scholarly journals DYN3D and CTF Coupling within a Multiscale and Multiphysics Software Development (Part I)

Energies ◽  
2021 ◽  
Vol 14 (16) ◽  
pp. 5060
Author(s):  
Sebastian Davies ◽  
Dzianis Litskevich ◽  
Ulrich Rohde ◽  
Anna Detkina ◽  
Bruno Merk ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Software Development ◽  
Energy Production ◽  
Nuclear Reactor ◽  
Journal Article ◽  
Reactor Core ◽  
Reactor Physics ◽  
Fuel Pin ◽  
The Uk ◽  
Physical Phenomena

Understanding and optimizing the relation between nuclear reactor components or physical phenomena allows us to improve the economics and safety of nuclear reactors, deliver new nuclear reactor designs, and educate nuclear staff. Such relation in the case of the reactor core is described by coupled reactor physics as heat transfer depends on energy production while energy production depends on heat transfer with almost none of the available codes providing full coupled reactor physics at the fuel pin level. A Multiscale and Multiphysics nuclear software development between NURESIM and CASL for LWRs has been proposed for the UK. Improved coupled reactor physics at the fuel pin level can be simulated through coupling nodal codes such as DYN3D as well as subchannel codes such as CTF. In this journal article, the first part of the DYN3D and CTF coupling within the Multiscale and Multiphysics software development is presented to evaluate all inner iterations within one outer iteration to provide partially verified improved coupled reactor physics at the fuel pin level. Such verification has proven that the DYN3D and CTF coupling provides improved feedback distributions over the DYN3D coupling as crossflow and turbulent mixing are present in the former.

scholarly journals CTF and FLOCAL Thermal Hydraulics Validations and Verifications within a Multiscale and Multiphysics Software Development

Energies ◽  
2021 ◽  
Vol 14 (5) ◽  
pp. 1220
Author(s):  
Sebastian Davies ◽  
Ulrich Rohde ◽  
Dzianis Litskevich ◽  
Bruno Merk ◽  
Paul Bryce ◽  
...  
Keyword(s):  
Software Development ◽  
Turbulent Mixing ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Journal Article ◽  
Thermal Hydraulics ◽  
Reactor Physics ◽  
Wide Range ◽  
The Uk ◽  
Simulation Codes

Simulation codes allow one to reduce the high conservativism in nuclear reactor design improving the reliability and sustainability associated with nuclear power. Full-core coupled reactor physics at the rod level are not provided by most simulation codes. This has led in the UK to the development of a multiscale and multiphysics software development focused on LWRS. In terms of the thermal hydraulics, simulation codes suitable for this multiscale and multiphysics software development include the subchannel code CTF and the thermal hydraulics module FLOCAL of the nodal code DYN3D. In this journal article, CTF and FLOCAL thermal hydraulics validations and verifications within the multiscale and multiphysics software development have been performed to evaluate the accuracy and methodology available to obtain thermal hydraulics at the rod level in both simulation codes. These validations and verifications have proved that CTF is a highly accurate subchannel code for thermal hydraulics. In addition, these verifications have proved that CTF provides a wide range of crossflow and turbulent mixing methods, while FLOCAL in general provides the simplified no-crossflow method as the rest of the methods were only tested during its implementation into DYN3D.

scholarly journals Experimental study of heat transfer through cooling water circuit in a reactor vault by using Al2O3 nano fluid

2018 ◽  
Vol 22 (2) ◽  
pp. 1149-1161 ◽  
Author(s):  
Maria Anish ◽  
Balakrishnan Kanimozh
Keyword(s):  
Heat Transfer ◽  
Temperature Distribution ◽  
Base Fluid ◽  
Nuclear Reactor ◽  
Characteristic Curve ◽  
Reactor Core ◽  
Cooling Water ◽  
Heat Transfer Process ◽  
Nano Fluid ◽  
Al2o3 Nanoparticle

The heat produced in the nuclear reactor due to fission reaction must be kept in control or else it will damage the components in the reactor core. Nuclear plants are using water for the operation dissipation of heat. Instead, some chemical substances which have higher heat transfer coefficient and high thermal conductivity. This experiment aims to find out how efficiently a nanofluid can dissipate heat from the reactor vault. The most commonly used nanofluid is Al2O3 nanoparticle with water or ethylene as base fluid. The Al2O3 has good thermal property and it is easily available. In addition, it can be stabilized in various PH levels. The nanofluid is fed into the reactor?s coolant circuit. The various temperature distribution leads to different characteristic curve that occurs on various valve condition leading to a detailed study on how temperature distribution carries throughout the cooling circuit. As a combination of Al2O3 as a nanoparticle and therminol 55 as base fluid are used for the heat transfer process. The Al2O3 nanoparticle is mixed in therminol 55 at 0.05 vol.% concentration. Numerical analysis on the reactor vault model was carried out by using ABAQUS and the experimental results were compared with numerical results.

scholarly journals Lagrangian Approach for the Study of Heat Transfer in a Nuclear Reactor Core Using the SPH Methodology

2019 ◽  
pp. 108-124
Author(s):  
F. Pahuamba-Valdez ◽  
E. Mayoral-Villa ◽  
C. E. Alvarado-Rodríguez ◽  
J. Klapp ◽  
A. M. Gómez-Torres ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Lagrangian Approach ◽  
Nuclear Reactor Core

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.

