Validation of new APOLLO3 neutron transport calculation models for Light Water Reactors using multigroup Monte Carlo simulations combined with a perturbative approach
For the past twelve years, CEA has been developing a deterministic multi-purpose neutron transport code, APOLLO3, which is starting to be used for reactor studies. A classical two-step APOLLO3 calculation scheme is based on a first stage of two-dimensional infinite lattice calculations in fine transport, generating multi-parameter cross-section libraries used in the second stage of 3D core calculations. In the case of a large power reactor, the core calculation requires approximations that can differ in accuracy, depending on the type of application.
The reference calculation schemes of the SHEM-MOC type and the industrial schemes of the REL2005 type, still in use at the lattice stage by CEA and its industrial partners, EDF and Framatome, were developed in the mid-2000s, based on the methods available in the APOLLO2.8 code. Since then, new methods have been implemented in the APOLLO3 code, which have been individually verified and validated, demonstrating their ability to improve the quality of results at the lattice stage. These include new self-shielding methods, subgroups and Tone, the use of surface line sources in flux calculations using the method of characteristics, flux reconstruction for burnup calculations and a new 383-group fine energy mesh.
The aim of this thesis is to define and validate two new lattice calculation schemes for LWR applications to be used in future calculation tools at CEA and its partners. The goal is to integrate all or part of the new calculation methods, while aiming for reasonable calculation times for the reference scheme, and compatible with fast-running routine usage for the industrial scheme. The calculation schemes implemented will be validated in 2D on geometries taken from the VERA benchmark. Validation will be carried out using an innovative approach involving continuous-energy or multi-group Monte Carlo calculations and a perturbation analysis.
Designing a fast reactor burnup credit validation experiment in the JHR reactor
The primary mission of the Jules Horowitz experimental nuclear Reactor (JHR) is to meet the irradiation needs of materials and fuels for the current nuclear industry and future generations. It is expected to start around 2032. The design of the first wave of experimental devices for RJH already includes specifications for GEN2 and 3 industrial constraints. On the other hand, the field of experiments essential to GEN4 Fast Breeder Reactor remains quite open in the longer term, while no fast-spectrum irradiation facility is currently available.
The objective of this thesis is to study the feasibility of integral experiments in the JHR or another light water reactor, for validation of the reactivity loss with innovative FBR fuels.
In the first part of this thesis, fission products (FPs) that contribute to the loss of reactivity in a typical FBR will be identified and ranked by importance. The second part is the activation measurement and evaluation of the capture cross section of stable FPs in a fast spectrum. It involves the design, specification, implementation and achievement of a “stable” FBR-FP target in the ILL reactor or in the CABRI reactor fuel recovery station (potentially with thermal neutron shields). The third and final part is the design of an experiment in the JHR to generate and characterize FBR FPs. This experiment should be sufficiently representative of fuel irradiation conditions in a FBR. The goal is to access the FP inventory by underwater spectrometry in the JHR and integral reactivity weighing before/after irradiation in CABRI or another available facility.
The thesis will be carried out in a team experienced in the physics and thermal-hydraulics characterization of the JHR. The candidate will be advised by several experts based in the department. The candidate will have the opportunity to promote his/her results before the nuclear industry partners (CEA, EDF, Framatome, Orano, Technicatome etc.).
Modeling and upscaling of sodium boiling flow within a 4th generation nuclear reactor core
The stabilized boiling in sodium is a subject that has been studied for many years at CEA in order to improve the validation of scientific calculation tools such as CATHARE3. Being able to reproduce properly this phenomena is a key safety related question for liquid metal liquid 4th generation reactors. When an unprotected loss of flow (ULOF) happens in the reactor and the safety measures are not deployed, the coolant can reach saturation, which can ultimately lead to a degradation of the subassembly. In order to avoid this situation, new fuel assembly designs provide negative neutronic feedback as the void fraction is generated. To understand how this void fraction evolves in the sub-assembly (within the rod bundle or the top plenum), the code requires a state of the art sodium modeling in terms of momentum, heat and mass transfer.
