Recent Progress in the Design of a Tomographic Device for Measurements of the Three-Dimensional Pin-Power Distribution in Irradiated Nuclear Fuel Assemblies

2010 ◽  
Vol 165 (2) ◽  
pp. 232-239 ◽  
Author(s):  
Tobias Lundqvist Saleh ◽  
Staffan Jacobsson Svärd ◽  
Ane Håkansson ◽  
A. Bäcklin
Keyword(s):  
Nuclear Fuel ◽  
Power Distribution ◽  
Recent Progress ◽  
Three Dimensional ◽  
Fuel Assemblies ◽  
Irradiated Nuclear Fuel

A Small PWR-Core Physical Calculation Based on PWR-Core Analysis Code CORAL

2021 ◽  
Author(s):  
Wen Yang ◽  
Lun Zhou ◽  
Junrong Qiu ◽  
Yun Tai
Keyword(s):  
Density Distribution ◽  
Power Distribution ◽  
Channel Model ◽  
Three Dimensional ◽  
Least Square ◽  
Object Oriented Programming ◽  
Core Analysis ◽  
Fuel Temperature ◽  
Fuel Assemblies ◽  
Calculation Results

Abstract Three dimensional PWR-core analysis code CORAL is developed by Wuhan Second Ship Design and Research Institute. This code provides basic functions including three-dimensional power distribution, fine power reconstruction, fuel temperature distribution, critical search, control rod worth, reactivity coefficients, burnup and nuclide density distribution, etc. CORAL employ nodal expansion method to solve neutron diffusion equation, and the least square method is used to achieve few group constants, and sub-channel model and one-dimensional heat transfer is used to calculate fuel temperature and coolant density distribution, and burnup distribution and nuclide nuclear density could be obtained by solving macro-depletion and micro-depletion equation. The CORAL code is convenient to update and maintain in consider of modular, object-oriented programming technology. In order to analyze the computational accuracy of the CORAL code in small PWR-core and its capability to deal with heterogeneous, calculation analysis are carried out based on the material and geometry parameters of the SMART core. The core has 57 fuel assemblies, with 8, 20 or 24 gadolinium rods arranged in the fuel assemblies. In this paper, a quantitative comparison and analysis of the small PWR problem calculation results are carried out. Numerical results, including effective multiplication factor, assembly power distribution and pin power distribution, all agree well with the calculation results of OpenMC or Bamboo at both hot zero-power (HZP) and hot full-power (HFP) conditions.

The application of synchrotron micro-computed tomography to characterize the three-dimensional microstructure in irradiated nuclear fuel

2020 ◽  
Vol 537 ◽  
pp. 152161
Author(s):  
Jonova Thomas ◽  
Alejandro Figueroa Bengoa ◽  
Sri Tapaswi Nori ◽  
Ran Ren ◽  
Peter Kenesei ◽  
...  
Keyword(s):  
Computed Tomography ◽  
Nuclear Fuel ◽  
Three Dimensional ◽  
Micro Computed Tomography ◽  
Irradiated Nuclear Fuel

scholarly journals Idea of an experimental technique for quantitative passive gamma emission tomography on irradiated nuclear fuel assemblies

2021 ◽  
Vol 7 ◽  
Author(s):  
Peter Jansson
Keyword(s):  
Image Reconstruction ◽  
Nuclear Fuel ◽  
Experimental Technique ◽  
Quantitative Information ◽  
Emission Tomography ◽  
Gamma Emission ◽  
Fuel Rods ◽  
Activity Distribution ◽  
Fuel Assemblies ◽  
Irradiated Nuclear Fuel

An idea is presented in which passive gamma emission tomography of irradiated nuclear fuel is developed to enable quantitative information of the spatial activity distribution of selected isotopes within the fuel rods of the assembly. The idea is based on using well-known calibration sources mounted in the measurement device during measurement. The image reconstruction would include the sources, thereby enable quantification of the activity distribution. Should the idea be proven viable, the outcome would be valuable to the global community dealing with characterisation of nuclear fuel in terms of safety, security, safeguards and fuel development.

