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Concept# Nuclear reactor core

Résumé

A nuclear reactor core is the portion of a nuclear reactor containing the nuclear fuel components where the nuclear reactions take place and the heat is generated. Typically, the fuel will be low-enriched uranium contained in thousands of individual fuel pins. The core also contains structural components, the means to both moderate the neutrons and control the reaction, and the means to transfer the heat from the fuel to where it is required, outside the core.
Water-moderated reactors

Inside the core of a typical pressurized water reactor or boiling water reactor are fuel rods with a diameter of a large gel-type ink pen, each about 4 m long, which are grouped by the hundreds in bundles called "fuel assemblies". Inside each fuel rod, pellets of uranium, or more commonly uranium oxide, are stacked end to end. Also inside the core are control rods, filled with pellets of substances like boron or hafnium or cadmium that readily capture neutrons. When the control rods are lo

Inside the core of a typical pressurized water reactor or boiling water reactor are fuel rods with a diameter of a large gel-type ink pen, each about 4 m long, which are grouped by the hundreds in bundles called "fuel assemblies". Inside each fuel rod, pellets of uranium, or more commonly uranium oxide, are stacked end to end. Also inside the core are control rods, filled with pellets of substances like boron or hafnium or cadmium that readily capture neutrons. When the control rods are lo

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Réacteur nucléaire

Un réacteur nucléaire est un ensemble de dispositifs comprenant du combustible nucléaire, qui constitue le « cœur » du réacteur, dans lequel une réaction en chaîne peut être initiée et contrôlée par

Combustible nucléaire

vignette|Modèle de l'atome.
Le combustible nucléaire est le produit qui, contenant des isotopes fissiles (uranium, plutonium…), fournit l'énergie dans le cœur d'un réacteur nucléaire en entretenant

Fission nucléaire

redresse=1.67|vignette|Schéma animé (simplifié) d'une fission nucléaire. Un atome (énorme rond rouge) est percuté par un neutron (point bleu). Celui-ci se scinde en deux atomes. La réaction émet d'aut

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ME-464: Introduction to nuclear engineering

This course is intended to understand the engineering design of nuclear power plants using the basic principles of reactor physics, fluid flow and heat transfer. This course includes the following: Reactor designs, Thermal analysis of nuclear fuel, Nuclear safety and Reactor dynamics

PHYS-443: Physics of nuclear reactors

In this course, one acquires an understanding of the basic neutronics interactions occurring in a nuclear fission reactor as well as the conditions for establishing and controlling a nuclear chain reaction.

PHYS-447: Reactor technology

Reactor core cooling, power limits and technological consequences due to fuel, cladding and coolant properties, main principles of reactor and power plant design including auxiliary systems are explained. System technology of most important thermal and fast reactor types is introduced.

Séances de cours associées (12)

DYN3D is a well-established Light Water Reactor (LWR) simulation tool and is being extended for safety analyses of Sodium cooled Fast Reactors (SFRs) at the Helmholtz-Zentrum Dresden-Rossendorf. This thesis focuses on the first stage of the development process, that is, the extension and application of DYN3D for steady-state and transient SFR calculations on reactor core level.
In contrast to LWRs, the SFR behavior is especially sensitive to thermal expansions of the reactor components. Therefore, a new thermal-mechanical module accounting for thermal expansions is implemented into DYN3D. At first step, this module is capable of treating two important thermal expansion effects occurring within the core, namely axial expansion of fuel rods and radial expansion of diagrid.
In order to perform nodal calculations with DYN3D, pre-generated homogenized few-group cross sections (XS) are necessarily needed. Therefore, prior to the development of thermal expansion models, a general methodology for XS generation is established for SFR nodal calculations based on the use of the Monte Carlo code Serpent.
The new methodological developments presented in this thesis are verified against the Monte Carlo solutions of Serpent. Two SFR cores are used for testing: the large oxide core of the OECD/NEA benchmark and a smaller core from the Phenix end-of-life tests. Finally, the extended DYN3D is validated against selected IAEA benchmark tests on the Phenix end-of-life experiments that contain both steady-state and transient calculations.
The contribution to the SFR-related developments at the HZDR, as presented in this thesis, makes it possible of performing steady-state and transient calculations for SFRs on reactor core level by using DYN3D. With this study, the basis of the next stage of DYN3D developments is established, that is, the up-scale of SFR analysis to system level can continue by coupling with a sodium capable thermal-hydraulic system code.

