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.

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.

scholarly journals Conceptual Study on Recriticality Prevention Core Having Duplex Pellets with Neutron Absorber in Outer Core in a Fast Reactor

2019 ◽  
Vol 2019 ◽  
pp. 1-6
Author(s):  
Toshio Wakabayashi ◽  
Makoto Takahashi ◽  
Naoyuki Takaki ◽  
Yoshiaki Tachi ◽  
Mari Yano
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Thermal Characteristics ◽  
Outer Core ◽  
Severe Accident ◽  
Monte Carlo Code ◽  
Core Concept ◽  
Neutron Absorber ◽  
The Core ◽  
Nuclear Data Library

In a fast reactor, we evaluated a new core concept that prevents severe recriticality after whole-scale molten formation in a severe accident. A core concept in which Duplex pellets including neutron absorber are loaded in the outer core has been proposed. Analysis by the continuous energy model Monte Carlo code MVP using the JENDL-4.0 nuclear data library revealed that this fast reactor core has large negative reactivity due to fuel melting at the time of a severe accident, so that the core prevents recriticality. Regarding the core nuclear and thermal characteristics, the loading of Duplex pellets including neutron absorber in the outer core caused no significant differences from the normal core without Duplex pellets.

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Massimo Sarotto ◽  
Gabriele Firpo ◽  
Anatoly Kochetkov ◽  
Antonin Krása ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Nuclear Data ◽  
Control Rod ◽  
The Core ◽  
Major Interest ◽  
Reactor Spectrum ◽  
Data Libraries ◽  
Nuclear Data Libraries ◽  
Sensitivity And Uncertainty Analysis

Abstract During the EURATOM FP7 project FREYA, a number of experiments were performed in a critical core assembled in the VENUS-F zero-power reactor able to reproduce the ALFRED lead-cooled fast reactor spectrum in a dedicated island. The experiments dealt with the measurements of integral and local neutronic parameters, such as the core criticality, the control rod and the lead void reactivity worth, the axial distributions of fission rates for the nuclides of major interest in a fast spectrum, the spectral indices of important actinides (238U, 239Pu, 237 Np) with respect to 235U. With the main aim to validate the neutronic codes adopted for the ALFRED core design, the VENUS-F core and its characterization measurements were simulated with both deterministic (ERANOS) and stochastic (MCNP, SERPENT) codes, by adopting different nuclear data libraries (JEFF, ENDF/B, JENDL, TENDL). This paper summarizes the main results obtained by highlighting a general agreement between measurements and simulations, with few discrepancies for some parameters that are discussed here. Additionally, a sensitivity and uncertainty analysis was performed with deterministic methods for the core reactivity: it clearly indicates that the small over-criticality estimated by the different codes/libraries resulted to be lower than the uncertainties due to nuclear data.

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.

Fast Reactor Core Seismic Analysis for Verification of Assessment Model Considering Deformation of Core Elements

2019 ◽  
Author(s):  
Shinichiro Matsubara ◽  
Akihisa Iwasaki ◽  
Kazuteru Kawamura ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Fast Reactor ◽  
Seismic Analysis ◽  
Evaluation Model ◽  
Reactor Core ◽  
Vertical Displacement ◽  
Core Element ◽  
Structure Interaction ◽  
The Core ◽  
Core Elements ◽  
Vertical Displacements

Abstract To design fast reactor (FR) core components, seismic response must be evaluated in order to ensure structural integrity. Thus, a core seismic analysis method has been developed to evaluate 3D core vibration behavior considering fluid structure interaction and vertical displacements (rising). The analysis code is verified by a series of vibration tests. The evaluation model to simulate the influence of core element deformation due to heat and irradiation were developed and the analysis of the seismic test was performed. And the evaluation model was verified by comparing the seismic test and analysis results. A fast reactor core consists of hundreds of core elements, which lengthen due to thermal expansion and swelling. So, the core elements are self-standing on the core support structure and not restrained in the axial direction. When the vertical seismic excitation surpasses gravitational acceleration, it is necessary to consider vertical displacements and horizontal displacements of the core elements simultaneously. This 3-D vibration behavior is affected by the fluid loads from ambient coolant and the interference of surrounding structures. To solve this, the influential factors to vibration behaviors due to the structure and fluid body, including fluid structure interaction, are extracted and the 3-D reactor core group vibration analysis code (REVIAN-3D) is developed. Core elements are deformed due to thermal expansion and irradiation, and are interfered with surrounding elements each other. The interference increases the frictional force acting on the core element and reduce the vertical displacement (rising) of the core element during the earthquake. To evaluate this reduction of rising, the evaluation model simulating this deformation is incorporated in REVIAN-3D. In this study, the analysis of the vibration test was carried out using the new incorporated evaluation model. As the deformation of mock-up increases, the vertical displacement (rising) decreases, and when the initial interference due to deformation exceeds the threshold, no rising occurs. This trend agreed well between the vibration test and analysis. It is verified that the new incorporated evaluation model simulates the test result enough.