To improve the qualification of the CATHARE3 code for such situations, the doctoral student will implement CFD models allowing a better understanding of the boiling mechanisms in sodium-cooled subassemblies. New CFD models, such as large interface modelling, wall boiling, heat and mass exchange at the interface will be applied, yielding detailed information on local variables. Subsequently, this detailed information will be transferred to the 1D system code during an upscaling operation. Once this information is properly gathered and transferred, new models will be developed and implemented into the system code. Finally, these new models will be confronted to experimental data in a validation exercise over the CATHARE code validation database. Ultimately, the aim is to increase the confidence in the CATHARE3 1-D simulation tool for predicting the specific physics of sodium boiling during an unprotected loss of flow transient.
The doctoral student will be based in a research unit on innovative nuclear systems at CEA/IRESNE Cadarache, in a dynamic and international environment. Travel to CEA-Saclay and EDF-Chatou is planned during the thesis, as well as participation in international conferences.
Sensitivity calculation in deterministic neutronics: development of methodologies for the lattice phase.
Deterministic neutronics calculations usually rely on a two-step approach, called lattice and core steps. In the first one, the multigroup cross-sections are reduced (condensed over a few energy groups and homogenized over assembly-size regions) using a small subset of the whole system geometrical model (typically, a single subassembly representative of a repeated pattern) in order to reduce the dimensionality of the core calculation step. When those reduced cross-section sets are used for core sensitivity analyses, the impact of the lattice step is usually neglected. For some quantities of interest, this can lead to important discrepancies between the computed sensitivities and the actual ones, since lattice transport calculations are key for carrying the fine-energy local neutron spectrum information and resonance self-shielding effects. There can be an additional concern when those sensitivity calculations are used to provide feedback on nuclear data evaluations, or in the case of similarity studies. In order to address this issue, several approaches are available, such as direct calculations or perturbation theory studies, each representing different trade-offs in terms of cost or complexity.
The goal of this PhD is therefore to explore the state of the art of the domain, ranging from the most brute force approach to the ones based on perturbation theory, with the possibility to propose new methodologies. The implementation of the chosen methodologies in new generation codes (such APOLLO3) will allow eventually to improve the accuracy of sensitivity calculation.
The doctoral student will be based in a reactor physics research unit at CEA/IRESNE in Cadarache, which hosts many students and interns. Post-graduation perspectives include research in nuclear R&D labs and industry.
Methodology for studying the deployment of a fleet of innovative nuclear reactors driven by grid needs and constraints
Power grids are to a society what the blood system is to the human body: the providers of electrical energy essential to the daily life of all the organs of society. They are highly complex systems that have to ensure balance at all times between consumer demand and the power injected onto its lines, via mechanisms on different spatial and temporal scales.
The aim of this thesis is to develop a methodology for optimizing the deployment of innovative nuclear reactors in power grids, adapted to their specific needs and constraints. This approach should be applicable to a wide variety of grids, from island to continental scale, and to various levels of penetration and technologies of Variable Renewable Energies (VREs). Network constraints will need to reflect stability requirements in the short term (location and capacity of inertial reserves, participation in ancillary services), medium term (controllability and load following), and long term (seasonal availability and load factor of generation resources). Innovative nuclear reactors can be of any technology, and are characterized by macroscopic parameters such as load ramp-up/down kinetics, partial power levels, time before restart, cogeneration capacities, etc., as well as the technical and economic data required for dispatching. The aim is then to be able to draw up a profile (i.e. location, power, kinetics) of nuclear reactor fleets guaranteeing stabilized operation of power grids despite a high VREs penetration rate. Two main contributions are expected:
- Academic contribution: to propose an innovative methodology for optimizing the deployment of large-scale energy systems comprising innovative nuclear reactors, by integrating both the physics of power grids and their operational constraints;
- Industrial contribution: develop recommendations for the optimal deployment of innovative nuclear reactors in power systems incorporating VREs, taking into account aspects such as reactor power and inertia, location, reserve requirements for system services, load-following capability and availability.
