scholarly journals Multiobjective Core Reloading Pattern Optimization of PARR-1 Using Modified Genetic Algorithm Coupled with Monte Carlo Methods

2021 ◽  
Vol 2021 ◽  
pp. 1-13
Author(s):  
Nadeem Shaukat ◽  
Ammar Ahmad ◽  
Bukhtiar Mohsin ◽  
Rustam Khan ◽  
Salah Ud-Din Khan ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Nuclear Reactor ◽  
Computational Cost ◽  
Reactor Core ◽  
Reactor Physics ◽  
Irradiation Site ◽  
Loading Pattern ◽  
Intelligent Technique ◽  
Safety Constraints ◽  
The Given

In order to maximize both the life cycle and efficiency of a reactor core, it is essential to find the optimum loading pattern. In the case of research reactors, a loading pattern can also be optimized for flux at an irradiation site. Therefore, the development of a general-use methodology for core loading optimization would be very valuable. In this paper, general-use multiobjective core reloading pattern optimization is performed using modified genetic algorithms (MGA). The developed strategy can be applied for the constrained optimization of research and power reactor cores. For an optimal reactor core reloading design strategy, an intelligent technique GA is coupled with the Monte Carlo (MC) code SuperMC developed by the FDS team in China for nuclear reactor physics calculations. An optimal loading pattern can be depicted as a configuration that has the maximum keff and maximum thermal fluxes in the core of the given fuel inventory keeping in view the safety constraints such as limitation on power peaking factor. The optimized loading patterns for Pakistan Research Reactor-1 (PARR-1) have been recommended using the implemented strategy by considering the constraint optimization, i.e., to maximize the keff or maximum thermal neutron flux while maintaining low power peaking factor. It has been observed that the developed intelligent strategy performs these tasks with a reasonable computational cost.

Preliminary Core Design and Optimization Analysis of Travelling Wave Reactor

2014 ◽  
Vol 1070-1072 ◽  
pp. 357-360
Author(s):  
Dao Xiang Shen ◽  
Yao Li Zhang ◽  
Qi Xun Guo
Keyword(s):  
Monte Carlo ◽  
Rational Design ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Travelling Wave ◽  
Optimization Analysis ◽  
Design And Optimization ◽  
Uranium Fuel ◽  
Enriched Uranium ◽  
Ignition Zone

A travelling wave reactor (TWR) is an advanced nuclear reactor which is capable of running for decades given only depleted uranium fuel, it is considered one of the most promising solutions for nonproliferation. A preliminary core design was proposed in this paper. The calculation was performed by Monte Carlo method. The burning mechanism of the reactor core design was studied. Optimization on the ignition zone was performed to reduce the amount of enriched uranium initially deployed. The results showed that the preliminary core design was feasible. The optimization analysis showed that the amount of enriched uranium could be reduced under rational design.

scholarly journals Variational methods and speed-up of Monte Carlo perturbation computations for optimal design in nuclear systems

2019 ◽  
Vol 34 (3) ◽  
pp. 211-221
Author(s):  
Zafar Koreshi ◽  
Hamda Khan ◽  
Muhammad Yaqub
Keyword(s):  
Monte Carlo ◽  
Variational Methods ◽  
Nuclear Reactor ◽  
Critical Mass ◽  
Reactor Core ◽  
Radiation Shielding ◽  
Optimization Simulation ◽  
Nuclear Systems ◽  
Optimal Mix ◽  
Optimal Material

Seeking optimal material distribution in a nuclear system to maximize a response function of interest has been a subject of considerable interest in nuclear engineering. Examples are the optimal fuel distribution in a nuclear reactor core to achieve uniform burnup using minimum critical mass and the use of composite materials with an optimal mix of constituent elements in detection systems and radiation shielding. For such studies, variational methods have been found to be useful but, they have been used for standalone analyses often restricted to idealized models, while more elaborate design studies have required computationally expensive Monte Carlo simulations ill-suited to iterative schemes for optimization. Such an inherent disadvantage of Monte Carlo methods changed with the development of perturbation algorithms but, their efficiency is still dependent on the reference configuration for which a hit-and-trial approach is often used. In the first illustrative example, this paper explores the computational speedup for a bare cylindrical reactor core, achievable by using a variational result to enhance the computational efficiency of Monte Carlo design optimization simulation. In the second example, the effect of non-uniform material density in a fixed-source problem, applicable to optimal moderator and radiation shielding, is presented. While applications of this work are numerous, the objective of this paper is to present preliminary variational results as inputs to elaborate stochastic optimization by Monte Carlo simulation for large and realistic systems.