Effect of Deformation of Core Elements of Fast Reactor Core to the Seismic Response

2019 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Kazuteru Kawamura ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Thermal Expansion ◽  
Fast Reactor ◽  
Seismic Analysis ◽  
Three Dimensional ◽  
Reactor Core ◽  
Axial Direction ◽  
Vertical Displacement ◽  
Analysis Model ◽  
The Core ◽  
Core Elements

Abstract Core elements of a fast reactor are self-standing on the core support structure and not restrained in the axial direction. When the earthquake occurs, it is necessary to consider vertical behavior and horizontal displacement of the core elements simultaneously. In the core seismic analysis, a three dimensional core vibration behavior was evaluated by considering fluid structure interaction, collision with adjacent core elements and vertical displacement and verified by a series of vibration tests. But the evaluation had a assumption of straightness of each core elements which may be bowed due to thermal expansion and swelling under restraint of the horizontal direction between the upper pad and lower structure (Entrance Nozzle). If the core elements are deformed in its plant operation, they may push each other against its adjacent core elements. The large horizontal interference forces may work to decrease the vertical displacement of the core elements. In this study, to grasp and estimate the behavior under the deformed core elements under the earthquake motion, a three dimensional seismic analysis model consist of all of core elements with consideration of the effect of deformed core elements were prepared, analyzed and verified by hexagonal-matrix tests with 37 core elements and single row mock-up models with 7 core elements. These test results show that the rising displacements decrease with increased deformation and no rising occurs when the deformations exceed a threshold. In this paper, the effect of bending deformation due to thermal expansion and swelling on the rising displacement of the core elements was shown by seismic experiments.

Analysis of Seismic Experiment of Hexagonal Multi Bundle Model Using Core Assembly Mock-Up

2020 ◽  
Author(s):  
Akihisa Iwasaki ◽  
Shinichiro Matsubara ◽  
Hidenori Harada ◽  
Tomohiko Yamamoto
Keyword(s):  
Fast Reactor ◽  
Three Dimensional ◽  
Reactor Core ◽  
Vertical Direction ◽  
Horizontal Displacement ◽  
Vertical Displacement ◽  
The Core ◽  
Vibration Behavior ◽  
Seismic Experiment ◽  
Core Elements

Abstract The fast reactor core is composed of hundreds of core elements that stand independently on the core support plate, but does not have support to constrain vertical displacement in order to avoid effects such as thermal elongation. When the earthquake occurs, the group vibration behavior is shown, including the rising of core elements in vertical direction, the collision with adjacent core elements in horizontal direction, and the fluid structure interaction. The three dimensional core group vibration analysis code (REVIAN-3D) was constructed to evaluate them. In the case of fast reactor cores in Japan, the horizontal displacement of core elements at the outermost periphery is restricted by the core former (core barrel). However, since there is no core former in fast reactors other than Japan and the boundary conditions are different from those in Japan, the vibration behavior also differs. In this study, to grasp and estimate the group vibration behavior with and without a core former under the earthquake motion, seismic experiment of hexagonal multi bundle model using core assembly mock-up was conducted [1]. These test results show that the horizontal displacements are larger and impact force between pads of core assembly mock-up is smaller without the core former. In this paper, the analysis was verified by group vibration tests with and without a core former.

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 NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE

2016 ◽  
Vol 18 (1) ◽  
pp. 29 ◽  
Author(s):  
Tukiran Surbakti ◽  
Tagor Malem Sembiring
Keyword(s):  
Key Words ◽  
Power Density ◽  
Research Reactor ◽  
Safety Margin ◽  
Reactor Core ◽  
Safety Analysis ◽  
High Density ◽  
Point Of View ◽  
The Core ◽  
Test Fuel

Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini akan dibahas analisis keselamatan uji bahan bakar mini U7Mo-Al dan U6Zr-Al ditinjau dari segi neutronik. Perhitungan dilakukan dengan menggunakan program komputer WIMSD-5B dan Batan-3DIFF. Hasil analisis menunjukkan bahwa kedua bahan bakar uji dapat diiradiasi dengan derajat bakar < 70 % selama 12 siklus operasi tanpa melampaui batas keselamatan neutronik. Kerapatan panas bahan bakar uji U7Mo-Al lebih besar dari bahan bakar U6Zr-Al.  Kata kunci: Bahan bakar mini, analisis neutronik, teras reaktor,  analisis keselamatan

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