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.).
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.
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.
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.
Measurement and evaluation of the energy dependence of delayed neutron data from 239Pu
This PhD proposal aims to measure and characterize the delayed neutron emissions from the fission of 239Pu. This actinide is involved in various reactor concepts, and the nuclear data available remains insufficient, particularly with fast neutrons. The project has a strong experimental focus, with multiple measurement campaigns at the MONNET electrostatic accelerator from JRC Geel, in which the candidate will actively participate.
The first phase focuses on the intercomparison of the neutron flux measurement methods (dosimetry, fission chamber, long-counter detector and recoil proton scintillator) which will be confronted to Monte-Carlo simulations of neutron emission from charged particle interactions (D+T, D+D, p+T). This work will ensure proper neutron flux characterization, a crucial step for the project.
Next, the candidate will replicate the delayed neutron measurements for ²³8U using an existing target in order to verify the results from a 2023 experimental campaign.
Finally, the candidate will measure the delayed neutron yields and group abundances for ²³?Pu in a neutron energy range from 1 to 8 MeV. The objective is to produce an energy-dependent evaluation, integrated into an ENDF file, to be tested on reactor calculations (beta-eff, power transients, absorber efficiency calibration, etc.). These measurements will complement a thermal spectrum study conducted at ILL in 2022, forming a coherent model for ²³?Pu from 0 to 8 MeV.
This project will contribute to the OECD/NEA's JEFF-4 nuclear data file, addressing a strong demand from the nuclear industry (highlighted by the IAEA) to improve the precision of multiplicity measurements and delayed neutron kinetic parameters, thus enhancing reactor safety and reducing safety margins.
Integral measurement of fission products capture cross-section using a combination of oscillation and activation techniques
This thesis is proposed as part of the POSEIDON (Fission Product Oscillation Experiments for Improving Depletion Calculations) project that deals with the integral measurement of the neutron capture and scattering cross-sections of the main fission products contributing to the reactivity loss in irradiated fuel. It consists of measuring the reactivity effect of separated isotope samples using a pile oscillation device, coupled with neutron activation measurements, in three different core spectral configurations : thermal, PWR and epithermal.
Part of the work will be done at CEA IRESNE in Cadarache and part at the Research Center of the Czech Republic, CV Rez. The PhD student will be involved in testing and optimizating the oscillation device that is currently being designed, as well as performing the measurements in the LR-0 Czech experimental reactor. The work at Cadarache will be on the analysis of the measurements with MC simulation tools. Functionalities needed for data analysis will require additional developments of the codes by the student.
The expected impact is a better prediction of the reactivity loss in reactor cores as a function of burn-up. Indeed, even with the most recent international nuclear data libraries, there is an important bias in the estimation of this reactivity loss.
The PhD student will develop competences in experimental and theoretical neutronics. Following job opportunities include R&D laboratories and nuclear industry.