Energy Group optimization for forward and inverse problems in nuclear engineering: application to downwell-logging problems (original) (raw)
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Simplified transport models for neutron energy spectra evaluations
2019
An important issue in the design of nuclear reactor cores is the determination of multigroup cross sections. This task is accomplished by means of a combined procedure involving a space homogenization and an energy averaging to obtain group-collapsed data that can well reproduce the neutronic interactions in the real multiplying system. For this purpose, it is necessary to determine accurate neutron spectra (distribution in the energy domain of the neutron flux) that are used as weights in the collapsing process. In this thesis the energy aspects are analysed. The neutron transport equation is solved through the use of some simplifications: first of all the asymptotic theory is used in order to eliminate the space dependence of the problem and reduce it to a single parameter; then some methods to approximate the collision density and, consequently, the scattering integral are used. In particular, the classic Fermi continuous slowing down approach, the Groeling-Goertzel and the Selen...
Annals of Nuclear Energy, 2006
In this article, we extend the one-speed multi-layer models to neutron reflection and transmission developed in our earlier work (de Abreu, M.P., 2005. Multi-layer models to neutron reflection and transmission for whole-core transport calculations, Annals of Nuclear Energy 32, 215) to multigroup transport theory. We begin by considering a two-layer boundary region, and we develop for such a region discrete ordinates models to the diffuse reflection and transmission of neutrons for multigroup nuclear reactor core problems with anisotropic scattering. We perform numerical experiments to show that our models to neutron reflection and transmission can be used to replace efficiently and accurately two nonactive boundary layers in whole-core transport calculations. We conclude this article with an inductive extension of our two-layer results to a boundary region with an arbitrary number of layers.
Progress in Nuclear Well Logging Modeling Using Deterministic Transport Codes
Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.
Computation of Continuous-Energy Neutron Spectra with Discrete Ordinates Transport Theory
Nuclear Science and Engineering, 1995
A procedure is presented to obtain a continuous-energy representation of the neutron spectrum using one-dimensional discrete ordinates calculations with a combination of multigroup (MG) and pointwise (PW) nuclear data. This provides the capability of determining the fine-structure energy distribution of the angular flux and flux moments within the resonance range as well as the smoother spectrum in the high-and thermal-energy ranges. A new method called a submoment expansion is developed to accurately calculate the Legendre moments of the elastic scatter source for the PW transport calculation, and the coupling between the MG and PW calculations is discussed in detail. The continuous-energy flux spectra can be utilized as problem-dependent weighting functions to process self-shielded MG cross sections for reactor physics and/or criticality safety analysis. This calculational method has been implemented in a new PW transport code called CENTRM that can be executed as a module in the AMPX and SCALE computer code packages. An example application using ENDF/B-VI cross-section data to analyze critical benchmarks is presented.
VETOR - Revista de Ciências Exatas e Engenharias, 2021
In this work, we present the most recent numerical results in a nodal approach, which resulted in the development of a new numerical spectral nodal method, based on the spectral analysis of the multigroup, isotropic scattering neutron transport equations in the discrete ordinates ($S_N$) formulation for fixed-source calculations in non-multiplying media (shielding problems). The numerical results refer to simulations of typical problems from the reactor physics field, in rectangular two-dimensional Cartesian geometry, X,YX, YX,Y geometry, and are compared with the traditional Diamond Difference ($DD$) fine-mesh method results, used as a reference, and the spectral coarse-mesh method Green's function ($SGF$) results.
The objective of this work is to compare the performances of Monte Carlo and deterministic transport methods for analysis of well logging problems. Several typical oil well logging geometries were analysed using the MCNP, DORT, TORT and EVENT codes. These include a 14 MeV pulsed neutron source, two NaI or 3 He detectors, and cylindrical as well as rectangular casings. Calculated neutron and gamma spectra were compared at two detector positions. Several cross section libraries were used in the deterministic calculations. This allows to validate applicability of the available codes and libraries, and to an optimal procedure for well logging applications.
2020
A new approach for the development of a coarse-mesh numerical spectral nodal method is presented in this paper. This method, referred to as the Spectral Deterministic Method – Constant Nodal (SDM–CN), is based on a spectral analysis of the multigroup X,Y-Geometry, linearly anisotropic scattering neutron transport equations in discrete ordinates ( SN )formulation for fixed-source calculations in non-multiplying media. In this paper we present typical model problems to illustrate the accuracy and the efficiency for coarse-mesh energy multigroup SN calculations of the SDM-CN method. The numerical results obtained are compared with the traditional fine-mesh Diamond Difference (DD) method and the results obtained by DOT–II and TWOTRAN codes. The numerical results are also compared with the spectral nodal method, spectral Green’s function (SGF).
Life Cycle Reliability and Safety Engineering, 2020
Neutron transport codes are an integral part of reactor physics calculation. The freely available lattice code DRAGON results from an effort to unify inside a single computer code various well-established numerical techniques and calculation methodologies which are commonly used to solve the neutron transport equation. It is of utmost importance for the user community, both from safety and operation point of view, that the codes being utilised for neutronic calculations maintain a high degree of confidence in their predictions. Benchmark problems are designed to test the capability of a neutronic code by comparing the results obtained from the code with well-established results, either from experimentation or from other validated neutronic codes. After PWRs, BWRs, and PHWRs are two of the most popular types of nuclear reactors currently in use worldwide. Consequently, the ability to perform accurate neutronic calculation involving these lattice types can be deemed as a necessary requirement in most modern lattice codes. In this work, we will study two benchmark problems based on the aforementioned reactor lattice types. Using the lattice code DRAGON and subsequently comparing the results with available published solutions, we aim to ascertain the capability of DRAGON to effectively simulate both of these two types of lattices with fresh and burnt fuel.
Brazilian Journal of Radiation Sciences, 2020
In this paper, we propose a new deterministic numerical methodology to solve the one-dimensional linearized Boltzmann equation applied to neutron shielding problems (fixed-source), using the transport equation in the discrete ordinates formulation (SN) considering the multigroup theory. This is a hybrid methodology, entitled Modified Spectral Deterministic Method (SDM-M), that derives from the Spectral Deterministic Method (SDM) and Diamond Difference (DD) methods. This modification in the SDM method aims to calculate neutron scalar fluxes with lower computational cost. Two model-problems are solved using the SDM-M, and the results are compared to the coarse-mesh methods SDM, Spectral Green's Function (SGF) and Response Matrix (RM), and the fine-mesh method DD. The numerical results were obtained in the programming language JAVA version 1.8.0_91.
2020
A new approach for the development of a numerical method of spectral nodal class for the solution of multigroup, anisotropic slab geometry, discrete ordinates transport problems with fixed-source is analyzed in this paper. The method, denominated Spectral Deterministic Method (SDM), is based on the spectral analysis of the neutron transport equations in the formulation of discrete ordinates (SN). The unknowns in the methodology are the cell-edge, and cell average angular fluxes, the numerical values computed for these quantities concur with the analytic solution of the discrete ordinate’s equation. Numerical results are given and compared with the traditional finemesh DD method, the spectral nodal method, spectral Green’s function (SGF) and the FN method to illustrate the method’s numerical accuracy.