This thesis presents a systematic study on the merits and limitations on using pin-by-pin resolution and transport theory based approaches for nuclear core design calculations. Starting from the lattice codes and an optimal cross section generation scheme, it compares different methods, transport approximations, and spatial discretizations used in pin-by-pin homogenized codes.
It is in the interest of nuclear power plant operators to employ more heterogeneous core loadings in order to improve the fuel utilization and decrease the amount of spent fuel. This necessarily increases the requirements on the accuracy of the computation tools used for the core design and safety analysis. One possibility is employing 3D core solvers with higher spatial resolution, e.g. pin-cell wise.
The comparison of several lattice codes indicates that already the proper generation of diffusion coefficients and higher-order scattering moments for pin-cell geometry is not straightforward. Out of three available lattice codes, two generated unphysical diffusion coefficient when using the inscatter approximation, while the last code was not able to provide the higher scattering moments.
Several few-assembly test cases with either high neutron leakage effects or MOX/uranium interfaces showed that the quality of the results of diffusion and SP3 solvers depends critically on the choice of the diffusion coefficient: in the outscatter approximation, it can cause major deviations, while the inscatter seems overall more adequate. Regarding the transport solvers, results obtained with the SN solver DORT showed very good performance for MOX/uranium interfaces, but major deviations occurred for problems with large power gradients. On the other hand, the MOC solver nTRACER showed in all cases small average error, but 2 - 3 % error around the interface of different assemblies.
A study on spatial discretization indicated that the finite difference method applied on pin-cells does not properly capture the big flux changes between MOX and uranium fuel, while the nodal expansion method is more accurate but too slow. It was suggested to use the finite difference method with finer mesh in the outer assembly pin-cells, which increases the required computation time by only 50 % and decreases the pin power errors below 1 % with respect to lattice code results.
Due to some problems which were observed with the available diffusion/SP3 solvers, a new SP3 solver was implemented in the DORT-TD platform. Several core tests showed that the SP3 pin-by-pin solver can significantly outperform the state-of-the-art nodal solver SIMULATE-5, in particular for reactor cores with inserted control rods.
Finally, the pin-by-pin solvers were coupled to a depletion solver. For that, an accurate and fast interpolation routine had to be implemented. The obtained results of full-cycle depletion with complex core loading showed very good performance in comparison to a heterogeneous transport-based fine-group calculations.

The current doctoral research is focused on the development and validation of a coupled computational tool, to combine the advantages of computational fluid dynamics (CFD) in analyzing complex flow fields and of state-of-the-art system codes employed for nuclear power plant (NPP) simulations. Such a tool can considerably enhance the analysis of NPP transient behavior, e.g. in the case of pressurized water reactor (PWR) accident scenarios such as Main Steam Line Break (MSLB) and boron dilution, in which strong coolant flow asymmetries and multi-dimensional mixing effects strongly influence the reactivity of the reactor core, as described in Chap. 1. To start with, a literature review on code coupling is presented in Chap. 2, together with the corresponding ongoing projects in the international community. Special reference is made to the framework in which this research has been carried out, i.e. the Paul Scherrer Institute's (PSI) project STARS (Steady-state and Transient Analysis Research for the Swiss reactors). In particular, the codes chosen for the coupling, i.e. the CFD code ANSYS CFX V11.0 and the system code US-NRC TRACE V5.0, are part of the STARS codes system. Their main features are also described in Chap. 2. The development of the coupled tool, named CFX/TRACE from the names of the two constitutive codes, has proven to be a complex and broad-based task, and therefore constraints had to be put on the target requirements, while keeping in mind a certain modularity to allow future extensions to be made with minimal efforts. After careful consideration, the coupling was defined to be on-line, parallel and with non-overlapping domains connected by an interface, which was developed through the Parallel Virtual Machines (PVM) software, as described in Chap. 3. Moreover, two numerical coupling schemes were implemented and tested: a sequential explicit scheme and a sequential semi-implicit scheme. Finally, it was decided that the coupling would be single-phase and isothermal, leaving to future work the extension to more complex cases. The development work itself is presented in Chap. 3, together with a generic consideration of code-coupling issues and the discussion of a few verification cases. After the basic development and verification of the coupled tool, an experiment was devised for its initial validation. The employed experimental set-up, presented in Chap. 4, features a double T-junction, connected to a recirculation loop and instrumented with wire-mesh sensors to measure the concentration of a tracer injected into the flow. The main aim of this experiment has been to challenge the coupled tool with the transport of a tracer in a steady-state flow field. The experimental results, the CFX and TRACE stand-alone simulations, and the CFX/TRACE coupled simulations are compared with each other for validation purposes, as well as for a clear demonstration of the improvements that one can achieve by using a coupled tool. The simulations, at the same time, indicated the occurrence of strong "numerical diffusion" effects in the TRACE simulations, these being found to result from weaknesses in the numerical discretization adopted in the code for the solute tracking equation. Accordingly, as described in Chap. 5, a third-order upwind scheme for the numerical discretization, namely QUICKEST-ULTIMATE, has been implemented in TRACE to replace the original first-order upwind scheme. The mathematical derivation of the new scheme is presented, together with certain verification and validation tests. In particular, the improvements over the original TRACE scheme are shown in the context of the coupled CFX/TRACE simulations of the double T-junction experiment. Finally, a second phase of experimental validation was devised for the coupling. To this end, certain qualification tests for the new FLORIS facility at PSI have been used, as presented in Chap. 6. This second facility features a scaled-down, simplified, two-dimensional vertical slice of a BWR vessel. The aim of this second mixing experiment has been, on the one hand, to challenge the momentum equation coupling in the context of the transport of a tracer in a transient flow field, and, on the other hand, to test the performance of the coupled tool for the case of a more complex geometry. Once again, comparisons have been made between experimental results, CFX and TRACE stand-alone simulations, and CFX/TRACE coupled simulations, employing the QUICKEST-ULTIMATE discretization where possible. As before, it is clearly demonstrated that the coupled tool yields much better results than the stand-alone codes. Furthermore, it has been found to be sufficiently robust for being extended to more advanced applications, such as the analysis of PWR transients in which strong reactivity feedback effects occur in the context of complex coolant flow phenomena.