Applying image analysis techniques to tomographic images of irradiated nuclear fuel assemblies

2016 ◽  
Vol 96 ◽  
pp. 223-229 ◽  
Author(s):  
Anna Davour ◽  
Staffan Jacobsson Svärd ◽  
Peter Andersson ◽  
Sophie Grape ◽  
Scott Holcombe ◽  
...  
Keyword(s):  
Image Analysis ◽  
Nuclear Fuel ◽  
Tomographic Images ◽  
Analysis Techniques ◽  
Fuel Assemblies ◽  
Irradiated Nuclear Fuel ◽  
Image Analysis Techniques

scholarly journals The Three-­Dimensional Neutron-­Physical Model of Spent Nuclear Fuel Storage System

2021 ◽  
Vol 20 ◽  
pp. 51-59
Author(s):  
О. R. Trofymenko ◽  
◽  
І. M. Romanenko ◽  
М. І. Holiuk ◽  
C. V. Hrytsiuk ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Storage Facility ◽  
Spent Fuel ◽  
Wide Range ◽  
Fuel Assemblies ◽  
Fuel Storage

The management of spent nuclear fuel is one of the most pressing problems of Ukraine’s nuclear energy. To solve this problem, as well as to increase Ukraine’s energy independence, the construction of a centralized spent nuclear fuel storage facility is being completed in the Chornobyl exclusion zone, where the spent fuel of Khmelnytsky, Rivne and South Ukrainian nuclear power plants will be stored for the next 100 years. The technology of centralized storage of spent nuclear fuel is based on the storage of fuel assemblies in ventilated HI-STORM concrete containers manufactured by Holtec International. Long-term operation of a spent nuclear fuel storage facility requires a clear understanding of all processes (thermohydraulic, neutron-physical, aging processes, etc.) occurring in HI-STORM containers. And this cannot be achieved without modeling these processes using modern specialized programs. Modeling of neutron and photon transfer makes it possible to analyze the level of protective properties of the container against radiation, optimize the loading of MPC assemblies of different manufacturers and different levels of combustion and evaluate biological protection against neutron and gamma radiation. In the future it will allow to estimate the change in the isotopic composition of the materials of the container, which will be used for the management of aging processes at the centralized storage of spent nuclear fuel. The article is devoted to the development of the three-dimensional model of the HI-STORM storage system. The model was developed using the modern Monte Carlo code Serpent. The presented model consists of models of 31 spent fuel assemblies 438MT manufactured by TVEL company, model MPC-31 and model HISTORM 190. The model allows to perform a wide range of scientific tasks required in the operation of centralized storage of spent nuclear fuel.

scholarly journals Validation of Pin Power Calculations Using DYN3D on MIDICORE Benchmark

2019 ◽  
pp. 5-12
Author(s):  
O. Kuchyn ◽  
I. Ovdiienko ◽  
V. Khalimonchuk ◽  
M. Ieremenko
Keyword(s):  
Power Distribution ◽  
Three Dimensional ◽  
Power Calculation ◽  
Light Water Reactors ◽  
Distribution Calculation ◽  
Fuel Assemblies ◽  
Burnable Absorber ◽  
Water Reactors ◽  
Radial Reflector ◽  
Effective Multiplication

Three-dimensional code DYN3D is widely used for the calculation of steady states and transients in light water reactors with hexagonal fuel assemblies like VVER. The capability of pin-by-pin power calculation is implemented in the code through an intranodal power reconstruction approach. The calculations of pin power distribution using DYN3D were performed for AER MIDICORE benchmark for the validation of given extension and developed cross-section library. MIDICORE VVER-1000 core periphery power distribution benchmark was proposed on the 20th Symposium of AER. It is a 2D calculation benchmark based on the VVER-1000 core cold state geometry taking into account the geometry of explicit radial reflector. The main issue of MIDICORE benchmark is to provide the reference solution for the validation of pin-by-pin power distribution at the VVER-1000 core periphery calculated by few-group diffusion codes. Various 3D neutron kinetics nodal solvers HEXNEM1, HEXNEM2 and HEXNEM3 are used in DYN3D for neutron flux distribution calculation. The AER MIDICORE benchmark was solved using all solvers implemented in DYN3D with regard to the three most representative fuel assemblies. Considered fuel assemblies are placed both in the inner part and in the peripheral part of the core, and contain the pin with integrated gadolinium burnable absorber. This paper provides results of comparing the effective multiplication factor, assembly-wise power distribution and pin-by-pin power distribution calculated by DYN3D with benchmark data.

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