The PhD student will be based in an innovative nuclear systems research unit. At the intersection of the study of nuclear reactor dynamics, power system physics and optimization, this energetics thesis will offer the PhD student the opportunity to develop in-depth knowledge of tomorrow's energy systems and the issues associated with them.
Impact of power histories on the decay heat of spent nuclear fuel
Decay heat is the energy released by the disintegration of radionuclides present in spent fuel. Precise knowledge of its average value and range of variations is important for the design and safety of spent fuel transport and storage systems. Since this information cannot be measured exhaustively, numerical simulation tools are used to estimate the nominal value of decay heat and quantify its variations due to uncertainties in nuclear data.
In this PhD, the aim is to quantify the variations in decay heat induced by reactor operating data, particularly power histories, which are the instantaneous power of fuel assemblies during their residence in the core. This task presents a particular challenge as the input data are no longer scalar quantities but time-dependent functions. Therefore, a surrogate model of the scientific computing tool will be developed to reduce computation time. The global modeling of the problem will be carried out within a Bayesian framework using model reduction approaches coupled with multifidelity methods. Bayesian inference will ultimately solve an inverse problem to quantify uncertainties induced by power histories.
The doctoral student will join the Nuclear Projects Laboratory of the IRESNE institute at CEA Cadarache. He/she will develop skills in neutron simulation, data science, and nuclear reactors. He/she will be given the opportunity to present his/her work to various audiences and publish it in peer-reviewed journals.
Experimental and theoretical studies of the fission fragment excitation energy and angular momentum generation
The discovery of nuclear fission in 1939 profoundly changed our understanding of nuclear physics. The fission reaction is the splitting of heavy nuclei, such as uranium 235, into two lighter nuclei, together with the release of a large amount of energy. Many years of research have led to the development of nuclear fission models, from which evaluated nuclear data files are derived. These files are essential inputs to reactor simulations; yet, their quality needs to be improved.
This PhD thesis aims to study the generation of angular momentum and the excitation energy of fission fragments from both experimental and theoretical standpoints. These studies will not only improve our understanding of the underlying process and our models, but also enhance the predictive power of simulation tools, particularly those used to predict gamma heating in reactors. Part of the work will involve finalizing the analysis of data acquired as part of a recent thesis. The student will take part in complementary experimental campaigns at the nuclear reactor of the Institut Laue-Langevin (ILL), using the LOHENGRIN spectrometer to measure isomeric ratios and the kinetic energy distributions of fission fragments.
The doctoral student will be based in a nuclear and reactor physics unit. He/she will develop skills in nuclear physics, data analysis, and computer programming. The programming languages used will be C++ and Python. Professional perspectives include academic research, R&D organisations, nuclear industry, and possibly also data scientist positions.
Modeling of Critical Heat Flux Using Lattice Boltzmann Methods: Application to the Experimental Devices of the RJH
The Lattice Boltzmann Methods (LBM) are numerical techniques used to simulate transport phenomena in complex systems. They allow for the modeling of fluid behavior in terms of particles that move on a discrete grid (a "lattice"). Unlike classical methods, which directly solve the differential equations of fluids, LBM simulates the evolution of distribution functions of fluid particles in a discrete space, using propagation and collision rules. The choice of the lattice in LBM is a crucial step, as it directly affects the accuracy, efficiency, and stability of the simulations. The lattice determines how fluid particles interact and move within space, as well as how the discretization of space and time is performed.
LBM methods exhibit natural parallelism properties, as calculations at each grid point are relatively independent. Although classical CFD methods based on the solution of the Navier-Stokes equations can also be parallelized, the nonlinear terms can make parallelism more difficult to manage, especially for models involving turbulent flows or irregular meshes. Therefore, LBM methods allow, at a lower computational cost, to capture complex phenomena. Recent work has shown that it is possible, with LBM, to reproduce the Nukiyama cooling curve (boiling in a vessel) and thus accurately calculate the critical heat flux. This flux corresponds to a mass boiling, known as the boiling crisis, which results in a sudden degradation of heat transfer.
