Study of neutronic characteristics for the VVER reactor assembly with five various burnable poisons using MCNPX code

2020 ◽  
Vol 129 ◽  
pp. 103512
Author(s):  
Sonia M. Reda ◽  
Nourhan A. Elkhawas ◽  
Esmat A. Amin
Keyword(s):  
Vver Reactor ◽  
Mcnpx Code ◽  
Reactor Assembly ◽  
Neutronic Characteristics

scholarly journals Simulate the effect of integral burnable absorber on the neutronic characteristics of a PWR assembly

2018 ◽  
Vol 4 (4) ◽  
pp. 287-293 ◽  
Author(s):  
Ahmed Galahom
Keyword(s):  
Power Distribution ◽  
Three Dimensional ◽  
Published Data ◽  
Pressurized Water Reactor ◽  
Three Dimensional Model ◽  
Pressurized Water ◽  
Mcnpx Code ◽  
Absorber Material ◽  
Burnable Absorber ◽  
Neutronic Characteristics

This article examines the effect of an integral burnable absorber (IBA) on the neutronic characteristics of Pressurized Water Reactor (PWR) to provide possible improvements for the fuel management. MCNPX code was used to design a three dimensional model for PWR assembly. The designed model has been validated by comparing the output data with a previously published data. MCNPX code was used to analyze the radial thermal neutron flux and the radial power distribution through PWR assembly with and without IBA. Gadolinium is burnable absorber material that was used in the IBA rods. The gadolinium element suppressed the power in the regions where they were distributed. The existence of IBA rods has a large effect on the Kinf. This effect decreases gradually with burnup due to the degradation of gadolinium. The gadolinium isotopes degradation was analyzed with burnup. Different numbers of IBA rods were investigated to optimize the suitable number that can be used in the PWR assembly. The gadolinium effect on the concentration of 135Xe and 149Sm resulting from the fission process was analyzed.

scholarly journals In Silicon Monte Carlo Simulation Trials for Investigation of V2O5 Reinforcement Effect on Ternary Zinc Borate Glasses: Nuclear Radiation Shielding Dynamics

Materials ◽  
2021 ◽  
Vol 14 (5) ◽  
pp. 1158
Author(s):  
Huseyin O. Tekin ◽  
Shams A. M. Issa ◽  
Gokhan Kilic ◽  
Hesham M. H. Zakaly ◽  
Mohamed M. Abuzaid ◽  
...  
Keyword(s):  
Effective Atomic Number ◽  
Mean Free Path ◽  
Radiation Shielding ◽  
Nuclear Radiation ◽  
Zinc Borate ◽  
Linear Attenuation Coefficient ◽  
Buildup Factor ◽  
Glass Composites ◽  
Mcnpx Code ◽  
Negative Effect

In the current study, promising glass composites based on vanadium pentoxide (V2O5)-doped zinc borate (ZnB) were investigated in terms of their nuclear-radiation-shielding dynamics. The mass and linear attenuation coefficient, half-value layer, mean free path, tenth-value layer, effective atomic number, exposure-buildup factor, and energy-absorption-buildup factor were deeply simulated by using MCNPX code, Phy-X PSD code, and WinXcom to study the validation of ZBV1, ZBV2, ZBV3, and ZBV4 based on (100−x)(0.6ZnO-0.4B2O3)(x)(V2O5) (x = 1, 2, 3, 4 mol%) samples against ionizing radiation. The results showed that attenuation competencies of the studied glasses slightly changed while increasing the V2O5 content from 1 mol% to 4 mol%. The domination of ZnO concentration in the composition compared to B2O3 makes ZnO substitution with V2O5 more dominant, leading to a decrease in density. Since density has a significant role in the attenuation of gamma rays, a negative effect was observed. It can be concluded that the aforementioned substitution can negatively affect the shielding competencies of studied glasses.

Radiation Shielding Analysis for the Reactor Assembly of the SMART Reactor

2008 ◽  
Vol 45 (sup5) ◽  
pp. 82-85
Author(s):  
Gyuhong Roh ◽  
Ha-Yong Kim ◽  
Kang-Seog Kim ◽  
Kyo-Youn Kim ◽  
Chung-Chan Lee
Keyword(s):  
Radiation Shielding ◽  
Reactor Assembly

scholarly journals Benchmarking of Siemens Linac in Electron Modes: 8-14 MeV Electron Beams

2018 ◽  
Vol 8 (2) ◽  
Author(s):  
H Dowlatabadi ◽  
A A Mowlavi ◽  
M Ghorbani ◽  
S Mohammadi ◽  
F Akbari
Keyword(s):  
Monte Carlo ◽  
Radiation Treatment ◽  
Electron Beams ◽  
Dose Profile ◽  
Depth Dose ◽  
Mcnpx Code ◽  
Gamma Index ◽  
Simulation Results ◽  
Mc Simulation ◽  
Good Agreement