Depletion Calculation for a Nodal Reactor Physics Code

2010 ◽  
Author(s):  
Antonio Carlos Marques Alvim ◽  
Fernando Carvalho da Silva ◽  
Aquilino Senra Martinez
Keyword(s):  
Diffusion Equation ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Expansion Method ◽  
Design System ◽  
Pressurized Water Reactor ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Nodal Method ◽  
Orthogonal Polynomial Expansion

This paper deals with an alternative numerical method for calculating depletion and production chains of the main isotopes found in a pressurized water reactor. It is based on the use of the exponentiation procedure coupled to orthogonal polynomial expansion to compute the transition matrix associated with the solution of the differential equations describing isotope concentrations in the nuclear reactor. Actually, the method was implemented in an automated nuclear reactor core design system that uses a quick and accurate 3D nodal method, the Nodal Expansion Method (NEM), aiming at solving the diffusion equation describing the spatial neutron distribution in the reactor. This computational system, besides solving the diffusion equation, also solves the depletion equations governing the gradual changes in material compositions of the core due to fuel depletion. The depletion calculation is the most time-consuming aspect of the nuclear reactor design code, and has to be done in a very precise way in order to obtain a correct evaluation of the economic performance of the nuclear reactor. In this sense, the proposed method was applied to estimate the critical boron concentration at the end of the cycle. Results were compared to measured values and confirm the effectiveness of the method for practical purposes.

scholarly journals Advances in Monte-Carlo code TRIPOLI-4®’s treatment of the electromagnetic cascade

2018 ◽  
Vol 170 ◽  
pp. 01008
Author(s):  
Davide Mancusi ◽  
Alice Bonin ◽  
François-Xavier Hugot ◽  
Fadhel Malouch
Keyword(s):  
Monte Carlo ◽  
Nuclear Reactor ◽  
Thick Target ◽  
Monte Carlo Code ◽  
Reactor Physics ◽  
Validation And Verification ◽  
Electron Positron ◽  
A Charge ◽  
Charge Deposition ◽  
Nuclear Instrumentation

TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4 in the electron/positron/photon sector. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

The Super Equivalence Method in Monte Carlo Based Homogenization

2014 ◽  
Author(s):  
Mancang Li ◽  
Kan Wang ◽  
Dong Yao
Keyword(s):  
Monte Carlo ◽  
Cross Sections ◽  
Reactor Core ◽  
Surface Current ◽  
Reaction Rates ◽  
Reactor Physics ◽  
Sph Method ◽  
Equivalence Theory ◽  
Spe Method ◽  
Improved Accuracy

The general equivalence theory (GET) and the superhomogenization method (SPH) are widely used for equivalence in the standard two-step reactor physics calculation. GET has behaved well in light water reactor calculation via nodal reactor analysis methods. The SPH was brought up again lately to satisfy the need of accurate pin-by-pin core calculations. However, both of the classical methods have their limitations. The super equivalence method (SPE) is proposed in the paper as an attempt to preserve the surface current, the reaction rates and the reactivity. It enhances the good property of the SPH method through reaction rates based normalization. The concept of pin discontinuity factors are utilized to preserve the surface current, which is the basic idea in the GET technique. However, the pin discontinuity factors are merged into the homogenized cross sections and diffusion coefficients, thus no additional homogenization parameters are needed in the succedent reactor core calculation. The eigenvalue preservation is performed after the reaction rate and surface current have been preserved, resulting in reduced errors of reactivity. The SPE has been implemented into the Monte Carlo method based homogenization code MCMC, as part of RMC Program, under developed in Tsinghua University. The C5G7 benchmark problem have been carried out to test the SPE. The results show that the SPE method not only suits for the equivalence in Monte Carlo based homogenization but also provides improved accuracy compared to the traditional GET or SPH method.