The critical heat flux is a crucial issue for the Jules Horowitz Reactor, as experimental devices (DEX) are cooled by water in either natural or forced convection. Therefore, to ensure proper cooling of the DEX and the safety of the reactor, it is essential to ensure that, within the studied parameter range, the critical heat flux is not reached. It must therefore be determined with precision.
In the first part of the study, the student will define a lattice to apply LBM methods on an RJH device in natural convection. The student will then consolidate the results by comparing them with available data. Finally, exploratory calculations in forced convection (from laminar to turbulent flow) will be conducted.
Innovative modeling for multiphysics simulations with uncertainty estimates applied to sodium-cooled fast reactors
Multiphysics modeling is crucial for nuclear reactor analysis, yet uncertainty propagation across different physical domains—such as thermal, mechanical, and neutronic behavior—remains underexplored due to its complexity. This PhD project aims to address this challenge by developing innovative methods for integrating uncertainty quantification into multiphysics models.
The key objective is to propose optimal modeling approaches tailored to different precision requirements. The project will explore advanced techniques such as reduced-order modeling and polynomial chaos expansion to identify which input parameters most significantly impact reactor system outputs. A key aspect of the research is the comparison between "high-fidelity" models, developed using the CEA reference simulation tools, and "best-estimate" models designed for industrial use. This comparative analysis will highlight how these errors propagate through different models and simulation approaches.
The models will be validated using experimental data from SEFOR, a sodium-cooled fast reactor. These experiments provide valuable benchmarks for testing multiphysics models in realistic reactor conditions. This research directly addresses the growing need for reliable, efficient modeling tools in the nuclear industry, aiming to improve reactor safety and performance.
The candidate will work in a dynamic environment at the CEA, benefiting from access to advanced simulation resources and opportunities for collaboration with other researchers and PhD students. The project offers the possibility of presenting results at national and international conferences, with strong career prospects in nuclear reactor design, safety analysis, and advanced simulation.
Building a new effective nuclear interaction model and propagating statistical errors
At the very heart of any « many-body » method used to describe the fundamental properties of an atomic nucleus, we find the effective nucleon-nucleon interaction. Such an interaction should be capable of taking into account the nuclear medium effects. In order to obtain it, one has to use a specific fitting protocol that takes into account a variety of nuclear observables such as radii, masses, the centroids of the giant resonances or the properties of the nuclear equation of state around the saturation density.
A well-known model of the strong interaction is the Gogny model. It is a linear combination of coupling constants and operators, plus a radial form factor of the Gaussian type [1]. The coupling constants are determined via a fitting protocol that typically uses the properties of spherical nuclei such as 40-48Ca, 56Ni, 120Sn and 208Pb.
The primary goal of this thesis is to develop a consistent fitting protocol for a generic Gogny interaction in order to access some basic statistical information, such as the covariance matrix and the uncertainties on the coupling constants, in order to be able to perform a full statistical error propagation on some selected nuclear observables calculated with such an interaction [2].
After having analysed the relations between the model parameters and identified their relative importance on how well observables are reproduced, the PhD candidate will explore the possibility of modifying some terms of the interaction itself such as the inclusion of a real three-body term or beyond mean-field effects.
The PhD candidate will work within a nuclear physics group at CEA/IRESNE Cadarache. The work will be done in close collaboration with CEA/DIF. Employment perspectives are in academic research and nuclear R&D labs.
[1] D. Davesne et al. "Infinite matter properties and zero-range limit of non-relativistic finite-range interactions." Annals of Physics 375 (2016): 288-312.
[2] T. Haverinen and M. Kortelainen. "Uncertainty propagation within the UNEDF models." Journal of Physics G: Nuclear and Particle Physics 44.4 (2017): 044008.