Introduction: Radiation therapy using electron beams is a promising method due to its physical dose distribution. Monte Carlo (MC) code is the best and most accurate technique for forespeaking the distribution of dose in radiation treatment of patients.Materials and Methods: We report an MC simulation of a linac head and depth dose on central axis, along with profile calculations. The purpose of the present research is to carefully analyze the application of MC methods for the calculation of dosimetric parameters for electron beams with energies of 8–14 MeV at a Siemens Primus linac. The principal components of the linac head were simulated using MCNPX code for different applicators. Results: The consequences of measurements and simulations revealed a good agreement. Gamma index values were below 1 for most points, for all energy values and all applicators in percent depth dose and dose profile computations. A number of states exhibited rather large gamma indices; these points were located at the tail of the percent depth dose graph; these points were less used in in radiotherapy. In the dose profile graph, gamma indices of most parts were below 1. The discrepancies between the simulation results and measurements in terms of Zmax, R90, R80 and R50 were insignificant. The results of Monte Carlo simulations showed a good agreement with the measurements. Conclusion: The software can be used for simulating electron modes of a Siemens Primus linac when direct experimental measurements are not feasible.

scholarly journals Investigation of Coolant Local Hydrodynamics in the Mixed Core of the VVER Reactor

2020 ◽  
Vol 63 (2) ◽  
pp. 151-162
Author(s):  
S. M. Dmitriev ◽  
A. V. Gerasimov ◽  
A. A. Dobrov ◽  
D. V. Doronkov ◽  
A. N. Pronin ◽  
...  
Keyword(s):  
Cross Sections ◽  
Experimental Studies ◽  
Reactor Core ◽  
Cross Flow ◽  
Coolant Flow ◽  
Vver Reactor ◽  
Computational Hydrodynamics ◽  
Cross Section Area ◽  
Fuel Assemblies ◽  
Flow Pressure

The article presents the results of experimental studies of the local hydrodynamics of the coolant flow in the mixed core of the VVER reactor, consisting of the TVSA-T and TVSA-T mod.2 fuel assemblies. Modeling of the flow of the coolant flow in the fuel rod bundle was carried out on an aerodynamic test stand. The research was carried out on a model of a fragment of a mixed core of a VVER reactor consisting of one TVSA-T segment and two segments of the TVSA-T.mod2. The flow pressure fields were measured with a five-channel pneumometric probe. The flow pressure field was converted to the direction and value of the coolant velocity vector according to the dependencies obtained during calibration. To obtain a detailed data of the flow, a characteristic cross-section area of the model was selected, including the space cross flow between fuel assemblies and four rows of fuel rods of each of the TVSA fuel assemblies. In the framework of this study the analysis of the spatial distribution of the projections of the velocity of the coolant flow was fulfilled that has made it possible to pinpoint regularities that are intrinsic to the coolant flowing around spacing, mixing and combined spacing grates of the TVSA. Also, the values of the transverse flow of the coolant caused by the flow along hydraulically nonidentical grates were determined and their localization in the longitudinal and cross sections of the experimental model was revealed. Besides, the effect of accumulation of hydrodynamic flow disturbances in the longitudinal and cross sections of the model caused by the staggered arrangement of hydraulically non-identical grates was determined. The results of the study of the coolant cross flow between fuel assemblies interaction, i.e. between the adjacent TVSA-T and TVSA-T mod.2 fuel assemblies were adopted for practical use in the JSC of “Afrikantov OKB Mechanical Engineering” for assessing the heat engineering reliability of VVER reactor cores; also, they were included in the database for verification of computational hydrodynamics programs (CFD codes) and for detailed cell-based calculation of the reactor core.

A Multi-Reactor Assembly for Screening of Diesel Particulate Filters

2006 ◽  
Author(s):  
Athanasios G. Konstandopoulos ◽  
Dimitrios Zarvalis ◽  
Evdoxia Kladopoulou ◽  
Ioannis Dolios
Keyword(s):  
Diesel Particulate ◽  
Diesel Particulate Filters ◽  
Particulate Filters ◽  
Reactor Assembly

Estimation of Spring-Back of Single Torus Inner Vessel Sector by Finite Element Analysis

2016 ◽  
Vol 852 ◽  
pp. 588-594
Author(s):  
Gagan Gupta ◽  
V. Balasubramaniyan
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Technology Development ◽  
Cold Pool ◽  
Spring Back ◽  
Element Analysis ◽  
Finite Element Software ◽  
Reactor Assembly ◽  
Design Requirements ◽  
Inner Vessel

Inner vessel in reactor assembly of sodium cooled fast reactor separates hot and cold pool sodium. The shape of inner vessel is optimized with reduced upper & lower shell diameters and toroidal redan for future Fast Breeder Reactor (FBR). This results in higher buckling strength and reduced thickness and hence reduced weight. To achieve the intricate toroidal shape with specified dimensional tolerances, a comprehensive technology development exercise was carried out successfully for the manufacture of inner vessel 30° sector. The achieved profile of the redan meets the specified dimensions and other design requirements. Spring-back observed in the sector was small. To verify the developmental exercise results, a finite element analysis (FEA) of forming of inner vessel sector was performed on finite element software ABAQUS. In this paper, FEA results and spring back are discussed. Spring back assessed is maximum at the center and relatively lower towards the edges for the redan with the chosen radius of 5980 mm.

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