On the Use of Serpent Monte Carlo Code to Generate Few Group Diffusion Constants (original) (raw)
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Generation of Few-Group Diffusion Theory Constants by Monte Carlo Code McCARD
Nuclear Science and Engineering, 2012
The purpose of this paper is to present the Monte Carlo (MC) method augmented by the B 1 spectrum to generate few-group diffusion theory constants, to assess their qualification in terms of the core depletion analysis, and thus to validate the MC method implemented into the Seoul National University MC code, McCARD, as a few-group diffusion theory constant generator. To do so, two-step core neutronics analyses are conducted for two types of power reactors, pressurized water reactors and very high temperature gas-cooled reactors, by the McCARD/MASTER code system in which McCARD is used as a MC few-group constant generation code and MASTER as a deterministic core analysis code. The two-step calculations for the effective multiplication factors and assembly power distributions of the two types of power reactor cores by McCARD/MASTER are compared with the reference calculations from McCARD, the nuclear design report, or measurements. By showing excellent agreement between McCARD/MASTER and the reference neutronics analyses for the two types of power reactors, it is concluded that the MC method implemented in McCARD can generate few-group diffusion theory constants that are well qualified for high-accuracy two-step core neutronics calculations.
Use of Monte Carlo code MCS for multigroup cross section generation for fast reactor analysis
Nuclear Engineering and Technology, 2021
Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MW th metal-fueled SFR (MET-1000) and a 3600 MW th oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core k eff , power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core k eff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.
Macroscopic Cross Sections Generation by Monte Carlo Code MCS for Fast Reactor Analysis
Epj Web of Conferences, 2021
Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the codeto-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.
Annals of Nuclear Energy, 2018
AZNHEX is a novel 3D neutron diffusion code for nuclear core analysis with hexagonal-z geometry, and it is part of the Mexican project on the development of domestic nuclear analysis software ''AZTLAN Platform". Currently, the code is under development but some important steps have been made in the verification procedure of the code. A composite nodal method for prismatic hexagons has been developed and applied to solve the neutron diffusion equations for cores built by the union, side by side, of hexagonal prisms. A standard finite element method is used starting by the strong form to get the corresponding weak form. The verification and validation process of the AZNHEX code started with some academic exercises. In the case of this paper, in order to move to much more realistic scenarios, the OECD/NEA UAM Sodium Fast Reactor Benchmark has been used together with the SERPENT code for further verification. One of the exercises of the OECD/NEA Sodium Fast Reactor Benchmark is focused on the neutronic characterization of global parameters (k eff , sodium void worth, Doppler, etc.) and feedback coefficient evaluation. This Benchmark is intended to confirm the ability of participants and their neutronic codes to provide generally consistent results when analyzing SFR core characteristics and thus, it represents a very good exercise for the AZNHEX development team. The differences in k eff of AZNHEX versus the well validated deterministic codes DYN3D and PARCS are 70 pcm and 113 pcm respectively when using the same set of XS previously generated with the stochastic code SERPENT. Also, the radial power distribution compared between the deterministic codes over the main diagonal of the core presented very good agreement among them. In other exercises, differences in the order of 200 pcm, in the worst cases, were found when comparing integral parameters with SERPENT. The AZNHEX results give the confidence to keep developing the code in order to convert it into the standard domestic tool for hexagonal-z geometry core analysis. Future developments on the code will be focused on the implementation of discontinuity factors and the thermal expansion effects in an operating core as well as the time dependent implementation.
Neutron Multigroup Homogenized Cross Section Determination with the Monte Carlo Method
2012
In this modern age of powerful computers and availability of large computer clusters it is common to use a Monte Carlo method as a neutronic solver of reasonably large reactor systems, like research reactors. It is the only approach capable of giving detail insight of neutron transport phenomena in complex geometries. Due to its high demand of computer power and memory, the method has been mainly used in criticality calculations. One of the possible ways of using Monte Carlo method is for the generation of neutron homogenized multigroup cross sections, which are later used in deterministic codes to provide neutron solution on a coarse mesh. These types of applications are implemented in the Monte Carlo code SERPENT. One is based on a simple homogenization method with volume and flux weighting (FVH) of cross sections and the other is based on the B1 method. While the first method suffers in the presence of the strong absorbers the second method is applicable only for the cases with f...
SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo, 2014
Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.
Monte Carlo Codes for Neutron Buildup Factors
2018
The point-kernel method is a widely used practical tool for gamma-ray shielding calculations. However, application of that method for neutron transport simulations is very limited. The accuracy of the method strongly depends on the accuracy of buildup factors used in the calculations. Buildup factors are usually obtained using appropriate computer codes, either based on discrete ordinates transport method or Monte Carlo approach. Since these codes put strong demands on computer resources, they are applied on a limited number of shielding configurations and an attempt is made to use these results and formulate an empirical expression for buildup factors estimation. Due to high physical complexity of neutron transport through shielding material it is very hard to perform parameterisation in order to establish adequate empirical formula. Existing formulas are very limited and are usually applicable to a narrow neutron energy range for few commonly used shielding materials, mostly in mo...
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.
ARRC: A random ray neutron transport code for nuclear reactor simulation
Annals of Nuclear Energy, 2018
A massively parallel implementation of a recently developed technique for numerically integrating the transport equation, The Random Ray Method (TRRM)[1], is applied to several large reactor benchmark problems. The implementation, which is part of a new development called The Advanced Random Ray Code (ARRC), is one of the first parallel implementations of TRRM. Our goal is to better understand the accuracy and performance characteristics of TRRM on massive scale problems, and to provide community software that facilitates further algorithmic development and potentially its application to a broader class of problems. Key features of ARRC include extreme memory efficiency, domain decomposition, a task based parallel structure, and the ability to efficiently utilize Single Instruction Multiple Data (SIMD) vector units. These attributes lead to efficient performance on modern high performance computer (HPC) architectures, enabling the detailed simulation of reactor cores in three dimensions.
MUDICO-2D is intended for two-dimensional neutronic and thermalhydraulic coupling core calculations for research reactors applications. This code uses the fine mesh finite difference box scheme method and inner/outer iterative scheme to solve a set of multigroup two-dimensional, time independent neutron diffusion equations. A common first perturbation formulae's are applied to evaluate kinetic parameters which are prompt neutron lifetime and effective delayed neutron fractions. The thermalhydraulic equations are solved analytically on a representative fuel cell (cylindrical or plate geometry). The model used to evaluate coolant thermalhydraulic parameters ignores coolant boiling and assumes the coolant to go only once through the reactor core in the axial direction. Validation results are reported for the International Atomic Energy Agency two-dimensional benchmark core research reactor.