Burnup and neutronic analysis for UOX and MOX fuel assemblies in VVER reactor

2021 ◽  
Vol 164 ◽  
pp. 108586
Author(s):  
Heba K. Louis ◽  
Esmat Amin
Keyword(s):  
Vver Reactor ◽  
Mox Fuel ◽  
Fuel Assemblies

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.

scholarly journals HEXAGONAL PWR-CORE MODELING AND SIMULATION WITH APPLICATION OF NECP-BAMBOO

2021 ◽  
Vol 247 ◽  
pp. 06017
Author(s):  
Cheng Zhang ◽  
Liangzhi Cao ◽  
Yunzhao Li ◽  
Guowei Hua
Keyword(s):  
Monte Carlo ◽  
Modeling And Simulation ◽  
Monte Carlo Code ◽  
Constructive Solid Geometry ◽  
Modeling And Simulations ◽  
Nodal Method ◽  
Solid Geometry ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
1D Model

In this paper, the modeling and simulation of the PWRs loaded with hexagonal fuel assemblies has been implemented with the NECP-Bamboo code. NECP-Bamboo, consisting of a 2D lattice code named Bamboo-Lattice and a 3D steady-state core code named Bamboo-Core, was primitively designed for the PWRs loaded with the rectangular fuel assemblies. As the capability extension for PWRs with hexagonal fuel assemblies, four aspects of improvement have been implemented in NECP-Bamboo. Firstly, the Constructive Solid Geometry (CSG) has been implemented in Bamboo-Lattice for the lattice modeling. Secondly, the explicit modeling of the reflector assembly has been applied to provide more reliable few-group constants, compared with the conventional 1D model for the reflector assembly. Thirdly, the assembly-homogenization capability has been extended to the hexagonal assembly. Fourthly, the diffusion solver in Bamboo-Core based on the Variational Nodal Method (VNM) has been extended to handle hexagonal geometry. With application of the capability-extended NECP-Bamboo, the modeling and simulations for the VVER-1000 benchmark loaded with MOX fuel has been implemented. It can be observed that the numerical results provided by NECP-Bamboo can agree well with corresponding results by the Monte-Carlo code.

Long-Term In-Package Spent Fuel Criticality Calculations

2003 ◽  
Author(s):  
Olivier Wantz ◽  
Olivier Smidts ◽  
Alain Dubus ◽  
R. Beauwens
Keyword(s):  
Multiplication Factor ◽  
Spent Fuel ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Neutron Multiplication

This paper presents MCNP criticality calculations for both UOX and MOX disrupted fuel assemblies canisters systems in the reference Belgian disposal concept and one of its variant. We examine the influence of different parameters (water moderation and geometry alteration) on the neutron multiplication factor, keff. In all the studied cases, the reference concept does not present criticality risks. The variant concept sometimes presents criticality risks. The present results only concern fresh UOX and MOX fuel assemblies. Further developments of this work will include irradiated (UOX and MOX) fuels.

Mixed Reload Design Using MOX and UOX Fuel Assemblies

2002 ◽  
Author(s):  
J. Ramo´n Rami´rez Sa´nchez ◽  
R. T. Perry
Keyword(s):  
Conceptual Design ◽  
Cycle Length ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
Fuel Assemblies ◽  
Current Cycle ◽  
Full Power ◽  
Control Rods ◽  
Mixed Load

As part of the studies involved in plutonium utilization assessment for a Boiling Water Reactor, a conceptual design of MOX fuel was developed, this design is mechanically the same design of 10×10 BWR fuel assemblies but different fisil material. Several plutonium and gadolinium concentrations were tested to match the 18 months cycle length which is the current cycle length of LVNPP, a reference UO2 assembly was modeled to have a full cycle length to compare results, an effective value of 0.97 for the multiplication factor was set as target for 470 Effective Full Power days for both cycles, here the gadolinium concentration was a key to find an average fisil plutonium content of 6.55% in the assembly. A reload of 124 fuel assemblies was assumed to simulate the complete core, several load fractions of MOX fuel mixed with UO2 fresh fuel were tested to verify the shutdown margin, the UO2 fuel meets the shutdown margin when 124 fuel assemblies are loaded into the core, but it does not happen when those 124 assemblies are replaced with MOX fuel assemblies, so the fraction of MOX was reduced step by step up to find a mixed load that meets both length cycle and shutdown margin. Finally the conclusion is that control rods losses some of their worth in presence of plutonium due to a more hardened neutron spectrum in MOX fuel and this fact limits the load of MOX fuel assemblies in the core, this results are shown in this paper.

Analysis of minor actinide recycling using MOX fuel assemblies in a Boiling Water Reactor

2018 ◽  
Vol 120 ◽  
pp. 8-26 ◽  
Author(s):  
Gustavo Alonso ◽  
Eduardo Martinez ◽  
J. Ramón Ramírez ◽  
Rogelio Castillo ◽  
Alejandro Castillo
Keyword(s):  
Boiling Water ◽  
Minor Actinide ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Mox Fuel ◽  
Fuel Assemblies

scholarly journals Hydrodynamics of coolant in VVER reactor core with fuel assemblies of various designs

2020 ◽  
Vol 1683 ◽  
pp. 022004
Author(s):  
S M Dmitriev ◽  
A V Gerasimov ◽  
A A Dobrov ◽  
D V Doronkov ◽  
V E Liskova ◽  
...  
Keyword(s):  
Reactor Core ◽  
Vver Reactor ◽  
Fuel Assemblies ◽  
Hydrodynamics Of Coolant
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