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.

Multi-Scale Modelling of Nuclear Reactor Core Graphite

2013 ◽  
Vol 05 (01) ◽  
pp. 1350004 ◽  
Author(s):  
G. E. Smith ◽  
P. E. J. Flewitt ◽  
H. Schlangen
Keyword(s):  
Nuclear Reactor ◽  
Reactor Core ◽  
Volume Percentage ◽  
Multi Scale ◽  
Carbon Dioxide Gas ◽  
Scale Modelling ◽  
Filler Particles ◽  
Radiolytic Oxidation ◽  
The Uk

Magnox and Advanced Gas Cooled reactors operated in the UK are cooled by carbon dioxide gas. The graphite moderator bricks in the reactor core lose mass and become more porous during service due to radiolytic oxidation caused by energy deposition, mainly gamma radiation. The microstructure of these graphites comprises filler particles embedded in a graphitised matrix which contains porosity arising from the manufacturing process and subsequent radiolytic oxidation. Computer models have been developed to be representative of the graphite microstructure and these results are used as an input to multi-scale finite element models. These idealised models provide descriptions of the key features that contribute to an understanding of the relationship between the role of filler particles and the amount of porosity on both load-displacement (stress-strain) and fracture characteristics. Two models are considered to span the length-scales: (i) up to millimetre dimensions and (ii) centimetre dimensions. The results are considered with respect to (i) the model length-scale, (ii) the role of the volume percentage of porosity and (iii) the size and distribution of filler particles.

A Rapid Computational Method for Reducing the Search Space Required in Nuclear Reactor Optimization

2010 ◽  
Author(s):  
Andrew G. Osborne ◽  
Mark R. Deinert
Keyword(s):  
Nuclear Reactor ◽  
Search Space ◽  
Computational Method ◽  
High Fidelity ◽  
Reactor Physics ◽  
Dimensional Parameter ◽  
Fuel Cycles ◽  
Reactor Optimization ◽  
Fuel Pin ◽  
Physics Model

Reactor optimization is central to increasing the efficiency of nuclear fuel cycles and critical for making meaningful comparisons between different design options. Optimization algorithms work by generating trial parameter sets which can be used as inputs to reactor physics models. Unfortunately, many reactor physics codes require substantial CPU time, making optimization of large parameter sets impractical. We have developed a method for finding optima within an N-dimensional parameter space using a fast, flexible reactor physics model that is capable of performing fuel burnup calculations on the order of once per second. Global optima found in this way can then be verified using a high fidelity reactor physics code. We demonstrate our approach by considering a simple fuel pin pitch optimization for a light-water reactor, and we find our code executes in 5 minutes. Repeating this approach using a high-fidelity Monte Carlo simulation requires approximately 15 days of runtime by contrast.

In-Vessel Corium Behavior and Steam Explosion Risks: A Review of Experimental Approaches

2017 ◽  
Author(s):  
Pei Shen ◽  
Wenzhong Zhou
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Radioactive Material ◽  
Severe Accident ◽  
Reactor Accident ◽  
Experimental Approaches

Although no one would like to see, a severe nuclear reactor accident may result in reactor core melting, the fuel melt dropping into water in the reactor vessel, and then interacting with coolant into steam explosion. Steam explosion is a result of very rapid and intense heat transfer and violent interaction between the high temperature melt and low temperature coolant. The timescale for heat transfer is shorter than that for pressure relief, resulting in the formation of shock waves and/or the production of missiles at a later time during the expansion of coolant steam explosion. Steam explosion may endanger the reactor vessel and surrounding structures. During a severe reactor accident scenario, steam explosion is an important risk, even though its probability to occur is pretty low, since it could lead to large releases of radioactive material, and destroy the containment integrity. This study provides a comprehensive review of vapor explosion experiments, especially the most recent ones. In this review, fist, small to intermediate scale experiments related to premixing, triggering and propagation stages are reviewed and summarized in tables. Then the intermediate to large scale experiments using prototypic melt are reviewed and summarized. The recent OECD/SERENA2 project including KROTOS and TROI facilities’ work is also discussed. The studies on steam explosion are vital for reactor severe accident management, and will lead to improved reactor safety.

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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