then used to determine an equivalent stiffness for a solid hex cylinder that is to have the same deformation as the given hex-can. The entire planar core region is then analyzed as a homogeneous medium of the equivalent stiffness. The method was applied to the core confinement design for a fast reactor. The thermoelastic solution was then applied to a relatively simpler configuration than the actual case to give a measure of the lateral pressure. The available friction forces for various lift configurations were then obtained. The gas pressure necessary to overcome the minimum friction force thus resulted. In addition, using the lateral pressure, the safety margin of the wall thickness of the hex-can for stability failures was determined.

The Analysis and Realization of Interpreter Language in Nuclear Industry

2014 ◽  
Vol 989-994 ◽  
pp. 2097-2100
Author(s):  
Zheng Zhang ◽  
Hai Bo He ◽  
Hao Liang Lu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Semantic Analysis ◽  
Reactor Core ◽  
Nuclear Industry ◽  
Logical Language ◽  
Grammatical Analysis ◽  
Actual Operating ◽  
Comparison Results

In order to satisfy the calculation requirements of nuclear power plant operating in different conditions, the integration and combination of reactor core computation modules have been proposed. By writing logical language instructions, and then read by interpreter, the engineering designers can make grammatical analysis, lexical analysis, semantic analysis and information extraction. In Linux system environment, the interpreter can fulfill computational tasks based on the actual operating parameters of nuclear power plant. The comparison results indicate that the calculated results obtained by the interpreter language are correct. Therefore, it also demonstrates that the interpreter language is valid.

LBLOCA Analysis of CPR1000 NPP With Advanced Accumulator

2013 ◽  
Author(s):  
Hongwei Hu ◽  
Jianqiang Shan ◽  
Junli Gou ◽  
Bo Zhang ◽  
Haitao Wang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Low Pressure ◽  
Point Of View ◽  
Primary Coolant ◽  
Pressure Drops ◽  
The Core ◽  
Injection Pump ◽  
Model Combining ◽  
Large Flow

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K)[1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.

Analysis of Dynamic Stress in the Reactor Core Barrel due to the Pressure Pulsations

2015 ◽  
Author(s):  
Igor Orynyak ◽  
Iaroslav Dubyk ◽  
Anatolii Batura
Keyword(s):  
Nuclear Power ◽  
Dynamic Stress ◽  
Reactor Core ◽  
Pressure Pulsations ◽  
Power Station ◽  
Coolant Pump ◽  
The Core ◽  
Calculation Results ◽  
Full Power ◽  
Reactor Pressure

This article presents vibrations analysis of the reactor core barrel caused by pressure pulsations induced by the main coolant pump. For this purpose, the calculations of the pressure distribution in the annulus between the core barrel and the reactor pressure vessel, bounded above by a separating ring were performed. Using transfer matrix method is obtained the solution of two-dimensional problem of pressure pulsations in the annulus between reactor core barrel and reactor vessel. The calculation results are compared with the pulsation pressure measurements made at commissioning unit 2 of the South Ukraine Nuclear Power Station. The distribution of pressure over the height of core barrel was obtained, which makes possible to estimate its strength for variant deformation of the core barrel as a beam, and in the case of deformation of the core barrel as a shell. The calculation results are used to assess the reliability of core barrel pre-load, which clamps the core barrel flange in place at the top, at full power operating.

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