Towards Development of Uncertainty Library for Nuclear Reactor Core Simulation

2018 ◽  
Author(s):  
Hany S. Abdel-Khalik ◽  
Dongli Huang ◽  
Ondrej Chvala ◽  
G. Ivan Maldonado
Keyword(s):  
Cross Sections ◽  
Degrees Of Freedom ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Sensitivity Study ◽  
Reactor Physics ◽  
Rigorous Approach ◽  
Uncertainty Space ◽  
Core Simulation

Uncertainty quantification is an indispensable analysis for nuclear reactor simulation as it provides a rigorous approach by which the credibility of the predictions can be assessed. Focusing on propagation of multi-group cross-sections, the major challenge lies in the enormous size of the uncertainty space. Earlier work has explored the use of the physics-guided coverage mapping (PCM) methodology to assess the quality of the assumptions typically employed to reduce the size of the uncertainty space. A reduced order modeling (ROM) approach has been further developed to identify the active degrees of freedom (DOFs) of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. In the current work, a sensitivity study, based on the PCM and ROM results, is applied to identify a suitable compressed representation of the uncertainty space to render feasible the quantification and prioritization of the various sources of uncertainties. While the proposed developments are general to any reactor physics computational sequence, the proposed approach is customized to the TRITON-NESTLE computational sequence, simulating the BWR lattice model and the core model, which will serve as a demonstrative tool for the implementation of the algorithms.

scholarly journals Advances in the treatment of the electromagnetic cascade in the TRIPOLI-4® Monte-Carlo code

2018 ◽  
Author(s):  
Davide Mancusi
Keyword(s):  
Monte Carlo ◽  
Nuclear Reactor ◽  
Thick Target ◽  
Monte Carlo Code ◽  
Reactor Physics ◽  
Validation And Verification ◽  
Electron Positron ◽  
A Charge ◽  
Charge Deposition ◽  
Nuclear Instrumentation

TRIPOLI-4® is a Monte-Carlo particle-transport code developed at CEA-Saclay (France) that is employed in the domains of nuclear-reactor physics, criticality-safety, shielding/radiation protection and nuclear instrumentation. The goal of this paper is to report on current developments, validation and verification made in TRIPOLI-4® in the treatment of electron/positron/photon transport. The new capabilities and improvements concern refinements to the electron transport algorithm, the introduction of a charge-deposition score, the new thick-target bremsstrahlung option, the upgrade of the bremsstrahlung model and the improvement of electron angular straggling at low energy. The importance of each of the developments above is illustrated by comparisons with calculations performed with other codes and with experimental data.

Review of Mathematical Models for Pebble Dynamics

2008 ◽  
Author(s):  
Wenping Hu ◽  
Shengyao Jiang ◽  
Xingtuan Yang
Keyword(s):  
Mathematical Models ◽  
Nuclear Reactor ◽  
Model Simulation ◽  
Reactor Core ◽  
Computational Time ◽  
Reactor Physics ◽  
Pebble Bed ◽  
Reactor Technology ◽  
Spot Model ◽  
Dem Model

Pebble-bed nuclear reactor technology, with a reactor core typically composed of spherical pebbles draining very slowly in a continuous refueling process, is currently being revived around the world. But the dense slow pebble flow in the reactor, which has an important impact on reactor physics, is still poorly understood. Under such circumstance, this article studies mathematical models which are potential to research the pebbles motion in the pebble-bed reactor, including void model, spot model and DEM model. The fundamental principles of these models are introduced, the success and deficiency of each model is briefly analyzed. Theoretically, it’s expected that spot model and DEM model may be more practical to apply on studying the pebble dynamics. Though, spot model still needs to be refined based on further experimentation, and more research is necessary to solve the problem of huge computational time in order to make the DEM model simulation technique a really practical notion.

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